ML18331A026

Enclosure 3 - Revised Ltp Pages

ML18331A026
Person / Time
Site:	La Crosse  File:Dairyland Power Cooperative icon.png
Issue date:	11/15/2018
From:	LaCrosseSolutions
To:	Office of Nuclear Material Safety and Safeguards
References
LC-2018-0075
Download: ML18331A026 (23)

Text

Enclosure 3 to LC-2018-0075 Revised LTP Pages (22 pages total)

La Crosse Boiling Water Reactor License Termination Plan Revision 1

• Decreases a survey unit area classification (i.e., impacted to not impacted, Class 1 to Class 2; Class 2 to Class 3; or Class 1 to Class 3 without providing NRC a minimum 14 day notification prior to implementing the change in classification,

• Increases the DCGLs and related minimum detectable concentrations (for both scan and fixed measurement methods),

• Increases the radioactivity level, relative to the applicable DCGL at which an investigation occurs,

• Changes the statistical test applied to one other than the Sign test, or

• Increases the Type I decision error.

• Change the approach used to demonstrate compliance with the dose criteria (e.g., change from demonstrating compliance using DCGLs to demonstrating compliance using a dose assessment that is based on final concentration data).

• Change parameter values or pathway dose conversion used to calculate the dose, such that the resultant dose is lower than in the approved LTP and if a dose assessment is being used to demonstrate compliance with the dose criteria.

The contact for LTP information, including any submitted changes and updates, is:

Gerard P. van Noordennen Vice President, Regulatory Affairs LaCrosseSolutions, LLC S4601 State Road 35 Genoa, WI. 54632-8846 (224) 789-4025 gpvannoordennen@energysolutions.com 1.7. References

1. Letter from Dairyland Power Cooperative to the Nuclear Regulatory Commission, Application for Order Approving License Transfer and Conforming Administrative License Amendments, dated October 8, 2015.

2. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.179, Standard Format and Content of License Termination Plans for Nuclear Power Reactors, Revision 1 - June 2011.

3. U.S. Nuclear Regulatory Commission NUREG-1700, Revision 1, Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans - April 2003.

4. U.S. Nuclear Regulatory Commission NUREG-0191, Final Environmental Statement related to Operation of the La Crosse Boiling Water Reactor by Dairyland Power Cooperative - April 1980.

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La Crosse Boiling Water Reactor License Termination Plan Revision 1

5. Letter from Dairyland Power Company to the Nuclear Regulatory Commission,

Subject:

Planning for ISFSI, LAC-14029, March 10, 2008.

6. www.wunderground.com.

7. www.mvp.wc.usace.army.mil/projects/Lock 8.

8. Dairyland Power Cooperative, LaCrosse Boiling Water Reactor (LACBWR)

Decommissioning Plan, revised November 2003.

9. http://www.city-data.com/city/Genoa-Wisconsin.html.

10. 2008 Vernon County Tax Assessment Rolls.

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

12. EnergySolutions LC-RS-PN-164017-001, LACBWR Radiological Characterization Survey Report for June thru August 2015 Field Work - November 2015.

13. U.S. Nuclear Regulatory Commission, NUREG-1575, Revision 1, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), August 2000.

14. U.S. Nuclear Regulatory Commission NUREG-0586, Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, Supplement 1, Volume 1 -

November 2002.

15. Dairyland Power Cooperative, LACBWR Decommissioning Plan and Post-Shutdown Decommissioning Activities Report (D-Plan/PSDAR), Revision - March 2014.

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La Crosse Boiling Water Reactor License Termination Plan Revision 1 characterization data collected and an assessment of the characterization results, the characterization survey is considered adequate to demonstrate that it is unlikely that significant quantities of residual radioactivity have gone undetected.

As indicated in Chapter 3 of this License Termination Plan (LTP), the Reactor Building and the Waste Gas Tank Vault (WGTV) will be demolished and removed to a depth of at least 3 feet below grade. All other impacted LACBWR buildings, structures and components, other than the following structures, will be demolished and removed in their entirety. The impacted above grade structures that will remain are:

• LACBWR Administration building

• G-3 Crib House

• LACBWR Crib House

• Transmission Sub-Station Switch House

• G-1 Grib House

• Barge Wash Break Room

• Back-up Control Center

• Security Station None of the buildings and structures associated with the Genoa 3 Fossil Station (G-3) are expected to be radiologically impacted. Therefore, the structures associated with G-3 will remain intact and functional for G-3 power operations. The G-3 Crib House is classified as impacted due to its location in an impacted soil survey unit. The impacted above grade structures that will remain are not expected to contain residual radioactivity and will be subjected to FSS.

The major sub-grade structures that will be backfilled and remain at license termination are the basements of the Reactor Building and Waste Gas Tank Vault (WGTV) located below the 636 foot elevation (3 feet below grade). The Waste Treatment Building (WTB), Turbine Building (including the Turbine Sump and Turbine Pit), the Piping and Ventilation Tunnel, Reactor/Generator Plant, and the Chimney Foundation will all be removed in their entirety.

In the Reactor Building, all internal structural surfaces, systems and components will be removed. All internal concrete will be removed to expose the steel liner, which will also be removed, leaving only the remaining structural concrete outside the liner below the 636 foot elevation (i.e., concrete bowl below 636 foot elevation, concrete pile cap and piles.) In the WGTV, the remaining structure will consist of the floors and foundation walls as well as concrete piling cap and piles below the 636 foot elevation. The structural surfaces that will remain at LACBWR following the termination of the license are constructed of steel reinforced concrete which will be covered by at least 3 feet of soil and physically altered to a condition which would not allow the remaining backfilled structures to be plausibly occupied.

LTP Chapter 1, section 1.3.1 describes the site license boundary. The fenced area of the Independent Spent Fuel Storage Installation (ISFSI) Facility as located in Lot 7 of the site (and shown in Figure 2-1) is excluded from the scope of the LTP.

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La Crosse Boiling Water Reactor License Termination Plan Revision 1 0.061 pCi/g. Based on the fact that no other plant derived ROC was detected, it was concluded that the elevated direct alpha measurements were due most likely to the presence of coal ash material on the roof. Following the second characterization effort, the roof materials were entirely replaced during a roof repair of the G-3 buildings.

Direct and removable contamination surveys were performed of representative high personnel traffic areas of the G-3 Coal Plant. Scan measurements were performed using the Ludlum Model 2360 instrument with 43-93 detector. Approximately 5% of the concrete surfaces in the areas of interest were scanned, including floors and lower wall surfaces. The average scan MDC ranged from 2,540 dpm/100cm2 to 2,617 dpm/100cm2. No observed scan measurement exceeded the MDC during the course of this survey. Ten (10) locations were chosen at random, primarily from the main lobby area and the maintenance area. The results of all direct measurements for gross activity was less than the MDC of 74 dpm/100cm2 for alpha and 875 dpm/100cm2 for beta-gamma. No removable contamination was identified at concentrations greater than MDC by the analysis of the smear samples.

