ML20129J936

Enclosure 1 - Columbia Fuel Fabrication Facility Evaluation in Support of 10 CFR 20.2003 Request for Alternate Waste Disposal

ML20129J936
Person / Time
Site:	Westinghouse
Issue date:	05/08/2020
From:	Westinghouse
To:	Office of Nuclear Material Safety and Safeguards
Shared Package
ML20129J934	List: ML20129J935 ML20129J936 ML20129J937 ML20129J938 ML20129J939
References
LTR-RAC-20-45
Download: ML20129J936 (31)

Text

Enclosure 1 to LTR-RAC-20-45 Date: May 8, 2020 Enclosure 1 COLUMBIA FUEL FABRICATION FACILITY EVALUATION IN SUPPORT OF 10 CFR 20.2002 REQUEST FOR ALTERNATE WASTE DISPOSAL DOCKET NO. 70-1151

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to LTR-RAC-20-45 Date: May 8, 2020

1.0 INTRODUCTION

Westinghouse Electric Company, LLC (Westinghouse) requests U.S. Nuclear Regulatory Commission (NRC) authorization for alternate disposal of specified low-activity waste containing special nuclear material (SNM) from the Columbia Fuel Fabrication Facility (CFFF),

License No. SNM-1107. The authority of 10 CFR 20.2002, and the exemptions requested herein from the requirements in 10 CFR 30.3 and 10 CFR 70.3 for byproduct material and SNM would allow Westinghouse to transfer the specific waste for disposal at the US Ecology Idaho, Inc.

(USEI) disposal facility located near Grand View, Idaho. The USEI disposal facility is a Subtitle C Resource Conservation and Recovery Act (RCRA) hazardous waste disposal facility permitted by the State of Idaho to receive radioactive waste that is not licensed or exempted from licensing by the NRC.

The dose evaluation for this request for alternate disposal was performed using US Ecologys NRC-Approved Site Specific Dose Assessment Methodology (SSDA) for USEI. The SSDA provides a consolidated dose assessment framework for all occupational, transportation, and post-closure dose receptors required in 10 CFR 20.2002(d) - Analyses and procedures to ensure that doses are maintained ALARA and within the dose limits in this part. The information provided in this enclosure as well as the Technical Evaluation Report documents and Safety Evaluation Report produced by the NRC serve to satisfy the requirements in 10 CFR 20.2002(a),

(b), and (c). The NRC approved the SSDA for use on August 24, 2015 (ADAMS Accession No. ML15125A364, provided as Enclosure 2).

Characteristics and operating parameters of the USEI disposal site are summarized in Section 2 of this Enclosure. Environmental conditions at the USEI site are well-documented in previous submittals to the NRC, including the Westinghouse Hematite Decommissioning Project (Docket

1. 70-00036) and the Humboldt Bay Nuclear Power Plant Decommissioning Project (Docket #50-133).

A description of the material to be disposed is included in Sections 3 and 4. The material description includes physical and chemical properties of the material important to risk evaluation and the proposed conditions of waste disposal. Results of the SSDA dose evaluation are summarized in Section 5 for all occupational and transportation workers as well as postulated members of the public based on USEIs ResRad model (Ver. 6.5) and Inadvertent Intruder Scenarios described in NUREG-0782, Draft Environmental Impact Statement on 10 CFR Part 61 Licensing Requirements for Land Disposal of Radioactive Waste and NUREG/CR-4370, Update of Part 61 Impacts Analysis Methodology - Methodology Report. Enclosure 3 contains the Waste Acceptance Criteria (WAC) set forth in USEIs permit issued by the Idaho Department of Environmental Quality. The SSDA Data Input Screens with the project inputs for the CFFF waste is provided in Enclosure 4. The conclusion confirms doses to workers and members of the public will be well below NRC limits.

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to LTR-RAC-20-45 Date: May 8, 2020 2.0 DISPOSAL SITE CHARACTERISTICS The USEI site is located in the Owyhee Desert of southwestern Idaho. It is at the end of Lemley Road, approximately 17 kilometers (10.5 miles) northwest of Grand View, (Owyhee County)

Idaho. Grand View has a population of approximately 340. Owyhee County is a ranching and agricultural area of approximately 19,900 square kilometers (7,678 square miles). The county is sparsely populated, with an average population of 0.5 people per square kilometer (1.4 people per square mile per Reference 1).

This region has an arid climate with an average annual precipitation rate of 7.4 inches. The USEI site is located on a 1.6 kilometer (1 mile) wide plateau. Maximum surface relief on the facility is 27 meters (90 feet) and the mean surface elevation is 790 meters (2,600 feet) above sea level. The nearest residence is 1.6 kilometers (1 mile) southwest of the site. There are no other land uses in the immediate vicinity of the site.

The operational performance characteristics of the USEI site have been reviewed by the NRC and determined to be protective within the NRCs less than a few millirem (mrem) per year policy for Alternate Disposal Requests first stated in NRC Regulatory Issue Summary (RIS) 2004-08, Results of the License Termination Rule Analysis, and reaffirmed in SECY-07-0060, Basis for Justification and Approval Process for 10 CFR 20.2002 Authorizations and Options for Change. The NRC has previously granted USEI 10 CFR 70.17 special nuclear material and 10 CFR 30.11 byproduct material exemptions for purposes of disposal of various licensee waste streams. Two key documents are referenced from previous NRC submittals:

• Hazardous Waste Facility Siting License Application for Cell 16 (American Geotechnics, dated June 30, 2006); This document describes USEIs environmental setting and was accepted by the Idaho Department of Environmental Quality (IDEQ) as part of the 2005 siting process, which resulted in IDEQ approval (December 6, 2006) of USEIs request to expand its landfill operations. (ADAMS Accession No. ML100320540 - Attachment 7)

• Summary of Hydrogeologic Conditions and Groundwater Flow Model for US Ecology Idaho Facility, Grand View, Idaho (Eagle Resources, dated January 13, 2010. This document provides a detailed description of USEIs site geology and hydrogeology. (ADAMS Accession No. ML101170554 - Exhibit B)

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to LTR-RAC-20-45 Date: May 8, 2020

3.0 DESCRIPTION

OF CFFF LEGACY WASTE 3.1 EAST LAGOON The East Lagoon is a treatment/settling pond that is approximately 160 x 130. The East Lagoon receives liquid inputs such as effluent from the Deionized Water Building (primarily from regeneration water from resin beds) and rainwater from containment areas such as the chemical tank farm. The East lagoon also provides extra capacity for overflow from other lagoons or for containment in the event of a spill or emergency. Current East Lagoon operations are regulated under Westinghouses National Pollutant Discharge Elimination System (NPDES) permit for the Columbia site. However, based on past wastewater treatment area operations and the age of the East Lagoon liner, a Wastewater Treatment Area Operable Unit (OU) was established under the South Carolina Department of Health and Environmental Control (SCDHEC) Consent Agreement signed on February 26, 2019. As part of the Consent Agreement, the East Lagoon is scheduled for closure and remediation.

The East Lagoon contains approximately 3 to 4 feet of radiologically contaminated sludge. The East Lagoon was originally lined in the early 1980s. With this liner still in place, it is assumed it may have lost some integrity. Therefore, there is the possibility of soil contamination under and around the East Lagoon due to leaching.

Approximately 45,000 ft3 of sludge, soil and debris will be generated from the closure of the East Lagoon. The waste from the East Lagoon being considered under this request is contaminated with SNM (low enriched uranium (<5 wt% U-235) and the fission product Technetium-99 (Tc-99). The SNM and Tc-99 contaminated wastes were generated from plant operations during the fabrication of nuclear fuel. Tc-99 is present in the process due to uranium feed that originated from sources of recycled uranium or down-blended high enriched uranium.

