Technical Evaluation Report

For the U.S. Department of Energy West 

Valley Draft Waste Incidental to 

Reprocessing Evaluation for the 

Concentrator Feed Makeup Tank and the 

Melter Feed Hold Tank 

Final Report 

U.S. Nuclear Regulatory Commission 

Office of Federal and State Materials 

 and Environmental Management Programs 

Washington, DC 20555-0001 

October 2012 
i 

TABLE OF CONTENTS 

EXECUTIVE SUMMARY .............................................................................................................. iii

ABBREVIATIONS/ACRONYMS ................................................................................................... v

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

1.1 Background ........................................................................................................ 1-1

1.2 Waste Incidental to Reprocessing Evaluation .................................................... 1-4

1.3 Previous NRC Reviews ...................................................................................... 1-5

1.4 The NRC Review Approach ............................................................................... 1-5

2.0 THE WASTE HAS BEEN PROCESSED TO REMOVE KEY RADIONUCLIDES 

TO THE MAXIMUM EXTENT THAT IS TECHNICALLY AND ECONOMICALLY 

PRACTICAL ................................................................................................................... 2-1

2.1 The DOE’s Characterization of Waste Inventory ................................................ 2-1

2.2 NRC Evaluation of Waste Inventory ................................................................... 2-3

2.3 DOE Selection of Key Radionuclides ................................................................. 2-6

2.4 NRC Evaluation of Key Radionuclides ............................................................... 2-7

2.5 Removal to the Maximum Extent Technical and Economically Practical ........... 2-7

3.0 THE WASTE WILL BE MANAGED TO MEET SAFETY REQUIREMENTS 

COMPARABLE TO THE PERFORMANCE OBJECTIVES OF 10 CFR PART 61, 

SUBPART C ................................................................................................................... 3-1

3.1 Protection of Individuals during Operations ........................................................ 3-1

3.2 Waste Form Stability .......................................................................................... 3-3

4.0 THE WASTE WILL NOT EXCEED CLASS C CONCENTRATION LIMITS AND 

WILL BE MANAGED IN ACCORDANCE WITH DOE REQUIREMENTS AS LLW ........ 4-1

4.1 The DOE’s Waste Classification Approach ........................................................ 4-1

4.2 NRC Evaluation of the DOE’s Waste Classification Approach ........................... 4-3

5.0 CONCLUSIONS ............................................................................................................. 5-1

5.1 NRC Review: Waste has been Processed to Remove Key Radionuclides 

to the Maximum Extent that is Technically and Economically Practical ............. 5-1

5.2 NRC Review: Waste will be Managed to Meet Safety Requirements 

Comparable to the Performance Objectives of 10 CFR 61, Subpart C .............. 5-2

5.3 NRC Review: Waste will not Exceed Class C Concentration Limits and 

will be Managed in Accordance with DOE Requirements as LLW ..................... 5-3

6.0 CONTRIBUTORS .......................................................................................................... 6-1

7.0 REFERENCES ............................................................................................................... 7-1 
ii 

LIST OF TABLES 

Table 1-1: Applicable NRC Review Guidance and DOE-M 435.1-1 Criteria ............................ 1-5

Table 2-1: Total Activity Estimates as of October 2004 (DOE, 2012b) ..................................... 2-2

Table 2-2: Radionuclide Scaling Factors (Ratios to Cs-137) (DOE, 2011b) ............................. 2-6

Table 2-3: Key Radionuclides ................................................................................................... 2-6

Table 2-4: Vessel Flushing Effectiveness in Terms of Estimated Cs-137 Removal ............... 2-10

Table 3-1: Comparison of NRC, DOE, and Texas Radiological Protection Standards ............. 3-2

Table 4-1: Concentrator Feed Makeup Tank Waste Concentration Results ............................. 4-4

Table 4-2: Melter Feed Holder Tank Waste Concentration Results.......................................... 4-5

LIST OF FIGURES

Figure 1-1: HLW Storage and Vitrification Storage Facility ....................................................... 1-2

Figure 1-2: Vitrification Process Flow [Adapted from DOE (2011a)] ......................................... 1-3

 
iii 

EXECUTIVE SUMMARY 

In June 2012, the U.S. Department of Energy (DOE) submitted the Draft Waste Incidental to 

Reprocessing Evaluation for the Concentrator Feed Makeup Tank and the Melter Feed Hold 

Tank to the U.S. Nuclear Regulatory Commission (NRC) for consultative review. The DOE’s 

evaluation assesses whether the Concentrator Feed Makeup Tank (CFMT) and the Melter Feed 

Hold Tank (MFHT) at the West Valley Demonstration Project (WVDP) meet the waste incidental 

to reprocessing (WIR) criteria of Section II.B (2) (a) in DOE-Manual 435.1-1 (DOE-M 435.1-1), 

Radioactive Waste Management, which accompanies DOE Order 435.1-1. Demonstration that 

the criteria in DOE-M 435.1-1 are met allows DOE to dispose of the used components (UC) 

from both CFMT and MFHT offsite as low-level waste (LLW). This Technical Evaluation Report 

(TER) presents information on the DOE’s evaluation process, the applicable review criteria, and 

the NRC’s review approach, as well as the NRC’s analysis and conclusions with respect to 

whether there is technical sufficiency to demonstrate that the DOE’s proposed approach can 

meet the criteria of DOE-M 435.1-1 for determining that waste is not high-level waste (HLW). 

The NRC staff has performed a technical review to assess whether the draft evaluation is 

technically sufficient to demonstrate that the UC meet the criteria in DOE M 435.1-1. The NRC 

has conducted this consultative review at the request of the DOE in accordance with 

Interagency Agreement DE-EM0001931. 

Per the specific requests of the interagency agreement (IA), the NRC’s review was focused on 

assessing whether the methodology that the DOE employed contained sound technical 

assumptions, analyses, projections, and conclusions. The NRC employed the relevant review 

procedures in Chapter 3, 6, and 7 of NUREG-1854 (NRC, 2007), to complete its review. The 

NRC's review focused on the following general topics, as they relate to the criteria in 

DOE-M 435.1-1: 

• Waste characterization, 

• Waste form stability, 

• Waste classification, 

• Removal of radionuclides to the maximum extent technically and economically practical, 

and 

• Operational radiation protection. 

Section II.B (2)(a) of DOE Manual 435.1-1 states the following: 

II. B. Waste Incidental to Reprocessing. 

Waste resulting from reprocessing spent nuclear fuel that is determined to be incidental to 

reprocessing is not high-level waste, and shall be managed under DOE’s regulatory authority 

in accordance with the requirements for transuranic waste or low-level waste, as appropriate. 

When determining whether spent nuclear fuel reprocessing plant wastes shall be managed as 

another waste type or as high-level waste, either the citation or evaluation processes 

described below shall be used: 

(1) Citation. Waste incidental to reprocessing by citation includes spent nuclear fuel 

reprocessing plant wastes that meet the description included in the Notice of Proposed 

Rulemaking (34 FR 8712) for proposed Appendix D, 10 CFR Part 50, Paragraphs 6 and 7. 

These radioactive wastes are the result of reprocessing plant operations, such as, but not 
iv 

limited to: contaminated job wastes including laboratory items such as clothing, tools, and 

equipment. 

(2) Evaluation. Determinations that any waste is incidental to reprocessing by the evaluation 

process shall be developed under good record-keeping practices, with an adequate quality 

assurance process, and shall be documented to support the determinations. Such wastes 

may include, but are not limited to spent nuclear fuel reprocessing plant wastes that: 

(a) Will be managed as low-level waste and meet the following criteria: 

• Have been processed, or will be processed, to remove key radionuclides to the 

maximum extent that is technically and economically practical; and 

• Will be managed to meet safety requirements comparable to the performance 

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

• 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-M 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 LLW as set out in §61.55, Waste Classification; or will meet alternative 

requirements for waste classification and characterization as DOE may 

authorize. 

Based on the information provided by DOE and its associated contractor, West Valley 

Environmental Services, LLC, in the draft evaluation dated June 20, 2012, and letter dated 

September 20, 2012 (RAI response), the NRC staff has concluded that the DOE’s draft 

evaluation is technically sufficient to demonstrate that the UC meet the NRC-reviewed portions 

of the criteria in DOE M-435.1-1. 

DOE plans to ship the packaged UC to a suitable offsite LLW disposal facility, such as the 

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

Control Specialists (WCS) facility in Texas for disposal. 

In addition to the NRC, the DOE has solicited review and comment from State officials and 

members of the public on the draft WIR evaluation and associated documentation. The DOE 

will make a final determination of whether the UC are or are not HLW after consideration of the 

NRC’s comments and any State and public comments on DOE’s draft evaluation. 

If DOE decides to dispose of the UC at the WCS facility, the component waste packages would 

be disposed of as Class C LLW in the Federal Facility Waste Disposal Facility. The NRC’s 

review does not address the long-term performance of the site ultimately selected for disposal of 

the UC. Therefore, the NRC's review will not address the long-term performance or long-term 

stability of this disposal site, but will focus only on the general topics listed above. Also noted in 

the IA, the NRC’s review does not include independent review of the key radionuclide list 

because the decontamination method used and the other methods considered for additional 

decontamination were bulk cleaning methods that remove all radionuclides present in similar 

proportions; that is, the methods do not target specific radionuclides. 
v 

ABBREVIATIONS/ACRONYMS 

AEC Atomic Energy Commission 

BTP Branch Technical Position on Concentration Averaging 

CFMT Concentrator Feed Makeup Tank 

CFR Code of Federal Regulations 

CNWRA Center for Nuclear Waste Regulatory Analysis 

DOE U.S. Department of Energy 

DOE-EM U.S. Department of Energy – Environmental Management program 

EIS Environmental Impact Statement 

GTCC Greater than Class C 

HLW High-Level Waste 

IA Interagency Agreement 

LLW Low-Level Waste 

MFHT Melter Feed Hold Tank 

NNSS Nevada National Security Site Area 5 Radioactive Waste Management Site 

NFS Nuclear Fuel Services 

NRC U.S. Nuclear Regulatory Commission 

NYSERDA New York State Energy Research And Development Authority 

PA Performance Assessment 

PUREX Plutonium - Uranium Extraction 

RAI Request for Additional Information 

SOW Statement of Work 

TER Technical Evaluation Report 

TRU Transuranic Waste 

UC Used Components 

WCS Waste Control Specialist 

WIR Waste Incidental to Reprocessing 

WMA Waste Management Area 

WNYNSC Western New York Nuclear Service Center 

WVDP West Valley Demonstration Project 

WVNS West Valley Nuclear Services Company 
1-1 

1.0 INTRODUCTION 

1.1 Background 

The WVDP is located in western New York State, about 50 km (30 miles) south of Buffalo, New 

York. The WVDP facilities occupy a security-fenced area of about 0.676 km2

 (167 acres) within 

the 13.51-km2

 (3,338-acre) Western New York Nuclear Service Center (WNYNSC) located 

primarily in the town of Ashford in northern Cattaraugus County. 

