ML17178A252

Boiling Water Reactor - License Termination Plan Request for Additional Information Letter

ML17178A252
Person / Time
Site:	La Crosse  File:Dairyland Power Cooperative icon.png
Issue date:	08/04/2017
From:	Bruce Watson Reactor Decommissioning Branch
To:	Sauger J EnergySolutions, EnergySolutions
MGVaaler NMMS/DUWP/RDB 415-3178 T-8I06
References
CAC L53134
Download: ML17178A252 (45)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 4, 2017 Mr. John Sauger Executive Vice President Chief Nuclear Officer Reactor D & D EnergySolutions 2701 Deborah Avenue Zion, IL 60099

SUBJECT:

LA CROSSE BOILING WATER REACTOR - REQUEST FOR ADDITIONAL INFORMATION REGARDING THE LACROSSE SOLUTIONS LICENSE TERMINATION PLAN AND ASSOCIATED CONFORMING LICENSE AMENDMENT (CAC NO. L53134)

Dear Mr. Sauger:

By letter dated June 27, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16200A095), as supplemented by letter dated December 1, 2016 (ADAMS Accession No. ML16347A026), LaCrosseSolutions, LLC (LS, the licensee) requested U.S. Nuclear Regulatory Commission (NRC) approval of the License Termination Plan (LTP) for the La Crosse Boiling Water Reactor (LACBWR), as well as an associated conforming license amendment. The LACBWR LTP provides the details of the plan for characterizing, identifying, and remediating the remaining residual radioactivity at the LACBWR site to a level that will allow the site to be released for unrestricted use in the future. The LACBWR LTP also describes how the licensee will confirm the extent and success of remediation through radiological surveys, provide financial assurance to complete decommissioning, and ensure the environmental impacts of the decommissioning activities are within the scope originally envisioned in the associated environmental documents.

The proposed conforming amendment would amend Possession Only License No. DPR-45 for LACBWR to reflect the approval of the LACBWR LTP when the review and approval process is completed by the NRC staff. Specifically, License Condition 2.C.(5), which approves the LACBWR LTP and establishes the criteria for determining when changes to the LTP require prior NRC approval, would be added to the license. The LTP will become a supplement to LACBWRs other decommissioning documents and will be implemented by the licensee to complete decommissioning activities at the LACBWR site. Once decommissioning is complete, a separate request will be made to the NRC by the licensee to terminate the LACBWR license.

In order to complete its ongoing review of the LACBWR LTP, the NRC staff requests additional information as specified in the Enclosure. The requested information was discussed in part with your staff during a conference call with NRC staff on March 22, 2017, and April 25, 2017.

Additional administrative and reference documents were submitted to the NRC as a result of these conference calls in support of the ongoing review; however, the staff has determined that additional information is needed to support completion of the LTP review. In order to assist the NRC staff in continuing its review of the subject request, please respond to this request for additional information within 60 days of the date of this letter.

J. Sauger If you have any questions, please contact Marlayna Vaaler, the LACBWR Project Manager, at (301) 415-3178 or via e-mail at marlayna.vaaler@nrc.gov.

Sincerely,

/RA/

Bruce A. Watson, CHP, Chief Reactor Decommissioning Branch Division of Decommissioning, Uranium Recovery, and Waste Programs Office of Nuclear Material Safety and Safeguards Docket Nos.: 50-409 and 72-046 License No.: DPR-45 cc: La Crosse Boiling Water Reactor Service List

ML17178A252 OFFICE NMSS/RDB/PM NMSS/DUWP/LA NMSS/RDB/HP NMSS/RDB/HG NAME MVaaler CHolston SGiebel RFedors DATE 03/16/2017 06/29/2017 06/13/2017 05/02/2017 OFFICE NMSS/PAB/PA NMSS/ERB/ER NMSS/RDB/BC NMSS/RDB/PM NAME LParks JQuintero BWatson MVaaler DATE 05/4/2017 02/27/2017 08/04/2017 0802/2017 La Crosse Boiling Water Reactor Service List:

Ken Robuck Jeffery Kitsembel Group President Disposal and Electric Division Decommissioning Wisconsin Public Service Commission EnergySolutions P.O. Box 7854 299 South Main Street, Suite 1700 Madison, WI 53707-7854 Salt Lake City, UT 84111 Paul Schmidt, Manager John Sauger Radiation Protection Section Executive VP and Chief Nuclear Officer Bureau of Environmental and Occupational Health Reactor DD Division of Public Health EnergySolutions Wisconsin Department of Health Services 2701 Deborah Avenue P.O. Box 2659 Zion, IL 60099 Madison, WI 53701-2659 Gerard van Noordennen Barbara Nick VP Regulatory Affairs President and CEO EnergySolutions Dairyland Power Cooperative 2701 Deborah Avenue 3200 East Avenue South, Zion, IL 60099 La Crosse, WI 54602-0817 Joseph Nowak Cheryl Olson, ISFSI Manager General Manager La Crosse Boiling Water Reactor LaCrosseSolutions Dairyland Power Cooperative S4601 State Highway 35 S4601 State Highway 35 Genoa, WI 54632-8846 P.O. Box 817 Genoa, WI 54632-8846 Dan Shrum Senior VP Regulatory Affairs Lane Peters, Site Manager EnergySolutions La Crosse Boiling Water Reactor 299 South Main Street, Suite 1700 Dairyland Power Cooperative Salt Lake City, UT 84111 S4601 State Highway 35 Genoa, WI 54632-8846 Russ Workman General Counsel Thomas Zaremba EnergySolutions Wheeler, Van Sickle and Anderson, S.C.

299 South Main Street, Suite 1700 44 East Mifflin Street, Suite 1000 Salt Lake City, UT 84111 Madison, WI 53703 George Kruck, Chairman John E. Matthews Town of Genoa Morgan, Lewis & Bockius LLP S5277 Mound Ridge Road 1111 Pennsylvania Avenue, NW Genoa, WI 54632 Washington, DC 20004 Regional Administrator U.S. NRC, Region III 2443 Warrenville Road Lisle, IL 60532-4352

REQUEST FOR ADDITIONAL INFORMATION LICENSE TERMINATION PLAN REVIEW LA CROSSE BOILING WATER REACTOR LACROSSESOLUTIONS, LLC DAIRYLAND POWER COOPERATIVE DOCKET NO. 50-409 By letter dated June 27, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16200A095), as supplemented by letter dated December 1, 2016 (ADAMS Accession No. ML16347A026), LaCrosseSolutions, LLC (LS, the licensee) requested U.S. Nuclear Regulatory Commission (NRC) approval of the License Termination Plan (LTP) for the La Crosse Boiling Water Reactor (LACBWR), as well as an associated conforming license amendment. The LACBWR LTP provides the details of the plan for characterizing, identifying, and remediating the remaining residual radioactivity at the LACBWR site to a level that will allow the site to be released for unrestricted use in the future. The LACBWR LTP also describes how the licensee will confirm the extent and success of remediation through radiological surveys, provide financial assurance to complete decommissioning, and ensure the environmental impacts of the decommissioning activities are within the scope originally envisioned in the associated environmental documents.

The proposed conforming amendment would amend Possession Only License No. DPR-45 for LACBWR to reflect the approval of the LACBWR LTP when the review and approval process is completed by the NRC staff. Specifically, License Condition 2.C.(5), which approves the LACBWR LTP and establishes the criteria for determining when changes to the LTP require prior NRC approval, would be added to the license. The LTP will become a supplement to LACBWRs other decommissioning documents and will be implemented by the licensee to complete decommissioning activities at the LACBWR site. Once decommissioning is complete, a separate request will be made to the NRC by the licensee to terminate the LACBWR license.

In order to complete the NRC staffs ongoing review of the LACBWR LTP, the information contained in the following requests for additional information (RAIs) is necessary.

Dose Assessment Topics (Identified by PAB)

PAB-1 Additional information is required to evaluate the dose contribution from the Insignificant Radionuclides given the uncertainty in mixture percentages.