3.1.1 EAST LAGOON RADIOLOGICAL CHARACTERIZATION In preparation for closure of the East Lagoon and the need to dispose of the East Lagoon materials, a sampling campaign was undertaken. A sampling work plan was developed by Westinghouse to characterize sludge. The sampling work plan was submitted to SCDHEC for approval prior to sampling, and SCDHEC representatives were present during the sampling to collect split samples.

This plan established methods for obtaining sludge samples for radiological and chemical analysis to evaluate subsurface conditions of the East Lagoon. The characterization plan was written to ensure the proper collection, handling, documentation, and evaluation of sludge samples in support of the CFFF environmental goals. East Lagoon sludge sampling was designed to follow EPA guidance, as advised in EPA Region 4 Operating Procedure SESDPROC-200-R3, Sediment Sampling. The East Lagoon sludge was measured to be up to approximately 3 to 4 feet in depth and covered by a shallow layer of water. Based on this understanding, the method selected to sample the East Lagoon was to use a Sludge Push Probe, with an internal acetate sleeve that is capable of collecting sediment in vertical columns.

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to LTR-RAC-20-45 Date: May 8, 2020 A systematic grid strategy was selected for sampling. The grid sampling ensured that the sludge is fully and uniformly represented in the collected data. The East Lagoon was sub-divided into 15 separate grids using east-west and north-south transecting lines. In addition to the systematic samples, biased samples were also collected at locations where site processes discharge to the East Lagoon. At locations where the biased samples and systematic samples were co-located, a single sample location was used for representation of that specific grid and the process inlet. All samples were analyzed for Uranium, Tc-99, Fluoride, Nitrate, and Ammonia. In addition, a more extensive parameter list was analyzed in three samples. The more extensive analyses include the full Toxicity Characteristic Leaching Parameters (TCLP) list and the TCL/TAL except for pesticides and herbicides (in both TCLP and TCL) since these parameters are not potentially present in the sludge. All chemical sample results were found to be well below Hazardous Waste levels, and do not affect the planned disposal routes.

A review of the analytical data indicated that the sample at Location #16 was higher for Tc-99 relative to the results for the other sample locations. To verify and validate the Tc-99 results for the East Lagoon sludge, additional sampling in close proximity to Location #16 was performed.

The results indicated diminishing Tc-99 concentrations extending from Location #16, confirming that the result for Location #16 is an isolated elevated area and is not representative of the entirety of the East Lagoon. Nevertheless, to be conservative, the Location #16 Tc-99 data has been utilized for radiological characterization of the East Lagoon.

Table 3.1 provides a summary of the radiological sample data from the East Lagoon sampling campaign. Enclosure 5 contains the laboratory analytical reports for the East Lagoon samples.

Figure 3.1 shows the East Lagoon sampling grid, along with the approximate location of all 16 samples collected from the East Lagoon.

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Enclosure 1 to LTR-RAC-20-45 Date: May 8, 2020 Table 3.1 East Lagoon Radiological Summary Dry Weight Corrected Conc. (pCi/g)

Analyte (pCi/g)

Sample Description Grid #  % Moisture Sample ID Moisture Corrected U- U-U-235 U-238 Tc-99 U-235 U-238 Tc-99 234 234 Location #1 1 0.91 100819-01 370 13 45 0 33 1 4 0 Location #2 2 0.709 100819-03 103 4 15 0 30 1 4 0 Location #3 2 0.541 100819-06 488 35 62 0 224 16 28 0 Location #4 3 0.475 100919-01 155 6 21 0.3 81 3 11 0.3 Location #5 4 0.776 100819-04 883 33 126 0 198 7 28 0 Location #6 5 0.694 100919-02 456 18 59 0 140 5 18 0 Location #7 6 0.781 100919-03 1,698 69 240 0 372 15 53 0 Location #8 7 0.673 100919-04 309 12 53 0 101 4 17 0 Location #9 8 0.778 100919-05 8,086 339 1,296 0 1,795 75 288 0 Location #10 9 0.79 100819-05 13,373 565 2,055 5.19 2,808 119 432 5.19 Location #11 10 0.8 100919-06 10,698 440 1,710 8.23 2,140 88 342 8.23 Location #12 11 0.715 100919-07 4,466 192 798 7.5 1,273 55 227 7.5 Location #13 12 0.76 100919-08 4,373 178 783 5.16 1,049 43 188 5.16 Location #14 13 0.752 100919-09 6,282 246 1,077 6.55 1,558 61 267 6.55 Location #15 14 0.787 100919-10 15,177 634 2,535 3.06 3,233 135 540 3.06 Location #16 15 0.524 100919-11 1,045 42 162 164 497 20 77 164 AVERAGE: 4,248 177 690 13 971 41 158 13 Note: For Tc-99, values in Italics are reported as positive, but less than the Reporting Limit of 50 pCi/g. Negative values are reported as zero.

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to LTR-RAC-20-45 Date: May 8, 2020 Figure 3.1 East Lagoon Radiological Sampling Grid

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to LTR-RAC-20-45 Date: May 8, 2020 3.2 CALCIUM FLUORIDE PILE In addition, CFFF intends to dispose of approximately 50,400 ft3 of solid CaF2 sludge dredged from the Calcium Fluoride Lagoons and subsequently placed in a storage pile. The CaF2 sludge was generated as a waste from uranium recovery waste treatment process. The CaF2 waste being considered under this request is contaminated with SNM (low enriched uranium {<5wt%

U-235}).

3.2.1 CALCIUM FLUORIDE PILE RADIOLOGICAL CHARACTERIZATION As the Calcium Fluoride Lagoons were dredged on numerous occasions over time, solid CaF2 material was collected and radiological sampling was performed. CaF2 piles identified during dredging operations to be under an average concentration of 30 pCi/g Total U were segregated, and the material was dispositioned as originally approved by NRC in License Amendment 8 to SNM-1107, dated December 9, 1997. Any CaF2 piles identified during dredging operations to be over an average concentration of 30 pCi/g Total U were segregated and stored on site.

Approximately 50,400 ft3 was segregated, and 164 radiological samples for Total Uranium were collected from this material over the course of several years. Table 3.2 below provides a summary of the radiological sample data from the historic CaF2 sampling. Enclosure 6 contains the full radiological sample data set.

Table 3.2 Calcium Fluoride Pile Summary of Historical Sample Results Sample Range (Total U) U-235 Sample Range (3.88 wt% U-235)

Minimum 11.3 pCi/g Minimum 0.4 pCi/g Average 54.1 pCi/g Average 1.7 pCi/g Maximum 175.5 pCi/g Maximum 5.6 pCi/g Isotopic U Activity (based on average concentration) Isotopic U Mass U-234 45.6 pCi/g U-234 16.7 g U-235 1.7 pCi/g U-235 1,841 g U-238 6.7 pCi/g U-238 45,874 g 47,732 g Total U

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to LTR-RAC-20-45 Date: May 8, 2020 3.3 UF6 CYLINDERS In addition, CFFF intends to dispose of up to 526 obsolete UF6 Cylinders, which represents a disposal volume of approximately 23,000 ft3 prior to downsizing. The UF6 Cylinders are transportation containers that are no longer in service. The UF6 Cylinders are solid form (steel),

approximately 6 feet in length and 2.5 feet in diameter. The UF6 Cylinders are empty and upon last use had previously been through the UF6 Cylinder internal wash/rinse process prior to being placed into storage for pending disposal. The UF6 Cylinders will be downsized to eliminate void space prior to packaging for shipment off-site for disposal. While emptied and cleaned to the standard mentioned, the UF6 Cylinders are internally contaminated with SNM.