To date, the West Valley site in West Valley, New York is the first and only commercial 

reprocessing plant to operate in the United States. From 1966 to 1972, Nuclear Fuel Services 

(NFS) reprocessed 640 metric tons of spent fuel under an Atomic Energy Commission (AEC) 

license. Approximately four years after shutting down, NFS returned control of the facilities to 

the site owner, New York State Energy Research and Development Authority (NYSERDA). 

Operations at the facility resulted in approximately 2.3 million liters (600,000 gallons) of liquid 

high-level waste (HLW), which was stored below ground in carbon-steel tanks. 

Figure 1-1 shows the perimeter of the WV site and the relationship of the underground tanks to 

the Vitrification Facility, where the Vitrification Melter was housed. Tank 8D-2 stored 

approximately 2.12 million liters (560,000 gallons) of neutralized PUREX wastes, which were 

composed of a bottom sludge layer containing insoluble hydroxides and other salts that 

precipitated out of solution, covered by a layer of alkaline liquid (supernatant) rich in sodium 

nitrate. Tank 8D-2 wastes were neutralized by adding sodium hydroxide to the nitric acid-based 

stream during the reprocessing of uranium fuel using the PUREX process. 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. In 

addition, neutralizing the wastes reduced the possibility of corrosion of the carbon steel tank. 

Therefore, the HLW was not homogeneous but was comprised of supernatant (liquid) and 

sludge (solids). Tank 8D-4 stored approximately 46,000 liters (12,000 gallons) of acidic highlevel radioactive liquid waste produced in reprocessing thorium-enriched uranium fuel using the 

THOREX process; this waste was stored without being neutralized. 

In 1980, the West Valley Demonstration Project Act (WVDP Act) was passed. The WVDP Act 

made DOE responsible for solidifying the liquid HLW stored in underground tanks, disposing of 

the waste created by solidification, and decontaminating and decommissioning the facilities 

used during the process. Under the WVDP Act, the DOE entered into a Cooperative Agreement 

with NYSERDA (DOE and NYSERDA, 1981) that established the framework for cooperative 

implementation of the WVDP Act. Under the agreement, the DOE was provided exclusive use 

and possession of a portion of the West Valley site. A supplement to the Cooperative 

Agreement between the two agencies set forth special provisions for the preparation of a joint 

Environmental Impact Statement (EIS) [DOE and NYSERDA, 1981]. 

In 1981, the DOE and the NRC entered into a Memorandum of Understanding that established 

specific agency responsibilities and arrangements for informal review and consultation by the 

NRC (DOE and NRC, 1981). Because NYSERDA holds the license and title to the West Valley 

site, the NRC put the technical specifications of the license in abeyance to allow the DOE to 

carry out the responsibilities of the WVDP Act. 
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1-4 

1.2 Waste Incidental to Reprocessing Evaluation 

The DOE is evaluating whether the CFMT and the MFHT waste packages at the WVDP meet 

the DOE’s waste incidental to reprocessing criteria of DOE-M 435.1-1, Radioactive Waste 

Management, in order to dispose of the UC offsite as LLW. DOE plans to ship the packaged 

CFMT and the MFHT components to either the Nevada National Security Site (NNSS, formerly 

the Nevada Test Site) in Nevada or the Waste Control Specialists (WCS) facility in Texas for 

disposal as LLW. 

Consistent with the Department‘s policy (as per DOE-M 435.1-1), the DOE has reached out to 

the NRC, as well as State officials and members of the public, to solicit review and comment on 

a draft WIR evaluation. The DOE will make a final determination of whether the CFMT and the 

MFHT UC are or are not HLW after considering comments provided by the NRC (as 

documented in this TER), the public, and any State. 

Section II.B (2)(a) of DOE-M 435.1-1 states the following: 

II. B. Waste Incidental to Reprocessing. 

Waste resulting from reprocessing spent nuclear fuel that is determined to be incidental 

to reprocessing is not high-level waste, and shall be managed under DOE’s regulatory 

authority in accordance with the requirements for transuranic waste or low-level waste, 

as appropriate. When determining whether spent nuclear fuel reprocessing plant wastes 

shall be managed as another waste type or as high-level waste, either the citation or 

evaluation processes described below shall be used: 

(1) Citation. Waste incidental to reprocessing by citation includes spent nuclear fuel 

reprocessing plant wastes that meet the description included in the Notice of Proposed 

Rulemaking (34 FR 8712) for proposed Appendix D, 10 CFR Part 50, Paragraphs 6 and 

7. These radioactive wastes are the result of reprocessing plant operations, such as, but 

not limited to: contaminated job wastes including laboratory items such as clothing, tools, 

and equipment. 

(2) Evaluation. Determinations that any waste is incidental to reprocessing by the 

evaluation process shall be developed under good record-keeping practices, with an 

adequate quality assurance process, and shall be documented to support the 

determinations. Such wastes may include, but are not limited to; spent nuclear fuel 

reprocessing plant wastes that: 

(a) Will be managed as low-level waste and meet the following criteria: 

• Have been processed, or will be processed, to remove key radionuclides 

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

• Will be managed to meet safety requirements comparable to the 

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

Performance Objectives; and 

• 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-M 435.1-1], provided the waste will be incorporated in 

a solid physical form at a concentration that does not exceed the 
1-5 

applicable concentration limits for Class C LLW as set out in §61.55, 

Waste Classification; or will meet alternative requirements for waste 

classification and characterization as DOE may authorize. 

1.3 Previous NRC Reviews 

In 2011, DOE consulted with the NRC during development of the West Valley Demonstration 

Project Draft Waste Incidental to Reprocessing Evaluation for the Vitrification Melter

(DOE, 2011a). NRC completed this review under a similar IA with the DOE (DOE, 2010). NRC 

submitted its review to DOE in a TER on September 30, 2011 (NRC, 2011a). 

1.4 The NRC Review Approach 

NRC’s review of the draft WIR evaluation was performed in accordance with the specific 

requests as described in the Interagency Agreement (IA), DE-EM0001931. To avoid confusion, 

this IA was very specific in its requests in an attempt to avoid any overlap with analogous 

programs either between NRC and DOE-EM, or other NRC/DOE interactions at the West Valley 

site. 

Table 1-1: Applicable NRC Review Guidance and DOE-M 435.1-1 Criteria 

General Review Topic NUREG-1854 Chapter 435.1 

Criterion

Waste characterization Chapter 3: Radionuclide Removal and 

Concentration Limits 1 

Waste form stability Chapter 7: Site Stability, Waste Stability, 

and Facility Stability 2 and 3 

Waste classification Chapter 3: Radionuclide Removal and 

Concentration Limits 3 

Removal of radionuclides to the 

maximum extent technically and 

economically practical 

Chapter 3: Radionuclide Removal and 

Concentration Limits 3 

Operational radiation protection Chapter 6: Protection of Individuals 

During Operations 2 

 

Pursuant to DOE’s specific request, the NRC has reviewed the draft evaluation for the CFMT 

and MFHT components at the WVDP site to provide a technical opinion on whether the draft 

evaluation is technically sufficient to demonstrate that the WVDP CFMT and MFHT UC meet the 

criteria in DOE-M 435.1-1, which accompanies DOE-O 435.1-1. 
1-6 

The NRC’s review of the draft WIR evaluation focused on assessing whether the methodology 

the DOE employed was based on sound technical assumptions, analyses, and projections, and 

resulted in acceptable conclusions. The NRC employed the relevant review procedures in 

Chapter 3, 6, and 7 of NUREG-1854 (NRC, 2007), to complete its review of the general topics 

identified in Table 1-1. The NRC reviewed the general topics identified in Table 1-1 as they 

relate to the criteria in DOE-M 435.1-1 (followed by applicable chapter of NUREG-1854). 

The DOE plans to dispose of the CFMT and MFHT UC off-site, at a LLW disposal site such as 

the NNSS or WCS, neither of which are licensed or otherwise regulated by the NRC (but are 

instead licensed and regulated by DOE — in the case of NNSS — or by an NRC Agreement 

State, the State of Texas in the case of WCS). As per the IA, the NRC’s review does not 

address the long-term performance of the disposal site or the sufficiency of the waste 

acceptance criteria for the potential disposal facilities being considered for disposal of the CFMT 

and MFHT UC. Therefore, the NRC's review will not address the long-term performance or 

long-term stability of this disposal site, but will focus only on the general topics listed above. 

The IA also notes that the NRC’s review does not include independent review of the key 

radionuclide list because the decontamination method used and the methods considered for 

additional decontamination were bulk-cleaning methods that remove all radionuclides present in 

similar proportions; that is, the methods do not target specific radionuclides. 
2-1 

2.0 THE WASTE HAS BEEN PROCESSED TO REMOVE KEY 

RADIONUCLIDES TO THE MAXIMUM EXTENT THAT IS TECHNICALLY 

AND ECONOMICALLY PRACTICAL 

This section of the NRC staff’s technical review covers Section II.B(2)(a)(1) of DOE-M 435.1-1– 

that the waste has been processed, or will be processed, to remove key radionuclides to the 

maximum extent that is technically and economically practical. 

2.1 The DOE’s Characterization of Waste Inventory 

The DOE estimates the amount of residual material in the CFMT and the MFHT after 

decontamination to be approximately 3.59 TBq (97 Ci) and 3.81 TBq (103 Ci) respectively. The 

inventory of the residual material, as estimated in 2004, is described in the waste 

characterization (WMG, 2011) and the standardized waste profile (CHBWV, 2011a). 

The DOE’s characterization is based on measured gamma dose rates using a Ludlum Model 

1337 Geiger Mueller shielded detector probe (WMG, 2011) and analytical results from samples. 

The measured gamma dose rates were attributed solely to Cs-137 to estimate Cs-137 activity. 

OAO-GCCP-A geometry models were developed for each tank to determine a dose-to-curie 

conversion factor (DCF) for Cs-137, while the amounts of other radionuclides were estimated 

using scaling factors based on analytical results from the laboratory samples. RADmanTM

software was applied to calculate the activities of other radionuclides using the scaling factors 

and Cs-137 activity. 