Basis: The NRC staff considers radionuclides and exposure pathways that contribute no greater than 10 percent of the dose criteria established for decommissioning (i.e., 2.5 millirem per year (mrem/yr)) to be insignificant contributors. NUREG-1757, Consolidated Decommissioning Guidance, Volume 2, Characterization, Survey, and Determination of Radiological Criteria, Appendix O, Enclosure

Lessons Learned and Questions and Answers to Clarify License Termination Guidance and Plans, Question 2, states that it is incumbent on the licensee to have adequate characterization data to support and document the determination that some radionuclides may be deselected from further detailed consideration in planning the Final Status Survey (FSS). Radionuclides that are undetected may also be considered insignificant, as long as the minimum detectable concentrations (MDCs) are sufficient to conclude that the dose contribution is less than 10 percent of the dose criterion (i.e., with the assumption that the radionuclides are present at the MDC concentrations).

LACBWR Reference RS-TD-31319-00, Revision 1, describes how the radionuclide mixture fractions for the initial suite radionuclides were determined by calculating the mean concentration of each radionuclide in the 12 cores collected from the first 1.27 cm (0.5 in) of concrete...A mean mixture fraction was then calculated for each radionuclide by dividing the mean activity for each radionuclide by the summation of all mean activities. The maximum percentage insignificant contribution (IC) dose determined by the licensee assuming the mean concentration of the 12 cores is three percent. Of the 12 cores, 6 are from the Reactor Building, 3 are from the Piping Tunnel, and 3 are from the Waste Treatment Building (WTB). In an email to the NRC on April 25, 2017, the licensee informed the staff that the WTB and the Piping Tunnel will now be removed as part of decommissioning. Therefore, additional justification is necessary as to why the average of the 12 concrete core samples is still representative of the end state of the site.

Page 6-40 of the LTP states that to account for any additional, unspecified variability and to provide confidence that [hard to detect] (HTD) analyses that may occur during Continuing Characterization will not result in an IC dose exceeding the selected IC dose, a margin will be applied to the maximum individual IC percentage calculated using the Table 6-3 [Initial Suite of Potential Radionuclides and Mixture Fractions]

mixture by increasing the percentage to 5 percent. of RS-TD-31319-001, Revision 1, shows the insignificant radionuclide dose contribution (in terms of mrem/yr) assuming the concentration present in each of the 20 pucks analyzed from the 12 concrete cores. RS-TD-31319-001 states that these IC doses are calculated separately for each core using the [derived concentration guideline levels] (DCGLs) applicable to the basement where the core was collected and are a direct representation of IC dose from the data. The maximum, mean, and 95 percent [upper confidence limit] (UCL) of the 20 individual core IC doses were, 0.395 mrem/yr, 0.043 mrem/yr, and 0.060 mrem/yr, respectively. All of these values are below the maximum IC dose of 0.752 mrem/yr calculated using the mean mixture fraction (Table 15).

In Attachment 6 of RS-TD-31319-001, Revision 1, the concentration in the pucks is converted from picocuries per gram (pCi/g) to picocuries per square meter (pCi/m2),

but the conversion factor is not fully explained. The conversion factor is 2.35x104, which appears to be derived from the concrete density of 2.35 grams per cubic centimeter (g/cm3), or 2.35 grams per square centimeter (g/cm2) per unit depth (where 1 centimeter (cm) is equal to 0.393 inch (in)), converted to g/m2. Note that

the 12 cores used to find the mean mixture fractions represented the first 1.27 cm (0.5 in) of concrete, and that additional cores were taken which found contamination to a depth in the range of 1.27 cm - 5.08 cm (0.5 in to 2 in). If one were to assume that contamination exists deeper than 1 cm into the concrete in the buildings where the initial cores were taken, the conversion factor would be impacted.

In addition, there appears to be enough variation in the mixture fractions to warrant further investigation. For example, in Sample No. B1001101-CJ-FC-005-CV 0-1/2, taken from the Reactor Building, there were just as many or more radionuclides above the MDC (bold values in the table below) than for the other Reactor Building Samples, but this sample had a lower Cesium-137 (Cs-137) value. This results in mixture fractions that are quite different from the average mixture percentages without placing undue emphasis on the MDC levels. If one were to calculate the relative dose of the insignificant radionuclides in this sample, the percentage is 4 percent, which is greater than the most limiting value calculated, but is still less than the 5 percent margin imposed by the licensee.