The UF6 Cylinders will be transported to the USEI site by means of trucks separate from the Aggregated Waste shipments.

3.3.1 UF6 CYLINDERS RADIOLOGICAL CHARACTERIZATION To facilitate the radiological characterization of the UF6 Cylinders to be disposed of in this request, fifteen (15) UF6 Cylinders were selected to be cut in half to allow access to the internal surfaces of the UF6 Cylinders to perform radiological survey and characterization. These 15 UF6 Cylinders were randomly selected from the population of UF6 Cylinders segregated for disposal. A characterization sampling plan was developed by Westinghouse to guide the handling, cutting, inspection, and radiological survey of each UF6 Cylinder.

After each UF6 Cylinder was cut in half, a visual inspection was performed to determine if any scale or product buildup was present. No scale or buildup was identified on any of the 15 UF6 Cylinders identified for inspection. Next, a radiological survey was performed. Each UF6 Cylinder half was surveyed internally and externally with a gamma sensitive NaI 2x2 probe.

Then the internal UF6 Cylinder halves were scanned using an alpha sensitive frisker, and smeared to determine removable activity.

Table 3.3 below provides a summary of the radiological sample data from the UF6 Cylinder inspection and sampling. Enclosure 7 contains the full radiological sample data set.

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to LTR-RAC-20-45 Date: May 8, 2020 Table 3.3 UF6 Cylinder Radiological Data Summary UF6 Cylinder Physical Dimensions weight 1480 lbs 671,316 g length 85 inches 215.9 cm diameter 30 inches 76.2 cm radius 15 inches 38.1 cm wall thickness 0.6 inch internal SA cm2 60,804 Summary of 15 UF6 Cylinder Surveys Internal Gamma Scan Internal Alpha Scan Removable Alpha Removable Beta gcpm dpm/100cm2 dpm/100cm2 dpm/100cm2 Min 25,000 9,000 79 25 Mean 70,000 69,000 2,859 2,090 Max 125,000 254,000 11,279 6,719 Average U Activity: 1.9E-05 Curies per cylinder

@ assumed 5% max EU:

U-234: 1.55738E-05 curies = 0.003 grams = 23.2 pCi/g U-235: 8.59413E-07 curies = 0.4 grams = 1.3 pCi/g U-238: 2.5384E-06 curies = 7.6 grams = 3.8 pCi/g grams Total U 8.0 per cylinder

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to LTR-RAC-20-45 Date: May 8, 2020 3.4 ADDITIONAL SOILS As indicated in Section 3.1, the CFFF site is required to complete remediation and closure of the current East Lagoon. As a prudent measure it is planned that additional soil waste will be generated from the remediation effort.

It is expected that some contamination will exist in the soil underlying the East Lagoon liner given the long operating history of the Lagoon and the potential for a liner system leak.

However, the CFFF site is prohibited from performing soil sampling to verify contamination levels with the East Lagoon liner in its current configuration. Therefore a typical radiological characterization campaign cannot be undertaken.

3.4.1 ADDITIONAL SOIL RADIOLOGICAL CHARACTERIZATION It can be inferred that if the underlying soil is contaminated that the source of the contamination would be the East Lagoon sludge, via a leak in the liner. Based on operational experience and process knowledge there is no reason to expect underlying soil to have higher concentrations than what is in the East Lagoon sludge. Rather, site operations history indicates that the underlying soil should have only minor contamination. This submittal makes the conservative assumption that a similar volume of material must be removed from the underlying soil to what is physically in the East Lagoon. This is viewed as a conservative assumption from both an activity and volume perspective. The additional soil volume requested in this submittal is meant to cover contingencies in waste volume for the East Lagoon and underlying soil should the soil be found to be contaminated.

Upon removal of the liner, the newly exposed underlying soil area will be subjected to a preliminary radiological survey and a 16 point grid based systematic sampling. Additional bias samples will be taken for any hotspots identified by the preliminary and/or based on visual inspection. The following analytical will be run for each sample, at a minimum: Isotopic Uranium and Tc-99 by liquid scintillation.

Soil identified as additional waste soil will be removed and will either be directly packaged for disposal at USEI, aggregated with other lower-contaminated soil, CaF2, or East Lagoon material to meet the criteria in Table 4.2, or will be shipped to a licensed Class A Low Level Radioactive Waste disposal facility if necessary based on radiological concentrations. No soil exceeding the USEI WAC will be shipped to USEI.

For the purpose of completing the Additional Soil radiological characterization, data from the East Lagoon sludge will be utilized for additional soil generated during the East Lagoon remediation effort.

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to LTR-RAC-20-45 Date: May 8, 2020

4.0 DESCRIPTION

OF WASTE FOR DISPOSAL 4.1 WASTE AGGREGATION FOR DISPOSAL The radiological characterization data provided in Section 3.0 indicates that each individual portion of the CFFF Legacy Waste described above could be individually packaged and shipped to USEI such that it would meet the sites waste acceptance criteria. Nevertheless, the East Lagoon remediation plan and waste shipment plans include stabilization of a significant portion of the CFFF Legacy waste for shipment and co-mingling of waste streams to ensure concentration limits are met and to maximize transport efficiency.

Based upon the WAC demonstration (as described below) an aggregated waste mix design was developed for remediation of the East Lagoon. At various times throughout the process of remediation of the East Lagoon, the crew will segregate portions of the sludge material and stabilize that material using a 10% by volume mix of Type 2 Portland cement stored in super sacks. Once the Portland cement has been applied to the sludge material, an excavator will mix the material into a homogenous mixture and let these interim mixes cure for approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Utilizing Portland cement to stabilize the waste will add approximately 12,500 ft3 to the total waste volume shipped to USE. Once the Portland cement/sludge material has been cured, the crew will add approximately 40% CaF2, mix the material a second time and load it into the onsite haul truck for transportation to the material storage area. The resultant aggregated waste will then be packaged for shipment to USEI.

4.1.1 WAC DEMONSTRATION FOR AGGREGATED WASTE To demonstrate that the aggregated waste will meet the physical waste acceptance criteria at USEI (e.g. no free liquids), waste acceptance criteria bench scale testing was performed by USEI personnel at the CFFF site. US Ecology conducted the full scope of this study on site.

US Ecologys objective of the treatability study was to determine the ideal waste mixture, and determine several factors including:

1. Ensure the waste material maintains a pH between 2-12;

2. Ensure the waste material passes paint filter and paint shaker tests;

3. Maximize the potential utilization of the existing CaF2 as a stabilizing reagent;

4. Identify the most effective reagent while reducing the overall weight of the waste;

5. Minimize off gassing and release of noxious odors during mixing driven by organics.

The results of the bench scale testing demonstrated that the ideal mixture of waste consisted of approximately 50% East Lagoon sludge, 40% CaF2, and 10% Portland cement. This mixture met all of the physical characteristics necessary for shipment to the USEI waste disposal facility.

The final waste mixture was also sampled in the same manner as the East Lagoon sludge, and the radiological results of the Waste Profile sample are provided in Table 4.1.

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Enclosure 1 to LTR-RAC-20-45 Date: May 8, 2020 Table 4.1 Waste Profile Sample Radiological Results Dry Weight Moisture Corrected Sample  % Analyte (pCi/g)

Sample ID Conc. (pCi/g)

Description Moisture U-234 U-235 U-238 Tc-99 U-234 U-235 U-238 Tc-99 Waste Profile 0.443 101019-01 216 8 57 1.18 121 5 32 1.18 4.1.2 AGGREGATED WASTE CHARACTERIZATION All waste destined for disposal at USEI will be excavated as necessary, mixed and handled as described above, and moved to a Material Staging Area (MSA). Once the final waste mixture has been relocated to the MSA, a minimum of 1 (one) radiological sample (Iso-U, and Tc-99) will be collected from each approximate 100 cubic yards of soil like material prior to packaging.