Nine dose rate measurements (although ten were recorded) of the CFMT (WVNSCO, 2004c) 

were averaged to estimate Cs-137 activity. Five analytical samples were utilized in 

characterizing the CFMT: 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 

in July 2003 after completion of vitrification (WMG, 2011). Each of the batch samples was 

analyzed nine times and the average was taken to calculate a representative value for each of 

the four batch samples. The July 2003 liquid sample was analyzed three times. The DOE 

originally calculated the geometric mean of seven values (4 batch sample values and the three 

analyses of the liquid sample) in determining the scaling factors, which the DOE later revised as 

discussed in the following section, treating the 3 liquid analyses as a single sample. 

Nine dose rate measurements (although twelve were recorded) of the MFHT (WVNSCO, 2004c) 

were averaged to estimate Cs-137 activity. Six analytical samples were utilized in 

characterizing the MFHT: the same four batch sample values used for characterizing the CFMT 

plus two glass shard samples taken from the two evacuated canisters used to remove molten 

glass from the vitrification melter (WMG, 2011). The DOE analyzed the two glass shards three 

times each and averaged the values, the results of which are documented in WMG (2004a). 

The DOE calculated the geometric mean of six values (4 batch sample values and two glass 

samples) in determining the scaling factors. 
2-2 

Table 2-1: Total Activity Estimates as of October 2004 (DOE, 2012b) 

Nuclide CFMT 2 MFHT Melter 

Activity (Ci) 1 Activity (Ci) 1 Activity (Ci) 1

C-14 NA 3.98x10-4 2.12x10-2

K-40 NA 1.54x10-3 8.19x10-2

Mn-54 NA 1.67x10-3 8.57x10-2

Co-60 4.1x10-3 1.58x10-3 8.33x10-2

Ni-63 NA 1.89x10-2 1.01 

Sr-90 3.9 5.34 2.47x10+2

Zr-95 NA 3.72x10-2 1.65 

Tc-99 1.8x10-3 8.34x10-4 1.11x10-2

Cs-137 9.5x10+1 9.71x10+1 4.31x10+3

Eu-154 5.6x10-2 3.18x10-2 1.21 

Np-237 5.6x10-5 7.26x10-5 6.20x10-3

Pu-238 6.9x10-3 9.19x10-3 6.84x10-1

Pu-239 1.5x10-3 2.28x10-3 1.59x10-1

Pu-240 1.5x10-3 1.74x10-3 1.21x10-1

Pu-241 1.4x10-2 5.88x10-2 3.12 

Pu-242 NA NA NA 

Am-241 3.7x10-2 4.33x10-2 3.00 

Am-242m NA NA NA 

Am-243 3.3x10-4 3.93x10-4 3.50x10-2

Cm-242 3.9x10-4 3.42x10-4 7.33x10-2

Cm-243 3.2x10-3 2.84x10-4 1.68x10-2

Cm-244 3.2x10-3 7.36x10-3 4.35x10-1

Cm-245 NA NA NA 

Cm-246 NA NA NA 

Th-228 NA 7.79x10-4 4.09x10-2

Th-230 NA 6.84x10-6 3.65x10-4

Th-232 1.9x10-6 7.53x10-6 4.01x10-4

U-232 1.0x10-4 9.40x10-4 5.01x10-4

U-233 3.2x10-5 3.86x10-4 NA 

U-234 3.2x10-5 1.84x10-4 9.81x10-3

U-235 NA 7.07x10-6 3.76x10-4

U-236 NA 2.12x10-5 NA 

U-238 4.8x10-6 4.23x10-5 2.25x10-3

TOTAL 99 103 4569 

1

 Conversion to Bq from curies, multiply by 0.37 TBq/Ci. 2

 From WMG 2011 as of October 1, 2004 as revised (Kurasch, 2012). 
2-3 

Using the gamma dose rates and the scaling factors, the DOE estimated that the total activity 

remaining in the MFHT as of February 9, 2004 was 3.81 TBq (103 Ci), 95 percent of which was 

Cs-137. The total activity in the CFMT was 3.68 TBq (99 Ci), 96 percent of which was Cs-137. 

The activities of the other radionuclides were inferred from Cs-137 using scaling factors. The 

DOE’s estimate of the inventories of the vessels (also listed in Table 2-2 and 2-4 of the draft 

WIR evaluation) along with the inventory of the Melter (DOE, 2011a) for comparison purposes 

are replicated in Table 2-1. 

2.2 NRC Evaluation of Waste Inventory 

In this part of the review, the NRC staff evaluated the physical and chemical form of the waste, 

previous inventory estimates, data quality objectives, homogeneity, sampling plan for 

characterization, volume and mass estimates, and uncertainties. 

The NRC staff has reviewed information about waste generation and treatment activities and 

finds that the predicted physical and chemical forms of radionuclides, as described by DOE, are 

consistent with the properties of contributing waste streams and treatment processes. 

The NRC staff reviewed historical inventories and compared those to the radionuclides 

considered for waste classification purposes. NRC staff notes that some of the radionuclides 

listed in the inventory are denoted with an “NA.” DOE clarified that certain transuranics that 

were listed in the sludge/supernatant inventory (Pu-242, Am-242m, Cm-245, Cm-246), were 

present in such small quantities in the sludge/supernatant, so they were also listed in the final 

residual inventory. However, since they are present in small quantities, inclusion of these 

transuranics does not significantly impact the sum of fractions for waste classification as 

discussed in Section 4.2. 

The NRC staff compared the inventory of the Vitrification Melter to that of each of the vessels to 

identify any differences between the lists of radionuclides reported. The NRC staff noted in its 

review that certain radionuclides (C-14, K-40, Mn-54, Ni-63, Zr-95, Th-228, Th-230, U-235, and 

U-236) are listed with concentrations in the Melter and the MFHT, but not in the CFMT. Since 

the same material (with few exceptions) passed through both vessels and was eventually 

formed into HLW glass (in the Melter), NRC staff would have expected the list of radionuclides 

to be the same. During its review, the NRC asked for additional information regarding the 

differences in the radionuclides that are listed in the MFHT and the CFMT. DOE clarified that 

the MFHT contains several radionuclides not included in the inventory for the CFMT because 

the glass shard samples that were used in characterizing the MFHT were analyzed for these 

radionuclides while the liquid samples used for the CFMT were not analyzed for these 

radionuclides. Records describing the basis for the selected samples (and why certain 

radionuclides were analyzed while others were not) were not available, but the CFMT liquid 

sample and the glass shards were likely chosen to characterize each vessel since they were the 

last materials contained in the CFMT and the MFHT respectively. 

DOE stated that while different sample analytical data sets were used in characterizing the two 

vessels based on the judgment of the site contractor when the vessels were characterized in 

2004, it would also be reasonable to conclude that the radionuclide distributions in these 

vessels are the same of those in the Vitrification Melter. Since the scaling factors used for the 

Melter are slightly higher, utilizing the Melter scaling factors as opposed to the scaling factors 

developed for the vessels would be more conservative. In order to determine whether use of 

the scaling factors developed for the Vitrification Melter would produce significantly different 
2-4 

results, DOE calculated the residual radioactivity in each vessel using the Vitrification Melter 

scaling factors (WMG, 2004a) combined with the Cs-137 estimates for the CFMT and MFHT in 

the WMG characterization report (WMG, 2011). The results were as follows: (1) the CFMT 

estimate would be 3.77 TBq (102 Ci) compared to 3.68 TBq (99.4 Ci) and (2) the MFHT 

estimate would be 3.81 TBq (103 Ci) in both cases. Because the difference in this comparison 

is not significant, the NRC staff finds the list of radionuclides presented in the inventory in 

Table 2-1 acceptable. 

The NRC staff evaluated the sampling analysis and data quality objectives to verify that relevant 

data quality objectives were met. The DOE states that data were validated in accordance with 

the requirements of the Characterization Management Plan for the Facility Characterization 

Project (Michalczak, 2004), which includes data quality objectives. Furthermore, the 2004 

characterization was reviewed by the site contractor, WVNSC, and DOE incorporated 

recommendations resulting from this review. The NRC staff concluded that the approaches 

taken for data verification as outlined in the Characterization Management Plan for the Facility 

Characterization Project were appropriate, because it follows generally accepted practices. 

Regarding homogeneity of the waste inventory, the NRC staff notes that the processes used in 

both tanks kept the material relatively homogeneous. The CFMT received pretreated batches of 

liquid HLW. Although there was some variation from batch to batch, each new batch was mixed 

with the heel of the previous batch and the combined contents were sampled and treated until it 

met certain specifications consistent with those required in the vitrification process. The MFHT 

held the HLW slurry for delivery to the Melter and included an agitation system that was used to 

maintain homogeneity of the slurry. DOE attributes slight variations in the dose rates, which are 

due to the primary gamma emitter Cs-137, at different locations on the outside of each of the 

vessels due to small variations in the amounts of Cs-137 in the vessel interior. This is 

supported by photographic evidence provided in Chapter 4 of the draft WIR evaluation 

(DOE, 2012b). However, the general consistency of the of the dose rates measured on the 

sides of the vessels (1.17 R/h to 2.25 R/h for the MFHT) as shown in the survey record 

(WVNSCO 2004a) support the DOE’s overall assumption of generally uniform distribution of 

residual waste. 

During the review, NRC asked DOE to provide a technical basis for the number and location of 

dose rate measurements and to describe how the number and location of required samples are 

based on assumptions about the heterogeneity of the residual waste within the tanks after 

cleaning. DOE explained that the “survey plans provided for nine measurements along the side 

of each vessel at intervals of approximately one foot, along with four measurements on the top 

of each vessel (if possible)” (DOE, 2012c). The sampling plan was based on the assumption 

that the residual contamination left after flushing was uniformly distributed on the vessel interior 

surfaces based on the results of visual inspections. DOE acknowledged that this assumption is 

somewhat contradictory to the simplifying assumption used to determine cleaning effectiveness 

(that prior to flushing, the material covered only the upper one-third of the vessels). DOE further 

clarified by stating that it is reasonable to expect that some amount of residual material was 

present in the lower two-thirds of the vessels prior to flushing. DOE explained that only one 

measurement was taken on top of the CFMT and 3 on top of the MFHT due to limited 

accessibility. Only nine side measurements of each vessel were used in the dose 

concentrations due to the assumptions of the QAD geometry model (DOE, 2012c). Given this 

information, along with technical drawings of the vessels showing the inaccessibility of areas on 

the top of the vessels, the NRC staff finds the number and location of dose rate measurements 
2-5 

used for determining the Cs-137 activity acceptable because it follows the plans detailed in the 

Radiation and Contamination Survey Report (WVNSCO, 2004c). 