2.3.6. Impacted Structures and Systems The decommissioning approach for LACBWR requires the demolition and removal of all impacted Class 1 buildings, structures, systems and components to a depth of at least 3 feet below grade. In addition, all systems and exposed metal below 3 feet below grade will also be removed. The accepted elevation for grade at LACBWR is the 639 foot elevation. The only Class 1 structures that will remain and be subjected to FSS are the remaining reinforced concrete walls and floors of the Reactor Building that will be exposed by the removal of the interior concrete and steel liner, the remaining reinforced concrete walls floors of the WTB, the remaining reinforced concrete walls and floors of the WGTV, and the remainder of the Piping and Ventilation Tunnels, Reactor/Generator Plant basement, the one foot thick portion of the Chimney Foundation, the Turbine sump and the Turbine pit. Consequently, all systems and components and structural surfaces above the 636 foot elevation will be remediated, disassembled and/or demolished, segregated by waste classification and disposed of as clean demolition debris, clean salvage or radioactive waste. No extensive characterization was or will be performed of equipment, systems or structures that will be removed prior to the performance of FSS.

Radiological surveys of the interiors of structures at LACBWR are routinely performed to ensure compliance with 10 CFR 20 requirements regarding the posting of areas and to identify radiological conditions for the implementation of controls for the protection of workers in these areas. The radiological information from these surveys will provide the basis for the disassembly and removal of systems and the demolition of impacted structures at the site. When remediation has adequately reduced radiological conditions to levels suitable for controlled demolition, the impacted structures will be demolished, packaged and properly disposed of as waste.

After commodity removal is complete, the structures that will remain at license termination, i.e.,

3 feet below grade, will be re-surveyed to determine the concentrations of the residual radioactivity and the extent of additional remediation required, if any, to meet the unrestricted use criteria.

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La Crosse Boiling Water Reactor License Termination Plan Revision 1 Table 2-19 2014 Groundwater Monitoring Results (pCi/L)

JUNE 2014 MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW-DW3 DW4 DW5 DW7 B11R B11AR B2 B3 200-A 200-B 201-A 201-B 202-A 202-B 203-A 203-B 204-A 204-B Gross Alpha 1.26E-01 -1.33E-01 4.44E+00 -3.78E-01 0.00E+00 7.33E-01 -2.13E+00 6.46E-01 4.04E+00 -2.26E-01 3.84E-01 -1.04E+00 1.06E+00 4.07E-01 1.74E+00 -2.29E-01 7.58E-01 1.27E+00 Gross Beta 1.48E+00 3.13E+00 1.93E+00 0.00E+00 3.59E+00 -2.70E-01 1.12E+01 4.14E+00 8.66E+00 -1.79E+00 7.26E+00 -1.04E+00 4.34E+00 1.49E+00 2.27E+00 5.53E+00 2.69E+00 2.57E-01 H-3 1.05E+02 1.59E+02 1.94E+02 1.23E+02 2.45E+02 1.61E+02 1.60E+02 1.94E+02 5.23E+01 1.24E+02 7.03E+01 1.05E+02 3.36E+02 1.06E+02 2.79E+02 2.79E+02 1.06E+02 3.49E+01 C-14 2.79E+00 0.00E+00 -3.97E+00 -5.08E+00 9.06E+00 6.83E+00 1.94E+00 -2.05E+00 9.26E+00 4.84E+00 4.84E+00 4.51E+00 0.00E+00 4.70E+00 4.54E+00 4.55E+00 4.66E+00 1.30E+01 Fe-55 -2.00E+01 -3.12E+01 -2.66E+01 -3.86E+01 -3.19E+01 -3.75E+01 -9.20E+00 -4.39E+01 1.93E+01 -4.16E+01 -3.20E+01 1.58E+01 -3.26E+00 -4.54E+01 -4.95E+01 -4.96E+01 -4.03E+01 -3.98E+00 Ni-59 1.83E+01 3.96E+01 -4.75E+01 2.60E+01 -4.16E+00 2.53E+00 -7.13E+00 3.15E+01 -2.18E+01 3.32E+01 -1.67E+01 -7.11E+01 2.49E+01 -1.05E+01 -2.71E+01 -4.66E+00 -6.22E+00 -2.47E+01 Co-60 1.02E+00 3.94E-01 2.50E+00 -2.42E+00 8.76E-01 -1.94E+00 -1.39E+00 -1.13E+00 1.01E+00 -1.27E+00 -6.04E-02 1.73E+00 -8.91E-01 1.02E+00 3.48E-02 -9.55E-01 1.91E-01 -2.31E-01 Ni-63 -1.78E+00 -6.14E+00 -7.40E+00 -4.38E+00 -3.57E+00 0.00E+00 -7.94E+00 1.50E+00 -1.97E+00 -4.69E+00 1.82E+00 -1.91E+00 -1.94E+00 -3.73E+00 -3.70E+00 -3.80E+00 0.00E+00 -1.97E+00 Sr-90 6.09E-01 8.99E-02 8.99E-02 1.78E-02 7.34E-01 6.11E-01 6.52E-01 6.12E-01 9.86E-01 9.98E-01 6.86E-02 1.02E+00 1.12E+00 6.05E-01 1.17E+00 8.23E-01 2.01E+00 6.11E-01 Nb-94 4.40E-01 1.43E-01 -1.26E+00 7.49E-01 -6.99E-01 1.53E+00 1.77E+00 6.65E-01 1.08E+00 8.47E-01 3.73E-01 1.80E+00 -3.55E-01 1.04E-01 1.59E+00 1.50E+00 2.26E-01 1.28E+00 Tc-99 -8.28E+00 -7.37E+00 -8.26E+00 -9.36E+00 -5.52E+00 -8.46E+00 -7.41E+00 -8.10E+00 3.55E+00 5.08E+00 3.92E+00 1.17E+00 3.88E-01 2.73E+00 4.36E+00 4.17E+00 6.95E+00 6.31E+00 Cs-137 -5.97E-01 -3.47E-01 -7.45E-01 -1.83E+00 1.24E+00 1.84E+00 1.37E+00 2.14E+00 1.77E-01 -3.64E-01 3.10E-01 -6.21E-01 -4.64E-01 2.86E+00 1.48E+00 1.39E+00 -3.04E-01 1.93E+00 Eu-152 9.48E+00 2.34E+00 -7.43E+00 -5.51E+00 -9.93E-01 1.48E+00 1.07E-01 1.42E+01 9.71E+00 -1.16E+01 4.15E+00 1.12E+01 4.11E+00 1.08E+01 -9.59E-01 4.99E+00 2.12E+00 7.68E+00 Eu-154 -5.24E+00 3.16E+00 1.35E+00 -4.73E+00 1.63E+00 -2.61E+00 2.36E+00 1.69E+00 5.77E-01 5.60E-01 -2.39E+00 -2.43E+00 -1.12E+00 -1.18E+00 1.93E+00 -3.69E+00 2.36+00 6.06E-01 Eu-155 -2.67E+00 1.46E+00 -1.54E-01 -1.07E+02 1.60E+00 -4.47E+00 1.60E+00 -3.17E+00 -3.63E+00 -3.44E+00 -2.92E+00 -2.60E+00 -1.59E+00 -3.49E+00 -4.91E+00 4.05E+00 -4.34E-01 1.66E+00 Pu-238 2.83E-02 -4.77E-02 -2.22E-02 8.22E-02 0.00E+00 4.09E-02 5.40E-02 8.29E-02 -1.38E-02 1.51E-02 5.33E-02 -5.02E-02 -8.46E-03 1.48E-01 -5.67E-02 -5.87E-02 -5.12E-02 -9.28E-03 Pu-239/240 -9.61E-02 3.36E-02 3.13E-02 4.07E-01 -2.54E-02 -3.39E-02 3.40E-02 1.46E-01 2.68E-02 -1.57E-02 -6.86E-03 2.36E-02 -2.27E-02 -4.54E-02 -2.27E-02 -3.95E-02 -2.49E-02 6.28E-02 Pu-241 3.51E+00 2.70E+00 -8.39E+00 2.62E+00 -5.72E+00 8.38E-01 0.00E+00 -7.39E-01 0.00E+00 4.49E+00 -4.85E-01 -1.42E+00 2.20E+00 2.72E+00 4.07E+00 1.03E+01 4.07E+00 -1.89E+00 Am-241 5.39E-02 3.42E-02 4.65E-03 -1.01E-02 1.22E-02 8.37E-02 3.98E-03 2.28E-02 1.72E-01 3.71E-01 9.46E-02 3.99E-03 1.16E-01 3.92E-02 2.94E-02 7.64E-02 1.56E-01 8.24E-02 a Bold values indicate concentration greater than MDC. Italicized values indicate reported values that were less than MDC.