Each sample will be analyzed by an offsite laboratory to confirm that the material meets the USEI WAC prior to shipment offsite.

Table 4.2 Aggregated Waste Characterization Waste Volume (ft3) Mass(g) U-234 (pCi/g) U-235 (pCi/g) U-238 (pCi/g) Tc-99 (pCi/g)

CaF2 50,400 2.28E+09 45.6 1.7 6.7 0 East Lagoon 44,776 2.23E+09 971 41 158 13 Underlying Soils 44,776 2.23E+09 971 41 158 13 Portland Cement 12,500 5.10E+08 0 0 0 0 Totals Weighted Average Aggregated 152,452 7.26E+09 611.4 25.5 99.2 8.0 Material 4.2 UF6 CYLINDER DISPOSAL The UF6 Cylinders will be transported to USEI by lined 50 cubic yard/23 ton capable aluminum end-dump trucks. UF6 Cylinders will be cut in half and will be sampled via direct alpha scan measurement, and smear sample surveys for on-site analysis prior to loading for shipment.

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to LTR-RAC-20-45 Date: May 8, 2020 5.0 RADIOLOGICAL ASSESSMENT As described in the following exposure scenarios, the dose equivalent for the Maximally Exposed Individual (MEI) has been demonstrated to not exceed a few mrem per year. The standard of a few mrem per year to a member of the public is set forth in NRC RIS 2004-08, Results of the License Termination Rule Analysis. The NRC has clarified in the Guidance For The Review Of Proposed Disposal Procedures And Transfers Of Radioactive Material Under 10 CFR 20.2002 and 10 CFR 40.13(a) final draft that a few mrem per year should be understood as less than 5 mrem/year. The transportation workers and USEI workers are treated as members of the public because the USEI site, while permitted by the State of Idaho under RCRA to accept certain radioactive materials, is not licensed by the NRC.

External exposure assessments in the SSDA were performed using MicroShield Code, Version 7.02. Evaluations of potential external and internal dose hazards are discussed in the sections that follow while all inputs to the SSDA workbook are provided in Enclosure 4. A summary of total estimated doses for all transporters, as well as USEI workers performing surveying, handling, treatment and disposal tasks on the CFFF waste is provided in Table 5.1.

Since the UF6 Cylinders will be transported to USEI by means of trucks separately from the East Lagoon/CaF2 material, a separate SSDA model was run for the UF6 Cylinders. Dose results are summed for any job functions that are shared between the two models. In addition, post closure dose scenarios are summed between the two models to report the overall post closure dose.

Results are discussed below.

5.1 TRANSPORT DOSE TO THE PUBLIC All materials will be transported by truck or a combination of truck and rail to the USEI facility in Grand View, ID. All conveyances will be verified to comply with DOT external loose surface contamination limits prior to shipment. Therefore, transport will not pose the potential for internal dose to the drivers or other members of the public. All loads will meet the DOT requirements for packaging.

East Lagoon and CaF2 aggregated material will be loaded into 9 cubic yard/10 ton IP-1 bags and staged for transport. Once the bag is ready to be shipped, it will be lifted into a standard 50 cubic yard / 22 ton capacity aluminum end-dump truck. Two bags will be placed in each truck. The truck will then be tarped and proceed 5 miles to the railyard where the bags will be lifted into lined gondola railcars. Accounting for the estimated volume of the material, 355 truckloads will be required to haul bagged material to the railyard. Modeled doses to the truck drivers for this process are reported in Table 5.1.

Transportation dose with respect to the gondola cars is expected to be very low. Calculated exposure rates 1 meter from the surface of the rail car would be 6.91E-4 mrem/year. In order for a member of the public to receive a dose greater than a few mrem, they would have to stand within 1 meter of the car for over 7,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. This is a very unlikely scenario and not considered to be credible.

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to LTR-RAC-20-45 Date: May 8, 2020 The backend-dray portion of the transportation takes approximately 45 minutes from the Rail Transfer Facility (RTF). Truck transport is shared between 8 drivers. The dose model assumes the driver sits 0.6 meters from the material. Dose results are reported in Table 5.1.

The UF6 Cylinders will be transported to USEI by lined 50 cubic yards/23 ton capable aluminum end-dump trucks. The volume of the UF6 Cylinders is estimated to be ~857 cubic yards /368 tons. UF6 Cylinders will be cut in half before they are loaded onto the trucks. The distance from CFFF to the USEI disposal facility is approximately 2,520 miles. Assuming an average speed of 55 miles per hour, the trip is estimated to take 46.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Five drivers will be used to haul the estimated 20 loads to transport the entire volume of UF6 Cylinders. Each truck driver is expected to make 4 round trips over the course of the project. For dose modelling purposes, the driver sits approximately 4 meters from the surface of the vessel with approximately 0.25 inches of steel and aluminum shielding between him and the UF6 Cylinders. The shielding accounts for the aluminum end of the trailer and back wall of the truck cab. Because of the low average concentrations of radionuclides, external doses to the truck drivers are also very low. As a result, the dose to other members of the general public can reasonably be concluded to be much less.

The MEI for transportation dose to the public as described above is the Back-End Dray driver with a max calculated dose of 1.96E-02 mrem/year.

5.2 USEI WORKER DOSE ASSESSMENT External dose rates in the SSDA are calculated using dose-to-source ratios (DSR) developed with the Micro Shield Code, Version 7.02. A total dose rate for all nuclides present is calculated by summing the contributions from the individual nuclides. Specifics for these templates used for each job function can be found in the Technical Basis Document for the SSDA. In addition, below are summaries of each USEI worker function and the assumptions used in performing the dose calculations.

Internal Doses are calculated using Dose Conversion Factors from Federal Guidance Report 11 for all of the radionuclides present. The SSDA uses a dust loading fraction of 2.3E-04 g/m3, and a standard man breathing rate of 1.2 m3/hr for light work. (ICRP, 2004). A total dose rate for all nuclides present is calculated by summing the contributions from the individual nuclides.

Based on the SSDA dose modeling performed, the MEI is the RTF Excavator Operator with a calculated dose of 7.19E-1 mrem/year. This dose is low and well within the few millirem requirement. Results for the below described functions are reported in Table 5.1 below.

Gondola Railcar Surveyor Upon receipt at USEIs RTF, the gondolas will be surveyed and screened prior to transloading the material to trucks and transporting to USEI Site 2 for direct disposal. Approximately 10 minutes is required to perform a survey of each gondola. Based on current practice, a surveyor is assumed to stand at a distance of one meter from the gondola during the survey, with four surveyors sharing the task.

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to LTR-RAC-20-45 Date: May 8, 2020 RTF Excavator Operator All transloading of material are done within a containment building employing a 24,000 cubic feet per minute (cfm) filtration system. An excavator positioned on top of a bridge platform above the railcar will transfer the material into end-dump trucks. During off-loading operations the excavator operator remains in the cab that pulls air through a filtration system.

For dose modeling it is assumed off-loading of a gondola car can take up to 45 minutes. The operator sits approximately 2 meters from the material. Two excavator operators share these activities.

Gondola Railcar Cleanout Once a railcar is off-loaded, USEI personnel will remove any residual material inside of the railcars with shovels and brooms. This operation normally takes 10 minutes to complete. Four personnel share this task. The dose rate is modeled at 30cm from a 1/2 layer of waste material.