However, if one were to also account for Neptunium-237 (Np-237) at the average MDC for the 12 cores (0.044 pCi/g), the insignificant contributor percentage would be greater than 5 percent; in fact, it would be closer to 10 percent. The NRC staff acknowledges that the total insignificant dose contribution calculated using the concentrations for this core (4.688x10-2 mrem/yr) is much less than 1.25 mrem/yr.

However, assuming that this same mixture percentage may exist in other areas of the Reactor Building, but scaled up to concentrations that would in total be equivalent to 25 mrem/yr, the Insignificant Contributor assignment of 5 percent may not adequately account for the dose contribution of the insignificant radionuclides.

There also appears to be differences between the buildings where the samples were taken that warrant further explanation and investigation. For example, samples from the Reactor Building and the WTB had more positive detections for insignificant radionuclides and higher Cs-137 levels than the Piping Tunnel samples.

Furthermore, there was one sample from the WTB that had significantly greater Cs-137 levels than the any other sample in the WTB, Reactor Building or Piping Tunnel (Sample No. B1002101-CJ-FC-002-CV 0-1/2). It is difficult to tell if this sample is unduly influencing the average mixture percentages given that the overall number of cores (12) is low. For example, if this sample is removed and the mean mixture fractions are recalculated, the maximum insignificant contribution is approximately 4 percent not including Np-237, and 8.1 percent including Np-237.

Concrete Core Sample No. B1001101-CJ-FC-005-CV 0-1/2 (Reactor Building)

Reactor Building Isotope Activity Activity Groundwater Drilling Spoils Excavation (pCi/g) (pCi/m2) (mrem/yr) (mrem/yr) (mrem/yr)

H-3 2.266 5.32E+04 5.97E-05 1.38E-10 1.01E-07 C-14 1.409 3.31E+04 8.36E-04 1.51E-10 4.96E-08 Fe-55 2.934 6.89E+04 4.03E-05 6.07E-11 6.84E-08 Ni-59 2.59 6.09E+04 3.76E-06 3.04E-11 2.37E-08 Co-60 39.572 9.30E+05 1.09E-01 3.92E-02 7.33E-01 Ni-63 213.849 5.03E+06 8.43E-04 6.38E-09 5.35E-06 Sr-90 29.256 6.88E+05 8.34E-01 5.12E-05 1.05E-03 Nb-94 0.531 1.25E+04 2.04E-04 3.59E-04 6.25E-03 Tc-99 0.562 1.32E+04 3.43E-04 1.49E-09 6.64E-05 Cs-137 58.907 1.38E+06 5.84E-02 1.42E-02 2.40E-01 Eu-152 0.972 2.28E+04 6.38E-05 4.53E-04 8.11E-03 Eu-154 3.887 9.13E+04 3.62E-04 1.92E-03 3.50E-02 Eu-155 0.35 8.23E+03 5.10E-06 6.00E-06 7.40E-05 Np-2371 0.044 1.03E+03 5.97E-02 3.78E-06 5.88E-04 Pu-238 0.68 1.60E+04 1.26E-02 6.23E-07 9.53E-05 Pu-239/240 0.489 1.15E+04 1.02E-02 4.94E-07 7.62E-05 Pu-241 7.226 1.70E+05 2.89E-03 1.02E-06 4.65E-05 Am-241 1.578 3.71E+04 1.77E-02 9.34E-06 3.40E-04 Am-243 0.047 1.12E+03 5.33E-04 3.87E-06 6.03E-05 Cm-243/244 0.103 2.43E+03 2.84E-04 5.19E-06 8.48E-05 Sum Dose 1.108 0.056 1.026 (mrem/yr)

Sum IC Dose 0.107 0.003 0.051 (mrem/yr)

Percent Relative 10% 5% 5%

IC Dose Percent IC Dose 0.4% 0.0% 0.2%

of 25 mrem Sum Dose w/o Np-237 1.048 0.056 1.025 (mrem/yr)

Sum IC Dose w/o Np-237 0.047 0.003 0.050 (mrem/yr)

Percent Relative IC Dose 4% 5% 5%

w/o Np-237 Percent IC Dose w/o 0.2% 0.0% 0.2%

Np-237 of 25 mrem 1Np-237 concentration is assumed to be at the average MDC for the 12 cores.