The NRC staff evaluated the statistical metric of radionuclide concentrations used to calculate 

inventories in the waste determination (e.g., mean, geometric mean) to ensure that the technical 

basis for the selection is adequate and the metric is properly calculated. As described in the 

previous section, the CFMT scaling factors were originally based on the geometric average of 

the four batch samples and the three analyses for the single liquid sample. The MFHT scaling 

factors are based on the geometric average of the same four batch samples along with two 

glass shard samples taken from the two evacuated canisters used to remove molten glass from 

the Vitrification Melter (WMG, 2011). During the review NRC asked DOE to explain why the 

liquid sample for CFMT was treated as three separate samples in the calculation of the 

geometric mean and to describe any impacts on overall conclusions if this liquid sample had 

been treated as one sample instead of three. In its response, DOE revised the estimate for the 

CFMT (as shown in Table 2-1), treating the liquid sample as one instead of three separate 

results. DOE also revised the density of the liquid sample to be 1.15 g/cm3

 (DOE, 2012c) and 

the revised estimate of 3.68 TBq (99.4 Ci) is reflected above in Table 2-1 (the original estimate 

was 3.57 TBq (96.5 Ci). The NRC staff notes that the revised estimate does not impact 

conclusions about the removal of radionuclides to the maximum extent technically and 

economically practical or the vessel waste classification. The DOE states that applying the 

geometric average is appropriate because it “takes into account the differing radionuclide 

distributions in the samples.” NRC staff notes that the geometric mean is usually reserved for 

comparing items with markedly different numeric ranges. Since all these samples are meant to 

represent the same population of data, it is not clear to the NRC staff why DOE chose to apply 

the geometric mean as opposed to a different statistic. The arithmetic mean or the 95th

percentile would have yielded slightly higher concentrations and also slightly higher scaling 

factors for most radionuclides. However, NRC staff evaluated the impact of applying a different 

statistical metric (i.e., the arithmetic mean or the 95th percentile value) and determined that use 

of a different statistic would not change the conclusions discussed in later Chapters of this 

review. During the review, the NRC staff requested additional justification regarding DOE’s use 

of the geometric mean. DOE clarified that because multiple results were available and they 

were consistent, geometric averaging was used to calculate the scaling factors citing that 

geometric averaging of scaling factors is a common practice throughout the commercial nuclear 

power industry in cases where more than one representative sample is available (DOE, 2012c). 

Given that use of alternative statistical metric (i.e., the arithmetic mean or the 95th percentile 

value) would not impact the overall conclusions, the NRC staff finds DOE’s averaging approach 

acceptable. 

In NRC’s previous review concerning the Melter (NRC, 2011), the DOE pointed to the similarity 

in the scaling factors for Sr-90, Np-237, Pu-238, and Am-241 of the glass samples to the batch 

samples. These data show that although the concentrations vary from batch to batch, the 

scaling factors in relation Cs-137 are similar. These data are reproduced in Table 2-2 and are 

relevant to this review since these are also the same batch samples that are utilized in the 

CFMT and MFHT characterization. The similarities in the scaling factors in Table 2-2 support 

the representativeness of the sample data for the residual material in the vessels. 
2-6 

Table 2-2: Radionuclide Scaling Factors (Ratios to Cs-137) (DOE, 2011b) 

Radionuclide Glass Sample Data Batch 74 Data Batch 75 Data 

Sr-90 5.73x10-2 4.00x10-2 7.47x10-2

Np-237 1.41x10-6 3.25x10-7 6.13x10-7

Pu-238 1.57x10-4 5.48x10-5 1.09x10-4

Am-241 6.84x10-4 2.19x10-4 3.28x10-4

The NRC staff evaluated the impact of uncertainties due to heterogeneity, sample variability, 

and analytical methods. The DOE estimates that the uncertainty (from all sources) associated 

with the radionuclide inventories of the CFMT and MFHT in Table 2-1 is bounded by +/-20 

percent concentration range. The high and low values are identified in the NNSS waste profile 

radiological technical basis document (CHBWV, 2011a) but are not propagated into the 

discussion of removal the maximum extent practical or the waste classification. The NRC staff 

notes that if a 20 percent upper bound is assumed, the conclusions regarding waste 

classification and removal to the maximum extent practical would remain the same as discussed 

in the following sections of this evaluation. 

2.3 DOE Selection of Key Radionuclides 

Section II.B of DOE-Guide 435.1-1, states the following regarding key radionuclide selection: 

“... it is generally understood that [the term] key radionuclides applies to those 

radionuclides that are controlled by concentration limits in §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 

those that are important to satisfying the performance objectives of 10 CFR Part 61, 

Subpart C [for near-surface radioactive waste disposal facilities].” 

In following this guidance, DOE identifies key radionuclides for the draft WIR evaluation based 

on the radionuclides listed in §61.55 and the PA for the disposal facilities (WCS, 2011). 

The key radionuclides, (also listed in Table 4-3 of the draft WIR evaluation) are replicated in 

Table 2.3. 

Table 2-3: Key Radionuclides 

Radionuclide §61.55 Long-lived 

Radionuclides 

§61.55 Short-lived 

Radionuclides 

Radionuclides 

Important to PA 

H-3 X 

C-14 X X 

Co-60 X 

Ni-59 X 

Ni-63 X 

Sr-90 X 

Nb-94 X 

Tc-99 X X 
2-7 

I-129 X X 

Cs-137 X 

Th-229 X 

U-233 X 

U-234 X 

U-238 X 

Np-237 X 

Pu-238 X 

Pu-239 X 

Pu-240 X 

Pu-241 X 

Pu-242 X 

Am-241 X 

Am-243 X 

Cm-242 X 

Cm-243 X 

Cm-244 X 

2.4 NRC Evaluation of Key Radionuclides 

The IA for this review specifically states that NRC’s scope of review does not include an 

independent review of the key radionuclide list because the cleaning technologies evaluated 

and applied were bulk-cleaning methods. Because the technologies that the DOE applied and 

the alternative technologies considered for removal of radionuclides did not target specific key 

radionuclides and instead removed bulk volume from the vessels, the DOE’s cleaning approach 

would be expected to remove all radionuclides in equal proportions, as opposed to only the key 

radionuclides. Therefore, the NRC staff does not have to make a determination on whether or 

not the key radionuclide list is appropriate in order to review other portions of the waste 

evaluation; the staff does not rely on this list for coming to a conclusion about removal to the 

maximum extent practical or waste classification. 

2.5 Removal to the Maximum Extent Technical and Economically 

Practical 

2.5.1 The DOE Selection of Waste Removal Technologies

The DOE considered a range of additional potential removal technologies documented in the 

Decommissioning Handbook (DOE, 1994). Approaches most relevant to the vessels included: 

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

blasting was eliminated due to potential interference with the vitrification process, and the other 

three methods are generally similar except for the water pressure used. DOE further evaluated 

high-pressure spray, which is a combination of the other three methods, along with two other 

options described below. 
2-8 

The DOE fully evaluated the following options for technical and economical practicality: 

(1) flushing vessel internals with water using high-pressure spray; (2) mechanical removal 

through ball milling; and (3) chemical decontamination. The selection criteria for technology 

selection included technology maturity, usefulness, cost per unit activity removed, limitations, 

and net social costs. The DOE determined that processing decontamination fluids with a 

mixture of glass formers, and use of evacuated canisters were both technically and 

economically practical, but that ball grinding was not practical. Specifically, ball grinding 

produced a fine glass powder which could lead to embedded contamination within the metal. 

Although improvements to the process could have been made, the improvements would have 

been so extensive that the resulting technology would no longer be considered proven; this 

would make its use technically impractical. Chemical decontamination was shown to be 

effective in testing, but it was deemed impractical because the chemicals would have been 

incompatible with the requirements for an acceptable glass mixture in the vitrification process. 

Since the resulting flush solutions would be fed to the Melter and transferred to the evacuated 

canisters, this made the chemical decontamination approach unacceptable due to technical 

impracticality (WVNSCO, 2001). 

The DOE made a simplifying assumption that residual material coated the upper one-third of 

each vessel based in part on physical and operational attributes associated with the two 

vessels. The liquid levels were generally limited to the nominal lower two-thirds of both the 

CFMT and MFHT. During operation, the upper portion of the vessels were periodically 

impacted by the liquid waste material due to splashing resulting from additional waste being 

added to the mixture, and the swirling induced by the mixer blades. When the vessels were 

refilled with new batches, the material in the lower two-thirds of the vessel would be 

continuously agitated, but the material in the upper portion of the vessel would have the 

opportunity to dry and harden. Evidence of residual material in the upper portion of the vessels 

was provided in photographic images showing a buildup of dried slurry around the head 

stiffeners. While the simplifying assumption was that residual material covered only the upper 

one-third of the vessels prior to flushing, the DOE also stated that it is reasonable to expect that 

some contamination was present on the lower two-thirds of the vessels before flushing and that 

some lesser amount would remain after flushing. The consistency of the dose rates measured 

on the sides of the vessels confirms that both the upper and lower portions of the vessel were 

uniformly clean after flushing (WVNSCO 2004c). High-pressure spray, consisting of using water 

spray to clean the inner surfaces of the vessels, was determined to be the preferred technology. 

Because the bottom two thirds of the vessels were expected to be relatively clean, the flushing 

focused on the upper third of each vessel. Water at a pressure of approximately 68.95 bar 

(1,000 psi) was delivered at 95 to 130 Lpm (25 to 35 gpm) through a rotary spray nozzle that 

had been inserted through one or more openings in the vessel heads. The CFMT and MFHT 

were each flushed three times, with each flush taking about 60 minutes. The external surfaces 

of the vessels were decontaminated with high-pressure demineralized water from a spray wand. 

2.5.2 The NRC Evaluation of Waste Removal Technologies

The NRC staff has reviewed DOE’s selection for radionuclide removal technologies and has 

concluded that DOE evaluated a reasonable range of technologies. The selection process was 

based on relevant sources of information and the selection criteria were reasonable. Of the 

selection criteria (technology maturity, usefulness, cost per unit activity removed, limitations, 

and net social costs), technology maturity and limitations were applied in narrowing the 

selection. During the review, the NRC staff asked the DOE to provide additional information 

regarding the impracticality of chemical decontamination for the MFHT given that the CFMT 
2-9 

underwent an indirect chemical cleaning after it had been disconnected from the system. 

