ML19261A329

Final Status Survey Final Report - Phase 1

ML19261A329
Person / Time
Site:	La Crosse  File:Dairyland Power Cooperative icon.png
Issue date:	08/29/2019
From:	Yetter R LaCrosseSolutions
To:	Office of Nuclear Material Safety and Safeguards
References
LC-2019-0040
Download: ML19261A329 (61)

Text

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La Crosse Boiling Water Reactor FINAL STATUS SURVEY FINAL REPORT - PHASE 1 August 2019

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 2

Prepared By:

R. Yetter III Date:

8/28/19 Survey Design Specialist Reviewed By:

P. Hollenbeck Date:

Radiological Engineer Reviewed By:

R. Yetter Date:

8/28/19 Director of Radiological Site Closure Approved By:

S. Zoller Date:

FSS Manager urvey Design Specialist Radiological Engineer 8/29/19 FSS Mana n ger 8/29/19

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 3

TABLE OF CONTENTS 1

Introduction............................................................................................................................. 6 1.1 Executive Summary......................................................................................................... 6 1.2 Phased Submittal Approach............................................................................................. 9 2

Final Status Survey Program Overview................................................................................ 11 2.1 Survey Planning............................................................................................................. 13 2.2 Survey Design................................................................................................................ 19 2.3 Survey Implementation.................................................................................................. 23 2.4 Survey Data Assessment................................................................................................ 24 2.5 Quality Assurance and Quality Control Measures......................................................... 24 3

Site Information.................................................................................................................... 25 3.1 Site Description.............................................................................................................. 25 3.2 Survey Unit Description................................................................................................. 29 3.3 Summary of Historical Radiological Data..................................................................... 31 3.3.1 Characterization Surveys........................................................................................ 32 3.3.2 Continuing Characterization................................................................................... 33 3.3.3 Remedial Action Support Surveys.......................................................................... 35 3.4 Conditions at the Time of Final Status Survey.............................................................. 37 3.5 Identification of Potential Contaminants........................................................................ 37 3.6 Radiological Release Criteria......................................................................................... 38 4

Final Status Survey Protocol................................................................................................. 38 4.1 Data Quality Objectives................................................................................................. 38 4.2 Survey Unit Designation and Classification.................................................................. 45 4.3 Background Determination............................................................................................ 45 4.4 Final Status Survey Sample Plans.................................................................................. 45 4.5 Survey Design................................................................................................................ 46 4.5.1 Determination of Number of Data Points............................................................... 46 4.5.2 Sample and Measurement Locations...................................................................... 47 4.6 Instrumentation............................................................................................................... 47 4.6.1 Detector Efficiencies............................................................................................... 48 4.6.2 Detector Sensitivities.............................................................................................. 48 4.6.3 Instrument Maintenance and Control...................................................................... 49 4.6.4 Instrument Calibration............................................................................................ 49 4.7 Survey Methodology...................................................................................................... 49

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 4

4.7.1 Scan Surveys........................................................................................................... 49 4.7.2 ISOCS Measurement Collection............................................................................. 51 4.7.3 Soil Sampling.......................................................................................................... 51 4.8 Quality Control Surveys................................................................................................. 52 5

Survey Findings.................................................................................................................... 52 5.1 Survey Data Conversion................................................................................................. 53 5.2 Survey Data Verification and Validation....................................................................... 54 5.3 Anomalous Data/Elevated Scan Results and Investigation........................................... 56 5.4 Evaluation of Number of Sample/Measurement Locations in Survey Units................. 56 5.5 Comparison of Findings with Derived Concentration Guideline Levels....................... 56 5.6 Description of ALARA to Achieve Final Activity Levels............................................. 58 5.7 NRC/Independent Verification Team Findings............................................................. 59 6

Summary............................................................................................................................... 59 7

References............................................................................................................................. 60 8

Appendices............................................................................................................................ 61

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 5

LIST OF TABLES Table 1-1 Phase 1 Survey Units...............................................................................................................................7 Table 2-1 Base Case and Operational DCGLs for Soil (1)......................................................................................17 Table 2-2 Base Case and Operational DCGLs for the Structure Basements (1)......................................................17 Table 2-3 Typical Final Status Survey Unit Areas.................................................................................................18 Table 2-4 Dose Significant Radionuclides and Mixture (1)....................................................................................20 Table 2-5 Recommended Survey Coverage for FSS (1)..........................................................................................21 Table 2-6 Adjusted Minimum Number of ISOCS Measurements per FSS Unit (1)................................................22 Table 4-1 FSS Investigation Levels.......................................................................................................................40 Table 4-2 Sr-90 to Cs-137 Surrogate Ratios (1)......................................................................................................40 Table 4-3 Action Levels for Phase 1 Survey Units................................................................................................42 Table 4-4 Number of Samples and Measurements for FSS...................................................................................47 Table 4-5 Recommended Scan Coverage..............................................................................................................50 Table 4-6 Summary of Total Area Scanned...........................................................................................................51 Table 5-1 Basic Statistical Properties of Phase 1 Open Land Systematic/Random Sample Populations...............54 Table 5-2 Basic Statistical Properties of Phase 1 Structure Basement Systematic Measurement Populations......54 Table 5-3 Mean Base Case SOF and Dose Contribution.......................................................................................58 LIST OF FIGURES Figure 1-1 Phase 1 Survey Unit Locations................................................................................................................8 Figure 2-1 Characterization/License Termination Group Organizational Chart......................................................12

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 6

1 Introduction 1.1 Executive Summary The purpose of this Phase 1 Final Status Survey (FSS) Final Report is to provide a summary of the survey results and overall conclusions which demonstrate that the La Crosse Boiling Water Reactor (LACBWR) facility, or portions of the site, meets the 25 mrem/yr release criterion as established in Nuclear Regulatory Commission (NRC)

Regulation 10 CFR 20.1402, Radiological Criteria for Unrestricted Use. This Phase 1 FSS Final Report encompasses open land areas, excavations and structure basements. The FSS results provided herein demonstrate that any residual radioactivity results in a Total Effective Dose Equivalent (TEDE) to an average member of the critical group (AMCG) that does not exceed 25 mrem/yr and that the residual radioactivity has been reduced to levels that are as low as reasonably achievable (ALARA). The release criterion is translated into site-specific Derived Concentration Guideline Levels (DCGLs) for assessment and summary.

This report documents that FSS activities were performed consistent with the guidance provided in the LACBWR License Termination Plan (LTP) (Reference 1); NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) (Reference 2);

LC-QA-PN-001, Final Status Survey Quality Assurance Project Plan (QAPP) (Reference 3); LC-FS-PR-002, Final Status Survey Package Development (Reference 4); LC-FS-PR-015, Final Status Survey for Structures (Reference 5); LC-FS-PR-010, Isolation and Control for Final Status Survey (Reference 6); LC-FS-PR-008, Final Status Survey Data Assessment (Reference 7); as well as various station implementing procedures.

This FSS Final Report has been written consistent with the guidance provided in NUREG-1757, Vol. 2, Consolidated Decommissioning Guidance Characterization, Survey, and Determination of Radiological Criteria - Final Report (Reference 8); MARSSIM; and the requirements specified in LC-FS-PR-009, Final Status Survey Data Reporting (Reference 9).

To facilitate the data management process, FSS Final Reports incorporate multiple Survey Unit Release Records. Release Records are complete and unambiguous records of the as-left radiological status of specific survey units. Sufficient data and information are provided in each Release Record to enable an independent re-creation and evaluation at some future time of both the survey activities and the derived results.

This Phase 1 Final Report specifically addresses four (4) sub-grade excavation (SGE) survey units, three (3) open land (OLA) survey units, and two (2) structure basement (STB) survey units. This report contains a compilation of all nine (9) Release Records that are within the Phase 1 scope. Table 1-1 provides a listing of all the survey units addressed in

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 7

this report, along with their classifications and size. Figure 1-1 depicts the locations of the survey units in relation to the LACBWR site as well as survey unit boundaries.

All FSS activities essential to data quality have been implemented and performed under approved procedures. Trained individuals, using properly calibrated instruments and laboratory equipment that are sensitive to the suspected contaminants, performed the FSS of the Phase 1 survey units. The survey data for all Phase 1 survey units demonstrate that the dose (TEDE) from residual radioactivity is less than the maximum annual dose (TEDE) which corresponds to the release criterion for license termination for unrestricted use specified in 10 CFR 20.1402 and support the release of these areas from the 10 CFR 50 license. Additionally, the ALARA requirement of 10 CFR 20.1402 has been satisfied.

Table 1-1 Phase 1 Survey Units Survey Unit Type Survey Unit Description Class Size (m2)

L1-SUB-DRS SGE RCA North Area Excavation 1

1,125 L1-SUB-TDS SGE TB, Sump, Pit Diesel Excavation 1

1,186 L1-SUB-LES SGE LSA, Eat Shack, Septic Excavation 1

1,336 L1-010-101C SGE Waste Treatment Building Excavation 1

88 L2-011-102 OLA Area South of LACBWR Site Enclosure (LSE) Fence 2

2,258 L2-011-103 OLA G-3 Crib House Surrounding Area 2

2,445 L3-012-102 OLA Transmission Switch Yard 3

11,711 B1-010-004 STB Waste Gas Tank Vault Basement 1

311 B1-010-001 STB Reactor Building Basement 1

512 As stated above, this report contains only the results of the FSS that address the dose due to soil and structures. The dose from groundwater will be calculated using the Groundwater Exposure Factors presented in Chapter 6 of the LTP. Because the industrial use scenario does not include irrigation the only exposure pathway from groundwater is potable water from an onsite well. The results of the dose calculation for groundwater will be presented in a separate final report.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 8

Figure 1-1 Phase 1 Survey Unit Locations

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 9

1.2 Phased Submittal Approach To minimize the incorporation of redundant historical assessment and other FSS program information, and to facilitate potential phased releases from the current license, FSS Final Reports will be prepared in a phased approach. LaCrosseSolutions estimates that a total of three (3) FSS Final Reports will be generated and submitted to the NRC during the decommissioning project.

Release of Non-Impacted Open lands On June 27, 2016, LaCrosseSolutions submitted a request (ADAMS Accession No. ML16181A068) to release a portion of the LACBWR site from the 10 CFR 50 license (DPR-45) in accordance with 10 CFR 50.83, Release of Part of a Power Reactor Facility or Site for Unrestricted Use, and 10 CFR 100, Reactor Site Criteria. A report was generated for the request that addressed the release of 88 of the 165 acres that comprise the LACBWR site. That report contains a summary of the final assessment performed as well as a summary of the characterization surveys performed of these non-impacted open-land survey units. LaCrosseSolutions reviewed and assessed the subject property to ensure that the radiological condition of these land areas will have no adverse impact on the ability of the site, in aggregate, to meet the 10 CFR 20, Subpart E, Radiological Criteria for License Termination. The five (5) survey units incorporated within the report are classified as non-impacted, and as such, no statistical tests, scan measurements, static measurements, or elevated measurement comparisons are required. The release of the non-impacted areas from the license(s) was approved by the NRC on April 12, 2017.

Phase 2 and 3 Final Status Survey Final Reports As discussed above, LaCrosseSolutions anticipates two (2) additional FSS Final Report submittals. The schedule and identity of survey units included in each of the remaining submittals were developed based on a review of the demolition and FSS schedule, as well as in consideration of NRC review requirements. The demolition schedule, including the cleanup of demolition debris to allow survey access, is dynamic and subject to continued refinement. With potential changes in the decommissioning schedule, it is possible that interim submittals will be filed with the NRC with the goal of providing Survey Unit Release Records as soon as possible to support review and the potential release of the survey units.

The Phase 2 FSS Final Report will encompass the seven (7) above grade buildings that will remain on site and the seven (7) buried pipe survey units. The Phase 3 FSS Final Report will encompass all remaining open land areas and sub-grade excavations, as well as groundwater. Table 1-2 provides a list of survey units within each submittal phase.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 10 Table 1-2 Listing of Survey Units within each FSS Final Report Phased Submittal Survey Unit Survey Unit Description Class Phase L4-012-103 G-3 Coal Plant Grounds Non-Impacted L4-012-105 Coal Pile Grounds Non-Impacted L4-012-106 Capped Ash Impoundment Grounds Non-Impacted L4-012-107 Grounds East of Highway 35 Non-Impacted L4-012-108 Hwy 35/Railroad Right of Way Grounds Non-Impacted L1-SUB-DRS RCA North Area 1

1 L1-SUB-TDS TB, Sump, Pit, Diesel 1

1 L1-SUB-LES LSA, Eat Shack, Septic 1

1 L1-010-101 C Waste Treatment Building 1

1 L2-011-102 Area South of LSE Fence 2

1 L2-011-103 LACBWR Crib House, Surrounding Area 2

1 L3-012-102 Transmission Switch Yard 3

1 B1-010-001 Reactor Building 1

1 B1-010-004 Waste Gas Tank Vault 1

1 B2-010-101 LACBWR Crib House 2

2 B2-010-102 G-3 Crib House 2

2 B2-010-103 LACBWR Administration Building 2

2 B3-012-101 Back-up Control Center 3

2 B3-012-102 Transmission Sub-Station Switch House 3

2 B3-012-103 G-1 Crib House 3

2 B3-012-104 Barge Washing Break Room 3

2 B3-012-109 Security Station 3

2 S1-011-102 Circulating Water Discharge Pipe 1

2 S2-011-103 Circulating Water Intake Pipe 2

2 S2-011-103 A De-Icing Line 2

2 S2-011-103 B Low Pressure Service Water 2

2 S3-012-109 A Storm Drain 1 (South of BCC) 3 2

S3-012-109 B Storm Drain 2 (48 line to the river) 3 2

S2-011-101 A Storm Drain 3 (10 PVC East of Admin.)