Path Forward:

a. Explain the basis for the conversion factor of 2.35x104 used to convert pCi/g to pCi/m2 assumed in Attachment 6 of RS-TD-31319-001, Revision 1, and how it incorporates an appropriate depth of contamination assumed for the concrete.

b. Evaluate the range of mixture percentages in the 12 concrete cores (20 pucks) in order to determine whether a conservative mixture percentage with regard to the HTD radionuclides is applied. Assess the relative contribution to dose of the insignificant radionuclides (including Np-237) assuming the mixture fractions from the individual cores, as shown in the example above. In cases where the percent Relative IC Dose exceeds the assumed 5 percent IC Dose Contribution, explain how the surveys that will be conducted will ensure that the IC dose will remain below 1.25 mrem/yr; otherwise, increase the IC contribution assumption.

c. Provide additional justification for why the average of the core samples taken from the Piping Tunnel, WTB, and Reactor Building is representative of what will remain on site at the end state of decommissioning given that the WTB and Piping Tunnel will now be removed. Evaluate the statistical differences in mixture fractions of the initial suite of radionuclides between the buildings.

Determine the potential insignificant dose contribution (including Np-237) using a mixture fraction for data specific to the Reactor Building and any other structure for which continuing characterization data may now be available. This evaluation should focus on the differences in samples between buildings where insignificant radionuclides were positively detected. The licensee should discuss how the assumed IC dose of 5 percent is protective for all basements and structures which will remain, whether it would be more appropriate to increase this assumption for all basements, and/or whether it is appropriate to identify different IC contributions for different areas of the LACBWR site.

PAB-2 Additional information is needed to evaluate whether Neptunium-237 should be considered a radionuclide of concern (ROC).

Basis: NUREG-1757, Volume 2, Appendix O, Question 2, states that it is incumbent on the licensee to have adequate characterization data to support and document the determination that some radionuclides may be deselected from further detailed consideration in planning the FSS. From the information presented in the LTP, it is unclear whether Np-237 should be included in the initial suite of radionuclides and/or the radionuclides of concern. Np-237 appears on Page 2-49 of the LACBWR LTP, Table 2-7 (Initial Suite of Radionuclides); however, it is not listed in Table 6-3 (Initial Suite of Potential Radionuclides and Mixture Fractions).

Np-237 also appears in the Historical Site Assessment (HSA) in the list of the initial suite of radionuclides and in Reference LC-RS-PN-164017-001. The HSA does not include Np-237 in the list of radionuclides considered for discounting. However, the HSA identifies Np-237 as an additional radionuclide in NUREG/CR-3474, Low-Level Radioactive Waste Classification, Characterization, and Assessment: Waste Streams and Neutron-Activated Metals, to be considered in the initial suite.

From the description of the process described in RS-TD-31319-001, Revision 1, it is not clear why Np-237 would not be included in the initial suite of radionuclides.

Radionuclides with half lives of two or more years identified in NUREG/CR-4289, Residual Radionuclide Contamination Within and Around Commercial Nuclear Power Plants, as being present in boiling water reactors (BWRs) were compared with the list of radionuclides in NUREG/CR-3474, and those which were not already in the initial suite were added Np-237 was identified in samples from the Dresden Generating Station (BWR) in NUREG/CR-4289 and its half-life is greater than 2 years, so it seems that Np-237 should have been included in the initial suite of radionuclides during this step for the LACBWR site.

It appears that the licensee is relying on the activities estimated for the spent fuel inventory from LACBWR, as listed in Table 5 of RS-TD-31319-001 (LAC-TR-138 Report, Spent Fuel Inventory Radionuclides), to discount Np-237. This reference states that the radioactivity inventory in the 333 spent fuel assemblies was performed by using the computer program Fact 1 and hand calculations performed by Dr. S. Raffety (Nuclear Engineer) during July of 1987. Activity in the fuel assemblies hardware is based on neutron activation of this hardware. All activity values have been decay corrected to January 1988. The associated table shows that Np-237 was less than 0.01 percent of the total activity. However, the licensee does not adequately explain why the LACBWR spent fuel inventory would be representative of the end state of the facility during decommissioning. In addition, this approach is not consistently applied to all radionuclides, since several of the radionuclides that were also below 0.01 percent relative activity were included in the initial suite for other reasons (e.g., tritium (H-3), Niobium-94 (Nb-94), Europium-154 (Eu-154), and Americium-243 (Am-243)).