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La Crosse Boiling Water Reactor License Termination Plan Revision 1 In preparation for the building demolition, an enclosure was erected on the face of the Reactor Building around the roll up doors that were installed to remove the reactor vessel. The engineered enclosure is approximately 80 feet by 80 feet by 90 feet. The enclosure is equipped with an HEPA ventilation system and a vestibule to allow waste containers to be loaded and removed from inside the enclosure.

Solutions utilized excavators with proper attachments to demolish and size-reduce all interior structures, components and concrete from within the enclosure. All system, components and concrete will be removed from the steel liner and loaded out through the Reactor Building Containment Tent and Waste Transfer Cell. The thermal shield was segmented using a hydraulic hammer into manageable pieces for special packaging as mixed waste.

Following the completion of Contamination Verification Surveys (CVS), the Reactor Containment Tent and Waste Transfer Cell were mechanically demolished utilizing excavators with appropriate attachments. The Reactor Building Liner was demolished utilized a combination of thermal cutting and mechanical methods.

After commodity removal and steel liner is complete, a radiological assessment will be performed on the exposed concrete to ensure that any individual ISOCS measurement will not exceed the Operational DCGL during Final Status Survey (FSS) (see section 5.4.1). If unacceptable contamination is identified on the liner, then decontamination activities will be conducted until levels are met. The remaining structural concrete below the 636 foot elevation (i.e., concrete bowl below 636 foot elevation, concrete pile cap and piles) will remain and be subjected to an FSS in accordance with LTP Chapter 5 Demolition debris resulting from the demolition within the Reactor Building will be treated as low-level radioactive waste, loaded into appropriate containers and trucked to a rail trans-load facility in Winona, MN where the waste container will be transferred to a rail car and then shipped to the EnergySolutions disposal site in Clive, Utah.

3.3.2. Turbine Building and Turbine Office Building The Turbine Building housed the steam turbine and generator, main condenser, electrical switchgear, and other pneumatic, mechanical and hydraulic systems and equipment. A 30-ton traveling bridge crane with a 5 ton auxiliary hoist capacity spans the Turbine Building. The crane has access to major equipment items located below the floor through numerous hatches in the main floor. The Turbine Building is 105 feet by 79 feet and 60 feet tall.

The Turbine Office Building contained offices, the Control Room, locker room facilities, laboratory, shops, counting room, personnel change room, decontamination facilities, heating, ventilating and air conditioning equipment, rest rooms, storeroom, and space for other plant services. In general, these areas were separated from power plant equipment spaces. The Turbine Office Building is 110 feet by 50 feet and 45 feet tall.

Commodity removal has been completed in the Turbine Building utilizing cutting tools and mechanical means to dismantle radioactive piping and components. Band saws and reciprocating saws will be the primary methods used. System pieces and waste will be sized to 3-10

La Crosse Boiling Water Reactor License Termination Plan Revision 1 3.4.5. Project Milestones Table 3-4 lists the current schedule for the remaining decommissioning activities.

Table 3-4 General Project Milestones Date Milestone Q2/2016 Submit LTP to NRC Q2/2016 License Transfer Complete Q3/2016 Mobilization Complete Q3/2017 Stack Demolition Complete Q1/2019 LTP Approval by NRC Q4/2018 Component Removal Complete Q1/2019 Building Demolition Complete Q2/2019 Transportation and Disposal Complete Q4/2019 Site Remediation Complete Q4/2019 FSS Complete Q2/2019 Site Restoration Complete Submit Remaining FSS Reports and Request to Reduce Licensed Area to Q4/2019 ISFSI Q2/2018 Submit License Transfer to Dairyland Amendment Request to NRC Q2/2019 License Transfer to Dairyland Approved by NRC Q2/2020 Request to Reduce Licensed Area to the ISFSI Approved by NRC Q2/2020 License Transfer to Dairyland Note: Circumstances can change during decommissioning. If Solutions determines that the decommissioning cannot be completed as outlined in this schedule, Solutions will provide an updated schedule to the NRC.