2 2

S2-011-101 B Storm Drain 6 (RCA to SD3) 2 2

S2-011-102 A Storm Drain 4 (West river to Corridor) 3 2

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 11 Survey Unit Survey Unit Description Class Phase S2-011-102 B Storm Drain 5 (East Access Road to Corridor) 3 2

L1-010-101 Reactor Building, WTB, WGTV, Ventilation Stack Grounds 1

3 L1-010-102 Turbine Building, Turbine Office Building, 1B Diesel Generator Building Grounds 1

3 L1-010-103 LSA Building, Maintenance Eat Shack Grounds 1

3 L1-010-104 North LSE Grounds 1

3 L1-010-105 North Interim Debris Storage Area 1

3 L1-010-106 North Loading Area 1

3 L1-010-107 Outside East LSE Area 1

3 L2-011-101 Area North of LSE Fence 2

3 L2-011-104 G3 Crib House, Circ. Water Discharge Land 2

3 L3-012-101 North End of Licensed Site 3

3 L3-012-109 Plant Access, ISFSI Haul Road Grounds 3

3 L1-SUB-CDR Stack, Pipe Tunnel, RPGPA 1

3 L1-SUB-TDS A Eastern Portion TB, Sump, Pit, Diesel 1

3 L1-SUB-TDS B RPGPA Area 1

3 2

Final Status Survey Program Overview The FSS Program consists of the methods used in planning, designing, conducting, and evaluating FSS at the LACBWR site to demonstrate that the premises are suitable for unrestricted use in accordance with the criteria for decommissioning in Title 10 CFR 20, Subpart E. Final Status Surveys serve as key elements to demonstrate that the TEDE to an AMCG from residual radioactivity does not exceed 25 mrem/yr, and that all residual radioactivity at the site is reduced to levels that are ALARA.

To implement the FSS

Program, LaCrosseSolutions established the Characterization/License Termination (C/LT) Group, within the Radiation Protection division, with sufficient management and technical resources to fulfill project objectives.

The C/LT Group is responsible for the safe completion of all surveys related to characterization and final site closure. Approved site procedures and detailed technical support documents (TSD) direct the FSS process to ensure consistent implementation and adherence to applicable requirements. Figure 2-1 provides an organizational chart of the C/LT Group.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 12 Figure 2-1 Characterization/License Termination Group Organizational Chart

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 13 2.1 Survey Planning The development and planning phase was initiated in 2015 by the EnergySolutions Technical Support Document RS-TD-313196-003, La Crosse Boiling Water Reactor Historical Site Assessment (HSA) (Reference 10) and the initiation of the characterization process. The characterization process is iterative and will continue until, in some cases, the time of completing FSS. The HSA consisted of a review of site historical records regarding plant incidents, radiological survey documents, and routine and special reports submitted by Dairyland Power to various regulatory agencies. Along with these assessments, interviews with current and past site personnel, reviews of historical site photos, and extensive area inspections were performed to meet the following objectives:

x Develop the information necessary to support FSS design, including the development of Data Quality Objectives (DQOs) and survey instrument performance standards.

x Develop the initial radiological information to support decommissioning planning, including building decontamination, demolition, and waste disposal.

x Identify any unique radiological or health and safety issues associated with decommissioning.

x Identify the potential and known sources of radioactive contamination in systems, surface or subsurface soils, groundwater, and on structures.

x Divide the LACBWR site into manageable survey units for survey and classification purposes.

x Determine the initial classification of each survey unit as non-impacted or impacted.

Impacted survey units are further designated as Class 1, 2, or 3, as defined in MARSSIM.

DQOs are qualitative and quantitative statements derived from the DQO process that clarify technical and quality objectives, define the appropriate type of data, and specify the tolerable levels or potential decision errors used as the basis for establishing the quality and quantity of data required to support inference and decisions. This process, described in MARSSIM and Procedure LC-FS-PR-002, Final Status Survey Package Development, is a series of graded planning steps found to be effective in establishing criteria for data quality and guiding the development of FSS Sample Plans. DQOs developed and implemented during the initial phase of planning directed all data collection efforts.

The DQO approach consists of the following seven steps:

x State the Problem - This step provides a clear description of the problem, identification of planning team members (especially the decision makers), a conceptual model of the hazard to be investigated, and the estimated resources required to perform the survey. The problem associated with FSS is to determine whether a given survey unit meets the radiological release criterion of 10 CFR 20.1402.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 14 x Identify the Decision - This step consists of developing a decision statement based on a principal study question (i.e., the stated problem) and determining alternative actions that may be taken based on the answer to the principle study question. Alternative actions identify the measures to resolve the problem. The decision statement combines the principal study question and alternative actions into an expression of choice among multiple actions. For the FSS, the principal study question is: Does residual radioactive contamination present in the survey unit exceed the established DCGL values?

The alternative actions may include no action, investigation, resurvey, remediation, and reclassification.

x Identify Inputs to the Decision - The information required depends on the type of media under consideration (e.g., soil, water, or concrete) and whether existing data are sufficient or if new data are needed to make the decision. If the decision can be based on existing data, then the source(s) will be documented and evaluated to ensure reasonable confidence that the data are acceptable. If new data are needed, then the type of measurement (e.g., scan, direct measurement, or sampling) will need to be determined.

x Define the Study Boundaries - This step includes identification of the target population of interest, the spatial and temporal features of that population, the time frame for collecting the data, practical constraints, and the scale of decision making. In FSS, the target population is the set of samples or direct measurements that constitute an area of interest. The medium of interest is specified during the planning process.

The spatial boundaries include soil depth, area dimensions, contained water bodies, and natural boundaries. Temporal boundaries include activities impacted by time-related events including weather conditions, season, operation of equipment under different environmental conditions, resource loading, and work schedule.

x Develop a Decision Rule - This step develops the binary statement that defines a logical process for choosing among alternative actions. The decision rule is a clear statement using the If...then... format and includes action level conditions and the statistical parameter of interest.

x Specify Tolerable Limits on Decision Errors - This step incorporates hypothesis testing and probabilistic sampling distributions to control the decision errors during data analysis. Hypothesis testing is a process based on the scientific method that compares a baseline condition (the null hypothesis) to an alternative condition (the alternative hypothesis). Hypothesis testing rests on the premise that the null hypothesis is true and that sufficient evidence must be provided to reject it.

x Optimize the Design for Obtaining Data - The final step in the DQO process leads to the development of an adequate survey design. By using an on-site analytical laboratory and, as is the case for structure basements, a Canberra In-Situ Object Counting System (ISOCS), sampling and analysis processes are designed to provide near real-time data assessment during implementation of field activities and FSS.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 15 Gamma and beta scans provide information on soil areas and concrete surfaces that have residual radioactivity greater than background and allow appropriate selection of biased sampling and measurement locations. This data will be evaluated and used to refine the scope of field activities to optimize implementation of the FSS design and ensure the DQOs are met.

As stated, the primary objective of the DQO process was to demonstrate that the level of residual radioactivity found in the soils in the land area survey units, and on structure surfaces, including any areas of elevated activity, was equal to or below the site-specific DCGLs that correspond to the 25 mrem/yr release criterion.

Surface soil is defined as the first 15 cm layer of soil, and FSS for surface soil will be performed on the first 15 cm. However, for conservatism, and to ensure efficient implementation of FSS, the surface soil dose assessment assumed a depth of 1 m from the surface. A standard surface soil contamination thickness of 15 cm would result in lower dose (i.e., higher DCGL). Using a 1 m thickness reduces the potential for delays or unnecessary remediation if contamination with a thickness somewhat greater than 15 cm is encountered. For the land area survey units addressed in this final report, no soil contamination was identified with a thickness greater than 1m, and therefore, additional dose modeling was not required. There is low potential for significant subsurface contamination to remain in the End State with a geometry comprised of a clean soil layer over a contaminated soil layer at depth.

Standard methods for RESRAD parameter selection and uncertainty analysis are used consistent with guidance in NUREG-1757. The AMCG is the Industrial Worker.

The ISOCS has been selected as the primary instrument that will be used to perform FSS of basement structures. LaCrosseSolutions TSD LC-FS-TSD-001, Use of ISOCS for FSS of End State Sub Structures at LACBWR (Reference 11) was developed to describe the method and source term geometry assumption that will be used to determine the ISOCS efficiency calibration. Direct beta measurements taken on the concrete surface will not provide the data necessary to determine the residual radioactivity at depth in concrete and therefore, would have to be augmented with core sampling. The ISOCS was selected as the instrument of choice to perform the FSS of basement structures for the following reasons:

x The surface area covered by a single ISOCS measurement is large (a nominal range of 10-30 m2) which essentially eliminates the need for scan surveys.

x Access for ISOCS measurements can be more readily accomplished remotely and does not require extensive and prolonged contact with structural surfaces that would be necessary to perform scan surveys using beta instrumentation.

x ISOCS measurements will provide results that can be used directly to determine total activity with depth in concrete.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 16 x One of the most significant advantages of the ISOCS system in the FSS application is the analytical capability to perform comprehensive uncertainty analysis of various potential source term geometries (depth and areal distribution). After an ISOCS measurement is collected, it can be tested against a variety of geometry assumptions to address uncertainty in the source term geometry if necessary. This uncertainty analysis could potentially be used to generate a clearly conservative result using an efficiency based on a noticeably conservative geometry to resolve questions without additional core samples or measurements.

For the FSS of the WGTV and Reactor Building basements, a majority of the ISOCS measurements were collected using the circular plane geometry although varied geometries were necessary due to the structural anomalies encountered within the survey units. The sump, located in the north-west corner of the WGTV basement floor, required the use of a rectangular plane geometry. ISOCS geometries specifically employed within the Phase 1 structure basement survey units are provided as attachments in the respective Release Records.

Each radionuclide-specific Base Case DCGL is equivalent to the level of residual radioactivity (above background levels) that could, when considered independently, result in a TEDE of 25 mrem/yr to an AMCG. To ensure that the summation of dose from each source term is 25 mrem/yr or less after all FSS is completed, the Base Case DCGLs are reduced based on an expected, or a priori, fraction of the 25 mrem/yr dose limit from each source term. These reduced values are designated as Operational DCGLs and are then used as the DCGL for the FSS design of the survey unit (calculation of surrogate DCGLs, investigation levels, etc.). Details of the Operational DCGLs derived for each dose component and the basis for the applied a priori dose fractions are provided in LaCrosseSolutions TSD LC-FS-TSD-002, Operational Derived Concentration Guideline Levels for Final Status Survey (Reference 12).

Table 2-1 provides a listing for the soil Base Case (DCGLS) and Operational (OpDCGLS)

DCGLs, and Table 2-2 provides the Structure Basement Base Case (DCGLB) and Operational (OpDCGLB) DCGLs of the main Radionuclides of Concern (ROC) used for the FSS of the Phase 1 survey units.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 17 Table 2-1 Base Case and Operational DCGLs for Soil (1)

Radionuclide Base Case DCGL (DCGLS)

Operational DCGL (OpDCGLS)

(pCi/g)

(pCi/g)

Co-60 1.06E+01 3.83E+00 Sr-90 5.47E+03 1.97E+03 Cs-137 4.83E+01 1.74E+01 Eu-152 2.36E+01 8.51E+00 Eu-154 2.19E+01 7.89E+00 (1)

From Tables 5-5 and 5-6 of the LTP Table 2-2 Base Case and Operational DCGLs for the Structure Basements (1)

Radionuclide Reactor Building Base Case DCGL (DCGLB)

Reactor Building Operational DCGL (OpDCGLB)

WGTV Base Case DCGL (DCGLB)

WGTV Operational DCGL (OpDCGLB)

(pCi/m2)

(pCi/m2)

(pCi/m2)

(pCi/m2)

Co-60 5.16E+06 3.61E+05 4.10E+06 2.87E+05 Sr-90 1.45E+07 1.02E+06 6.40E+06 4.48E+05 Cs-137 2.17E+07 1.52E+06 1.76E+07 1.23E+06 Eu-152 1.19E+07 8.33E+05 9.69E+06 6.78E+05 Eu-154 1.10E+07 7.71E+05 8.97E+06 6.28E+05 (1)

From Tables 5-3 and 5-4 of the LTP The development of information to support decommissioning planning and execution was accomplished through a review of all known site radiological and environmental records.

Much of this information was consolidated in the HSA, LaCrosseSolutions TSD LC-RS-PN-164017-001, Characterization Survey Report (Reference 13), and in files containing copies of records maintained pursuant to Title 10 CFR 50.75(g) (1). These documents are discussed further in applicable sections of this report.

An initial objective of site characterization and assessment was to correlate the impact of a radiological event to physical locations on the LACBWR site and to provide a means to correlate subsequent survey data. To satisfy these objectives, the entire 164-acre site was divided into survey areas. Survey area size determination was based upon the specific area and the most efficient and practical size needed to bound the lateral and vertical extent of contamination identified in the area. Survey areas that have no reasonable potential for contamination were classified as non-impacted. These areas had no radiological impact

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 18 from site operations and are identified in the HSA. Survey areas with reasonable potential for contamination were classified as impacted.

Classification, as described in MARSSIM, is the process by which an area or survey unit is described according to its radiological characteristics and potential for residual radioactivity. Residual radioactivity could be evenly distributed over a large area, appear as small areas of elevated activity, or a combination of both. In some cases, there may be no residual radioactivity in an area or survey unit. Therefore, the adequacy and effectiveness of the FSS process depends upon properly classified survey units to ensure that areas with the highest potential for contamination receive a higher degree of survey effort.

The survey areas established by the HSA were further divided into survey units. The suggested surface area limits provided in MARSSIM were used to establish the initial set of survey units for the LTP. A survey unit is a portion of a structure or open land area that is surveyed and evaluated as a single entity following FSS. Survey units were delineated to physical areas with similar operational history or similar potential for residual radioactivity.

To the extent practical, survey units were established with relatively compact shapes, and highly irregular shapes were avoided unless the unusual shape was appropriate for the site operational history or the site topography. For identification, survey units were assigned a five-digit number that could be further modified by a letter for future divisions if needed (i.e., if the classification changes, then the corresponding survey unit size limitation changes). Physically, land survey unit boundaries were determined using commercially available mapping software with coordinates consistent with the Wisconsin State Plane North American Datum (NAD) 1983 coordinate system. Table 2-3 provides an outline for classification versus area size for survey units consistent with MARSSIM, Table 1.