RTF Truck Surveyor Once trucks are loaded, surveys will be performed and screened prior to the material being sent to the disposal site. Truck surveys take 5 minutes to perform. Surveyors are assumed to stand one meter from the truck or trailer during the survey. Four surveyors share this task.

Disposal Site Truck Surveyor Since the UF6 Cylinders are being transported directly to the disposal site, surveys will be performed there and not at the RTF. Modeling assumptions are the same for this function as they are for the RTF truck surveyor.

Cell Operator After delivery to the disposal cell, a bulldozer operator wearing a respirator within an enclosed cab, spreads and compacts the waste. For this dose scenario the deposited material is based on the volume of one gondola car. It is assumed that 15 minutes is needed to spread and compact the volume of material, equivalent to one gondola car. Two personnel share this responsibility.

It is important to note that this function will be shared between the two SSDA models, specifically, the East Lagoon/CaF2/Soils and UF6 Cylinders. Doses from both models are therefore summed to calculate total dose for the project.

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Enclosure 1 to LTR-RAC-20-45 Date: May 8, 2020 Table 5.1 Results of SSDA Dose Evaluation for CFFF Waste Project Total Total External External Project Minimum Waste Exposure Internal Dose per Total Internal Dose per % of Max Number of Contact Time Rate Dose Rate Distance Total No. of Worker Dose per Worker Worker Annual Function Workers (hr) (mrem/hr) (mrem/hr) (m) Repetitions (mrem) (mrem) (mrem) MEI Dose Legacy Waste Streams- East Lagoon/CaF2/Soils Front-End Dray Truck Drivers 4 0.09 9.99E-04 0.00E+00 0.6 355 8.06E-03 0.00E+00 8.06E-03 0.2%

Gondola Railcar Surveyors 4 0.16 6.91E-04 0.00E+00 1.0 71 1.96E-03 0.00E+00 1.96E-03 0.0%

Bulk/IMC Truck Surveyors (RTF) 4 0.08 7.82E-04 0.00E+00 1.0 209 3.27E-03 0.00E+00 3.27E-03 0.1%

RTF Excavator Operator 2 0.75 5.25E-04 2.65E-02 2.0 71 1.40E-02 7.05E-01 7.19E-01 14.4%

Gondola Railcar Cleanout 4 0.16 6.63E-04 2.65E-02 0.3 71 1.88E-03 7.52E-02 7.71E-02 1.5%

Back-End Dray Truck Drivers 8 0.75 9.99E-04 0.00E+00 0.6 209 1.96E-02 0.00E+00 1.96E-02 0.4%

Landfill Cell Operators 2 0.25 1.95E-04 2.65E-02 1.0 142 3.45E-03 4.70E-01 4.74E-01 9.5%

Legacy Waste Streams- UF6 Cylinders Long-Haul Direct Truck Drivers - Drive Time 5 45.45 5.64E-05 0.00E+00 0.6 20 1.03E-02 0.00E+00 1.03E-02 0.2%

Bulk/IMC Truck Surveyors (disposal site) 4 0.08 4.58E-05 0.00E+00 1.0 20 1.83E-05 0.00E+00 1.83E-05 0.0%

Landfill Cell Operators 2 0.25 1.30E-05 1.03E-03 1.0 8 1.30E-05 1.03E-03 1.04E-03 0.0%

Project Dose-Totals Long-Haul Direct Truck Drivers - Drive Time 1.03E-02 0.00E+00 1.03E-02 0.2%

Front-End Dray Truck Drivers 8.06E-03 0.00E+00 8.06E-03 0.2%

Gondola Railcar Surveyors 1.96E-03 0.00E+00 1.96E-03 0.0%

Bulk/IMC Truck Surveyors (RTF-highest dose reported) 3.27E-03 0.00+00 3.27E-03 0.1%

RTF Excavator Operator 1.40E-02 7.02E-01 7.19E-01 14.4%

Gondola Railcar Cleanout 1.88E-03 7.48E-02 7.71E-02 1.5%

Back-End Dray Truck Drivers 1.96E-02 0.00E+00 1.96E-02 0.4%

Landfill Cell Operators (summed) 3.46E-03 4.71E-01 4.75E-01 9.5%

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to LTR-RAC-20-45 Date: May 8, 2020 5.3 POST CLOSURE DOSE TO THE GENERAL PUBLIC USEIs RCRA permit requires that it demonstrate that no person will receive an annual dose exceeding 15 mrem for 1,000 years after closure of the facility. This standard is more restrictive than the annual 25 mrem total effective dose equivalent (TEDE) stated in 10 CFR 20.1402 for NRC license termination, as well as the limits for near surface disposal of low-level radioactive waste set forth in 10 CFR 61. RESRAD code Version 6.5 was used for modeling the Grand View site for potential long-term post-closure doses. A number of default parameters in the Grand View model have been replaced with site specific parameters consistent with the facilitys 2005 permit modification and a report prepared by its consultant (previously submitted to the NRC as part of a Request for Additional Information response for the exemption request for the Westinghouse Hematite project, Docket #070-00036, ML12135A301).

The SSDA contains a screening RESRAD model to assess the impact of the CFFF waste on the USEI site. The model is consistent with USEIs post-closure dose model included in the Part B RCRA permit, which assumes that all of the CFFF waste is distributed evenly within the contaminated zone (area = 88,221 m2, depth = 33.6 m). Screening in the SSDA means that ALL nuclides are evaluated at their peak dose-to-source ratio regardless of when it occurs. The radionuclide concentrations in Tables 3.1, 3.2 and 3.3 are automatically adjusted in the SSDA Workbook to reflect aggregation into the entire landfill volume, resulting in a concentration dilution factor of over 50. All other RESRAD code parameters remain the same. The results of the screening model show a maximum annual dose of 4.20E-01 mrem. Due to the very low dose projection from the screening model, a separate project-specific dose model was not necessary.

Three post-closure inadvertent intruder scenarios were also conducted using the framework from NUREG-0782, Draft Environmental Impact Statement on 10 CFR Part 61 Licensing Requirements for Land Disposal of Radioactive Waste, and NUREG/CR-4370, Volume 1, Update of Part 61 Impacts Analysis Methodology built into the SSDA. These scenarios include:

• Intruder Construction Scenario - An inadvertent intruder may excavate or construct a building on a disposal site following a breakdown in institutional controls. Under these circumstances, dust will be generated from the application of mechanical forces to the surface materials (soil, rock) through tools and implements (wheels, blades) that pulverize and abrade these materials. The dust particles generated may be then entrained by localized turbulent air currents and can thus become available for inhalation by the intruder. The intruder may also be exposed to direct gamma radiation resulting from airborne particulates and by working directly in the waste-soil mixture. The Construction Worker scenario uses the Air Uptake and Direct Gamma Exposure pathways to estimate a total dose to the intruder.

• Intruder Well Drilling Scenario - An intruder accesses the site and develops a well. The intruder is exposed to contaminated drill cuttings spread over the ground surface and contaminated airborne dust. The scenario presented in NUREG/CR- 4370 was modified to exclude consideration of exposure to cuttings in a mud pit due to the standard practices in the area around the waste site. The assumption that drill cuttings are spread over the ground will

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to LTR-RAC-20-45 Date: May 8, 2020 result in higher dose estimates than if the cuttings were assumed to be in a mud pit because of the decrease in the shielding factor. The driller is assumed to work on site for a period of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> and it is assumed that the contaminated layer is drilled through in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. As such, the driller is assumed to be exposed to the undiluted cuttings for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and to diluted material for the balance of the exposure duration. The dilution is calculated based on the ratio of the depth of the waste layer to the total well depth. No dilution in the USEI landfill is assumed.

The Well Driller scenario includes contributions from Internal and External dose to the intruder.