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La Crosse Boiling Water Reactor License Termination Plan Revision 1

• Barge Wash Break Room

• Back-up Control Center

• Security Station The site and public roads and railways that traverse through the site will also remain. None of the buildings and structures associated with the Genoa 3 Fossil Station (G-3) are expected to be radiologically impacted. Therefore, the structures associated with G-3 will remain intact and functional for G-3 power operations. The G-3 Crib House is classified as impacted due to its location in an impacted soil survey unit. The above grade structures listed above will be subjected to FSS using the acceptable screening values for building surface contamination from Table H.1 of Appendix H from NUREG-1757, Volume 2, Revision 1, Consolidated Decommissioning Guidance - Characterization, Survey, and Determination of Radiological Criteria, Final Report (4). Section N.1.5 of NUREG-1757 states that licensees who have remediated surface soil and surfaces to the NRC default screening criteria have remediated soil such that it meets the unrestricted use criteria in 10 CFR 20.1402, or if no residual radioactivity distinguishable from background, may be left at the site would not be required to demonstrate that these levels are ALARA. Therefore, there is no ALARA analysis for above grade structures in this Chapter.

All impacted systems, components and structural surfaces above the 636 foot elevation in Class 1 buildings will be removed during the decommissioning process and disposed of as a waste stream. Grade level at LACBWR is at the 639 foot elevation. The below-grade structural surfaces, or basements, that will remain at LACBWR following the termination of the license are solid concrete structures which will be covered by at least three 3 feet of soil and physically altered to a condition which would not allow the remaining structural surfaces, if excavated, to be realistically occupied. The concrete walls and floors of the basements will be remediated to levels that will provide high confidence that FSS measurements with ISOCS will not exceed radionuclide-specific DCGLs that represent the annual dose criterion for unrestricted use specified in 10 CFR 20.1402.

Examples of remediation techniques that may be used for the below grade structural surfaces include washing, wiping, pressure washing, vacuuming, scabbling, chipping, and sponge or abrasive blasting. Cost estimates for these techniques also include the amount of water generated and the cost to process, package and ship this waste. Concrete removal may include using machines with hydraulic-assisted, remote-operated, articulating tools. These machines have the ability to exchange scabbling, shear, chisel and other tool heads.

4.2.1.1. Scabbling and Shaving The principal remediation method expected to be used for removing contaminants from concrete surfaces is scabbling and shaving. Scabbling entails the removal of concrete from a surface by the high-velocity impact of a tool with the concrete surface which transforms the solid surface to a volumetric particulate which can be removed. One method of scabbling is a surface removal process that uses pneumatically operated air pistons with tungsten-carbide tips that fracture the concrete surface to a nominal depth of 0.125 inches at a nominal rate of about 130 ft2 (in accordance with NUREG/CR-5884, Volume 2, Revised Analyses of Decommissioning for the Reference Pressurized Water Reactor Power Station (5), section G.3.1) or 12.08 m2 per hour.

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La Crosse Boiling Water Reactor License Termination Plan Revision 1 Excavation dose applies only to the basement-specific area of concrete on the ground surface after excavation.

Table 4-2 Radionuclide Half-Life(s) and Decay Constant(s)

Half-Life Radionuclide (a)

(yrs) (yr-)

Co-60 5.27 E 00 1.31 E-01 Sr-90 2.91 E+01 2.38 E-02 Cs-137 3.00 E+01 2.31 E-02 Eu-152 1.35E+01 5.12E-02 Eu-154 8.80E+00 7.88E-02 (a) Dose significant ROC in accordance with TSD RS-TD-313196-001.

The actual dose from each scenario, assuming a summation of the dose from each scenario equaled 25 mrem/yr is presented in Table 4-3. Therefore, the dose values for each ROC from Table 4-3 were used to derive the AMCG (DOSEAMCG) variable in Equation 4-10 for each scenario in each basement.

Table 4-3 Dose for Individual Scenarios (DOSEAMCG)

Reactor Building WGTV GW DS EX GW DS EX (mrem/yr) (mrem/yr)

Co-60 1.07 0.27 23.67 1.65 0.27 23.14 Sr-90 24.83 0.00 0.13 24.92 0.00 0.07 Cs-137 2.74 0.28 21.96 2.89 0.28 21.89 Eu-152 0.11 0.30 24.59 0.11 0.30 24.62 Eu-154 0.15 0.29 24.55 0.14 0.29 24.59 4.4.2.9.2. ALARA Calculation The ALARA calculations performed to evaluate the concrete scabbling or shaving remediation activities are presented in Table 4-4. A result for the Conc/DCGL ratio that is less than one would justify remediation whereas a result greater than one would demonstrate that residual radioactivity is ALARA. The lowest Conc/DCGL ratio was calculated for the Excavation scenario in the WGTV at 1.33.

4.4.2.10. Conclusion Concrete structural surfaces below the 636 foot elevation will remain in place after license termination. The site dose contribution from remaining residual radioactivity remaining in these buried plant structures will be accounted for by the BFM. 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.

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La Crosse Boiling Water Reactor License Termination Plan Revision 1 derived HTD to surrogate ratio is less than the applicable HTD to surrogate ratio from TSD RS-TD-313196-001, Table 40, then no further action is required.

Using the appropriate scaling factors, the DCGL of the measured radionuclide (Cs-137) will be modified to account for the inferred radionuclide (Sr-90) according to the following equation from section 4.3.2 of MARSSIM:

Equation 5-1 DCGLHTD DCGLSUR =DCGLETD x Conc DCGLETD +DCGLHTD ConcHTD ETD where:

DCGLSUR = modified DCGL (or Basement Dose Factor) for surrogate ratio, DCGLETD = DCGL for easy-to-detect radionuclide, DCGLHTD = DCGL for the hard-to-detect radionuclide, ConcHTD = Ratio of the HTD or inferred radionuclide, and ConcETD = Ratio of the ETD or surrogate radionuclide.

5.2.10. Sum-of-Fractions 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 use and application of the unity rule is performed in accordance with section 4.3.3 of MARSSIM.

5.2.11. Dose from Groundwater Based upon the results of groundwater monitoring performed on the LACBWR site since 1987, when the reactor was permanently shutdown through the current period of active decommissioning, the dose from existing residual radioactivity in groundwater is expected to be low. However, in order to select operation DCGLs to perform final status surveys, a groundwater dose of 3.25 mrem has been assigned as discussed in 6.11.

5.2.12. Demonstrating Compliance with Dose Criterion The DCGLs for backfilled basements, soil, and buried piping for each ROC are presented in Tables 5-3, 5-5, and 5-7, respectively. These values are equivalent to the level of residual radioactivity in the media (above background) that could, when considered independently for each ROC, result in a TEDE of 25 mrem per year to the AMCG. For all media, the dose from the residual radioactivity from each ROC (radionuclide i) can be expressed as shown in the following equation:

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La Crosse Boiling Water Reactor License Termination Plan Revision 1 characterization survey is considered adequate to demonstrate that it is unlikely that significant quantities of residual radioactivity have gone undetected.