Table 2-3 Typical Final Status Survey Unit Areas Classification Area Type Suggested Area Class 1 Land Up to 2,000 m2 Structure Up to 100 m2 Class 2 Land 2,000 to 10,000 m2 Structure 100 to 1,000 m2 Class 3 Land No Limit Structure No Limit Prior to FSS, each survey units classification is reviewed and verified in accordance with the LTP and its implementing procedures. A classification change to increase the class (e.g., Class 2 to Class 1) may be implemented without notification to regulatory authorities.

However, a classification change to decrease the class (e.g., Class 1 to Class 2) may be implemented only after accurate assessment and approval from regulatory authorities as detailed in the LTP and its implementing procedures. Typically, reclassification occurs

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 19 after the evaluation of continuing characterization results or emergent data indicates a more restrictive classification is required. Final classification was performed in conjunction with the preparation of the FSS Sample Plans.

The Sample Plans reconcile legacy characterization data with more recent continuing characterization data to verify that the final classification is correct.

2.2 Survey Design Final Status Surveys for the LACBWR site are designed following LaCrosseSolutions procedures, the LTP, and MARSSIM guidance. FSS design utilizes the combination of traditional scanning surveys, systematic or random sampling protocols, and investigative or judgmental methodologies to evaluate survey units relative to the applicable release criteria for open land or structure sample plans.

LTP Chapter 2 provides detailed characterization data that describes current contamination levels in the basements and soils from the characterization campaign conducted from September 2014 through August 2015. The initial survey data for basements was based on core samples obtained from the walls and floors of the Reactor Building, Waste Treatment Building (WTB) and the balance of the basement structures (primarily the Piping Tunnels) at biased locations with elevated contact dose rates, contamination levels, and/or evidence of leaks/spills. During subsequent characterizations, additional cores were obtained from the Reactor Building and the WGTV. Surface and subsurface soil samples were taken in each impacted open land survey unit (including soil beneath and adjacent to basements) and analyzed for the presence of plant-derived radionuclides. EnergySolutions TSD RS-TD-313196-001, Radionuclides of Concern During LACBWR Decommissioning (Reference 14) evaluates the results of the concrete core analysis data from the Reactor Building, WTB, Piping Tunnels and WGTV and refines the initial suite of potential ROC by evaluating the dose significance of each radionuclide.

Insignificant dose contributors were determined consistent with the guidance contained in section 3.3 of NUREG-1757. In all soil and concrete scenarios, Cs-137, Co-60, Sr-90, Eu-152 and Eu-154 contribute nearly 100% of the total dose. The remaining radionuclides were designated as insignificant dose contributors and are eliminated from further detailed evaluation. Therefore, the final ROCs for LACBWR soil, basement concrete, and buried piping are Cs-137, Co-60, Sr-90, Eu-152 and Eu-154.

The results of surface and subsurface soil characterization in the impacted area surrounding LACBWR indicate that there is minimal residual radioactivity in soil. Based on the characterization survey results to date, LACBWR does not anticipate the presence of significant soil contamination in any remaining subsurface soil that has not yet been characterized. In addition, minimal contamination is expected in the buried piping that LACBWR plans to leave in place.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 20 The final suite of potential radionuclides and the mixture to be applied to soil and basements is provided in Table 2-4.

Table 2-4 Dose Significant Radionuclides and Mixture (1)

Radionuclide Soil % of Total Activity (normalized)

Reactor Building

% of Total Activity (normalized)

WGTV % of Total Activity (normalized)

Co-60 6.44E-02 7.41E-02 1.01E-02 Sr-90 9.81E-02 1.23E-01 1.94E-02 Cs-137 8.29E-01 7.96E-01 9.57E-01 Eu-152 5.49E-03 2.97E-03 9.56E-03 Eu-154 2.81E-03 4.04E-03 3.42E-03 (1)

From Table 5-2 of the LTP Characterization results determined that Co-60 and/or Cs-137 would be the primary ROC for the majority of survey design. All the FSS results provided in this report utilize Cs-137 as the primary ROC. Cs-137 and Co-60 characterization data for the survey units discussed in this report were used to determine the expected variability, number of samples required, and investigation levels for FSS design.

The dose contribution from each ROC is accounted for using the Sum-of-Fractions (SOF) to ensure that the total dose from all ROC does not exceed the dose criterion. The SOF or unity rule is applied to the data used for the survey planning, and data evaluation and statistical tests for soil sample analyses since multiple radionuclide-specific measurements will be performed or the concentrations inferred based on known relationships. The application of the unity rule serves to normalize the data to allow for an accurate comparison of the various data measurements to the release criteria. When the unity rule is applied, the DCGLW (used for the nonparametric statistical test) becomes one (1). The Base Case DCGLs (DCGLS and DCGLB, for soils and structure basements, respectively) are directly analogous to the DCGLW as defined in MARSSIM. The use and application of the unity rule will be performed in accordance with section 4.3.3 of MARSSIM.

Survey design objectives included a verification of the survey instruments ability to detect the radiation(s) of interest relative to the DCGL. As standard practice to ensure that this objective was consistently met, portable radiation detection instruments used in FSS were calibrated on a yearly frequency with a National Institute of Standards and Technology (NIST) traceable source in accordance with EnergySolutions procedures. Instruments were response checked before and after use. Minimum Detectable Count Rates (MDCR) were established and verified prior to FSS. Control and accountability of survey instruments were maintained and documented to assure quality and prevent the loss of data.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 21 Based upon classification, a percentage of the surface area in the survey unit was selected and scanned with portable gamma radiation detection instruments or the ISOCS.

Information obtained during the survey was automatically logged by the instrument for review and analysis. Sample and scan coordinates were identified using a random or systematic sample tool in Visual Sample Plan (VSP). For land areas, sample location coordinates were programmed into a Global Positioning System (GPS), then physically located and marked. For basements, structure dimensions and coordinates based on a local origin selected by the survey designer were used to identify measurement locations in the field. Investigational samples or measurements were collected at areas of elevated scan readings. All details and instructions were incorporated into the survey units FSS Sample Plan. The recommended survey coverage based on survey unit classification is provided below in Table 2-5.

Table 2-5 Recommended Survey Coverage for FSS (1)

Classification Surface Scans Soil Samples/Static Measurements Class 1 100%

Number of sample/measurement locations for statistical test, additional samples/measurements to investigate areas of elevated activity Class 2 10% to 100%,

Systematic and Judgmental Number of sample/measurement locations for statistical test Class 3 Judgmental Number of sample/measurement locations for statistical test (1) From Table 5-15 of the LTP A sufficient number of ISOCS measurements was obtained in the two (2) Class 1 sub-grade basement survey units to meet the 100% scan requirement. Additionally, the ISOCS Field-of-View (FOV) was overlapped to ensure that there were no un-surveyed corners and gaps.

The adjusted minimum number of ISOCS measurements in each of the basement FSS units to account for the overlap is provided in Table 2-6.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 22 Table 2-6 Adjusted Minimum Number of ISOCS Measurements per FSS Unit (1)

FSS Unit Classification Required Areal Coverage Adjusted # of ISOCS Measurements Adjusted Areal Coverage Adjusted Areal Coverage (m2)

(based on FOV ~16 m2)

(m2)

(% of Area)

Reactor Building basement Class 1 512 43 512 100%

WGTV basement Class 1 311 22 311 100%

(1) From Table 5-14 of the LTP Surface soil samples were collected to a depth of 0.15 meters below the top soil surface.

Leaves, rocks, roots, and other objects were excluded as much as possible from the sample.

If a surface soil sample measurement exceeded 75% of the Operational DCGL, then a subsurface soil sample was required from the same location. This scenario was not encountered during the FSS of any of the four (4) sub-grade excavation and three (3) open land survey units addressed in this report.

Designated soil samples were sent to an off-site laboratory for Hard-to-Detect (HTD) analysis. Laboratory DQO and analysis results were summarized in release records and reported as actual calculated results. Sample report summaries within the release records included unique sample identification, analytical method, radioisotope, result, uncertainty of two standard deviations, laboratory data qualifiers, units, and required Minimum Detectable Concentration (MDC).

Another consideration of survey design was the use of surrogates. In lieu of analyzing every sample for HTD radionuclides, the development and application of Surrogate Ratio DCGLs as described in MARSSIM, Section 4.3.2 was applied to estimate HTD radionuclides. Surrogate ratios allow for expedient decision making in characterization, remediation planning, or FSS design.

A surrogate is a mathematical ratio where an Easy-to-Detect (ETD) radionuclide (e.g., Cs-137) concentration is related to a HTD radionuclide (e.g., Sr-90) concentration. From the analytical data, a ratio is developed and applied in the survey scheme for samples taken in the area. Details and applications of this method are provided in section 5.2.9 of the LTP.

Some portion of the radioactivity found in the soil samples is certainly attributed to fallout or background. Due to the lack of significant activity revealed during background studies, assessments and characterization, it was determined that background subtraction would not be applied during the FSS of open land survey units.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 23 2.3 Survey Implementation Final Status Survey implementation of the Phase 1 survey units started in September 2017 and completed in April 2019. Implementation was the physical process of the FSS Sample Plan execution for a given survey unit. Each Sample Plan was assigned to a FSS Supervisor for implementation and completion, in accordance with LaCrosseSolutions procedures and the FSS QAPP. The tasks included in the implementation were:

x Verification and validation of personnel training as required by Training Department and Radiation Protection procedures.

x Monitoring instrument calibration as detailed in LC-RP-PG-003, Radiological Instrumentation Program (Reference 15) and LC-RP-PR-060, Calibration and Initial Set Up of the 2350-1 (Reference 16).

x Implementation of applicable operating and health and safety procedures.

x Implementation of isolation and control of the survey unit in accordance with LC-FS-PR-010, Isolation and Control for Final Status Survey.

x Determination of the number of samples and measurements required to meet DQOs as described in LC-FS-PR-002, Final Status Survey Package Development.

x Determination of sample and measurement locations and creation of survey unit maps displaying the locations in accordance with LC-FS-PR-002.

x Proper techniques for collecting and handling FSS samples in accordance with LC-FS-PR-004, Sample Media Collection for Site Characterization and Final Status Survey (Reference 17).

x Maintaining Quality Assurance/Quality Control requirements (i.e., replicate measurements or samples) in accordance with LC-QA-PN-001, Final Status Survey Quality Assurance Project Plan.

x Sample Chain-of-Custody maintained in accordance with LC-FS-PR-005, Sample Media Preparation for Site Characterization and Final Status Survey (Reference 18).

x Sample submission to approved laboratories in accordance with LC-FS-PR-012, Chain of Custody Protocol (Reference 19).

x Application of the DCGLs to sample and measurement results in accordance with the Data Quality Assessment (DQA) process as detailed in LC-FS-PR-008, Final Status Survey Data Assessment.

x Determination of investigation methodology and corrective actions, if applicable.

The FSS implementation and completion process resulted in the generation of field data and analysis data consisting of measurements taken with handheld radiation detecting equipment, observations noted in field logs, and radionuclide specific analyses. Data were stored electronically on the EnergySolutions common network with controlled accessibility.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 24 2.4 Survey Data Assessment Prior to proceeding with data evaluation and assessment, the assigned FSS Supervisor ensures consistency between the data quality and the data collection process and the applicable requirements.

The DQA process is an evaluation method used during the assessment phase of FSS to ensure the validity of FSS results and demonstrate achievement with the FSS Sample Plan objectives. A key step in the data assessment process converts all of the survey results to DCGL units, if necessary. The individual measurements and sample concentrations are compared to the Operational DCGL for evidence of small areas of elevated activity or results that are statistical outliers. When practical, graphical analyses of survey data that depicts the spatial correlation of the measurements was used.

For the survey units addressed by this final report, the survey data was evaluated using the Sign Test (as described in the LTP). The Sign Test is a one-sample statistical test that compares data directly to the release criteria. Combined with an effective sampling scheme, passing the Sign Test satisfies the release criteria. Selection of the Sign Test is prudent and conservative in the assumption that the radionuclides being considered are not present in background or are at levels at a small fraction of the applicable release criteria.

Furthermore, any background contribution (e.g., Cs-137 from global fallout) in the sample increases the likelihood of failing the survey unit, which is conservative. If the release criteria were exceeded or if results indicated the need for additional data points, appropriate further actions were implemented, usually through the issue of an addendum to the FSS Sample Plan.

2.5 Quality Assurance and Quality Control Measures Quality assurance and control measures were employed throughout the FSS process to ensure that all decisions were based on data of acceptable quality. Quality assurance and control measures were applied to ensure:

x The plan was correctly implemented.

x The DQA process was used to assess results.

x DQOs were properly defined and derived.

x All data and samples were collected by individuals with the proper training and in adherence to approved procedures and sample plans.

x All instruments were properly calibrated.

x All collected data was validated, recorded, and stored in accordance with approved procedures.

x All required documents were properly maintained.

x Corrective actions were prescribed, implemented and tracked, as necessary.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 25 Independent laboratories used for analysis of the samples collected during FSS maintain Quality Assurance Plans designed for their facility. LaCrosseSolutions reviews these plans, as required by the FSS QAPP, prior to selection of a laboratory for FSS sample analysis to ensure standards are acceptable.

The Characterization/License Termination Group has undergone surveillance by the EnergySolutions QA department on a consistent basis throughout the project at LACBWR.

The QA surveillances have scrutinized the LTP, C/LT procedures, sample plans, release records, and other C/LT records. The responses to the QA surveillances are captured in the Corrective Action Program (CAP).