• Intruder Driller Occupancy Scenario - An inadvertent intruder occupies the site upon which a well had been drilled through waste materials. The Driller Occupancy Scenario uses the same concentrations in the exhumed well cuttings as the Well Driller scenario. The Driller Occupancy scenario uses the Air Uptake and Direct Gamma Exposure pathways to estimate a total dose to the intruder.

To be more complete with respect to post closure dose modeling, additional intruder scenarios were considered. These inadvertent intruder scenarios are not held to the same post closure dose standard as USEIs RCRA permitted RESRAD model as the NRC allows up to a 500 mrem/year dose limit (NUREG-2175). With relation to the material of interest, the estimated inadvertent intruder doses for the three above scenarios were calculated to be 8.79 mrem for the Construction Scenario, 1.01 mrem for the Well Driller Scenario, and 1.68E-01 mrem for the Driller Occupancy Scenario, as reported in Table 5.2. Even though a higher dose is allowed for inadvertent intruder post closure scenarios, each of these estimated doses meet USEIs RCRA permit post closure dose limit of 15mrem. These models are very conservative by design as to have flexibility to be used as a general tool for various types of sites. These estimates are likely overly conservative and not realistic, which is especially the case for the construction scenario.

For example, USEI has a requirement that all radiological waste must be placed no closer than 3.6 meters from the top of the constructed cap. The Intruder Construction scenario assumes excavation for constructing a building up to 3.0 meters below the surface with the lower 1.0 meter consisting of waste. Realistically in this scenario, the waste will not even be disturbed by the construction activities.

Table 5.2 USEI SSDA Post Closure Results for CFFF Waste Project East Lagoon/CaF2/Soils UF6 Cylinders Project Total (mrem/yr) (mrem/yr) (mrem/yr)

USEI RESRAD Post-Closure Screening Dose 4.20E-01 4.87E-05 4.20E-01 Inadvertent Intruder Doses

1. Construction Scenario 8.67E+00 1.22E-01 8.79E+00

2. Well Driller Scenario 9.67E-01 3.70E-02 1.01E+00

3. Driller Occupancy Scenario 1.65E-01 2.66E-03 1.68E-01

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to LTR-RAC-20-45 Date: May 8, 2020 6.0 CRITICALITY SAFETY A Criticality Safety Assessment for the USEI site was performed as part of a prior alternate disposal application by the Westinghouse Hematite site. The Nuclear Criticality Safety Assessment of the US Ecology Idaho (USEI) Site for the Land Fill Disposal of Decommissioning Waste from the Hematite Site, Rev. 2 (NSA, 2011) verified that wastes containing U-235 may be sent to the USEI site for disposal since very large margins of safety had been incorporated into the normal operating conditions associated with these wastes and the probability for serious abnormal conditions is acceptably small. A maximum fissile concentration of 0.1 gram U-235 per liter of media was developed as an inherently safe concentration of SNM for the exhumed Hematite waste materials. This converts to an equivalent activity concentration of 216 pCi/g U-235 in soil (assuming a soil density of 1 g/cc).

To achieve the average activity concentration, the candidate waste will be aggregated as described in Section 4.0. It is intended to only utilize the waste described in this submittal for aggregation of waste. To ensure the activity of the candidate waste as packaged for shipment does not exceed an average activity concentration of 216 pCi/g U-235, the waste will be sampled to verify the average activity concentration is acceptable for disposal at USEI. USEI personnel will review the sample data to ensure acceptability of the waste for disposal prior to shipment to the USEI site.

Considering the characterization results of the candidate waste, USEIs WAC is the limiting factor as it would be exceeded before the 0.1 gram of U-235 per liter of media safety limit is reached.

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to LTR-RAC-20-45 Date: May 8, 2020

7.0 CONCLUSION

CFFF developed this request and related evaluation in consultation with USEI, including health physics personnel responsible for the Grand View disposal facilitys waste acceptance and radiological performance assessment programs. This assessment team performed a radiological dose assessment of the material to be shipped and determined that the potential dose to the workers involved in the transportation and placement of the material and to members of the general public after site closure is less than one mrem per year TEDE and that criticality safety margins are maintained. This dose is a small fraction of the NRC decommissioning limits for exposure to any member of the public of 25 mrem/year TEDE, and is well within the few mrem per year criterion that the NRC has established in RIS 2004-08.

8.0 REFERENCES

8.1 American Geotechnics, Hazardous Waste Facility Siting License Application Cell 16, Project No. 06B-C1202, June 30, 2006 (ML100320540 - Attachment 7) 8.2 Eagle Resources, Inc. Summary of Hydrogeologic Conditions and Groundwater Flow Model for US Ecology Idaho Facility, Grand View, Idaho. January 13, 2010 (ML101170554 - Exhibit B) 8.3 US Ecology Idaho, Inc. USEI Site B Permit No. IDD073114654 (2004) 8.4 U.S. Nuclear Regulatory Commission, Regulatory Issue Summary 2004-08, Results of the License Termination Rule Analysis. Office of Material Safety and Safeguards, May 28, 2004 8.5 U.S. Nuclear Regulatory Commission, Basis and Justification for Approval process for 10CFR20.2002 Authorizations and Options for Change. SECY-070060. Division of Waste Management and Environmental Protection, March 27, 2007 8.6 U.S. Nuclear Regulatory Commission, Guidance For The Reviews Of Proposed Disposal Procedures And Transfers Of Radioactive Material Under 10 CFR 20.2002 And 10 CFR 40.13(a), Division of Uranium Recovery, Decommissioning, And Waste Programs Guidance Document, October 16, 2017 8.7 Nuclear Safety Associates Nuclear Criticality Safety Assessment of the US Ecology Idaho (USEI) Site for the Land Fill Disposal of Decommissioning Waste from the Hematite Site, Rev. 2 NSA-TR-09-14 8.8 NUREG-2175, Guidance for Conducting Technical Analysis for 10 CFR Part 61, Nuclear Regulatory Commission, Washington, DC, March 2015 8.9 Federal Guidance Report No. 11: Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, EPA 520/1-88-020 September 1988

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to LTR-RAC-20-45 Date: May 8, 2020 Enclosure 2 Copy of Letter from L. Camper to J. Weismann approving use of USEI SSDA for 10 CFR 20.2002 Alternate Disposal Authorization Requests, August 24, 2015 (ML15125A364)

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to LTR-RAC-20-45 Date: May 8, 2020 August 24, 2015 Mr. Joseph J. Weismann, CHP Vice President of Radiological Programs and Field Services US Ecology, Inc.

Lakepointe Centre I 300 East Mallard Dr., Suite 300 Boise, ID 83706

SUBJECT:

US ECOLOGY, INC. - TECHNICAL EVALUATION REPORT OF US ECOLOGY IDAHOS PROPOSED METHODOLOGY SUPPORTING ALTERNATE WASTE DISPOSAL PROCEDURES IN ACCORDANCE WITH 10 CFR 20.2002 By letter dated June 14, 2013, US Ecology, Inc. (USEI) requested an exemption to receive and dispose of low-activity radioactive waste from Studsviks Processing Facility in Memphis, TN at USEI, a Resource Conservation and Recovery Act Subtitle-C hazardous and low-activity waste facility near Grand View, ID. USEI also requested that the U.S. Nuclear Regulatory Commission (NRC) review a newly developed Site-Specific Dose Assessment Methodology (SSDA). In a letter dated March 10, 2014, USEI withdrew the request to dispose of low-activity waste from Studsvik Processing Facility; however, USEI requested that the NRC continue to review the SSDA. USEI stated that this process provides a streamlined methodology for preparing and reviewing future 10 CFR 20.2002 alternate disposal requests (ADR) from USEI.