The soil (i.e., open land) survey units and survey unit classifications that will be used for the FSS of open land at LACBWR are presented in LTP Chapter 2, section 2.1.6 and Table 2-1. Basements that will be subjected to FSS and backfilled have also received characterization sufficient to understand the nature and extent of contamination. The initial survey units and survey unit classifications for structures, both above and below 636 foot elevation that were developed for characterization and decommissioning planning purposes are presented in LTP Chapter 2, section 2.1.6 and Table 2-2.

However, the FSS that will be applied to structures below 636 foot elevation uses a different design criterion that is not directly driven by the preliminary classifications selected for characterization.

Therefore, the preliminary survey unit boundaries and classifications will not apply to the FSS of structures (basements) below 636 foot elevation. See section 5.5 for the FSS design criteria for basement structure survey unit boundaries and the approach to determining survey area coverage.

5.3.3.4. Inaccessible or Not Readily Accessible Areas Section 2.4 describes areas where characterization surveys were deferred including soils under structures (to be surveyed as access is achieved), soils under concrete or asphalt coverings (to be surveyed when covering is removed), currently inaccessible concrete basement surfaces (WGTV interior surfaces) and the interiors of buried pipe that will remain. A majority of the areas where the initial characterization was deferred have since been surveyed and is discussed in detail below. As access is gained to areas that were previously inaccessible, additional characterization data will be collected as necessary, evaluated and stored with-other radiological survey data in a survey history file for the survey unit. In addition, as the decommissioning progresses, data from operational events caused by equipment failures or personnel errors which may affect the radiological status of a survey unit(s) will be captured. These events will be evaluated and, when appropriate, stored in the appropriate characterization survey package. This additional characterization data will be used in validating the initial classification and in planning for the FSS.

Areas where characterization surveys were deferred will be surveyed in accordance with LC-FS-PN-002, Characterization Survey Plan (14) as they become accessible or in some instances the areas will be incorporated into the FSS plan (such as soil beneath or adjacent to the WTB). In other areas (e.g., soil beneath slab-on-grade structures after the slab is removed), continuing characterization may be performed in accordance with LC-FS-PR-003, Radiological Assessments and Remedial Action Support Surveys (15). The scope, methods and adequacy of the surveys are summarized as follows:

WGTV interior structural surfaces:

Continuing characterization of the Waste Gas Tank Vault 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 5-18

La Crosse Boiling Water Reactor License Termination Plan Revision 1 (including two cores in the sump). The cores were 3 diameter to a depth of 6 and all cores were sliced into 1/2 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 (less than 10% of the dose limit) 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.

Underlying concrete in the Reactor Building basement:

Continuing characterization of the underlying concrete in the Reactor Building basement will be performed once all interior demolition is complete. The characterization survey will consist of a Radiological Assessment (considered a form of continuing characterization) of the underlying concrete to ensure that any individual ISOCS measurements will not exceed the Operational DCGLB during FSS.

The RA will consist of a beta-gamma scan over 100% of all accessible surfaces of the conrete and a minimum of 30 loose surface contamination samples will be obtained. Six (6) concrete core samples will be obtained at evenly distributed locations and an additional four (4) cores will be obtained at any areas of elevated activity identified during the scan survey. If no areas of elevated activity are identified during the scan, then the four core samples will be obtained at biased locations such as low points, cracks, or areas of discoloration. The core samples will be obtained by coring into the concrete to a depth of 6. All core samples will first be 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, will be analyzed for HTD radionuclides. In addition, a minimum of one sample beyond the 10% minimum will be selected at random, also for HTD radionuclide analysis.

Additionally, if levels of residual radioactivity in an individual sample exceed the Sum-of-Fractions (SOF) of 0.1 then the sample(s) will be analyzed for HTD radionuclides. All samples selected for HTD analysis during the RA will be analyzed for the full initial suite of radionuclides from Table 5-1 in the LTP.

Soil under the Turbine Building (suspect broken drain line):

On June 25, 2015, as part of a broader site characterization, five (5) locations were selected for angled coring to obtain soil from beneath the Turbine Building at the location of the broken drain lines.

GeoProbe technology was used to obtain the samples. At each of the 5 locations, samples were collected from the 10, 15 and 20 depths, for a total of fifteen (15) soil samples. The results are provided in LC-RS-PN-164017-001, 2015 Characterization Survey Report (16).

In February of 2018, the Turbine Building foundation was removed in its entirety, including all broken drain lines and adjacent soil. In the eastern portion of the excavation, a total of eight (8) soil samples 5-19

La Crosse Boiling Water Reactor License Termination Plan Revision 1 (including two cores in the sump). The cores were 3 diameter to a depth of 6 and all cores were sliced into 1/2 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 (less than 10% of the dose limit) 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.

Underlying concrete in the Reactor Building basement:

Continuing characterization of the underlying concrete in the Reactor Building basement will be performed once all interior demolition is complete. The characterization survey will consist of a Radiological Assessment (considered a form of continuing characterization) of the underlying concrete to ensure that any individual ISOCS measurements will not exceed the Operational DCGLB during FSS.

The RA will consist of a beta-gamma scan over 100% of all accessible surfaces of the concrete and a minimum of 30 loose surface contamination samples will be obtained. Six (6) concrete core samples will be obtained at evenly distributed locations and an additional four (4) cores will be obtained at any areas of elevated activity identified during the scan survey. If no areas of elevated activity are identified during the scan, then the four core samples will be obtained at biased locations such as low points, cracks, or areas of discoloration. The core samples will be obtained by coring into the concrete to a depth of 6. All core samples will first be 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, will be analyzed for HTD radionuclides. In addition, a minimum of one sample beyond the 10% minimum will be selected at random, also for HTD radionuclide analysis.

Additionally, if levels of residual radioactivity in an individual sample exceed the Sum-of-Fractions (SOF) of 0.1 then the sample(s) will be analyzed for HTD radionuclides. All samples selected for HTD analysis during the RA will be analyzed for the full initial suite of radionuclides from Table 5-1 in the LTP.

Soil under the Turbine Building (suspect broken drain line):

On June 25, 2015, as part of a broader site characterization, five (5) locations were selected for angled coring to obtain soil from beneath the Turbine Building at the location of the broken drain lines.