3 Site Information 3.1 Site Description The La Crosse Boiling Water Reactor was a 50-Megawatt Electric (MWe) BWR that is owned by Dairyland Power Cooperative (Dairyland). This unit, also known as Genoa 2, is located on the Dairyland Genoa site on the east shore of the Mississippi River about 1 mile south of the Village of Genoa, Vernon County, Wisconsin, and approximately 19 miles south of the city of La Crosse, WI. See Figure 3-1 for a map showing the site location.

The licensed site comprises a total of 163.5 acres which is owned by Dairyland, with LACBWR comprising only 1.5 acres. The site is accessed by a road on the south side of the plant, off of Highway 35. The prominent features on the site are shown in Figures 3-2 and 3-3.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 26 Figure 3-1 Site Regional Location

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 27 Figure 3-2 Site Overview

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 28 Figure 3-3 LACBWR Buildings

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 29 The site is licensed under Possession Only License No. DPR-45 with Docket Numbers of 50-409 for LACBWR and 72-046 for the Independent Spent Fuel Storage Installation (ISFSI). Key milestones during the life of the plant are:

x Allis-Chalmers, under contract with the AEC, designed, fabricated, constructed and performed startup of the LACBWR from 1962 to 1967, x

Dairyland entered into a contact to purchase steam from the nuclear plant to operate a turbine-generator for production of electricity: June 1962, x

Allis-Chalmers dockets application for construction: November 5, 1962, x

Initial Criticality achieved: July 11, 1967, x

Low power testing completed: September 1967, x

Provisional Operating authorization issued (DPRA-6): October 31, 1969, x

Provisional Operating License, DPRA-45 issued: August 28, 1973, x

LACBWR permanently shut down: April 30, 1987, x

Final reactor defueling was completed on June 11, 1987, and x

Completion of fuel loading into the ISFSI was completed on September 19, 2012.

The reactor was critical for a total of 103,287.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The 50 MWe generator was on line for 96,274.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. The total gross electrical energy generated was 4.047 Gigawatt-Hours (GWH). The unit availability factor was 62.9%.

The LACBWR unit consists of major buildings and structures such as the Reactor Building, Turbine Building, 1B Diesel Generator Building, Waste Treatment Building, Underground Gas Storage Tank Vault, Ventilation Stack, Low Specific Activity (LSA) building and others which are currently undergoing decommissioning (see Figure 3-3).

Intermittent systems dismantlement and metallic radioactive equipment has been removed since 2007, including the Reactor Pressure Vessel. The ISFSI, located south of the Genoa 3 fossil station, became operational in 2012 and holds five above-ground Dry Storage Casks with 333 spent fuel assemblies.

3.2 Survey Unit Description The following information is a description of each survey unit at the time of FSS from September 2017 until April 2019. During this period, the FSS of four (4) sub-grade excavation survey units, three (3) open land survey units, and two (2) structure basements were completed.

Survey Unit L1-SUB-DRS L1-SUB-DRS is an impacted Class 1 sub-grade excavation survey unit. The survey unit consists of the underlying soil post-removal of the Radiologically Controlled Area (RCA) roadway, rail lines, storm drains, High Pressure Service Water (HPSW) lines, Low

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 30 Pressure Service Water (LPSW) lines, and well water lines located in open land survey unit L1-010-104 West. The surface area of the survey unit is 1,125 m2.

Survey Unit L1-SUB-TDS L1-SUB-TDS is an impacted Class 1 sub-grade excavation survey unit. The survey unit consists of the underlying soil post-removal of the Turbine Building/Turbine Building Offices and 1B Diesel Generator Building, as well as associated system lines. The surface area of the survey unit is 1,185.5 m2.

This survey unit includes only the western portion of the original L1-SUB-TDS. The eastern portion of the original survey unit did not undergo FSS because of high background readings emanating from the Reactor Building and surrounding environs. The eastern portion of the survey unit (now identified as L1-SUB-TDS A) will undergo FSS under a separate sample plan and will be reported in a separate final report. Additionally, the Reactor Plant Generator Plant Area (RPGPA) Sump Area, also within the east portion of the original survey unit, experienced groundwater intrusion due to rising Mississippi River levels and has caused it to become inaccessible. The backfilled RPGPA Sump Area (L1-SUB-TDS B) will undergo FSS via use of GeoProbe technology using a separate sample plan and will also be reported in a separate final report.

Survey Unit L1-SUB-LES L1-SUB-LES is an impacted Class 1 sub-grade excavation survey unit within open land survey unit L1-010-103. The survey unit consists of the underlying soil post-removal of the Low Specific Activity (LSA) Building, Septic Tank, Maintenance Eat Shack Foundation, Oily Water Tank, Circulating Water Intake & Discharge piping, and Main Transformer Substation as well as associated system lines. The surface area of the survey unit is 1,336 m2.

Survey Unit L1-010-101C L1-010-101C is an impacted Class 1 sub-grade excavation survey unit within open land survey unit L1-010-101. The survey unit consists of the underlying soil post-removal of the WTB. The surface area of the survey unit is 87.8 m2.

Survey Unit L2-011-102 L2-011-102 is an impacted Class 2 open land survey unit located west of the RCA. The surface area of the survey unit is 2,258 m2.

Survey Unit L2-011-103 L2-011-103 is an impacted Class 2 open land survey unit located northwest of the LSE.

The surface area of the survey unit is 2,445 m2.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 31 Survey Unit L3-012-102 L3-012-102 is an impacted Class 3 open land survey unit. This survey unit includes the current LACBWR Site Switchyard and Transmission Sub-Station Switch House grounds, which are located on the north end of the LACBWR Site. The surface area of the survey unit is 11,711 m2.

Survey Unit B1-010-004 B1-010-004, the WGTV basement, is an impacted Class 1 structure basement survey unit.

At end-state, this survey unit consisted of the concrete floor, concrete walls up to 3 feet below grade (636 foot elevation), a concrete support column in the center of the floor that extends up to 3 feet below grade, and a 3-foot by 3-foot by 3-foot deep sump in the north-west corner. The total surface area of the survey unit is 311 m2.

Survey Unit B1-010-001 B1-010-001, the Reactor Building basement, is an impacted Class 1 structure basement survey unit. At end-state, this survey unit consisted of the Reactor Building floor and walls below grade (636 foot elevation). The total surface area of the survey unit is 512 m2.

3.3 Summary of Historical Radiological Data The site historical radiological data for this Phase 1 FSS Final Report at LACBWR incorporates the results of the HSA issued in November 2015, the initial characterization surveys completed in 2015 as well as subsequent continuing characterization surveys, and the Remedial Action Support Surveys (RASS).

Historical Site Assessment The HSA was a detailed investigation to collect existing information (from the start of LACBWR activities related to radioactive materials or other contaminants) for the site and its surroundings. The HSA focused on historical events and routine operational processes that resulted in contamination of plant systems, onsite buildings, surface and subsurface soils within the RCA. It also addressed support structures, open land areas and subsurface soils outside of the RCA but within the owner-controlled area. The information compiled by the HSA was used to establish initial area survey units and their MARSSIM classifications. This information was used as input into the development of site-specific DCGLs, remediation plans and the design of the FSS. The scope of the HSA included potential contamination from radioactive materials, hazardous materials, and other regulated materials.

The objectives of the HSA were to:

x Identify potential, likely, or known sources of radioactive and chemical contaminants based on existing or derived information.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 32 x Distinguish portions of the site that may need further action from those that pose little or no threat to human health.

x Provide an assessment of the likelihood of contaminant migration.

x Provide information useful to subsequent continuing characterization surveys.

x Provide an initial classification of areas and structures as non-impacted or impacted.

x Provide a graded initial classification for impacted soils and structures in accordance with MARSSIM guidance.

x Delineate initial survey unit boundaries and areas based upon the initial classification.

The survey units established by the HSA were used as initial survey units for characterization. Prior to characterization, survey unit sizes were adjusted in accordance with the guidance provided in MARSSIM section 4.6 for the suggested physical area sizes for survey units for FSS.

For the land area survey units of interest in this report, the HSA indicated that the presence of residual radioactivity in concentrations in excess of the unrestricted release criteria was not expected; however, two (2) of the open land survey units were designated as Class 2 due to decommissioning activities, and all four (4) sub-grade excavation survey units were designated Class 1 due the proximity in the LSE and nature of materials removed.

The HSA identified the WGTV and Reactor Building basements as Class 1 structures.

3.3.1 Characterization Surveys Site characterization of the LACBWR facility was performed in accordance with EnergySolutions procedure LC-FS-PN-002, Characterization Survey Plan (Reference 20),

which was developed to provide guidance and direction to the personnel responsible for implementing and executing characterization survey activities. The Characterization Survey Plan worked in conjunction with implementing procedures and survey unit specific survey instructions (sample plans) that were developed to safely and effectively acquire the requisite characterization data.

Characterization data acquired through the execution of the Characterization Survey Plan was used to meet three primary objectives:

x Provide radiological inputs necessary for the design of FSS.

x Develop the required inputs for the LTP.

x Support the evaluation of remediation alternatives and technologies and estimate waste volumes.

The final output of the initial site characterization was EnergySolutions GG-EO-313196-RS-RP-001, LACBWR Radiological Characterization Survey Report for October and November 2014 Field Work - November 2015 (Reference 21) and LaCrosseSolutions TSD LC-RS-PN-164017-001, Characterization Survey Report. These reports concluded that based on the results of the characterization surveys, it was expected that minimal residual

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 33 plant-derived radioactivity was present in the seven (7) land area survey units addressed by this report.

Additional discussion, the statistical quantities for Cs-137 and Co-60 from the Characterization Surveys, and a summary of off-site analysis for characterization samples are provided in each of the subject release records (See Appendices A1 through A9; Appendix A7 [FSS Release Record, Survey Unit L3-012-102] does not detail characterization data as no characterization was performed).

3.3.2 Continuing Characterization Previously inaccessible areas identified in LTP Section 2.4 were characterized during the continuing characterization process. In order to verify that the insignificant contributor (IC) dose does not change prior to implementing the FSS and to verify the HTD to surrogate radionuclide ratios used for the surrogate calculation are still valid, LACBWR committed to obtaining soil samples during continuing characterization and FSS. In addition to verifying IC dose and surrogate radionuclide ratios, in the case of the Reactor Building and WGTV, continuing characterization was performed to ensure that any individual ISOCS measurements would not exceed the Operational DCGLB during FSS.

Section 5.1 of the LTP states that the actual IC dose will be calculated for each individual sample result using the DCGLs from TSD RS-TD-313196-004, LACBWR Soil DCGL, Basement Concrete DCGL, and Buried Pipe DCGL, Table 35 (Reference 22), for soil and structures. If the IC dose calculated is less than the IC dose assigned for DCGL adjustment, then no further action is required. If the actual IC dose calculated from the sample result is greater than the IC dose assigned for DCGL adjustment, then a minimum of five (5) additional investigation samples will be taken around the original sample location. Each investigation sample will be analyzed by the on-site gamma spectroscopy system and sent for HTD analysis (full suite of radionuclides from LTP Table 5-1). As with the original sample, the actual IC dose will be calculated for each investigation sample. In this case, the actual calculated maximum IC dose from an individual sample observed in the survey unit will be used to readjust the DCGLs in that survey unit. If the maximum IC dose exceeds 10%, then the additional radionuclides that were the cause of the IC dose exceeding 10%

will be added as additional ROC for that survey unit. The survey unit-specific DCGLs used for compliance, the ROC for that survey unit, and the survey data serving as the basis for the IC dose adjustment are required to be documented in the release record for the survey unit.

Out of the survey units presented in this report, survey units L1-SUB-TDS, L1-010-101C, B1-010-004, and B1-010-001 underwent continuing characterization.

L1-SUB-TDS In February of 2018, the Turbine Building foundation was removed in its entirety, including all suspect broken drain lines and adjacent soil. As required by Section 5.3.3.4 of

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 34 the LTP, in the western portion of the excavation, a total of eight (8) samples were collected from the region beneath the broken drain lines, turbine sump, turbine pit, and condenser pit. Although part of continuing characterization, these samples were collected as judgmental samples during the FSS. Per Section 5.1 of the LTP, all samples were analyzed by the on-site gamma spectroscopy system. Gamma spectroscopy results revealed eight (8) samples above MDC for Cs-137, with a maximum concentration of 0.188 pCi/g.

Two (2) samples were above MDC for Co-60, with a maximum concentration of 0.257 pCi/g. No samples were identified with concentrations greater than MDC for any other ROC.

Although four (4) of the samples were required to be sent off-site for Sr-90 analysis, as a conservative measure, all were sent off-site for Sr-90 analysis. Additionally, seven (7) of the eight (8) samples were sent off-site for analyses for the full initial suite of ROC. An assessment of the results confirmed the calculated IC dose is unchanged prior to FSS and there is no change to the surrogate ratio.

L1-010-101C After total removal of the WTB, continuing characterization samples were collected during the FSS of the resultant excavation. In addition to the systematic samples collected during FSS, two (2) additional samples were collected for continuing characterization. These samples were sent off-site for HTD analyses for the full initial suite of ROC. An assessment of the results confirmed the calculated IC dose is unchanged prior to FSS and there is no change to the surrogate ratio.

B1-010-001 Continuing characterization of the underlying concrete in the Reactor Building basement was performed once demolition was complete to 3 feet below grade. The characterization survey consisted of a Radiological Assessment (RA) (considered a form of continuing characterization) of the underlying concrete to ensure that any individual ISOCS measurements would not exceed the Operational DCGLB during FSS. The RA consisted of a beta-gamma scan over 100% of all accessible surfaces of the concrete and a minimum of 30 loose surface contamination samples. Six (6) concrete core samples were obtained at evenly distributed locations and an additional four (4) cores were obtained at areas of elevated activity identified during the scan survey or biased locations such as low points, cracks, or areas of discoloration. The 3-inch diameter core samples were obtained by coring into the concrete to a depth of 6 inches. After slicing into 1/2-inch pucks, all core samples were analyzed by the on-site gamma spectroscopy system. Because an RA is a form of continuing characterization, 10% of all media samples collected in this survey unit, with a minimum of one sample, were required to undergo analysis for the full initial suite of radionuclides. In addition, a minimum of one (1) sample beyond the 10% minimum was selected at random, also for full suite radionuclide analysis.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 35 B1-010-004 Continuing characterization of the WGTV was performed in September of 2017. The scope of the survey was the interior concrete surfaces as well as the soil adjacent to and beneath the structure.