Sodium hydroxide was added to the CFMT in December 2003 as a means of sending the 

chemical to Tank 8D-4 to raise the pH of its liquid, thereby serving as an indirect chemical flush. 

The DOE stated while a chemical flush of the MFHT after shutdown of the vitrification system 

would have been technically practical; it would not have been economically practical. The costs 

DOE cited include development of procedures and work packages, performing the actual 

flushing, and managing the resulting wastes. Meanwhile, the data from visual inspections 

showed that the vessel had already been effectively decontaminated through the high-pressure 

sprays. Therefore, the costs of such an approach would have outweighed the benefits. 

2.5.3 The DOE’s Estimation of Waste Removal Efficiency 

This section describes the DOE estimates of removal efficiency through flushing the internals 

with high-pressure spray. The removal efficiency for the flushing is reported based on Cs-137 

activity. The DOE claims that other key radionuclides are removed in equal proportions due to 

the homogeneity of the glass waste, and therefore the removal efficiencies for other 

radionuclides would closely follow that of Cs-137. Flushing results are also supported by visual 

inspections using photographs taken with a camera placed inside the vessel. 

The amount of Cs-137 in the vessel prior to flushing assumed an average 0.63 cm (0.25 inch) 

thickness in the upper one-third of the vessels and the Cs-137 concentration in the material was 

originally assumed to be the same as that in the last batch of slurry sent to the CFMT before the 

vitrification system flushing began (batch 75), which had the second highest Cs-137 

concentration (429 MBq/cm3

 [1.16x104 μCi/cm3]) among the feed material (Kurasch, 2011). 

DOE revised this assumption as described in the following section of this TER during NRC’s 

review. 

Table 4-5 of the draft WIR evaluation provides a summary of the indirect and direct flushes 

performed on the CFMT and MFHT. DOE refers to the spray nozzles deployed and operated 

for two cycles as direct flushes. Indirect flushing occurred when fluids from Tank 8D-4 were 

transferred to the CFMT, or when fluids from the MFHT were sent back to the CFMT. In total, 

the CFMT received three separate two-cycle direct flushes and four indirect flushes over the 

course of 2002. In this same year, the MFHT received three separate two-cycle direct flushes, 

and three indirect flushes before the final airlifts from the Melter were completed on August 

14, 2002. In September 2002, the heel of material in the MFHT was transferred to the CFMT 

and the vessel was filled with water which was stirred and sent to the CFMT. Then, about one 

year later in December 2003, about 210 liters (55 gallons) of 50 percent caustic sodium 

hydroxide solution and 110 liters (30 gallons) of non-radioactive water was added to the CFMT, 

stirred for about 30 minutes and remained in the CFMT for about one month. In late January 

2004, the CFMT contents were stirred and sent to Tank 8D-4. Afterward, about 760 liters 

(200 gallons) of water were transferred to the CFMT, stirred, and sent to Tank 8D-4. 

Inspection after flushing using video technology showed essentially no visible deposits in 

comparison to pictures prior to flushing which showed substantial amounts of dried slurry in both 

vessels (WVNSCO, 2002c and 2002d). Additionally, dose rate measurements taken on 

July 15, 2002 after the high-pressure spray flushing had been completed of 0.08 Gy/h and 

0.22 Gy/h (8 R/h and 22 R/h) showed a reduction from the measurements taken pre-flush on 

May 1, 2002 of 2 Gy/h and 2.5 Gy/h (200 R/h and 250 R/h). 
2-10 

Table 2-4: Vessel Flushing Effectiveness in Terms of Estimated Cs-137 Removal 

Condition CFMT Remaining 

Inventory (Ci) (1)

CFMT 

Decontamination 

Factor 

MFHT 

Remaining 

Inventory (Ci) (1)

MFHT 

Decontamination 

Factor 

Before Flushing (2) 630 NA 540 NA 

After All Flushes 95.3 6.6 (3) 97.1 5.6 (4)

(1) Multiply by 3.7x1010 to obtain Bq. (2)

The activity in each vessel before flushing began was estimated in the following manner: (a) the residual 

material (dried slurry) coating observed on the vessel interior surfaces before flushing was assumed to 

average 0.64-cm (0.250-inch) thickness over the upper one-third of the vessels, based on pre-flush visual 

inspection results; and (b) the Cs-137 concentration in this material was assumed to be a representative, 

decay corrected concentration of 5.0x103 μCi/cm3

. (3) This decontamination factor is based on the best estimate Cs-137 activity. If a 20 percent greater upper 

bound estimate were to be used, the decontamination factor would be 5.5 rather than 6.6. 

(4) This decontamination factor is based on the best estimate Cs-137 activity. If a 20 percent greater upper 

bound estimate were to be used, the decontamination factor would be 4.6 rather than 5.6. 

2.5.4 NRC Evaluation of Waste Removal Efficiency 

As part of the review, NRC asked DOE to provide additional support for the activity estimates 

(curies) before flushing, including: the surface area and volume assumptions, further justification 

for assuming the upper third of the vessels is coated, and a technical basis for assuming the 

concentration of batch 75. In its response, DOE noted that the batch 75 Cs-137 concentration 

was 429 MBq/cm3

 (1.16x104 μCi/cm3) as given in the characterization report (WMG, 2011), but 

it was actually measured and reported as 429 MBq/g (1.16x104 μCi/g). Using a specific gravity 

of 1.33 to convert the units appropriately, the revised estimates were increased, yielding 

approximately 74 TBq (2,000 Ci) for the CFMT and approximately 59 TBq (1,600 Ci) for the 

MFHT. 

Furthermore, in assuming that the Cs-137 concentration in the material was the same as that in 

the batch with the second highest concentration among the feed material, DOE was biasing the 

result towards overestimating the amount of Cs-137 in the vessels prior to flushing. Since the 

material remaining in the vessels is characterized by using batches 72, 74 and 75, it would be 

appropriate to also use this combination of samples to characterize what was in the vessel prior 

to flushing. Overestimating the activity before flushing could inflate the reduction factor and is 

therefore not a conservative approach. DOE acknowledged this in the response to RAIs 

(DOE, 2012c), stating that the geometric mean of all the batches as opposed to only batch 75 is 

a better approximate of the material since it likely accumulated over time. Therefore, DOE 

revised their approach to use the geometric mean of Cs-137 concentrations for all batches 

(240 MBq [6.5x103 μCi/cm3

]). The revised values are reflected in Table 2-4. 

DOE also provided additional support for the removal efficiency by providing detailed 

photographic images, and pointing to the pre- and post-flushing dose rate measurements. NRC 

staff notes that in assuming that the upper one-third of the vessel is covered in residual material 

as opposed to some larger portion, DOE is potentially biasing the removal efficiency in a 

conservative fashion. (If DOE were to assume a larger area, the pre-flushing contamination 

value would have been higher.) DOE explained in the RAI responses that the one-third 

coverage assumption was based on the operational characteristics of the vessels, but also 
2-11 

acknowledged that there was likely residual contamination in the bottom two-thirds of the vessel 

prior to flushing as well (DOE, 2012c). While the revised concentration assumption results in a 

lower decontamination factor, the visual inspections, dose rate reductions, along with the 

conservative one-third coverage assumption are sufficient to support the conclusion regarding 

removal efficiency. During the review, NRC staff requested additional information on the 

impacts of uncertainty on the removal to the maximum extent practical. In response, DOE 

clarified that if the upper bound for the scaling factors were applied, the CFMT reduction factor 

would change from 93 percent to about 92 percent. The MFHT reduction factor would change 

from 91 percent to about 89 percent. The dose rate reduction would also be minimally 

impacted. Based on the insignificance of these differences, the NRC staff concludes that the 

DOE has appropriately considered the impacts of uncertainty in the inventory for the purposes 

of determining the cleaning technology effectiveness. 

In the NRC’s prior review concerning the Melter (NRC, 2011), the NRC staff asked the DOE 

whether it considered modifications to the flushing method to improve efficiency prior to using 

the evacuated canisters. The DOE states that another batch of decontamination solution was 

not used because of concerns over acid hindering the integrity of the system (Kocialski, 2003). 

The DOE also stated that a decision was made in 2002 that the flushing had been satisfactorily 

completed before the evacuated canisters were employed (DOE, 2011a). As described in the 

evaluation of waste removal technologies the NRC staff questioned DOE about the practicality 

of a chemical wash for the MFHT after the vitrification system had been shut down and DOE 

sufficiently described how an additional chemical wash would not have been economically 

practical (DOE, 2012c). 

In reviewing the HLW Processing Systems Flushing Operations Run Plan (WVNSCO, 2002b), 

as well as the noted concerns over the limited life of the Melter, the risk of its malfunction and 

impacts on the entire system, and the response to the RAI, the NRC staff concludes that 

ceasing removal activities for the MFHT and CFHT after the direct and indirect flushing activities 

that were carried out was reasonable (DOE, 2012c). 

2.5.5 DOE Estimation of Costs and Benefits of Additional Waste Removal 

Although DOE assumed that an additional flush would have removed 90 percent of the activity 

remaining in the vessel at conclusion of the flushing that was actually performed, DOE 

determined that the benefits of additional removal were limited. The low dose rates on the 

outside of the waste packages and compliance with the waste handling requirements at WVDP 

and the disposal facility would ensure protection of individuals during operation. Additionally, 

DOE concluded that since the impacts of disposal to the general population or to an intruder 

were negligible, a further reduction in radioactivity would not have been beneficial from the 

standpoint of potential doses. Additional removal, which would have incurred additional 

cleaning costs, could have also resulted in a cost savings of about $200,000 by using lighter 

weight steel and less shielding. However, the worker doses would have been about the same 

with this reduced shielding and the costs of additional waste removal would have been greater 

than the cost savings of $200,000. Uncertainty surrounding the potential benefits is not 

discussed in the DOE’s analysis. 

The costs included by the DOE in the draft WIR evaluation were primarily related to the cost of 

extending the shutdown date and continued operation of the facility. DOE conservatively did not 

quantify other costs such as capital costs specific to the cleaning technology, canister/container 

costs, the additional processing costs, shipping and disposal costs, as well as worker and public 
2-12 

exposure costs. The Vitrification Facility operating cost per month is determined to be $1.89M 

per month. An additional flush, assumed to require two weeks, would have cost approximately 

$1M (DOE, 2012b). Uncertainty surrounding the potential costs is not discussed in DOE’s 

analysis. 