GeoProbe technology was used to obtain the samples. At each of the 5 locations, samples were collected from the 10, 15 and 20 depths, for a total of fifteen (15) soil samples. The results are provided in LC-RS-PN-164017-001, 2015 Characterization Survey Report (16).

In February of 2018, the Turbine Building foundation was removed in its entirety, including all broken drain lines and adjacent soil. In the eastern portion of the excavation, a total of eight (8) soil samples 5-19

La Crosse Boiling Water Reactor License Termination Plan Revision 1 Chapters 2 and 5 for additional discussion and justification of the non-impacted classification of areas outside of the LSE.

For the basements that will remain, all systems and components will be removed. The backfilled structures are generally referred to as either basements or structures in this LTP Chapter. For the Reactor Building, only the concrete exterior to the steel liner will remain; all interior concrete and the steel liner will be removed. The WGTV basement is comprised of concrete only. All remaining concrete will be decontaminated as necessary to meet the 10 CFR 20.1402 unrestricted use criteria.

Table 6-1 Basements and Below Ground Structures to Remain in LACBWR End State Ground Surface Elevation is 639 feet AMSL Material Floor and Wall Floor Elevation Basement/Structure remaining Surface Area (m2) (feet AMSL)

Reactor Building Concrete 511.54 612 Waste Gas Tank Vault Concrete 310.56 621 The End State will also include a limited number of buried pipes, with the majority not associated with contaminated operational systems. The exception is the remaining portion of the Circulating Water Discharge pipe which was used as for liquid effluent discharge as well as circulating water discharge. The buried piping to remain in the LACBWR End State is listed in Table 6-2.

For the purpose of this LTP, buried piping is defined as that contained in soil. Typical commercial power plants also contain piping that penetrates walls or is embedded in concrete.

However, the design of the LACBWR plant precludes the need for penetrations or embedded piping and none are present.

The potential for significant surface or subsurface soil contamination at LACBWR is low based on the findings of the EnergySolutions Technical Support Document (TSD) RS-TD-313196-003, La Crosse Boiling Water Reactor Historical Site Assessment (HSA) (2) and the results of extensive characterization performed in 2014 (see LTP Chapter 2). There are indications of subsurface soil contamination under the Turbine Building based on positive groundwater monitoring results down gradient of suspected broken drain lines. However, GeoProbe samples collected under the Turbine Building in the vicinity of the suspect drain lines did not identify plant-derived radionuclides above background. Additional characterization of the subsurface soil was performed after the Turbine Building foundation was removed and the underlying soil exposed. Very low levels of plant-derived radionuclides were identified.

6-4

La Crosse Boiling Water Reactor License Termination Plan Revision 1 credit for radioactive decay. The residual radioactivity remaining in the backfilled basements is assumed to inadvertently mix with the mass of structural concrete removed during excavation which is generally consistent with the guidance in NUREG-1757, Appendix J, for addressing subsurface contamination. The calculation is performed using Excel spreadsheet and results in BFM Excavation DCGLs that are expressed in the same units as the DCGLs for the BFM Insitu scenario, i.e., pCi/m2. The fundamental driver of the BFM Excavation DCGL calculation is that the average concentration in the excavated concrete is limited such that the surface soil DCGLs are not exceeded.

BFM Excavation DCGLs were calculated separately for the Reactor Building and WGTV consistent with the BFM Insitu DCGL calculations.

6.5.2. Soil Derived Concentration Guideline Levels, in units of pCi/g, were developed for residual radioactivity in surface soils that correspond to the 25 mrem/yr dose criterion. RESRAD was used to perform the calculation.

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 implantation 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). In the unlikely event that soil contamination is identified at LACBWR with a thickness greater than 15 cm, additional dose modeling may be required if the conceptual model assumed a 15 cm contamination thickness. 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.

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.

In the unlikely event that geometries are encountered during continuing characterization or during FSS that are not bounded by the assumed 1 m soil contamination thickness, the discovered geometries will be addressed by additional modeling. LS will seek approval from the NRC before implementing the change.

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

6.5.3. Buried Piping With the exception of the portion of the Circulating Water Discharge Pipe, none of the buried piping to remain at LACBWR was associated with contaminated systems and therefore contamination potential is minimal (see Table 6-2). A buried piping dose assessment was conducted to develop DCGLs for pipe. There is no embedded piping present at LACBWR (i.e.,

embedded in concrete).

The conceptual model for the buried piping dose assessment is similar to the BFM and includes two scenarios: Insitu and Excavation. In the Insitu scenario the residual radioactivity on the internal surfaces of the pipe is assumed to instantaneously release and mix with a thin 2.54 cm layer of soil in an area equal to the internal surface area of the pipe. For the Excavation scenario, the soil mixing layer is 15 cm due to the extensive ground surface disturbance associated with 6-8

La Crosse Boiling Water Reactor License Termination Plan Revision 1 6.10.1. Source Term The source term for the BFM is the residual radioactivity remaining in the backfilled basement End State at the time of license termination. LTP Chapter 2 provides the characterization data for the basements that will remain. The dimensions and surface areas are provided in RS-TD-313196-002 Final LACBWR End State Basement Concrete Surface Areas, Volumes, and Void Spaces (14). The expected source term configurations and activity levels projected to remain in each basement are summarized below.

6.10.1.1. Reactor Building The Reactor Building is a right circular cylinder with a hemispherical dome and semi-ellipsoidal bottom. It has an overall internal height of 144 feet and an inside diameter of 60 feet, and it extends 26 feet 6 inches below grade level. The steel shell thickness is 1.16 inch, except for the upper hemispherical dome, which is 0.60 inch thick. The lowest floor elevation is at 612 foot elevation.

The total wall/floor surface area in the portion of the building to be backfilled, i.e., below 636 foot elevation, is 512 m2. T Remediation plans call for all below grade concrete interior to the steel liner to be removed exposing the steel liner. Subsequent to interior concrete removal, the remaining portion of the steel liner will be removed. The remaining structural concrete outside the liner and below the 636 foot elevation will remain. The remaining concrete bowl sits on an external support structure comprised of a concrete pile cap and piles. The pile cap and piles were isolated from reactor operations by the interior concrete, the steel liner and the exterior concrete bowl. There is no evidence of contamination leakage beyond the steel liner or the concrete bowl exterior to the liner, therefore the pile cap and piles are considered to be non-impacted areas.

Six 1.27 cm thick core slices were collected from the surface downward and shipped to an offsite laboratory for analysis. The cores were collected from biased locations as indicated by elevated survey measurements and represent the areas expected to contain the highest levels of contamination. As shown in Table 6-3, Cs-137 is the predominate radionuclide at 88% of the radionuclide mixture. The Cs-137 results from the six cores ranged from 66 pCi/g to 7,500 pCi/g with an average of 1,903 pCi/g.