The continuing characterization of the structure interior concrete surfaces consisted of a beta scan of 100% of all accessible surfaces augmented with a minimum of 30 loose surface contamination samples. Five (5) concrete core samples were obtained on the floor and wall surfaces and an additional three (3) concrete cores were biased towards areas of elevated activity identified during the scan survey (including two cores in the sump). The cores were 3-inch diameter to a depth of 6 inches, and all cores were sliced into 1/2-inch pucks to ascertain the depth of contamination. All core samples underwent gamma spectroscopy using the on-site laboratory, and all were sent to the off-site laboratory for HTD analysis of the full suite of radionuclides. An assessment of the results confirmed the calculated IC dose is unchanged (dose fraction less than 10%) prior to FSS.

The soil adjacent to and beneath the WGTV was also characterized as part of the sample plan. A gamma scan was performed over 100% of the safely accessible topside soil adjacent to the structure and four (4) surface soil samples were obtained. Additionally, four (4) soil samples were obtained beneath the WGTV floor at the locations of highest activity identified during the scan survey, at low points (e.g., sump) or areas that could act as conduits for contamination migration such as cracks. This was accomplished by coring through the concrete until soil was encountered. Like the concrete core samples, all eight of the soil samples underwent gamma spectroscopy using the on-site laboratory and all were sent to the off-site laboratory for HTD analysis of the full suite of radionuclides. An assessment of the results confirmed the calculated IC dose is unchanged prior to FSS and there is no change to the surrogate ratio.

3.3.3 Remedial Action Support Surveys RASSs were performed in survey units L1-SUB-DRS, L1-SUB-LES, L1-SUB-TDS, and L1-010-101 C after the excavations were complete and before FSS. The purpose of the RASSs were to ensure that residual radionuclide concentrations in the excavations were below the Operational DCGL for soil.

L1-SUB-DRS 100% of the surface area of soil in the excavation was scanned using a Ludlum Model 2350-1 data logger paired with a Ludlum Model 44-10 detector. The Alarm Set Point (ASP) was set at average background plus 3,525 cpm, corresponding to the Operational DCGL. Five (5) alarms were produced and verified during the RASS scan survey. Seven (7) judgmental soil samples were collected in the excavation at areas where scan alarms occurred. The seven (7) samples were analyzed using the on-site gamma spectroscopy system. Gamma spectroscopy revealed Cs-137 concentrations ranging between 5.30E-02

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 36 pCi/g and 8.13E+00 pCi/g. Co-60 concentrations ranged between 5.60E-02 pCi/g and 2.90E-01 pCi/g. Areas that alarmed, and where judgmental samples were collected, were bounded and remediated to levels below the Operational DCGL. Remediation was deemed sufficient as no further alarms were produced upon rescan of the elevated areas.

L1-SUB-LES 100% of the surface area of soil in the excavation was scanned using a Ludlum Model 2350-1 data logger paired with a Ludlum Model 44-10 detector. The ASP was set at average background plus 3,525 cpm, which is equivalent to the Operational DCGL. Sixteen (16) alarms occurred and verified during the RASS scan survey. Fifteen (15) judgmental soil samples were collected in the excavation at areas where scan alarms occurred. The fifteen (15) samples were analyzed using the on-site gamma spectroscopy system. Gamma spectroscopy revealed Cs-137 concentrations ranging between 5.30E-02 pCi/g and 2.93E+01 pCi/g. Co-60 concentrations ranged between 1.23E-01 pCi/g and 8.57E+01 pCi/g. Areas where alarms occurred and where judgmental samples were collected were bounded and remediated to levels below the Operational DCGL. Remediation was deemed sufficient as no further alarms were produced upon rescan of the elevated areas.

L1-SUB-TDS 100% of the surface area of soil in the excavation was scanned using a Ludlum Model 2350-1 data logger paired with a Ludlum Model 44-10 detector. The ASP was set using a gross background of 31,800 cpm and an MDCR of 1,906 cpm for a total of 33,706 cpm.

This value equates to a scan MDC for Cs-137 of 13 pCi/g, which is equivalent to 25% of the IC-adjusted DCGL for soil that was in use before the Operational DCGLs were established. One (1) alarm occurred and was verified during the RASS scan survey. In accordance with the RASS sample plan, five (5) judgmental soil samples were collected in the excavation at areas where scan alarms occurred or where significant increases in count rate were noted. The five (5) samples were analyzed using the on-site gamma spectroscopy system. Gamma spectroscopy revealed Cs-137 concentrations ranging between 3.94E-02 pCi/g and 6.74E-01 pCi/g. Co-60 concentrations ranged between 4.30E-01 pCi/g and 5.52E-01 pCi/g. Areas where alarms occurred, and where judgmental samples were collected, were bounded and remediated to levels below the Operational DCGL.

Remediation was deemed sufficient, as no further alarms occurred upon rescan of the elevated areas.

L1-010-101C During implementation of the FSS, elevated scan readings were identified near concrete foundation caps and residual concrete chunks that remained post-demolition of the WTB foundation. These elevated scan readings were verified by a Region III NRC Inspector who was performing confirmatory scan surveys. Count rates of up to 70,000 counts per minute (cpm) were identified in these areas. FSS Supervision determined that the elevated areas

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 37 needed to be remediated and the remaining concrete in the survey unit taken out. The work crew was instructed to continue taking out soil and concrete until count rates were consistently below 10,000 cpm. Gamma scans of the material in excavator buckets during removal of the material indicated count rates of up to 1,400,000 cpm. Approximately seventy (70) cubic yards of soil and concrete debris were removed from the excavation.

After completion of the remediation, three (3) soil samples were collected at the edges of the remediated area and analyzed using the on-site gamma spectroscopy system. Cs-137 was detected at concentrations greater than MDC in three (3) of the samples, with a maximum concentration of 2.30E-01 pCi/g. Co-60 was detected at concentrations above MDC in two (2) of the samples, with a maximum concentration of 2.71E-02 pCi/g. The results of these samples are far below the Operational DCGLs.

3.4 Conditions at the Time of Final Status Survey The open land survey units in this report had little to no disturbance occurring since the shut-down of LACBWR. In open land and excavation survey units, depending on the time of survey, the soil was damp from melted snow, and several small patches of un-melted snow were still present in the areas. Additionally, the open land survey units were sparsely covered in dry grass and bushes. The snow and vegetation were deemed sparse enough as to not be a considerable constraint for the collection of samples and scan measurements. At the time of FSS, the basement survey units were free of debris and free of water to the extent possible.

Prior to FSS, areas ready for survey were isolated and controlled under LC-FS-PR-010, Isolation and Control for Final Status Survey. This included posting of the area as well as notifications to site personnel. Permission must be obtained from C/LT staff to enter and work in these areas. Posting of the boundaries controls personnel access, a routine surveillance program monitors for any inadvertent personnel access with procedurally defined recovery and reporting protocols in the event of any impact to these survey units final status.

3.5 Identification of Potential Contaminants EnergySolutions TSD RS-TD-313196-001, Radionuclides of Concern During LACBWR Decommissioning established the basis for an initial suite of potential ROC for decommissioning. Industry guidance was reviewed as well as the analytical results from the sampling of various media from past plant operations. Based on the elimination of some of the theoretical neutron activation products, noble gases and radionuclides with a half-life of less than two years, an initial suite of potential ROC for the decommissioning of LACBWR was prepared (see Table 5-1 in the LTP).

Cs-137 deposition resulting from global fallout is thought to be the source of most of the Cs-137 encountered in samples collected in the open lands surrounding LACBWR.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 38 Geological deposition, regional concentrations and transport mechanisms are well documented and the subject of numerous publications and studies. However, as a conservative measure, Cs-137 resulting from fallout or background was not subtracted from analytical results for FSS at LACBWR.

In accordance with the LTP, a minimum of 10% of the systematic or random sample population, or any sample with a SOF of 0.1 or greater, was sent to an off-site laboratory for HTD ROC analysis. This process is done to verify the surrogate ratios between HTD ROC (Sr-90) and the ETD ROC (Cs-137) as established in the LTP. For soil samples with positive results for both an HTD ROC and the corresponding surrogate radionuclide, the HTD to surrogate ratio was derived. The maximum surrogate ratio was used during data assessment. Analyses for HTD radionuclides did not positively identify Sr-90 in any samples sent off-site; therefore, the HTD to surrogate ratio as established in the LTP was unchanged for all survey units within this report.

3.6 Radiological Release Criteria All FSSs for the survey units detailed in this report were conservatively designed to the Operational DCGLs for soil and structure basements, and all results were compared to these values. However, since the release criteria were based on the Base Case DCGL, surpassing the Operational DCGL did not disqualify a survey unit from meeting the release criteria provided that the data passed the Sign Test and the Base Case DCGL was not exceeded.

4 Final Status Survey Protocol 4.1 Data Quality Objectives The DQO process as outlined in Section 2 of this report was applied for each FSS Sample Plan and contains basic elements common to all FSS Sample Plans at LACBWR. An outline of those elements presented in the Sample Plans are as follows:

State the Problem The problem: To demonstrate that the level of residual radioactivity in a survey unit does not exceed the release criteria of 25 mrem/yr TEDE and that the potential dose from residual radioactivity is ALARA.

Stakeholders: The primary stakeholders interested in the answer to this problem are LaCrosseSolutions LLC, Dairyland Power Generation LLC, the Wisconsin Department of Health Services and the United States Nuclear Regulatory Commission (USNRC).

The Planning Team: The planning team consisted of the assigned Radiological Engineer or FSS Supervisor with input from other C/LT personnel as well as the Safety Department.

The primary decision maker was the FSS Supervisor with input from the C/LT Manager.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 39 Schedule: The approximate time projected to mobilize, implement, and asses an FSS unit was between four (4) and ten (10) days.

Resources: The following resources were necessary to implement an FSS Sample Plan:

x Radiological Engineer or FSS Supervisor to prepare the plan and evaluate data.

x FSS Supervisor to monitor and coordinate field activities.

x Graphics/GPS Specialist to prepare survey maps, layout diagrams, composite view drawings, and other graphics as necessary to support design and reporting.

x FSS Technicians to perform survey activities, collect survey measurement data, and collect media samples.

x ISOCS Subject Matter Expert (SME) to verify detector geometries and counting sequences.

x Chemistry/Analysis laboratory Staff to analyze samples as necessary.

x Heavy equipment operator and laborers to assist in the transport and positioning of the ISOCS.

Identify the Decision Principal Study Question: Are the residual radionuclide concentrations found in the soil and structure basements equal to or below site-specific DCGLs?

Alternate Actions: Alternative actions include failure of the survey unit, remediation, reclassification, and resurvey.

The Decision: If the survey unit fails to demonstrate compliance with the release criteria, then the survey unit is not suitable for unrestricted release. The DQA process is reviewed to identify the appropriate additional action or combination of actions.

Identify Inputs to the Decision Information Needed: The survey unit requiring evaluation of residual activity and its surface area. The HSA, characterization surveys, continuing characterization surveys, and RASSs were preliminary sources of information for FSS; however, the results were not sufficient to provide the current as-left radiological conditions. New measurements of sample media are needed to determine the concentration and variability for those radionuclides potentially present at the site at the time of FSS. For the land area survey units, gamma scans are needed to identify areas of elevated activity and soil samples are needed as compliance measurements. For structure survey units, ISOCS measurements are required to identify areas of elevated activity and to show compliance with the release criteria.

Historical Information: The classification as originally identified in the HSA and the verification of that classification during characterization. A summary of site processes or incidents that occurred in the survey unit.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 40 Radiological Survey Data: The current radiological survey data from characterization, RASS, RA, or turnover surveys. This information is used to develop a sample size for FSS.

Radionuclides of Concern: The ROC are presented in Table 2-4 of this final report.

Basis for the Action Level: The investigation and action levels for FSS were selected in accordance with Table 5-16 of the LTP, recreated below as Table 4-1.

Table 4-1 FSS Investigation Levels Classification Scan Investigation Levels Direct Investigation Levels Class 1

>Operational DCGL or >MDCscan if MDCscan is greater than Operational DCGL

>Operational DCGL Class 2

>Operational DCGL or >MDCscan if MDCscan is greater than Operational DCGL

>Operational DCGL Class 3

>Operational DCGL or >MDCscan if MDCscan is greater than Operational DCGL

>0.5 Operational DCGL During FSS, concentrations for the HTD ROC Sr-90 were inferred using a surrogate approach. As presented in the LTP, Cs-137 is the principle surrogate radionuclide for Sr-

90. The surrogate ratios for concrete core samples taken in the Reactor Building, Tunnel, and Waste Treatment Building were calculated in EnergySolutions TSD RS-TD-313196-001, Radionuclides of Concern During LACBWR Decommissioning. Table 4-2 below displays the surrogate ratios for soil, the WGTV, and the Reactor Building. These ratios are used in the surrogate calculations during FSS unless area specific ratios are determined by continuing characterization or FSS.

Table 4-2 Sr-90 to Cs-137 Surrogate Ratios (1)

Building or Area Sr-90/Cs-137 Surrogate Ratio WGTV 6.75E-02 Reactor Building 5.00E-01 Soil 5.02E-01 (1)

From Table 5-11 of the LTP.