The NRC staff also noted that the licensee calculated dose-to-source ratios (DSRs) for Np-237 for the soil model and basement fill model (BFM), and these are shown in the RESRAD output sheets. The Kd factor assigned for Np-237 in these RESRAD analyses is 5 cubic centimeters per gram (cm3/g), based on the mean sand soil Kd values (Sheppard and Thibault, Table A-1, 1990). The licensee analyzed for Np-237 in the 12 concrete cores, although it was not measured above the MDC in any of the cores. However, there are still onsite areas which the licensee has not characterized.

Path Forward:

a. Analyze the significance of Np-237 as a potential radionuclide of concern, as it was analyzed for the initial suite of radionuclides. This analysis should incorporate an appropriate Kd value for the radionuclide in accordance with the guidance from NUREG/CR-6697, Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0 Computer Codes. In other words, if the Kd is a sensitive parameter for Np-237, it should be assigned the 25th percentile value as opposed to the mean value.

b. Describe any additional commitments for sampling for Np-237 during continuing characterization activities and/or FSS that would be used to show that the dose contribution from Np-237 is not significant.

PAB-3 Provide additional support for why Europium-152 (Eu-152) and Europium-154 (Eu-154) are considered insignificant.

Basis: NUREG-1757, Volume 2, Appendix O, Question 2, states that it is incumbent on the licensee to have adequate characterization data to support and document the determination that some radionuclides may be deselected from further detailed consideration in planning the FSS. Table 13 of RS-TD-31319-001, Revision 1, states that the presence of Eu-152 in this sample [Reactor Biological Shield Activation Survey] as well as another concrete sample (10 CFR 61 [Licensing Requirements for Land Disposal of Radioactive Waste] analysis of the Lower Cavity Shield Block concrete) confirms that Eu-152 should be included in the Radionuclides of Concern. With a 13.6 year half-life, Eu-152 would still be a constituent in activated concrete sections at LACBWR that may be handled in any dismantlement activities in 2015. It is unclear why the licensee is not following the guidance established in this quote from RS-TD-31319-001, Revision 1.

The NRC staff notes that Eu-152 and Eu-154 were not found above MDC levels in the 12 concrete core samples. However, there are areas of the LACBWR site that have not yet been characterized, including the underlying concrete in the Reactor Building basement that will be exposed after liner removal. Chapter 2 of the LACBWR LTP states that most of the activated concrete will be removed from the Reactor Building. Specifically, after the thermal shield is removed, the remainder of the interior concrete will be removed, completely 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 below the 636 foot elevation (i.e., the concrete bowl below the 636 foot elevation, concrete pile cap and piles) will remain and be subjected to a FSS in accordance with LTP Chapter 5.

Since this concrete in the Reactor Building has yet to be characterized, the basis for not including Eu-152 and Eu-154 as radionuclides of concern is not clear.

Section 5.5.1.2 of the LACBWR LTP states that all contaminated concrete will be removed from the Reactor Building basement. Once the concrete and the liner are removed, it is anticipated that the residual radioactivity source term in the remaining Reactor Building basement structure will be minimal. Continuing Characterization of the Reactor Building Basement will be performed after concrete and liner removal to confirm assumption that minimal residual radioactivity will remain. The Continuing Characterization Plans and Reports will be provided for NRC information. It is unclear how this claim of entire removal of contaminated concrete is justified and how this assumption impacts the overall classification of the Reactor Building. (See RAI PAB-13 on basement classification and survey coverage).

Path Forward:

a. Describe continuing characterization plans for verifying that the doses from Eu-52 and Eu-154 are insignificant, or treat these as radionuclides of concern in appropriate areas during FSS activities at LACBWR.

PAB-4 Additional information is required to evaluate the dose from the Alternate Gardener Scenario.

Basis: NUREG-1757, Volume 2, Revision 1, Page 5-24 states that if the licensee evaluated scenarios based on reasonably foreseeable land uses, the licensee needs to provide either a quantitative analysis of or a qualitative argument discounting the need to analyze all scenarios generated from the less likely but plausible land uses.