There is no indication that contamination is present in the concrete to remain after the steel lineris removed, therefore no cores were collected from the concrete outside the steel liner. In addition, general cleanup of loose contamination on the steel liner (concrete dust) after demolition and removal of the internal concrete is expected for operational radiation protection purposes before the steel liner is removed. This reduces the already low potential for transfer of activity in contaminated dust from the steel liner to the underlying concrete during removal of the liner. Therefore, minimal source term is expected in the Reactor Building End State.

6.10.1.2. Waste Gas Tank Vault The WGTV is a 29 foot by 31 foot underground concrete structure with 14 feet high walls and 2 feet thick floors, walls, and ceiling located 3 feet bgs just outside of the WTB. The gas decay system routed main condenser gases through various components for drying, filtering, recombining, monitoring and holdup for decay in the WGTV. The vault floor is at 621 foot elevation. A small sump is present that extends to 618 foot elevation.

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La Crosse Boiling Water Reactor License Termination Plan Revision 1 Table 6-34 Surface Soil Area Factors Area Factors Radionuclide 2 2 1m 2m 5m2 10m2 100m2 Cs-137 9.44 5.56 3.07 2.04 1.19 Co-60 9.11 5.42 3.01 2.00 1.18 Sr-90 11.21 6.66 3.69 2.45 1.41 Eu-152 9.30 5.50 3.04 2.02 1.18 Eu-154 9.38 5.54 3.06 2.03 1.18 6.20. Buried Piping Dose Assessment and DCGL Buried piping is defined as below ground pipe located outside of structures and basements. This section describes the buried pipe dose assessment methods and provides the resulting DCGLs.

The calculations are performed by RESRAD and Excel spreadsheet as detailed in Reference (8).

6.20.1. Source Term and Radionuclide Mixture With the exception of a portion of the Circulating Water System Pipe, none of the buried piping to remain at LACBWR was associated with contaminated systems and therefore contamination potential is minimal. See Table 6-2 for list of buried pipe to remain. The High Pressure Service Water from LACBWR Crib House to G-3 and Well water piping for Well #3 are considered non-impacted because they only contacted clean river water or groundwater with no potential for contamination and will continue operation after license termination.

DCGLs for buried pipe were calculated for the initial suite radionuclides. To date, no characterization has been performed in buried piping due to the very low contamination potential. The radionuclide mixture is assumed to be the same as listed in Table 6-3. As discussed in LTP Chapter 5, if continuing characterization is performed for buried pipe and the results indicate that the buried piping dose could exceed 10% of the 25 mrem/yr dose criterion, then samples will be analyzed for HTD radionuclides and additional assessments performed.

6.20.2. Exposure Scenario and Critical Group The dose assessment approach was generally consistent with the guidance in NUREG-1757, Appendix J in that two exposure scenarios were considered; 1) assuming that the buried pipe is excavated and spread across the surface (Excavation scenario), and 2) assuming that the buried pipe remains in situ (Insitu scenario).

NUREG-1757, Appendix J states that it should be appropriate to use the arithmetic average of the radionuclide concentration in the analysis, including any interspersing clean soil. The buried piping at LACBWR is a minimum of 1 m below grade. The LACBWR buried pipe excavation conceptual model is more conservative than the NUREG-1757, Appendix J conceptual model in that no mixing is assumed to occur with the soil in the 1 m cover or the interspersing clean soil between pipes during excavation.

The conceptual models for the buried pipe Insitu and Excavation scenarios are similar to those developed for the BFM. In the Insitu scenario, the residual radioactivity on the internal surfaces 6-52

La Crosse Boiling Water Reactor License Termination Plan Revision 1 Table 6-38 Summed Buried Pipe DCGLs for ROCs adjusted for Insignificant Radionuclide Fractions Buried Pipe Group Circulating Water Discharge Pipe Radionuclide (dpm/100 cm2) (dpm/100 cm2)

Co-60 7.50E+04 7.75E+04 Sr-90 5.16E+05 7.55E+05 Cs-137 3.18E+05 3.30E+05 Eu-152 1.64E+05 1.67E+05 Eu-154 1.52E+05 1.56E+05 6.21. Existing Groundwater Dose There is low potential for significant groundwater contamination to be present although low concentrations have been identified in groundwater adjacent to suspected broken floor drains under the Turbine Building. Sampling in 1983 from a well located down gradient of the Turbine Building indicated positive groundwater contamination at relatively low concentrations.

In late 2012, five additional monitoring well pairs were installed to support site characterization and license termination. Results indicated lower groundwater contamination levels than found in 1983, predominantly H-3. See LTP Chapter 2 for a summary of characterization and HSA results prior to submittal of LACBWR LTP Revision 0. In December 2017, the groundwater sampling program identified elevated H-3 with a maximum concentration of 24,200 pCi/L identified in February 2018 (well MW-203A). A subsequent sample from well MW-203A, collected in April 2018, contained lower H-3 concentrations of 12,100 pCi/L indicating a downward trend.

As a result an investigation was initiated to identify the source of the contamination and the extent of contaminant migration from the source. The investigation identified the source as being a reactor building ventilation exhaust that was directed toward the ground surface where H-3 condensed. This investigation included the performance of a dye tracer test and the development of a numerical groundwater transport model that concluded that the maximum groundwater concentration of H-3 would have been approximately 60,000 pCi/L for a brief period of time (less than a few months) and that the concentrations continue to decline. Prior to and during this investigation period, final status surveys were conducted using operational DCGLs with a dose contribution from groundwater as 3.25 mrem as assigned in LC-FS-TSD-002 Rev 01. To ensure that this dose is bounding, a dose calculation using a maximum H-3 concentration of 60,000 pCi/L and dose from the maximum well using the IC radionuclides from Table 2-19 (June 2014 sample data) is performed.

To support this dose calculation, Groundwater Exposure Factors from the initial suite of radionuclides were calculated using Ingestion Dose Conversion Factors from Federal Guidance Report 11 Reference (8) directly with an assumed industrial worker AMCG drinking water intake rate of 327 L/yr. See Table 6-39.