This Technical Evaluation Report (TER) documents the NRC staffs technical review of the proposed methodology. Similar to a review of a 10 CFR 20.2002 exemption request, the NRC staff performed a technical review of the methodology and associated documents and evaluated the technical basis and assumptions incorporated into the calculations used by USEI. The NRC staff also used the methodology to evaluate a previously evaluated exemption request and compared the conclusions. Based on this review, the NRC staff considers the use of USEIs SSDA to be an appropriate method for evaluating future proposed disposals. The SSDA methodology can be used to satisfy the criteria in § 20.2002 (d); however, individual 20.2002 requests by USEI, or other licensees wishing to ship to USEI, must address the criteria in

§ 20.2002 (a), (b), or (c) separately.

In response to your initial request, the SSDA, the technical basis document, and the NRCs detailed TER are considered proprietary and will not be available for public review. However, a second, publicly-available TER was also developed to demonstrate how this process will satisfy the NRCs mission of protecting public health, safety, and the environment. The NRC would note that specific parameter values, in the necessary form, that have not always been included with historical submittals may need to be included in future submittals in order for the SSDA methodology to be used.

J. Weismann - 2 -

to LTR-RAC-20-45 Date: May 8, 2020 In accordance with 10 CFR 2.390 of the NRCs Agency Rules of Practice and Procedure, a copy of this letter will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records component of NRCs Agencywide Documents Access and Management System (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html.

Copies of both TERs are enclosed. Please contact Mr. Maurice Heath if you have any questions concerning the above. He can be reached at (301) 415-3137 or via email at Maurice.Heath@nrc.gov.

Sincerely,

/RA/

Larry W. Camper, Director Division of Decommissioning, Uranium Recovery, and Waste Programs Office of Nuclear Material Safety and Safeguards

Enclosures:

Technical Evaluation Report (Proprietary Version)

Technical Evaluation Report (Public Version)

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to LTR-RAC-20-45 Date: May 8, 2020 Enclosure 3 USEI Part B Permit EPA ID. No.: IDD073114654 Revision Date: July 28, 2016 Part C.3.2 WASTE ACCEPTANCE CRITERIA

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to LTR-RAC-20-45 Date: May 8, 2020 US Ecology Idaho, Inc.

EPA ID. No.: IDD073114654 Effective Date: July 28, 2016 C.3.2 Radioactive Material Waste Acceptance Criteria The following waste acceptance criteria are established for accepting radiological contaminated waste material that is not regulated under the Atomic Energy Act of 1954 (AEA), as amended. This may be accomplished by the following regulatory mechanisms; use of a general or specific exemption from regulation by the Nuclear Regulatory Commission (NRC) or an Agreement State; a Release from Radiological Control declaration by the Department of Energy (DOE); or a determination that 91(b) radioactive material is no longer regulated by the Department of Defense (DoD). Material may also be accepted if it is not regulated or licensed by the NRC or Agreement State or has been authorized for disposal by the IDEQ and is within the numeric waste acceptance criteria. Waste acceptance criteria are consistent with these restrictions.

The following five tables establish types and concentrations of radioactive materials that may be accepted. These tables are based on categories and types of radioactive material not regulated by the NRC, an Agreement State, the DOE, or the DoD for alternate disposal. The criteria are consistent with these restrictions and detailed analyses set forth in Waste Acceptance Criteria and Justification for FUSRAP Material, prepared by Radiation Safety Associates, Inc. (RSA) as subsequently refined, expanded and updated in Waste Acceptance Criteria and Justification for Radioactive Material, prepared by USEI.

Material may be accepted if the material has been specifically exempted from regulation by rule, order, license, license condition, letter of interpretation, or specific authorization under the following conditions:

Thirty (30) days prior to intended shipment of such materials to the facility, USEI shall notify IDEQ of its intent to accept such material and submit information describing the materials physical, radiological, and/or chemical properties, impact on the facility radioactive materials performance assessment, and the basis for determining that the material does not require disposal at a facility licensed under the AEA. The IDEQ will have 30 days from receipt of this notification to reject USEIs determination or require further information and review. No response by IDEQ within thirty (30) days following receipt of such notice shall constitute concurrence. IDEQ concurrence is not required for generally exempted material as set forth in Table C-4a.

Based on categories of waste described in the waste acceptance criteria, the concentration of the various radionuclides in the conveyance (e.g., rail car gondola, other container etc.) shall not exceed the concentration limits established in the WAC without the specific written approval of the IDEQ unless generally exempted as set forth in Table C-4a. Radiological surveys will be performed as outlined in Exempt Radiological Materials Procedure-01 (ERMP-01) to verify compliance with the WAC. If individual pockets of activity are detected indicating the limits may be exceeded, the RSO or RPS shall investigate the discrepancy and estimate the extent or volume of the material with the potentially elevated radiation levels. The RPS or RSO shall then make a determination on the compliance of the entire conveyance load with the appropriate WAC limits. If the conveyance is determined not to meet the limits, USEI will notify IDEQs RCRA Program Manager within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a concentration based exceedance of the facility WAC to evaluate and discuss management options. The findings and resolution actions shall then be documented and submitted to the IDEQ.

The radioactive material waste acceptance criteria, when used in conjunction with an effective radiation monitoring and protection program as defined in the USEI Radioactive Material Health and Safety Plan and Exempt Radioactive Materials Procedures provides adequate protection of human health and the environment. Included within this manual are requirements for USEI to submit a written summary report of all radioactive material waste receipts showing volumes and radionuclide concentrations and total activities disposed at the USEI site on a quarterly basis. The 4th quarter report of each year will also include an updated analysis of the cumulative impact on the facility performance assessment based upon the previous years waste receipt.

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to LTR-RAC-20-45 Date: May 8, 2020 These criteria and procedures are designed to assure that the highest potential dose to a worker handling radioactive material at USEI shall not exceed 400 mrem/year TEDE dose, and that no member of the public is calculated to receive a potential post closure dose exceeding 15 mrem/year TEDE dose, from the USEI program. TEDE is defined as the Total Effective Dose Equivalent, which equals the sum of external and internal exposures. The public dose limit during operation activities is limited to 100 mrem/yr TEDE dose. An annual summary report of environmental monitoring results will be submitted to IDEQ by June 1st for the preceding year.

Materials that have a radioactive component that meets the criteria described in Tables C-1 through C-4c and are RCRA regulated material will be managed as described within this WAP for the RCRA regulated constituents.

Table C-1: Unimportant Quantities of Source Material Uniformly Dispersed in Soil or Other Media**

Sum of Concentrations Maximum Concentration of Status of Equilibrium Parent(s) and all progeny Source Material present a Natural uranium in equilibrium with <500 ppm / 167 pCi/g (238U activity) < 3000 pCi/g progeny Refined natural uranium <500 ppm / 167 pCi/g (238U activity) < 2000 pCi/g Depleted Uranium <500 ppm / 169 pCi/g < 2000 pCi/g b Natural thorium <500 ppm / 55 pCi/g (232Th activity) < 2000 pCi/g 230 0.1 ppm / <2000 pCi/g Th (with no progeny)

Any mixture of Thorium and Sum of ratios < 1**** < 2000 pCi/g Uranium 238U, 235U, 234U, 234Th, 234mPa, 231Th

• Refined Uranium includes Table C-2: Naturally Occurring Radioactive Material Other Than Uranium and Thorium Uniformly Dispersed in Soil or Other Media**

Maximum Concentration of Sum of Concentrations of Parent and Status of Equilibrium All Progeny Present Parent Nuclide a 226Ra or 228Ra with progeny in bulk form 1 500 pCi/g 4500 pCi/g b 226Ra or 228Ra with progeny in reinforced 1500 pCi/g 13,500 pCi/g IP-1 containers 1 c 210Pb with progeny( Bi & 210Po) 1500 pCi/g 4500 pCi/g 40K 818 pCi/g N/A Any other NORM 3000 pCi/g

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to LTR-RAC-20-45 Date: May 8, 2020 1

Any material containing 226Ra greater than 222 pCi/g shall be disposed at least 6 meters from the external point on the completed cell.