The results of these analyses will be used by the staff to evaluate the degree of sensitivity of dose to overall scenario assumptions (and the associated parameter assumptions). The reviewer will consider both the magnitude and time of the peak dose from these scenarios. If peak doses from the less likely but plausible land use scenarios are significant, the licensee would need to provide greater assurance that the scenario is unlikely to occur, especially during the period of peak dose.

The LACBWR LTP states that the reasonably foreseeable scenario for the La Crosse site is industrial use, while residential land use is categorized as less likely but plausible in accordance with the NUREG-1757, Table 5.1 definitions.

Section 6.15 of the LACBWR LTP describes the alternate scenario calculations for a Resident Gardner (the less likely but plausible scenario).

The following scenarios were evaluated by the licensee:

1. Soil Conceptual Model for Resident Gardener with two source term assumptions:

a) Assuming the maximum concentrations from characterization (0.68 mrem/yr).

b) Applying Industrial Soil DCGLs and assuming a mixture percentage of ROCs so in total the concentrations are equivalent to 25 mrem/yr for industrial use (32.44 mrem/yr).

2. Waste Treatment Building Groundwater + WTB Drilling Spoils for Resident Gardner (Insitu) (0.091 mrem/yr).

3. Excavation of the WTB surface concentrations assuming that the radionuclides homogeneously mix with all the fill volume of the WTB for the Resident Gardener (43.38 mrem/yr).

The primary reasons the Resident Gardener calculations provided in the LACBWR LTP are not adequate are as follows:

• The impact of uncertainty in the radionuclide mixture fractions on dose is not evaluated.

• The parameter values are not fully explained, and the licensee did not perform a separate sensitivity analysis to define parameters or find new sensitive parameters.

• The calculations are only for the three ROCs, and not the entire suite of radionuclides.

• Doses are only estimated for the WTB basement, and no other basements.

Given that the WTB DCGLs were not the highest for all basements, and that the WTB will no longer remain, this is no longer relevant.

• The excavation of concrete with surface contamination unrealistically and non-conservatively assumes full mixing with the fill volume.

Reference RS-TD-313196-004 states that the alternate scenario dose assessments included the ROC at their respective mixture fractions as listed in Table 22.

However, Table 22 of RS-TD-313196-004 is titled Soil Area Factors. This is a typo and should be clarified to point to the correct table, which the NRC staff assumes is Table 6-20, titled ROC and Insignificant Radionuclide Mixture Fractions, of the LTP.

Reference RS-TD-313196-004, Attachment 14, provides the details of the Alternate Scenario Calculations. In Attachment 14, a table titled Calculation of Soil Concentrations Resulting in 25 mrem/yr in Industrial Scenario contains the following notes: Note 1: The concentrations in this column were determined in an iterative manner by adjusting the Cs137 concentration until the check calculation equals 25 mrem/yr. Note 2: [Strontium-90] (Sr90) value is mean Sr90/Cs137 ratio calculated from Sr/90 ratio in each of the 20 cores analyzed for HTD Reference TSD RSTD313196001, Revision 2. It appears that a hypothetical maximum concentration was applied based on the ratio of Sr-90/Cs-137 in the 12 concrete cores (20 pucks). However, the sensitivity of dose due to uncertainty in the mixture percentage is not evaluated.

Chapter 6 of the LACBWR LTP describes how the Resident Gardener soil dose was calculated with the RESRAD deterministic parameters used for the Industrial Use soil dose assessment with limited metabolic and behavioral parameter changes (from NUREG/CR-5512, Residual Radioactive Contamination from Decommissioning).

The parameters changes are:

• Contaminated Zone Thickness - 0.15 meter

• Inhalation Rate - 8400 cubic meters per year

• Fraction of Time Spent Indoors - 0.649

• Fraction of Time Spent Outdoors - 0.124

• Fruit, Vegetable, and Grain Consumption - 112 kilograms per year

• Leafy Vegetable Consumption - 21.4 kilograms per year

• Drinking Water Intake - 478 liters per year

• Depth of Roots - 1.22 meters Since the Resident Gardener Scenario involves different dose pathways, the sensitivity of parameters may not remain the same as in the Industrial Scenario.