2.5.6 NRC Evaluation of Costs and Benefits of Additional Waste Removal

The NRC staff evaluated the following key assumptions regarding benefits: 

• No benefit from reduced worker exposure 

• No benefit from reduced public exposure, including an intruder 

Even though additional removal activities are estimated to have potentially removed 90 percent 

of the residual activity, the additional removal does not translate into quantifiable benefits in 

terms of worker dose. Due to the already low doses resulting from the vessels without further 

removal, the NRC staff agrees that further radionuclide removal would not significantly reduce 

worker doses at the disposal facility. 

In terms of public dose, as stated in NUREG-1854, page 3-11, “The primary benefit of 

radionuclide removal is expected to be a reduction in the risk the waste will pose to the general 

public, including inadvertent intruders.... an analysis of the costs and benefits of additional 

radionuclide removal will depend in part on the performance assessment and inadvertent 

intruder analysis predictions” (NRC, 2007). NRC staff reviewed only those items requested by 

DOE and did not independently verify the doses provided and any corresponding benefit of 

avoided public or intruder exposure by additional radionuclide removal. However, assuming 

that the dose estimates provided by DOE are accurate, even a large percentage of additional 

radionuclide removal would have negligible impact on the long-term dose in comparison to the 

social costs. Also, because the costs of additional removal would be so high, the NRC staff 

concludes that the impact on dose from additional flushing would be minimal in comparison to 

the large costs of additional radionuclide removal. 

The NRC staff also evaluated the following key assumptions regarding costs: 

• The vitrification operating costs and time to complete a flush 

• Storage transport and disposal costs for HLW canisters 

Annual operating costs for the Vitrification Facility were estimated to be between $25M and 

$30M per year ($500,000 per week). The DOE assumed that creation of a single new canister 

would have taken about two weeks from start to end, thus costing $1.0M (DOE, 2012b). The 

NRC staff finds these assumptions regarding cost and timing of additional operation to be 

reasonable considering canisters 267 through 275 (containing the flush solutions of batches 76 

and 77) were processed over roughly 3 months from May 7, 2002 to August 14, 2002 

(Kurasch, 2011). An additional flush could have required longer than 2 weeks, including 

downtime should the Melter have failed. 

While the DOE does not attempt to quantify cost of HLW disposal in its summary cost benefit 

analysis, the NRC staff notes that the cost of storing, transporting, and disposing of a HLW 

canister to the repository could be substantial given that there is currently no disposal path for 
2-13 

HLW canisters. It is likely that the cost of disposal of additional HLW would be substantial 

enough to make uncertainty in DOE’s estimates of the operating costs relatively insignificant. 

Based on the discussion presented above, the NRC staff concludes that the costs and benefits 

considered are conservative and reasonable, and that the costs outweigh the benefits of 

additional removal activities. 
3-1 

3.0 THE WASTE WILL BE MANAGED TO MEET SAFETY 

REQUIREMENTS COMPARABLE TO THE PERFORMANCE 

OBJECTIVES OF 10 CFR PART 61, SUBPART C 

This section of the NRC staff’s technical review partially covers DOE Manual 435.1-1, 

Section II.B(2)(a)(2) – that waste will be managed to meet safety requirements that are 

comparable to the performance objectives set out in 10 CFR Part 61, Performance Objectives. 

The performance objectives listed in 10 CFR Part 61 are as follows: 

• §61.41 Protection of the general population from releases of radioactivity. 

• §61.42 Protection of individuals from inadvertent intrusion. 

• §61.43 Protection of individuals during operations. 

• §61.44 Stability of the disposal site after closure. 

According to the March 13, 2012 IA, DOE is requesting that the NRC staff focus its review on 

the requirements for operational radiation protection (§61.43) and the waste form stability 

(partial review of §61.44). 

3.1 Protection of Individuals during Operations 

10 CFR 61.43 states: 

“Operations at the land disposal facility must be conducted in compliance with 

the standards for radiation protection set out in 10 CFR 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 exposures as low as is reasonably achievable.” 

The NRC staff’s review for this section focused on DOE commitments to adhere to the 

appropriate regulations for protection of individuals during operations, and descriptions of how 

the regulations are implemented with respect to the draft WIR evaluation (DOE, 2012b). DOE 

has evaluated the protection of individuals during operations for disposal of the CFMT and 

MFHT at either NNSS or the WCS facility in Texas. 

Section 61.43 references 10 CFR Part 20, Standards for Protection Against Radiation, which 

contains similar radiological protection standards to those of the DOE’s for workers and the 

public, which can be found in 10 CFR Part 835, Occupational Radiation Protection, and 

DOE 5400.5, Radiation Protection of the Public and the Environment. The State of Texas 

regulations – found in Title 30 of the Texas Administrative Code, Chapter 336, Radioactive 

Substance Rules – also mirror §61.43. Table 3-1 shows that NRC, DOE, and the State of 

Texas regulations contain similar radiological protection standards. The NRC staff is not 

including a discussion about Nevada State Regulations because the NNSS is a Federal facility 

and thus isn’t regulated by the state of Nevada. NNSS operations are performed in accordance 

with DOE regulations. 
3-2 

Table 3-1: Comparison of NRC, DOE, and Texas State Radiological Protection Standards 

Crosswalk Topics DOE NRC TX 

(Texas Rule) 

Annual Air Emission Limit 

for Individual Member 

DOE Order 5400.5 

0.1 Sv (10 mrem) 

§20.1101(d) 

0.1 Sv (10 mrem) 

§336.304 

0.1 Sv (10 mrem) 

Annual TEDE for Adult 

Workers 

§835.202(a)(1) 

0.05 Sv (5 rem) 

§20.1201(a) 

0.05 Sv (5 rem) 

§336.305 

0.05 Sv (5 rem) 

Any Individual Organ or 

Tissue Annual Dose Limit 

for Adult Workers 

§835.202(a)(2) 

0.5 Sv (50 rem) 

§20.1201(a) 

0.5 Sv (50 rem) 

§336.305 

0.5 Sv (50 rem) 

Annual Dose Limit to the 

Lens of the Eye for Adult 

Workers 

§835.202(a)(3) 

0.15 Sv (15 rem) 

§20.1201(a) 

0.15 Sv (15 rem) 

§336.305 

0.15 Sv (15 rem) 

Annual Dose Limit to the 

Skin of the Whole Body 

and to the Skin of the 

Extremities for Adult 

Workers 

§835.202(a)(4) 

0.5 Sv (50 rem) 

§20.1201(a) 

0.5 Sv (50 rem) 

§336.305 

0.5 Sv (50 rem) 

Limit on Soluble Uranium 

Intake 

DOE Order 440.1A 

2.4 mg/week 

§20.1201(e) 

10 mg/week 

§336.305 

10 mg/week 

Dose Equivalent to 

Embryo/Fetus 

§835.206(a) 

5 mSv (0.5 rem) 

§20.1208(a) 

5 mSv (0.5 rem) 

§336.312 

5 mSv (0.5 rem) 

Dose Limit for Individual 

Member of the Public 

(Total Annual Dose) 

DOE Order 5400.5 

1 mSv (100 mrem) 

§20.1301(a) 

1 mSv (100 mrem) 

§336.313 

1 mSv (100 mrem) 

Dose Limit for Individual 

Members of the Public 

(Dose Rates in 

Unrestricted Areas) 

§835.602 

0.005 mSv/hr 

(0.05 mrem/hr) 

§20.1301(a) 

0.02 mSv/hr 

(2 mrem/hr) 

§336.313 

0.02 mSv/hr 

(2 mrem/hr) 

Dose Limits for Members 

of the Public with Access 

to Controlled Areas 

§835.208 

0.001 Sv (0.1 rem) 

§20.1301(b) 

0.001 Sv (0.1 rem) 

§336.313 

0.001 Sv (0.1 rem) 

As Low As Reasonably 

Achievable §835.2 §20.1003 §336.2 

As part of the Criterion 2 demonstration, DOE shows that the CFMT and MFHT meet the waste 

acceptance criteria for the NNSS (DOE, 2012a) and WCS (WCS, 2008). There is a 300 

plutonium equivalent gram (PE-g) limit for waste packages disposed of at NNSS. The NRC 

staff agrees with DOE’s assessment that the residual wastes remaining in the CFMT and MFHT 

meet this 300 PE-g safety requirement. DOE notes that a waste profile package for disposal of 
3-3 

the CFMT and MFHT has already been approved by NNSS contingent upon the outcome of the 

WIR evaluation. WCS is limited in the total waste volume and total activity that can be 

accepted. For example, WCS has a license limit of 0.23 million cubic meters (8.1 million cubic 

feet) of waste and 190 million gigabecquerels (5.1 million Curies) of waste. DOE shows that the 

CFMT and MFHT will have a negligible contribution to these license limits. 

3.2 Waste Form Stability 

10 CFR 61.44 states: 

“The disposal facility must be sited, designed, used, operated, and closed to 

achieve long term stability of the disposal site and to eliminate to the extent practicable 

the need for ongoing active maintenance of the disposal site following closure so that 

only surveillance, monitoring, or minor custodial care are required.”

As specified in the IA, the NRC staff’s review focuses on waste form stability and does not 

include a review of the disposal site stability. Review of waste form stability involves verification 

of the structural stability of the waste after site closure as well as verification that void spaces in 

the waste will not cause differential settling of the waste or provide preferential flow paths for 

infiltrating water. 

The following three characteristics of the waste form support the argument that the waste is in a 

stable form: the structural integrity of the UC; the Department of Transportation Industrial 

Package 2 (IP-2) shipping containers, which contain the UC; and the low-density cellular 

concrete that will fill the void spaces in the CFMT, MFHT, and associated packages. 

The CFMT is constructed of Hastelloy C-22, a nickel-chromium-molybdenum-tungsten alloy. 

The lower part of the CFMT exterior is covered with heat transfer coils formed of stainless steel 

piping covered with fiberglass insulation and stainless steel sheeting. The MFHT is constructed 

of stainless steel and is partially covered by a cooling jacket. 

NUREG-1854 directs the staff to consider waste stability, as set forth in §61.56(b). The CMFT 

and MFHT IP-2 packages are “capable of supporting a uniformly distributed load (compressive 

strength) of 3,375 pounds per square foot strength requirement of the Nevada Test Site Waste 

Acceptance Criteria (DOE/NV-325)” (WMG, 2004b; WMG, 2004c). In addition, NRC staff and 

contractors previously concluded that concrete can be formulated for a 500-year service life 

(NRC, 1989). Among other things, factors such as sulfate attack, freezing and thawing, and 

stress cracking were addressed in the analysis (NRC, 1989; NRC, 1990). 