For the FSS of the open land and excavation survey units in this report, as well as basement survey units, the surrogate Operational DCGL for Cs-137 was computed based on the ratios from Table 4-2.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 41 The equation for calculating a surrogate DCGL is as follows:

Equation 1

=

1

1

+

+

+

Where:

DCGLSur

=

Surrogate radionuclide DCGL DCGL2,3n

=

DCGL for radionuclides to be represented by the surrogate Rn

=

Ratio of concentration (or nuclide mixture fraction) of radionuclide n to surrogate radionuclide Using the Operational DCGLs presented in Table 2-1 and the ratio for soil from Table 4-2, the following surrogate calculation was performed:

Equation 2

() =

1

1 1.74+ 01()+

5.0201 1.97+ 03()

= 1.73+ 01 /

The surrogate Operational DCGL that was used for Cs-137 in open land survey units, for direct comparison of sample results to demonstrate compliance, was 1.73E+01 pCi/g.

Using the WGTV Operational DCGLs presented in Table 2-2 and the respective ratio from Table 4-2, the following surrogate calculation was performed:

Equation 3

() =

1

1 1.23+ 06()+

6.7502 4.48+ 05()

= 1.04+ 06 /

The surrogate Operational DCGL that was used for Cs-137 in the WGTV, for direct comparison of measurement results to demonstrate compliance, was 1.04E+06 pCi/m2.

Using the Reactor Building Operational DCGLs presented in Table 2-2 and the respective ratio from Table 4-2, the following surrogate calculation was performed:

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 42 Equation 4

()

=

1

1 1.52+ 06()+

5.0001 1.02+ 06()

= 8.71+ 05 /

The surrogate Operational DCGL that was used for Cs-137 in the Reactor Building, for direct comparison of measurement results to demonstrate compliance, was 8.71E+05 pCi/m2.

The action level for investigation in a Class 3 survey units was 50% of the Operational DCGL. The action level for investigation in Class 1 and Class 2 survey units was the Operational DCGL. The action levels for the Phase 1 survey units are presented in Table 4-3.

Table 4-3 Action Levels for Phase 1 Survey Units ROC Class 1 and Class 2 Action Levels - Open Land and Excavation Class 3 Action Levels - Open Land WGTV Action Levels Reactor Building Action Levels (pCi/g)

(pCi/g)

(pCi/m2)

(pCi/m2)

Co-60 3.83E+00(1) 1.92E+00(3) 2.87E+05(1) 3.61E+05(1)

Cs-137 1.73E+01(2) 8.66E+00(4) 4.21E+06(2) 8.71E+05(2)

Eu-152 8.51E+00(1) 4.26E+00(3) 6.78E+05(1) 8.33E+05(1)

Eu-154 7.89E+00(1) 3.95E+00(3) 6.28E+05(1) 7.71E+05(1)

(1)

Based on the Operational DCGL.

(2)

Based on the surrogate adjusted DCGL for Cs-137 while inferring Sr-90.

(3)

Based on 50% of the Operational DCGL.

(4)

Based on 50% of the surrogate adjusted DCGL for Cs-137 while inferring Sr-90.

Investigation Levels: Table 4-1 provides the investigation levels for the Phase 1 survey units.

Sampling and Analysis Methods to Meet the Data Requirements: Soil samples were collected down to a depth of 0.15 meters (6 inches) or, in some cases, to a depth of 12 inches (in the case of survey unit L3-012-102) and analyzed for gamma emitting radionuclides by the on-site gamma spectroscopy. The media consisted of soil and gravel as required to complete the FSS. For open land survey units, typically fourteen (14) random or systematic soil samples were required for FSS, with the exception of survey unit L3-012-102, which had twenty-eight (28).

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 43 The number of compliance measurements required for FSS in a structure basement was based on the FOV of the ISOCS, the total surface area of the survey unit, and the required survey coverage percentage based on classification. The WGTV basement required twenty-two (22) measurements and the Reactor Building basement required forty-three (43) measurements. Forty-five (45) compliance measurements in total were acquired in the Reactor Building basement.

The target MDC for measurements obtained using laboratory instruments was 10% of the applicable Operational DCGL. Measurement results with associated MDC that exceeded these values were accepted as valid data after evaluation by health physics supervision. The evaluation considered the actual MDC, the reported value for the measurement result, the reported uncertainty, and the fraction of the Operational DCGL identified in the sample.

As per the LTP Chapter 5, section 5.1, 10% of the volumetric samples (a minimum of 1) collected during FSS, or any sample with a SOF of 0.1 or greater (when compared to the Operational DCGLs) were sent off-site for HTD ROC analysis. These analyses aimed to reaffirm that the radionuclide mix was not significantly different than that assumed in the LTP. In addition, at 10% of the ISOCS measurement locations, the collection of a concrete core was required. In the cases of the WGTV and Reactor Building basements, this requirement was fulfilled during continuing characterization. In addition to verifying IC dose and surrogate radionuclide ratios, concrete cores were collected to ensure that any individual ISOCS measurements would not exceed the Operational DCGLB during FSS.

Table 5-15 of the LTP (recreated as Table 2-5 in this report) specifies the required scanning coverage for FSS units. For the Class 1 Phase 1 survey units, a scan coverage of 100% of the total area of the survey unit was selected. For Class 2 Phase 1 survey units, 25% scan coverage was selected. For the Class 3 Phase 1 survey unit, 10% scan coverage was selected. Walkover gamma scans in open land survey units were conducted with a Ludlum Model 2350-1 data logger instrument coupled to a Ludlum Model 44-10 detector at a scanning speed of 0.5 m/sec. Scanning in structure survey units was conducted with the ISOCS.

All activities fell under the FSS QAPP, which requires, among other things, the use of trained technicians, calibrated instruments, and approved procedures. In addition to these requirements, a minimum of 5% of the required number of samples and measurements were selected for QC evaluation. At least one (1) duplicate soil sample was collected in each open land survey unit for QC evaluation, and replicate scans were conducted on 5% of the scan locations chosen at random. In the structure basement survey units, 5% of the ISOCS measurements, with the locations selected at random, were replicated for QC evaluation.

Define the Boundaries of the Survey Boundaries of the Survey: The actual physical boundaries as stated for each survey unit.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 44 Temporal Boundaries: Estimated times and dates for the survey. Scanning and sampling in a survey unit was normally performed only during daylight and dry weather.

Constraints: The most common constraints were the weather, temperature, wet surfaces, debris, and vegetation in a survey unit.

Develop a Decision Rule Decision Rule: If any measurement data exceeded the action levels specified in the survey plans, alternative actions would be taken.

Specify Tolerable Limits on Decision Errors The Null Hypothesis: Residual radioactivity in the survey unit exceeds the release criteria.

Type I Error: 7KLV LV DOVR NQRZQ DV WKH ³' HUURU. This is the error associated with incorrectly concluding the null hypothesis has been rejected. In accordance with LTP VHFWLRQWKHHUURUZDVVHWDW

Type II Error: 7KLV LV DOVR NQRZQ DV WKH ³' HUURU. This is the error associated with incorrectly concluding the null hypothesis has been accepted. In accordance with LTP VHFWLRQWKHHUURUZDVVHWDW

The Lower Bound of the Gray Region (LBGR): The LBGR was set at 50% of the Operational DCGL. In using the unity rule, the Operational DCGL becomes one (1) and the LBGR is set as 0.5.

Optimize Design Type of Statistical Test: The Sign Test was selected as the non-parametric statistical test for FSS. The Sign Test is conservative as it increases the probability of incorrectly accepting the null hypothesis (i.e., the conclusion will be that the survey unit does not meet the release criteria) and does not require the selection or use of a background reference area.

Number of Measurements/Samples Required for the Sign Test: All Phase 1 survey units had a relative shift of three (3), which corresponds to fourteen (14) systematic or randomly located samples or measurements for use with the Sign Test. Three (3) exceptions are survey units L3-012-102, B1-010-001 and B1-01-004, which had relative shifts of 1.67, which corresponds to seventeen (17) samples or measurements for use with the Sign Test.

Further, because of the method of survey in structure basements, the number of measurements for use with the Sign Test was based on 100% coverage of the surface area in the survey unit and the FOV of the ISOCS. The locations of the samples and measurements were determined using the software Visual Sample Plan (VSP).

Number of Judgmental Samples and Locations: Typically, a minimum of one (1) judgmental sample or measurement was required in each Phase 1 survey unit, with the exception of survey unit L1-010-101C, which did not require any judgmental samples. The selection of additional judgmental samples or measurements was at the discretion of the

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 45 FSS Supervisor. Locations chosen for judgmental investigation were usually areas of interest (small piles, trenches, collection points, cracks, etc.). In survey unit L1-SUB-TDS, continuing characterization samples were collected as judgmental samples during FSS.

Number of Scan Areas and Locations: For the lone Class 3 open land survey unit subject to this report (L3-012-102, Transmission Switch Yard), because there were no areas with elevated contamination potential, a scan area was developed around each random sample location that, when totaled, equated to 10% of the total surface area of the survey unit. For the Class 1 and Class 2 open land survey units detailed in this report, scan areas were established based on the systematic grid. A 25% scan coverage was prescribed for Class 2 land areas, and a 100% scan coverage was prescribed for Class 1 land areas. 100% scan coverage was required for the Class 1 structure survey units, and the locations of the scan areas were based on the systematic layout of the ISOCS measurements.

Number of Samples for Quality Control: The implementation of quality control measures as referenced by LTP Chapter 5, section 5.9.3 and the QAPP included the collection of soil samples for split sample analysis or the collection of replicate measurements and surveys, as appropriate, at a frequency of 5% of the sample/measurement set. The locations for duplicate samples or measurements and replicate scan areas were selected randomly using a random number generator.

Power Curve: The Prospective Power Curve, developed using characterization data and MARSSIM Power 2000 or COMPASS software, showed adequate power for the survey design in each of the survey units.

4.2 Survey Unit Designation and Classification Procedure LC-FS-PR-006, Survey Unit Classification (Reference 23), defines the decision process for classifying an area in accordance with the LTP and MARSSIM. During the FSS of areas submitted for this Phase 1 Final Report, no survey units were reclassified.

4.3 Background Determination During FSS area scanning, ambient backgrounds were determined and the technician established the ASP based on the background for each specific scan area. Each survey unit Release Record (enclosed as appendices to this report) discusses scan area readings.

4.4 Final Status Survey Sample Plans The level of effort associated with planning a survey is based on the complexity of the survey and nature of the hazards. Guidance for preparing FSS plans is provided in procedure LC-FS-PR-002, Final Status Survey Package Development and LC-FS-PR-015, Final Status Survey for Structures. The FSS plan for open land survey units uses an integrated sample design that combines scanning surveys and media sampling. For

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 46 structure basements, scanning and compliance measurements are performed with the ISOCS.

4.5 Survey Design 4.5.1 Determination of Number of Data Points The number of soil samples for FSS was determined in accordance with procedure LC-FS-PR-002, Final Status Survey Package Development and MARSSIM. The relative shift

IRUWKHVXUYH\\XQLWGDWDVHWLVGHILQHGDVVKLIWZKLFKLVWKH8SSHU%RXQGDU\\RI

the Gray Region (UBGR), or the DCGL (SOF of 1), minus the Lower Bound of the Gray 5HJLRQ/%*562)RIGLYLGHGE\\VLJPDZKLFKLVWKHVWDQGDUGGHYLDWLRQRIWKH

data set used for survey design. As the calculated relative shift for all Phase 1 survey units was greater than three (3), a value of 3 was used, except for survey units L3-012-103, B1-010-001 and B1-010-004 in which the relative shift was 1.67. The sample size from Table 5.5 in MARSSIM that equates to and errors of 0.05 and a relative shift of three (3) is fourteen (14). A relative shift of 1.67 equates to a sample size of seventeen (17). As a conservative measure, a total of twenty-eight (28) samples were collected in survey unit L3-012-102. Further, one (1) sample location was added in survey unit L1-010-101C during survey design, for a total of fifteen (15) systematic samples. Because of the method of survey in structure basements, the number of measurements for use with the Sign Test was based on 100% coverage of the surface area in the survey unit and the FOV of the ISOCS. Twenty-two (22) and forty-three (43) compliance measurements were prescribed for the WGTV and Reactor Building basements, respectively. Two (2) ISOCS measurements were added for the Reactor Building FSS, for a total of forty-five (45) compliance measurements.

A breakdown of the number of soil samples and ISOCS measurements collected for the Phase 1 survey units is provided in Table 4-4.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 47 Table 4-4 Number of Samples and Measurements for FSS 4.5.2 Sample and Measurement Locations Locations of the samples and measurements were determined using the software Visual Sample Plan (VSP). For open land areas, VSP software imports a topographical map of the selected survey area and designates the sample locations with coordinates and bearings based on the Wisconsin State Plane System. For structure survey units, a structure drawing, with all relevant survey surfaces represented, was constructed in or imported into VSP. The VSP software designated the measurement locations with coordinates based on a local origin established on the drawing (an x, y system in meters). Pacific Northwest National Laboratory (PNNL) created VSP for the United States Department of Energy. For those locations where access was impractical or unsafe, alternate sample locations were generated and documented. Sample locations were identified using GPS coordinates and are consistent with the Wisconsin State Plane System. Once located, sample points were physically marked and added to the survey map.

4.6 Instrumentation Radiation detection and measurement instrumentation for performing FSS is selected to provide both reliable operation and adequate sensitivity to detect the ROC identified at the site at levels sufficiently below the Operational DCGL. Detector selection is based on detection sensitivity, operating characteristics, and expected performance in the field.

The DQO process includes the selection of instrumentation appropriate for the type of measurement to be performed (i.e., scan measurements, ISOCS, or sample analysis) that are calibrated to respond to a radiation field under controlled circumstances. Instruments are also evaluated periodically for adequate performance to established quality standards Survey Unit Random/Systematic Samples or ISOCS Measurements Judgmental Samples or ISOCS Measurements Investigational Samples or ISOCS Measurements L1-SUB-DRS 14 6

0 L1-SUB-TDS 14 10(1) 0 L1-SUB-LES 14 1

0 L1-010-101 C 15 0

0 L2-011-102 14 8

0 L2-011-103 14 2

5 L3-012-102 28 2

0 B1-010-004 22 1

0 B1-010-001 45 6

0 (1) 8 samples for continuing characterization were collected as judgmental samples during FSS.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 48 and ensure that they are sensitive enough to detect the ROC with a sufficient degree of confidence. For example, when determining instrument MDCR, an index of sensitivity (d) of 1.38 was used to provide a correct detection rate of 95% and a false positive rate of 60%.