Table C-3: Particle Accelerator Produced Radioactive Material Acceptable Material Activity or Concentration Any particle All materials shall be packaged in accordance with USDOT packaging requirements.

accelerator produced Any packages containing iodine or volatile radionuclides will have lids or covers radionuclide.

sealed to the container with gaskets. Contamination levels on the surface of the packages shall not exceed those allowed at point of receipt by USDOT rules. Gamma or x-ray radiation levels may not exceed 10 millirem per hour anywhere on the surface of the package. All packages received shall be directly disposed in the active cell. All containers shall be certified to be 90% full.

Average over conveyance or container. The use of the phrase over the conveyance or container is meant to reflect the variability on the generator side. The concentration limit is the primary acceptance criteria.

• • Unless otherwise authorized by IDEQ, other Media does not include radioactively contaminated liquid (except for incidental liquids in materials). See radioactive contaminated liquid definition (definition section of Part B permit).

• • • Conc. of U in sample + Conc. of Th in Sample < 1 Allowable conc. of U Allowable conc. of Th Table C-4a: NRC Exempted Products, Devices or Items Exemption 10 CFR Isotope, Activity or Product, Device or Item Concentration Part*

30.15 As listed in the regulation Various isotopes and activities as set forth in 30.15 30.14, Other materials, products or devices specifically exempted Radionuclides in concentrations consistent with 30.18 from regulation by rule, order, license, license condition, the exemption concurrence, or letter of interpretation 30.19 Self-luminous products containing tritium, 85Kr, 3H or 147Pm Activity by Manufacturing license 30.20 Gas and aerosol detectors for protection of life and property Isotope and activity by Manufacturing license from fire 30.21 Capsules containing 14C urea for in vivo diagnosis of humans 14C, one µCi per capsule 31.12 General License for certain items and self-luminous products As set forth in 31.12 and see containing Radium 226

1. 4 under Additional information below 40.13(a) Unimportant quantity of source material: see Table C-1 <0.05% by weight source material 40.13(b) Unrefined and unprocessed ore containing source material As set forth in rule

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to LTR-RAC-20-45 Date: May 8, 2020 40.13(c)(1) Source material in incandescent gas mantles, vacuum tubes, Thorium and uranium, various welding rods, electric lamps for illumination amounts or concentrations, see rules 40.13(c)(2) (i)Source material in glazed ceramic tableware <20% by weight (ii)Piezoelectric ceramic <2% by weight (iii) Glassware not including glass brick, pane glass, ceramic <10% by weight tile, or other glass or ceramic used in construction 40.13(c)(3) Photographic film, negatives or prints Uranium or Thorium 40.13(c)(4) Finished product or part fabricated of or containing tungsten <4% by weight thorium or magnesium-thorium alloys. Cannot treat or process content.

chemically, metallurgically, or physically.

40.13(c)(5) Uranium contained in counterweights installed in aircraft, Per stated conditions in rule.

rockets, projectiles and missiles or stored or handled in connection with installation or removal of such counterweights.

40.13(c)(6) Uranium used as shielding in shipping containers if Depleted Uranium conspicuously and legibly impressed with legend CAUTION RADIOACTIVE SHIELDING - URANIUM and uranium incased in at least 1/8 inch thick steel or fire resistant metal.

40.13(c)(7) Thorium contained in finished optical lenses <30% by weight thorium, per conditions in rule.

40.13(c)(8) Thorium contained in any finished aircraft engine part <4% by weight thorium, per conditions in rule.

containing nickel-thoria alloy.

Table C-4b: Materials Specifically Exempted by the NRC or NRC Agreement State Isotope, Activity or Exemption Materials Concentration*

10 CFR Byproduct material including production particle Byproduct material at accelerator material exempted from NRC or concentrations consistent 30.11** Agreement State regulation by rule, order, license, license condition or letter of interpretation may be with the exemption accepted as determined by specific NRC or Agreement State exemption.***

10 CFR Source material exempted from NRC or Agreement Source material at concentrations State regulation by rule, order, license, license consistent 40.14**

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to LTR-RAC-20-45 Date: May 8, 2020 condition or letter of interpretation may be accepted as with the exemption.

determined by specific NRC or Agreement State exemption.***

10 CFR 70.17 Special Nuclear Material (SNM) exempted from SNM at concentrations NRC regulation by rule, order, license, license consistent with the condition or letter of interpretation may be accepted as determined by specific NRC or exemption.

Agreement State exemption.***

• Sum of all isotopes up to a maximum concentration of 3,000 pCi/gm.

• • Alternate disposals authorized by Agreement States also require an NRC exemption for the purposes of disposal in the State of Idaho.

• • • Similar material not regulated or licensed by the NRC may also be accepted. Sum of all isotopes up to a maximum concentration of 3,000 pCi/gm. IDEQ shall be notified prior to the receipt of Special Nuclear Material not regulated or licensed by the NRC.

Table C-4c Material Released by Other Government Agencies Isotope, Activity or Exemption Materials Concentration*

US DOE Radioactive materials that have Radioactive materials at concentrations consistent been released or cleared from with the Release**

radiological control US DoD Radioactive materials determined not to be regulated under the Radioactive materials at AEA under authority granted to concentrations consistent the DoD in Section 91(b) of the with the Authorization**

AEA of 1954, as amended

• Sum or all isotopes up to a maximum of 3,000 pCi/gm.

• • May include byproduct materials, source materials and special nuclear material as defined in the AEA of 1954 as amended.

NORM and Particle Accelerator Produced Radioactive Material may also be accepted under Tables C.2 and C.3, as part of these Releases and Authorizations.

Additional Information for USEIs Waste Analysis Plan

1. US Ecology Idaho, Inc. (USEI) may receive contaminated materials or other materials as described in Tables C C-4b above. USEI may not accept for disposal any material that by its possession would require USEI to have a radioactive material license from the Nuclear Regulatory Commission (NRC).

2. Unless approved in advance by USEI and IDEQ, average activity concentrations may not exceed those concentrations enumerated in Tables C-1 and C-2. Additionally, for Tables C-1 and C-2, individual pockets of material may exceed the WAC for the radionuclides present as long as the average concentration of all radionuclides within the package or conveyance remains at or below the WAC and the highest dose rate measured on the outside of the unshielded package or conveyance does not exceed those action levels enumerated in ERMP-01.

3. Other items, devices or materials listed in Table C-4a, which are exempted in accordance with 10 CFR Parts 30, 40 or equivalent Agreement State regulations or 10 CFR Part 70 may be accepted at or below the activities (per device or item) or concentrations specified in those exemptions.

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to LTR-RAC-20-45 Date: May 8, 2020

4. 10CFR20.2008 authorizes disposal of certain byproduct material as defined in Section 11.e(3) and 11.e(4) of the Atomic Energy Act, as amended, at disposal facilities authorized to dispose of such material in accordance with any Federal or State solid or hazardous waste law, as authorized under the Energy Policy Act of 2005.

5. The generator of particle accelerator produced waste must specify that the waste meets applicable acceptance criteria.

6. In accordance with permit requirements, notification of any exceedance of the WAC will be provided to the RCRA Program Manager within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, in accordance with the permit.

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