Criterion 3 of DOE Manual 435.1-1, Section II.B(2)(a) states: 

“[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 lowlevel 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 part of the Criterion 3 demonstration, DOE shows that the UC are in a solid physical form. 

DOE describes in the draft WIR that “void spaces in both vessels and their waste containers 
3-4 

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”

(DOE, 2012b). 

The NRC staff concludes that the waste is in a stable form because all void spaces will be filled 

with low-density cellular concrete, which was previously determined to have a 500-year service 

life, and because the IP-2 packages containing the CFMT and MFHT meet the compressive 

strength requirements of the disposal facility. As previously noted, this review did not evaluate 

the durability of the waste form or packaging with respect to meeting the §61.41 or §61.42 

performance objectives. 
4-1 

4.0 THE WASTE WILL NOT EXCEED CLASS C CONCENTRATION 

LIMITS AND WILL BE MANAGED IN ACCORDANCE WITH DOE 

REQUIREMENTS AS LLW 

This section of the TER pertains to the third criterion in Section II.B(2)(a) of DOE-M 435.1-1: 

(i) that the CFMT and MFHT waste packages will be in a solid physical form, (ii) will not exceed 

Class C concentration limits, and (iii) will be managed in accordance with DOE requirements as 

LLW, as applicable. The DOE indicated that the focus of NRC’s review under this criterion 

should be waste form stability and classification. 

4.1 The DOE’s Waste Classification Approach 

Criterion 3 of DOE-M 435.1-1 Section II.B (2)(a) is reproduced below for ease of reference: 

“[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 

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 lowlevel waste as set out in [Code of Federal Regulations] §61.55, Waste Classification, or 

will meet alternative requirements for waste classification and characterization as DOE 

may authorize.” 

Table 4-1 presents the DOE’s waste classification results for the CFMT and MFHT, and the 

results support a conclusion that the vessels do not exceed the Class C concentration limits. 

To evaluate the DOE’s approach for determining the class of the vessels, the staff reviewed the 

waste classification requirements found in 10 CFR §61.55. Table 1 of §61.55 contains 

concentration limits for long-lived radionuclides including a specific class of radionuclides, 

alpha-emitting transuranic radionuclides with half-lives greater than 5 years. DOE listed several 

radionuclides that were members of this class (e.g., Np-237, Pu-238, Pu-239, Pu-240, Am-241, 

Am-243, Cm-243, and Cm-244) along with their corresponding contributions to the sum of 

fractions. Table 2 of §61.55 contains a list of concentration limits for relatively short-lived 

radionuclides. DOE also evaluated the contributions of short-lived radionuclides found in 

Table 2 to the sum of fractions. The sum of fractions approach used to determine waste 

classification for mixtures of radionuclides is described in §61.55(a)(7). Because the CFMT and 

MFHT contain a mixture of long- and short-lived radionuclides, DOE applied §61.55(a)(5) to 

determine waste classification. §61.55(a)(5) and (a)(7) are reproduced below for additional 

background information on the waste classification calculations. 

§61.55(a)(5), “Classification determined by both long- and short-lived 

radionuclides. 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 determined by the 

concentration of the radionuclides 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.” 
4-2 

§61.55(a)(7), “The sum of the fractions rule for mixtures of radionuclides. For 

determining classification for waste that contains a mixture of radionuclides, it is 

necessary to determine the sum of fractions by dividing each nuclide’s concentration by 

the appropriate limit and adding the resulting values. The appropriate limits must all be 

taken from the same column of the same table. The sum of the fractions for the column 

must be less than 1.0 if the waste class is to be determined by that column. Example: A 

waste contains Sr-90 in a concentration of 50 Ci/m3

 and Cs-137 in a concentration of 

22 Ci/m3

. Because the concentrations both exceed the values in Column 1, Table 2, 

they must be compared to Column 2 values. For Sr-90 fraction 50/150 = 0.33; for 

Cs-137 fraction, 22/44 = 0.5; the sum of the fractions = 0.83. Because the sum is less 

than 1.0, the waste is Class B.” 

In its statement of work, the DOE requested that the NRC evaluate the waste classification 

calculations against NRC’s guidance found in NUREG-1854, Chapter 3, Section 3.5.1.1, 

“Concentration Averaging” (NRC, 2007). The NRC’s guidance on concentration averaging 

considers three categories of averaging based on (1) physical homogeneity, (2) stabilization, 

and (3) site-specific (intruder) analysis considerations. 

Because residual radioactivity associated with the vessels is not physically homogeneous (not 

evenly distributed across the entire vessel packages), NRC’s guidance on concentration 

averaging (NRC, 1995) would also allow the DOE to consider site-specific factors in evaluating 

the risk to an inadvertent intruder under option (3). However, the DOE did not elect to use this 

option although it may have led to a smaller Class C sum of fractions, thereby providing 

additional confidence that the waste is not greater than Class C. 

The DOE elected to use a method that is consistent with option (2) stabilization, although it did 

not explicitly take full advantage of this option. For example, the DOE did not attempt to 

average residual radioactivity over the weight of stabilizing grout it is planning to use to fill void 

spaces in the vessels and their respective waste containers. However, the DOE may have 

made reasonable arguments on how at least a fraction of the stabilizing grout may have been 

necessary to encapsulate or assist with immobilization of contamination within the vessels. 

Radionuclide concentrations for purposes of waste classification for the CFMT are based on 

averaging the total activity of each radionuclide to the total weight of the vessels themselves. 

The weight does not include the weight of the shipping containers (i.e., shielded IP-2) or the 

weight of stabilizing grout used to fill voids in the vessels and voids between the vessels and 

their shipping containers. The total weight of the CFMT and MFHT used in the calculation was 

8,530 kg (18,810 lbs) and 10.800 kg (23,710 lbs) respectively. The waste volume is calculated 

using the vessel weight and the density of stainless steel. The volume used was1.07 m3

(37.6 ft3

) and 1.34 m3 (47.4 ft3) respectively. The resulting sums of fractions the DOE calculated 

based on using the average dose rates and geometric mean values of analytical data as 

described in the characterization report (WMG, 2011) and the related analysis (Kurasch, 2012) 

are shown in Table 4-1. 
4-3 

4.2 NRC Evaluation of the DOE’s Waste Classification Approach 

The NRC staff has evaluated DOE’s methodology for classifying waste and finds the approach 

an acceptable application of Category 2 of NRC staff’s guidance on concentration averaging 

found in NUREG-1854 (NRC, 2007) and consistent with Section 3.4 of NRC’s Branch Technical 

Position (BTP) (NRC, 1995). It is important to note that NRC’s guidance on concentration 

averaging in NUREG-1854 does not replace the BTP. Rather NUREG-1854 attempts to apply 

the concentration averaging guidance principles specifically to WIR applications, providing 

additional flexibility, where appropriate, consistent with the general principles of the BTP. NRC 

considers DOE’s waste classification calculations consistent with Section 3.4 of the NRC’s BTP 

entitled “Contaminated Materials.” Section 3.4 of NRC’s BTP provides for averaging over the 

total weight of displaced volume of the contaminated item with major void volumes subtracted 

from the envelope volume. 

In using the weight of the vessels and the density of steel to calculate the volume, DOE avoided 

averaging over large void volumes. This approach follows NRC guidance, which states that 

large void volumes should usually not be part of the concentration averaging process as 

indicated in Section 3.4 of NRC’s BTP. NRC also notes that DOE did not take credit for the 

weight of stabilizing grout used to encapsulate the residual radioactivity. Therefore, the NRC 

finds DOE’s approach to determining the concentration appropriate. 

As noted in Section 2.2 of this document, the DOE identifies high and low activity ranges that 

are plus or minus 20 percent, respectively of the final waste form concentrations shown in 

Table 4-1 and Table 4-2. Even taking this uncertainty into consideration, the inventories are still 

well within the fraction of Table 1 or Table 2 of §61.55. DOE also calculated the Class C 

concentration assuming the Melter scaling factors as a result of RAIs, and shows that the 

impact on the sum of fractions is insignificant (DOE, 2012c). Therefore, the NRC staff 

concludes that the DOE’s assessment that the CFMT and MFHT is Class C, considering 

uncertainty in the volume of the vessels and uncertainty in the inventory estimates therein, is 

technically sufficient. 

As noted in Section 2.2, the NRC staff reviewed historical inventories and compared those to 

the radionuclides considered for waste classification purposes. NRC staff notes that some of 

the radionuclides listed in the inventory are denoted with an “NA.” DOE clarified that certain 

transuranics with half-lives greater than 5 yrs that were listed in the sludge/supernatant 

inventory (Pu-242, Am-242m, Cm-245, Cm-246), were not originally considered in the waste 

classification because they were present in small quantities in the sludge/supernatant. For 

consistency, the DOE added these radionuclides to the waste classification table. Since they 

are present in small quantities, inclusion of these transuranics does not significantly impact the 

sum of fractions. 

The vessels may be transported to the Nevada National Security Site Area 5 Radioactive Waste 

Management Site for disposal. The vessel waste packages would be disposed of as LLW and 

managed in accordance with DOE requirements for LLW disposal found in Chapter IV of 

DOE-M 435.1-1. The waste form meets the site’s acceptance criteria and the waste profile has 

been formally approved by the potential disposal facility. The DOE may alternatively decide to 

ship the waste to the commercially operated WCS federal facility waste disposal facility in 

Texas. The State of Texas Class C concentration limits are consistent with §61.55 and the 

vessels would meet the Class C LLW concentration limits established in the Texas 

Administrative Code. 
4-4 

Based on the NRC’s review of DOE’s draft WIR evaluation and supporting references, the NRC 

considers the DOE’s conclusions to be adequate and reasonable such that the DOE can meet 

Criterion 3 of DOE-M 435.1-1 related to management of WIR as LLW. 