The field instrumentation will, to the extent practicable, use data logging to automatically record measurements to minimize transcription errors.

Specific implementing procedures control the issuance, use, and calibration of instrumentation used for FSS. The specific DQOs for instruments are established early in the planning phase for FSS activities, implemented by standard operating procedures and executed in the FSS Sample Plan.

4.6.1 Detector Efficiencies The Ludlum Model 2350-1 Data Logger coupled with the Ludlum Model 44-10 2" x 2" Sodium Iodide (NaI) Gamma Scintillation Detector was selected as the primary radiation detection instrumentation for performing scanning for FSS land surveys at LACBWR.

The ISOCS was selected as the instrument of choice to perform FSS in basement structures. The source term geometry, i.e., concentration depth profile and areal distribution of the residual radioactivity in structures, is required to generate efficiency curves (i.e.,

efficiency as a function of energy) for the ISOCS gamma spectroscopy measurements. The concrete cores obtained during characterization, as described in LTP Chapter 2, provide information regarding the distribution of activity with depth for each structure. The areal distribution is assumed to be uniform for the efficiency calibration.

4.6.2 Detector Sensitivities The evaluation of open land areas requires a detection methodology of sufficient sensitivity for the identification of small areas of potentially elevated activity. Scanning measurements are performed by passing a hand-held detector, primarily the Ludlum Model 44-10 NaI detector, in gross count rate mode across the land surface under investigation. The centerline of the detector was maintained at the ground to detector distance detailed in the sample plan and moved from side to side in a 1-meter wide pattern at a rate of 0.5 m/sec.

The audible and visual signals were monitored for detectable increases in count rate. An observed count rate increase resulted in further investigation to verify findings and to define the level and extent of residual radioactivity.

An a priori determination of scanning sensitivity was performed to ensure that the measurement system (including the surveyor) was able to detect concentrations of radioactivity at levels below the regulatory release limit. The specified performance level and surveyor efficiency was expressed in terms of scan MDCR. This sensitivity is the lowest count rate that can be reliably detected at any given background by the measurement system. The specified MDCR correlates to the targeted MDC.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 49 This approach represents the surface scanning process for land areas defined in NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions (Reference 24). The gamma scan MDCR is discussed in detail in EnergySolutions TSD RS-TD-313196-006, Ludlum Model 44-10 Detector Sensitivity (Reference 25) which examines the gamma sensitivity for a 5.08 by 5.08 cm (2 x 2) NaI detector to several radionuclide mixtures of Co-60 and Cs-137 using sand (SiO2) as the soil base. TSD RS-TD-313196-006 derives the MDCR for the radionuclide mixtures at various detector distances and scan speeds. The model in TSD RS-TD-313196-006 uses the same geometry configuration as the model used in MARSSIM.

TSD RS-TD-313196-006 provides MDCR values for the expected LACBWR soil mixture based on detector background condition, scan speed, soil depth (0.15 meters), soil density (1.6 g/cm3), and detector distance to the surface of interest.

4.6.3 Instrument Maintenance and Control Control and accountability of survey instruments were maintained to assure the quality and prevent the loss of data. All personnel operating radiological instruments, analysis equipment, measurement location equipment, etc., were qualified to operate any assigned equipment and recognize irregular results and indications.

4.6.4 Instrument Calibration Instruments and detectors were calibrated for the radiation types and energies of interest or to a conservative energy source. Instrument calibrations were documented with calibration certificates and/or forms and maintained with the instrumentation and project records.

Calibration labels were also attached to all portable survey instruments. Prior to using any survey instrument, the current calibration was verified and all operational checks were performed.

Instrumentation used for FSS was calibrated and maintained in accordance with approved LaCrosseSolutions site calibration procedures. Radioactive sources used for calibration were traceable to the NIST and were obtained in standard geometries to match the type of samples being counted. When a characterized high-purity germanium (HPGe) detector was used, suitable NIST-traceable sources were used for calibration, and the software set up appropriately for the desired geometry. If vendor services were used, these were obtained in accordance with purchasing requirements for quality related services, to ensure the same level of quality.

4.7 Survey Methodology 4.7.1 Scan Surveys The LTP specifies the minimum amount of scanning required for each survey unit classification as summarized in Table 4-5. The total fraction of scanning coverage is

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 50 determined during the DQO process with the amount, and location(s) based on the likelihood of finding elevated activity during FSS.

Table 4-5 Recommended Scan Coverage Area Classification Surface Scans Class 1 100%

Class 2 10% to 100%,

Systematic and Judgmental Class 3 Judgmental The following pertains to survey unit L3-012-102 only, as it is the only Class 3 open land survey unit in this Phase 1 report. LTP Chapter 5, Section 5.6.4.4 states that for Class 3 survey units, judgmental surface scans will typically be performed on areas with the greatest potential of contamination and that, for open land areas, this will include surface drainage areas and collection points. Section 5.6.4.4 further notes that in the absence of these features, the locations of these judgmental scans will be at the discretion of the survey designer. Review of historical information provided in the HSA, combined with the results of the walkdown of the survey unit in preparation for the FSS, did not indicate any areas for increased contamination potential in survey unit L3-012-102. Therefore, the survey designers used discretion to choose the scan locations. Because there were no areas with elevated contamination potential, the scan areas were selected at the random sample locations with sufficient scan area at each location to meet the required scan percentage as defined in the FSS survey design. This allowed for the scanned areas to be evenly distributed throughout the survey unit. Two (2) judgmental scan areas were added during the performance of the FSS.

For the Class 1 and Class 2 open land survey units detailed in this report, scan areas were established based on the systematic grid. Each established scan area was located with GPS, marked, and completely scanned. Areas with elevated readings were marked and investigational samples were collected. The probe was positioned as close to the ground as possible and was moved at a scan speed not to exceed 0.5 meters per second. Table 4-6 provides a summary of the area scanned during FSS.

The use of ISOCS for structure basement FSS essentially eliminates the need for scanning because of the instruments large FOV. For the Class 1 structure basements discussed in this report, the scan requirement for FSS is fulfilled by ensuring that 100% of the surface area of the survey unit is captured by the FOVs of the systematic ISOCS measurements.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 51 Table 4-6 Summary of Total Area Scanned Survey Unit Survey Unit Classification Area (m2)

Area Scanned (m2)

% Scan L1-SUB-DRS 1

1,125 1,125 100 L1-SUB-TDS 1

1,186 1,186 100 L1-SUB-LES 1

1,336 1,336 100 L1-010-101 C 1

88 88 100 L2-011-102 2

2,258 585 25 L2-011-103 2

2,445 617 25 L3-012-102 3

11,711 1,344(1) 11 B1-010-004 1

311 311 100 B1-010-001 1

512 512 100 (1) Includes judgmental scan areas.

During the scanning, the technician recorded data and observations in a Field Log. This log documented field activities and other information pertaining to the survey.

4.7.2 ISOCS Measurement Collection FSS using the ISOCS was performed in accordance with LaCrosseSolutions procedure LC-FS-PR-014, Operation of the Canberra LabSOCS/ISOCS Genie-2000 Portable Gamma Spectroscopy System (Reference 26).

The ISOCS detector was oriented perpendicular to the surface of interest. In most cases, the exposed face of the detector was positioned at a distance of three (3) meters above the surface. A plumb or stand-off guide attached to the detector was used to establish a consistent source to detector distance and center the detector over the area of interest. With the 90-degree collimation shield installed, this orientation corresponds to a nominal FOV of 28 m2.

A majority of the ISOCS measurements were collected using the circular plane geometry, although varied geometries were necessary due to encountered structural anomalies. The sump, located in the north-west corner of the WGTV basement floor, required the use of a rectangular plane geometry. ISOCS geometries specifically employed within the structure survey units are provided as attachments in the respective Release Records.

4.7.3 Soil Sampling In accordance with the FSS Sample Plan and applicable procedures, FSS technicians collected surface soil samples at locations specified by the survey design. Each sample location was documented, along with soil conditions and observations, and a chain of custody was developed to maintain sample integrity.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 52 4.8 Quality Control Surveys The method used for evaluating Quality Control (QC) split samples and replicate measurements collected in support of the FSS program is specified in the FSS QAPP. QC split and replicate data was assessed using criteria taken from the USNRC Inspection Manual, Inspection Procedure 84750, Radioactive Waste Treatment and Effluent and Environmental Monitoring (Reference 27). A minimum of 5% of the sample and measurement locations used in the FSS design were selected randomly for QC evaluation using a random number generator.

Most soil split samples taken for FSS were field duplicates; that is, samples obtained from one location, homogenized, divided into separate containers, and treated as separate samples. For structures, replicate ISOCS measurements were collected for QC compliance.

QC split samples and replicate measurements were used to assess errors associated with sample/measurement heterogeneity, sample/measurement methodology, and analytical procedures. It is desirable that when analyzed, there is agreement between the split samples and replicate measurements to their respective standard sample or measurement, resulting in data acceptance. If there was no agreement, the FSS Supervisor or FSS Manager evaluated the magnitude and impact on survey design, the implementation and evaluation of results, as well as the need to perform confirmatory sampling. If the FSS Supervisor or FSS Manager had determined that the discrepancy affected quality or was detrimental to the implementation of FSS, then a Condition Report (CR) would have been issued.

No CRs were issued, and there was either an acceptable agreement between standard and split samples/replicate measurements in every Phase 1 survey unit or, if not, no further action was deemed necessary (splits or replicates far below Operational DCGL).

To maintain the quality of the FSS, isolation and control measures were implemented prior to, during, and upon completion of FSS until there was no risk of recontamination or when the survey area will be released from the license. Following FSS, and until the area is released, a semi-annual surveillance will be performed on FSS completed survey units.

This includes an inspection of area postings, inspection of the area for signs of dumping or disturbance and some sampling from selected locations. In the event that isolation and control measures were compromised, a follow-up survey may be performed after evaluation.

5 Survey Findings Procedure LC-FS-PR-008, Final Status Survey Data Assessment, provides guidance to C/LT personnel to interpret survey results using the DQA process during the assessment phase of FSS activities.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 53 The DQA process is the primary evaluation tool to determine that data is of the right type, quality and quantity to support the objectives of the FSS Sample Plan. The five steps of the DQA process are:

x Review the Sample Plan DQOs and the survey design.

x Conduct a preliminary data assessment.

x Select the statistical test.

x Verify the assumptions of the statistical test.

x Draw conclusions from the data.

Data validation descriptors described in MARSSIM Table 9.3 were used during the DQA process to verify and validate collected data as required by the FSS QAPP.

5.1 Survey Data Conversion During the data conversion, the FSS Supervisor or FSS Manager evaluated raw data for problems or anomalies encountered during Sample Plan activities (sample collection and analysis, handling and control, etc.) including the following:

x Recorded data, x Missing values, x Deviation from established procedure, and x Analysis flags.

Once resolved, initial data conversion, which is part of preliminary data assessment was performed and consisted of converting the data into units relative to the release criteria (e.g., pCi/g) and calculating basic statistical quantities (e.g., mean, median, standard deviation). Tables 5-1 and 5-2 provide summaries of the basic statistical properties for Cs-137 and Co-60 in Phase 1 systematic or random sample/measurement populations.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 54 Table 5-1 Basic Statistical Properties of Phase 1 Land Systematic/Random Sample Populations Survey Unit Class Samples Radionuclide Statistical Summary (pCi/g)

Cs-137 Co-60 Max Mean Std. Dev.

Max Mean Std. Dev.

L1-SUB-DRS 1

14 9.49E-02 5.64E-02 1.62E-02 7.15E-02 4.56E-02 1.72E-02 L1-SUB-TDS 1

14 2.08E-01 8.68E-02 5.74E-02 1.13E-01 5.27E-02 2.30E-02 L1-SUB-LES 1

14 1.01E-01 4.82E-02 2.65E-02 7.83E-02 3.93E-02 2.73E-02 L1-010-101 C 1

15 1.41E+00 2.67E-01 3.38E-01 9.39E-02 5.51E-02 1.79E-02 L2-011-102 2

14 9.35E-02 5.63E-02 2.01E-02 7.08E-02 3.09E-02 2.15E-02 L2-011-103 2

14 1.66E-01 7.43E-02 4.13E-02 1.33E-01 3.55E-02 3.64E-02 L3-012-102 3

28 4.17E-01 9.78E-02 1.08E-01 5.78E-02 3.14E-02 1.64E-02 Table 5-2 Basic Statistical Properties of Phase 1 Structure Basement Systematic Measurement Populations Survey Unit Class ISOCS Measurements Radionuclide Statistical Summary (pCi/m2)

Cs-137 Co-60 Max Mean Std. Dev.

Max Mean Std. Dev.

B1-010-004 1

22 1.93E+05 5.77E+04 5.46E+04 1.43E+04 5.49E+03 4.84E+03 B1-010-001 1

45 7.03E+03 8.78E+02 1.32E+02 1.22E+03 3.90E+02 3.59E+02 Tables 5-3 and 5-4 provide summaries of the basic statistical properties for Cs-137 and Co-60 in Phase 1 judgmental and investigational sample/measurement populations.

Table 5-3 Basic Statistical Properties of Phase 1 Land Judgmental Sample Populations Survey Unit Class Samples Radionuclide Statistical Summary (pCi/g)

Cs-137 Co-60 Max Mean Std. Dev.

Max Mean Std. Dev.