6-56

La Crosse Boiling Water Reactor License Termination Plan Revision 1 Table 6-39 Ingestion Dose Conversion Factors and Groundwater Exposure Groundwater FGR 11 ING DCF, Nuclide Exposure Factors, mrem/pCi mrem/y per pCi/L H-3 6.40E-08 2.09E-05 C-14 2.09E-06 6.83E-04 Fe-55 6.07E-07 1.98E-04 Ni-59 2.10E-07 6.87E-05 Co-60 2.69E-05 8.80E-03 Ni-63 5.77E-07 1.89E-04 Sr-90 1.42E-04 4.64E-02 Nb-94 7.14E-06 2.33E-03 Tc-99 1.46E-06 4.77E-04 Cs-137 5.00E-05 1.64E-02 Eu-152 6.48E-06 2.12E-03 Eu-154 9.55E-06 3.12E-03 Eu-155 1.53E-06 5.00E-04 Pu-238 3.20E-03 1.05E+00 Pu-239 3.54E-03 1.16E+00 Pu-240 3.54E-03 1.16E+00 Pu-241 6.85E-05 2.24E-02 Am-241 3.64E-03 1.19E+00 The concentrations for samples that were identified as positive in each sampling event from 2014 is shown in the following tables (the bolded values from LTP Table 2-19). Also, the corresponding annual dose is shown at the bottom of each table without considering radioactive decay.

As shown in these tables, the maximum dose from the identified positive detections in 2014 was 0.471 mrem/y from well MW-DW7 from the June 2014 sampling event from Pu-239 and no identified H-3. The dose from the more recent H-3 concentration of 60,000 pCi/L is 1.26 mrem for a total groundwater dose of 1.73 mrem. Given that the assigned dose from groundwater for selection of the operational DCGL is 3.25 mrem, this leaves an additional dose margin of 1.52 mrem.

This analysis demonstrates that the dose assigned dose for groundwater is well bounded using the maximum plume concentrations including the potential contributions of the IC radionuclides from 2014 results.

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La Crosse Boiling Water Reactor License Termination Plan Revision 1 Table 6-40 June 2014 Well Concentrations (pCi/L) for Radionuclides Identified as Positive and the Corresponding Dose (mrem/yr)

MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW-Nuclide DW3 DW4 DW5 DW7 B11R B11AR MW-B2 MW-B3 200-A 200-B 201-A 201-B 202-A 202-B 203-A 203-B H-3 2.45E+02 3.36E+02 2.79E+02 2.79E+02 C-14 Fe-55 Ni-59 Co-60 Ni-63 Sr-90 1.12E+00 1.17E+00 Nb-94 Tc-99 5.08E+00 Cs-137 Eu-152 9.48E+00 1.42E+01 9.71E+00 1.12E+01 1.08E+01 Eu-154 Eu-155 Pu-238 Pu-239 4.07E-01 Pu-240 Pu-241 1.72E-01 9.71E-01 1.16E-01 Am-241

Dose, 2.01E-02 0.00E+00 0.00E+00 4.71E-01 5.13E-03 0.00E+00 0.00E+00 3.01E-02 2.44E-02 2.42E-02 0.00E+00 2.37E-02 6.18E-02 2.29E-02 6.03E-02 5.84E-03 mrem/yr 6-57a

La Crosse Boiling Water Reactor License Termination Plan Revision 1 Table 6-40 June 2014 Well Concentrations (pCi/L) for Radionuclides Identified as Positive and the Corresponding Dose (mrem/yr)

(Continued)

Radionuclid MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW-e 204-A 204-B DW5 B11R B11AR 200-A 200-B 201-A 201-B 202-A 202-B 203-A 203-B 204-A 204-B H-3 C-14 1.30E+01 Fe-55 Ni-59 Co-60 3.56E+00 3.67E+00 Ni-63 5.00E+00 Sr-90 2.01E+00 1.14E+00 Nb-94 Tc-99 6.95E+00 6.31E+00 Cs-137 3.97E+00 2.17E+01 Eu-152 9.40E+00 Eu-154 4.18E+00 Eu-155 Pu-238 Pu-239 1.41E-01 Pu-240 Pu-241 1.56E-01 2.29E-01 1.40E-01 2.51E-01 2.69E-01 Am-241 Dose, 1.00E-01 1.19E-02 6.48E-02 0.00E+00 0.00E+00 5.13E-03 4.08E-03 1.63E-01 1.99E-02 1.30E-02 3.79E-02 0.00E+00 0.00E+00 6.02E-03 4.39E-01 mrem/yr 6-57b

La Crosse Boiling Water Reactor License Termination Plan Revision 1 6.22. Demonstrating Compliance with Dose Criterion As discussed in section 6.5.6, the final demonstration of compliance with the dose criterion will be made through the summation of dose from each of the five media. The compliance dose will be calculated using Equation 6-11 after FSS has been completed in all survey units. Note that the acronym BcDCGL in Equation 6-11 is defined in Reference (4) as Base Case DCGL which is equivalent to the DCGLs developed in this Chapter. Different terminology was required in Reference (4) to distinguish the full DCGLs (which represent 25 mrem/yr) from the Operational DCGLs which represent a fraction of 25 mrem/yr (see Reference (4)). The acronym SOF in Equation 6-11 is defined as Sum of Fractions.

The Release Record for each FSS unit will be reviewed to determine the maximum mean dose from each of the five source terms (e.g. basement, soil, buried pipe, above grade buildings, and existing GW). The compliance dose must be less than or equal to 25 mrem/yr. The calculation of the compliance dose will be documented in the final FSS Report for the site.

A detailed description of the terms in Equation 6-11 and the method for calculating the dose for each term is provided in Reference (4).

Equation 6-11 Compliance Dose = (Max BcSOFBASEMENT + Max BcSOFSOIL + Max BcSOFBURIED PIPE +

BcSOFAG BUILDING + GW BcSOFBS OB + GW BcSOFBPS OBP + Max SOFEGW) x 25 mrem/yr where:

Compliance Dose = must be less than or equal to 25 mrem/yr, Max BcSOFBASEMENT = Maximum BcSOF (mean of FSS systematic results plus the dose from any identified elevated areas) for backfilled Basements, Max BcSOFSOIL = Maximum BcSOF (mean of FSS systematic results plus the dose from any identified elevated areas) for open land survey units, Max BcSOFBURIED PIPE = Maximum BcSOF (mean of FSS systematic results plus the dose from any identified elevated areas) from buried piping survey units, Max BcSOFAG BUILDING = Maximum BcSOF (mean of FSS systematic results plus the dose from any identified elevated areas) from above grade standing building survey units, GW BcSOFBS OB = Groundwater scenario dose from the Other Basement (OB) which is defined as the basement not used to generate the Max BcSOFBASEMENT term in Equation 1 GW BcSOFBPS OBP = Groundwater scenario dose from the Other Buried Pipe (OBP) which is defined as the buried pipe survey unit not used to generate the Max BcSOFBURIED PIPE term in Equation 1 Max SOFEGW = Maximum SOF from existing groundwater (EGW) 6-58