Table 4-1: Concentrator Feed Makeup Tank Waste Concentration Results 

Nuclide Activity Class C Limit 2 CFMT 

Concentration 2 Percent 

(Ci) 1 (Ci/m3) 1 (nCi/g) 1 (Ci/m3) 1 (nCi/g) 1 Table 1 Table 2

C-14 NA 8.00x100

K-40 NA 

Mn-54 NA 

Co-60 4.1x10-03 1.73x10-3

Ni-63 NA 7.00x10+2

Sr-90 3.9x10+00 7.00x10+3 1.05x100 0.052%

Zr-95 NA 

Tc-99 1.8x10-03 3.00x100 3.90x10-3 0.060% 

Cs-137 9.5x10+01 4.60x10+3 8.91x10+1 1.936%

Eu-154 5.6x10-02 4.83x10-2

Np-237 5.6x10-05 1.00x10+2 7.81x10-3 0.007% 

Pu-238 6.9x10-03 1.00x10+2 6.08x10-1 0.809% 

Pu-239 1.5x10-03 1.00x10+2 1.64x10-1 0.176% 

Pu-240 1.5x10-03 1.00x10+2 1.25x10-1 0.176% 

Pu-241 1.4x10-02 3.50x10+3 1.81x100 0.047% 

Pu-242 NA 1.00x10+2

Am-241 3.7x10-02 1.00x10+2 2.91x100 4.338% 

Am-242m NA 1.00x10+2

Am-243 3.3x10-04 1.00x10+2 2.99x10-2 0.039% 

Cm-242 3.9x10-04 2.00x10+4 1.15x10-2 0.000% 

Cm-243 3.2x10-03 1.00x10+2 1.82x10-2 0.375% 

Cm-244 3.2x10-03 1.00x10+2 4.71x10-1 0.375% 

Cm-245 NA 1.00x10+2

Sum of Fractions 6.5% 2.0% 

From WMG (2011b), the calculations are based on the CFMT weight. The activity estimates used were 

as of October 1, 2004; the activities are now somewhat lower due to radioactive decay. The weight 

used in the calculation was 8,530 kg (18,810 lbs) and the volume used was 0.07 m3

 (37.62 ft3). 

Table numbers refer to §61.55, Tables 1 and 2. (Table I and Table II to Appendix E to Rule §336.362 

of the Texas Administrative Code are identical to NRC‘s Table 1 and Table 2). 

1 Conversion to GBq from Ci, multiply by 37 GBq/Ci.

2 Conversion to Bq from nCi, multiply by 37 Bq/nCi
4-5 

Table 4-2: Melter Feed Holder Tank Waste Concentration Results 

Nuclide Activity Class C Limit 2 CFMT 

Concentration 2 Percent 

(Ci) 1 (Ci/m3) 1 (nCi/g) 1 (Ci/m3) 1 (nCi/g) 1 Table 1 Table 2

C-14 3.98x10-4 8.00x100 2.97x10-4 0.004% 

K-40 1.54x10-3 1.15x10-3

Mn-54 1.67x10-3 1.25x10-3

Co-60 1.58x10-3 1.18x10-3

Ni-63 1.89x10-2 7.00x10+2 1.41x10-2 0.002%

Sr-90 5.34x100 7.00x10+3 3.99x100 0.057%

Zr-95 3.72x10-2 2.78x10-2

Tc-99 8.34x10-4 3.00x100 6.22x10-4 0.021% 

Cs-137 9.71x10+1 4.60x10+3 7.25x10+1 1.575%

Eu-154 3.18x10-2

Np-237 7.26x10-5 1.00x10+2 6.72x10-3 0.007% 

Pu-238 9.19x10-3 1.00x10+2 8.51x10-1 0.851% 

Pu-239 2.28x10-3 1.00x10+2 2.11x10-1 0.211% 

Pu-240 1.74x10-3 1.00x10+2 1.61x10-1 0.161% 

Pu-241 5.88x10-2 3.50x10+3 5.44x100 0.156% 

Pu-242 NA 1.00x10+2

Am-241 4.33x10-2 1.00x10+2 4.01x100 4.009% 

Am-242m NA 1.00x10+2

Am-243 3.93x10-4 1.00x10+2 3.64x10-2 0.036% 

Cm-242 3.42x10-4 2.00x10+4 3.17x10-2 0.000% 

Cm-243 2.84x10-4 1.00x10+2 2.63x10-2 0.026% 

Cm-244 7.36x10-3 1.00x10+2 6.81x10-1 0.681% 

Cm-245 NA 1.00x10+2

Sum of Fractions 6.3% 1.6% 

From WMG (2011b), the calculations are based on the MFHT weight. The activity estimates used were 

as of October 1, 2004; the activities are now somewhat lower due to radioactive decay. The weight 

used in the calculation was 10.800 kg (23,710 lbs) and the volume used was 1.34 m3

 (47.42 ft3). 

Table numbers refer to §61.55, Tables 1 and 2. (Table I and Table II to Appendix E to Rule §336.362 of 

the Texas Administrative Code are identical to NRC‘s Table 1 and Table 2.). 

1 Conversion to GBq from Ci, multiply by 37 GBq/Ci. 

2 Conversion to Bq from nCi, multiply by 37 Bq/nCi. 
5-1 

5.0 CONCLUSIONS 

5.1 NRC Review: Waste has been Processed to Remove Key 

Radionuclides to the Maximum Extent that is Technically and 

Economically Practical

Overall Conclusion: Based on the NRC’s review of DOE’s draft WIR evaluation and supporting 

references, the NRC considers the DOE’s conclusions to be adequate and reasonable such that 

the DOE can meet the NRC-reviewed portions of Criterion 1 of DOE-M 435.1-1 related to the 

removal of key radionuclides. This is based on the following specific topical conclusions. 

Waste Inventory:

• The NRC staff finds the list of radionuclides presented in the inventory as presented in the 

draft WIR evaluation acceptable. 

• The NRC staff concludes that the approaches taken for data verification as outlined in the 

Characterization Management Plan for the Facility Characterization Project were 

appropriate. 

Key Radionuclides:

• The IA for this review specifically states that NRC’s scope of review does not include an 

independent review of the key radionuclide list because the cleaning technologies evaluated 

and applied were bulk-cleaning methods. It is not necessary for the NRC staff to make a 

determination on whether or not the key radionuclide list is appropriate in order to review 

other portions of the waste evaluation because the staff does not rely on this list for coming 

to a conclusion about removal to the maximum extent practical or waste classification. 

Removal to the Maximum Extent Technical and Economically Practical:

• The NRC staff concludes that DOE evaluated a reasonable range of technologies that 

included methods to remove volumes of waste. 

• The NRC staff concludes that the DOE has appropriately considered the impacts of 

uncertainty in the inventory as it relates to the radionuclide removal demonstration. 

• The NRC staff concludes that ceasing removal activities for the MFHT and CFHT after the 

direct and indirect flushing activities that were carried out was reasonable. 

• Due to the already low doses resulting from the vessels without further removal, the NRC 

staff agrees that further radionuclide removal would not significantly reduce worker doses at 

the disposal facility. 

• Assuming that the DOE’s dose estimates are accurate, even a large percentage of 

additional radionuclide removal would have negligible impact on the long-term dose in 

comparison to the social costs. 
5-2 

• Based on the discussion presented above, the NRC staff concludes that the costs and 

benefits considered are conservative and reasonable, and that the costs outweigh the 

benefits of additional removal activities. 

5.2 NRC Review: Waste will be Managed to Meet Safety 

Requirements Comparable to the Performance Objectives of 10 

CFR 61, Subpart C

Conclusion: Based on the NRC’s review of DOE’s draft WIR evaluation and supporting 

references, the NRC considers the DOE’s conclusions to be adequate and reasonable such that 

the DOE can meet the NRC-reviewed portions of Criterion 2 of DOE-M 435.1-1 related to Safety 

Requirements associated with the Performance Objectives of 10 CFR 61, Subpart C. This is 

based on the following specific topical conclusions. 

Protection of Individuals during Operations: 

• The NRC staff agrees with DOE’s assessment that the residual wastes remaining in the 

CFMT and MFHT meet the applicable safety requirements for protection of individuals 

during operations. 

Waste Form Stability:

• As specified in the IA, the NRC staff’s review focuses on waste form stability and does not 

include a review of the disposal site stability. 

• The NRC staff concludes that the waste is in a stable form due to the structural integrity 

afforded by the CFMT and MFHT, the IP-2 packages, and the low-density cellular concrete 

that will fill void spaces. As previously noted, this review does not evaluate the durability of 

the waste form or packaging with respect to meeting the §61.41 or §61.42 performance 

objectives. 
5-3 

5.3 NRC Review: Waste will not Exceed Class C Concentration 

Limits and will be Managed in Accordance with DOE 

Requirements as LLW

Conclusion: Based on the NRC’s review of DOE’s draft WIR evaluation and supporting 

references, the NRC considers the DOE’s conclusions to be adequate and reasonable such that 

the DOE can meet Criterion 3 of DOE-M 435.1-1 related to management of WIR as LLW. 

DOE’s Waste Classification Approach:

• The NRC staff has evaluated DOE’s methodology for classifying waste and finds the 

approach an acceptable application of Category 2 of NRC staff’s guidance on concentration 

averaging found in NUREG-1854 (NRC, 2007) and consistent with Section 3.4 of NRC’s 

Branch Technical Position (BTP) (NRC, 1995). 

• The NRC finds DOE’s approach in determining the concentrations appropriate. 

• The NRC staff concludes that the DOE’s assessment that the CFMT and MFHT is Class C, 

considering uncertainty in the volume of the vessels and uncertainty in the inventory 

estimates therein, technically sufficient. 
6-1 

6.0 CONTRIBUTORS 

Project Manager for the West Valley Draft WIR Evaluation Review: 

Nishka Devaser 

M.S., Nuclear Engineering, University of Tennessee 

B.S., Nuclear and Radiological Engineering, University of Tennessee 

Lead Technical Reviewer: 

Dr. Leah Spradley Parks 

Systems Performance Analyst 

Ph.D., Environmental Management, Vanderbilt University 

M.S., Environmental Engineering, Vanderbilt University 

B.S., Systems and Information Engineering, University of Virginia 

Technical Reviewer: 

Amy Hixon 

M.S., Environmental Engineering & Earth Science, Clemson University 

B.S., Chemistry, Radford University 
7-1 

7.0 REFERENCES 

To avoid confusion, the reference suffixes throughout the document and in this section were, 

when applicable, kept consistent with the suffix used in the draft WIR evaluation (DOE, 2012b). 

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34 FR 8712, Federal Register, Proposed Rulemaking for 10 CFR 50, Licensing and Production 

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7-2 

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7-3 

U.S. Department of Energy, Statement of Work for Technical Assistance Request to the US 

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States Department of Energy and New York State Energy Research and Development Authority 

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7-4 

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WVNSCO, CFMT Water Flush, Work Order VFS-63499, completed documentation. West 

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No. ML12194A498. 

WVNSCO, CFMT Water Flush Number 2, Work Order VFS-65670, completed documentation. 

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Accession No. ML12194A493. 

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No. ML12194A569. 

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Equipment, Radiation and Contamination Survey Report 123427. West Valley Nuclear Services 

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