L1-SUB-DRS 1

6 8.40E-02 4.71E-02 2.22E-02 6.23E-02 4.81E-02 1.22E-02 L1-SUB-TDS 1

10(1) 1.88E-01 8.57E-02 5.09E-02 2.57E-01 6.93E-02 6.98E-02 L1-SUB-LES 1

1 2.86E-02 2.86E-02 N/A 8.83E-03 8.83E-03 N/A L2-011-102 2

8 1.28E-01 6.84E-02 3.51E-02 7.74E-02 4.21E-02 2.47E-02 L2-011-103 2

7(2) 7.36E-02 6.00E-02 1.43E-02 7.42E-02 4.69E-02 1.84E-02 L3-012-102 3

2 7.20E-02 7.02E-02 2.55E-03 7.02E-02 6.61E-02 5.80E-03 (1)

Includes 8 continuing characterization samples collected as judgmental samples during FSS.

(2)

Includes 5 investigational samples.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 55 Table 5-4 Basic Statistical Properties of Phase 1 Structure Basement Judgmental Measurement Populations Survey Unit Class ISOCS Measurements Radionuclide Statistical Summary (pCi/m2)

Cs-137 Co-60 Max Mean Std. Dev.

Max Mean Std. Dev.

B1-010-004 1

1 1.31E+07 1.31E+07 N/A 3.85E+05 3.85E+05 N/A B1-010-001 1

6 8.58E+05 3.63E+05 3.67E+05 1.88E+04 1.05E+04 6.80E+03 5.2 Survey Data Verification and Validation Items supporting DQO sample design and data are reviewed for completeness and consistency. This includes:

x Classification history and related documents, x Site description, x Survey design and measurement locations, x Analytic method and detection limits and that the required analytical method(s) are adequate for the radionuclides of concern, x Sampling variability provided for the radionuclides of interest, x QC measurements have been specified, x Survey and sampling result accuracy have been specified, x MDC limits, x Field conditions for media and environment, and x Field records.

Documentation, as listed, is reviewed to verify completeness and that it is legible:

x Field and analytical results, x Chain-of-custody, x Field Logs, x Instrument issue, return and source check records, x Instrument downloads, and x Measurement results relative to measurement location.

After completion of these previously mentioned tasks, a Preliminary Data Assessment record was initiated. This record served to verify that all data are in standard units in relation to the DCGLs and requires the calculation of the statistical parameters needed to complete data evaluation which at a minimum, included the following:

x The number of observations (i.e., samples or measurements),

x The range of observations (i.e., minimum and maximum values),

x Mean, x Median, and

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 56 x Standard deviation.

In order to adequately evaluate the data set, consideration as additional options included the coefficient of variation, measurements of relative standing (such as percentile), and other statistical applications as necessary (frequency distribution, histograms, skew, etc.).

Finalization of the data review consisted of graphically displaying the data in distributions and percentiles plots.

5.3 Anomalous Data/Elevated Scan Results and Investigation In survey unit L2-011-103, elevated scan readings resulted in investigations, which included the collection of five (5) investigational soil samples. This process was documented in accordance with LC-FS-PR-008, Final Status Survey Data Assessment. All investigational samples had a SOF of less than 0.1 when compared to the Operational DCGL, and no further action was taken.

In survey unit B1-010-004, one (1) judgmental ISOCS measurement exceeded the OpDCGLB with a SOF of 14.1505. Compared to the DCGLB, the measurement had a SOF of 0.9893. The area-weighted SOF of the elevated judgmental measurement was added to the average systematic measurement SOF to develop the total SOF and overall dose attributed to the survey unit.

In survey unit B1-010-001, two (2) judgmental ISOCS measurements exceeded the OpDCGLB with SOFs of 1.0542 and 1.0734. Compared to the DCGLB, the measurements had SOFs of 0.074 and 0.0753. The area-weighted SOF of the elevated judgmental measurements was added to the average systematic measurement SOF to develop the total SOF and overall dose attributed to the survey unit.

5.4 Evaluation of Number of Sample/Measurement Locations in Survey Units An effective tool utilized to evaluate the number of samples collected in the sampling scheme is the Retrospective Power Curve generated by MARSSIM Power 2000 or COMPASS. The Retrospective Power Curve shows how well the survey design achieved the DQOs. For reporting purposes, all Release Records include a Retrospective Power Curve analysis indicating that the sampling design had adequate power to pass the FSS release criteria (i.e., an adequate number of samples was collected).

5.5 Comparison of Findings with Derived Concentration Guideline Levels The SOF or unity rule was applied to FSS data in accordance with the guidance provided in Section 2.7 of NUREG-1757, Vol. 2, and the LTP. This was accomplished by calculating a fraction of the Operational DCGL for each sample or measurement by dividing the reported concentration by the Operational DCGL. If a sample had multiple ROC, then the fraction of the Operational DCGL for each ROC was summed to provide a SOF for the sample. Surrogate Operational DCGLs for Cs-137, which take into account the

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 57 HTD radionuclide Sr-90, were calculated as part of the survey design for the FSS, but were only used to develop action levels. During data assessment, activities for Sr-90 were inferred based on the HTD ratios in Table 4-2 and compared to their respective DCGLs.

A Base Case SOF was calculated for each ROC by dividing the reported concentration by the Base Case DCGL. A Base Case SOF of one (1) is equivalent to the decision rule, meaning any measurement with a SOF of one (1) or greater, would not meet the 25 mrem/yr release criteria. The mean Base Case SOF was multiplied by 25 to establish the dose attributed to a survey unit.

In accordance with Section 5.5.4 of the LTP, for structures, the dose from elevated judgmental measurements (measurements exceeding SOF of 1 when compared to the OpDCGLB) is accounted for using an area-weighted approach. An area-weighted SOF is calculated and added to the average systematic measurement SOF. The product of this summation is then used to calculate the overall dose assigned to the basement survey unit.

The equation for calculating the area-weighted SOF is provided below.

Equation 5

=

+

x

where:

SOFB

=

SOF for structural surface survey unit within a Basement using Base Case DCGLs Mean ConcB ROCi

=

Mean concentration for the systematic measurements taken during the FSS of structural surface in survey unit for each ROCi Base Case DCGLB ROCi

=

Base Case DCGL for structural surfaces (DCGLB) for each ROCi Elev ConcB ROCi

=

Concentration for ROCi in any identified elevated area (systematic or judgmental)

SAElev

=

surface area of the elevated area SASU

=

adjusted surface area of FSS unit for DCGL calculation The Elevated Measurement Comparison (EMC) was not required for the survey units addressed by this report.

A summary of the SOF and dose contribution for each Phase 1 survey unit is provided in Table 5-5.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 58 Table 5-5 Mean Base Case SOF and Dose Contribution Survey Unit Mean Base Case SOF Dose (mrem/yr)

L1-SUB-DRS 0.0105 0.2620 L1-SUB-TDS 0.0125 0.3115 L1-SUB-LES 0.0100 0.2495 L1-010-101 C 0.0144 0.3597 L2-011-102 0.0095 0.2368 L2-011-103 0.0097 0.2416 L3-012-102 0.0090 0.2247 B1-010-004 0.0233(1) 0.5813 B1-010-001 0.0006(1) 0.0150 (1) Includes the area-weighted SOF from elevated judgmental measurements added to the average systematic measurement SOF.

5.6 Description of ALARA to Achieve Final Activity Levels Section N.1.5 of NUREG-1757, Vol. 2, states that For residual radioactivity in soil at sites that may have unrestricted release, generic analyses show that shipping soil to a low-level waste disposal facility is unlikely to be cost effective for unrestricted release, largely because of the high costs of waste disposal. Therefore, shipping soil to a low-level waste disposal facility generally does not have to be evaluated for unrestricted release. Section 4.4.1 of LTP Chapter 4 presents a simple ALARA analysis for the excavation and disposal of soils as low-level radioactive waste that confirms the statement in section N.1.5 of NUREG-1757, Vol. 2 that the cost of disposing excavated soil as low-level radioactive waste is clearly greater than the benefit of removing and disposing of soil with residual radioactivity concentrations less than the dose criterion. Since the cost is greater than the benefit, it is not ALARA to excavate and dispose of soils with residual radioactivity concentrations below the DCGL.

Section 4.4.2 of LTP Chapter 4 presents the ALARA analysis for basement structures. The ALARA analysis based on cost benefit analysis shows that further remediation of concrete beyond that required to demonstrate compliance with the 25 mrem/yr dose criterion is not justified.

Housekeeping and cleanup of survey units was completed prior to turnovers for FSS, and good housekeeping practices were employed during FSS. Good housekeeping practices and properly executed isolation and control in survey units mitigated any potential cross-contamnation and ensured that the reported residual activity levels were accurate and final.

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 59 5.7 NRC/Independent Verification Team Findings According to NRC Inspection Report (IR) 05000409/2017001 (DNMS) (Reference 28), on December 31, 2017, the NRC completed inspection activities at LACBWR, which included the observation of FSS and confirmatory surveys of survey units L1-010-101C and B1-010-004. The inspectors determined that the LaCrosseSolutions had properly remediated and performed FSS of survey unit B1-010-004. For survey unit L1-010-101C, the NRC inspectors identified a violation of 10 CFR 20.1501, for failure to perform necessary surveys during the demolition of the WTB foundation. After additional remediation, the inspectors concluded that the licensee had also properly performed FSS of survey unit L1-010-101C. See Appendix A4 for more details concerning remediation and FSS in survey unit L1-010-101C.

According to NRC IR Nos. 05000409/2018001 (DNMS)

(Reference 29) and 07200046/2018001 (DNMS) (Reference 30), on January 16, 2019, the NRC completed inspection activities at LACBWR, which included the observation of FSS and confirmatory surveys of survey units L1-SUB-DRS, L1-SUB-TDS, and L1-SUB-LES. The inspectors determined that the LaCrosseSolutions had properly remediated and performed FSS of the aforementioned survey units.

At the time of submittal of this report, the site has not received a report from the NRC detailing the results of the confirmatory surveys they performed in survey units B1-010-001, L2-011-102, and L2-011-103. While the NRC was on-site, though, nothing of concern was noted.

Survey unit L3-012-102 did not undergo a confirmatory survey by the NRC.

6 Summary FSS is the process used to demonstrate that the LACBWR basement structures and soil comply with the radiological criteria for unrestricted use specified in 10 CFR 20.1402. The purpose of the FSS Sample Plan is to describe the methods to be used in planning, designing, conducting, and evaluating the FSS.

The two radiological criteria for unrestricted use specified in 10 CFR 20.1402 are: 1) the residual radioactivity that is distinguishable from background radiation results in a TEDE to an AMCG that does not exceed 25 mrem/yr, including that from groundwater sources of drinking water, and 2) the residual radioactivity has been reduced to levels that are ALARA.

The survey units addressed in this Final Report have met the DQOs of the FSS Sample Plans developed and implemented for each. In each survey unit, all identified ROC were used for statistical testing to determine the adequacy of the survey unit for FSS, the sample data passed the Sign Test, and a Retrospective Power Curve showed that adequate power

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 60 was achieved. Each of the survey units were properly classified. In accordance with the LTP Section 5.10, the survey units meet the release criterion.

7 References

1. La Crosse Boiling Water Reactor License Termination Plan (LTP)

2. NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM)

3. LC-QA-PN-001, Final Status Survey Quality Assurance Project Plan (QAPP)

4. LC-FS-PR-002, Final Status Survey Package Development

5. LC-FS-PR-015, Final Status Survey for Structures

6. LC-FS-PR-010, Isolation and Control for Final Status Survey

7. LC-FS-PR-008, Final Status Survey Data Assessment

8. NUREG-1757, Vol. 2, Consolidated Decommissioning Guidance Characterization, Survey, and Determination of Radiological Criteria - Final Report

9. LC-FS-PR-009, Final Status Survey Data Reporting

10. RS-TD-313196-003, La Crosse Boiling Water Reactor Historical Site Assessment

11. LC-FS-TSD-001, Use of ISOCS for FSS of End State Sub Structures at LACBWR

12. LC-FS-TSD-002, Operational Derived Concentration Guideline Levels for Final Status Survey

13. LC-RS-PN-164017-001, Characterization Survey Report

14. RS-TD-313196-001, Radionuclides of Concern During LACBWR Decommissioning

15. LC-RP-PG-003, Radiological Instrumentation Program

16. LC-RP-PR-060, Calibration and Initial Set Up of the 2350-1

17. LC-FS-PR-004, Sample Media Collection for Site Characterization and Final Status Survey

18. LC-FS-PR-005, Sample Media Preparation for Site Characterization and Final Status Survey

19. LC-FS-PR-012, Chain of Custody Protocol

20. LC-FS-PN-002, Characterization Survey Plan

21. GG-EO-313196-RS-RP-001, LACBWR Radiological Characterization Survey Report for October and November 2014 Field Work - November 2015

FINAL STATUS SURVEY FINAL REPORT - PHASE 1 61

22. RS-TD-313196-004, LACBWR Soil DCGL, Basement Concrete DCGL, and Buried Pipe DCGL

23. LC-FS-PR-006, Survey Unit Classification

24. NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions

25. RS-TD-313196-006, Ludlum Model 44-10 Detector Sensitivity

26. LC-FS-PR-014, Operation of the Canberra LabSOCS/ISOCS Genie-2000 Portable Gamma Spectroscopy System

27. USNRC Inspection Manual, Inspection Procedure 84750, Radioactive Waste Treatment and Effluent and Environmental Monitoring

28. NRC Inspection Report 05000409/2017001 (DNMS)

29. NRC Inspection Report 05000409/2018001 (DNMS)

30. NRC Inspection Report 07200046/2018001 (DNMS) 8 Appendices A1 FSS Release Record, Survey Unit L1-SUB-DRS A2 FSS Release Record, Survey Unit L1-SUB-TDS A3 FSS Release Record, Survey Unit L1-SUB-LES A4 FSS Release Record, Survey Unit L1-010-101C A5 FSS Release Record, Survey Unit L2-011-102 A6 FSS Release Record, Survey Unit L2-011-103 A7 FSS Release Record, Survey Unit L3-012-102 A8 FSS Release Record, Survey Unit B1-010-004 A9 FSS Release Record, Survey Unit B1-010-001