Furthermore, the licensee only analyzes Cobalt-60 (Co-60), Cs-137, and Sr-90 as contaminants for the alternate scenario. This alternate scenario should be conducted with the full initial suite of radionuclides Reference RS-TD-313196-004, Revision 1, states that the WTB was used for the assessment since it is projected to contain the majority of residual radioactivity at

license termination. The WTB DCGLs do not represent the highest DCGLs proposed, and the WTB is no longer going to remain on site; therefore, the Alternative Scenario Dose should be assessed for each set of DCGLs proposed for the remaining basements (i.e., Reactor Building, Waste Gas Tank Vault (WGTV),

and Remaining Structures) to ensure the limiting dose is assessed.

The drilling spoils dose for the WTB is the (Maximum Concentration (pCi/g) divided by the Drilling Spoils DCGL for Industrial Use for the WTB (pCi/g per 25 mrem/yr))*(the ratio of the Adjusted Industrial Soil DCGL to the WTB Resident Gardener Soil Concentration corresponding to 25 mrem/yr). RS-TD-313196-004, Revision 1, Attachment 14, has the following basis for the Drilling Spoils adjustment:

The dose adjustment factor for the alternate scenario dose calculation for drilling spoils is based on the ratio of the LACBWR Adjusted Soil DCGL to the Resident Gardener Alternate Scenario concentration corresponding to 25 mrem/yr. This is conservative because the actual dose from the Resident Gardener scenario from drilling spoils would be less than the ratio of full area dose because the Resident Gardener area factors would be greater than the industrial scenario area factors due to the effect of source area on plant ingestion dose. Instead of multiplying by the ratio of the Soil DCGLs, a more transparent and defensible analysis would involve running the soil Resident Gardener RESRAD model with a reduced area of 0.457 m2 in order to find the concentrations equivalent to 25 mrem/yr. Then the maximum concentrations could be divided by this concentration to find the drilling spoils dose without the need to multiply by an adjustment ratio.

The licensee provides the following explanation for the evaluation of the excavation doses under the Resident Gardener Scenario: Because the BFM Excavation conceptual model is based on the limitation that the excavated concrete would not exceed the soil DCGLs, the maximum Resident Gardener dose from excavated concrete would be the same as that calculated for soil. However, the dose from concrete would be less since plants cannot be grown in concrete. The estimate of excavation doses for the Resident Gardener are non-conservative for the same reasons they are non-conservative for the Industrial Use Scenario. (See RAI PAB-7 on the Excavation Scenario.)

To derive the Excavation DCGLs, the licensee starts with the Surface Soil DCGL concentrations in pCi/g and finds the total Curies associated with the total volume of concrete in the basement. This activity is homogeneously distributed over the entire surface area of the basement. So, in essence, the intruder is assumed to excavate all the surfaces and total volume of concrete in a basement, and homogeneously mix it to obtain a concentration in pCi/g. The concrete is assumed to act like soil, which is a conservative assumption. However, this analysis dilutes, to an unrealistic extent, the source term. The licensee also performs a second excavation analysis where the intruder is assumed to excavate the fill material. The fill material is assumed to be at concentrations derived by homogeneously mixing the WTB Surface DCGLs for the Industrial Scenario across the entire fill volume. This is also non-conservative because realistically, the source term will not evenly distribute across the entire fill volume. Therefore, the excavation scenario for the Resident Gardner should be reevaluated using more reasonable dilution assumptions.

Path Forward:

a. Describe the technical basis for the hypothetical maximum concentrations assumed for the Resident Gardener Scenario. Analyze the sensitivity of the dose to the uncertainty in the radionuclide mixture fractions, and thus maximum concentrations, assumed. Alternatively, describe how the maximum concentrations assumed reflect the actual maximum concentrations per radionuclide that will be allowed to be left at the site at license termination.

b. Show that sensitive parameters for the Resident Gardener Scenario are defined according to the guidance in NUREG-1757 and NUREG/CR-6697. Perform a Sensitivity Analysis for the Resident Gardener Scenario to identify sensitive parameters and determine their appropriate values.

c. Reevaluate the Resident Gardener Groundwater Dose for each set of DCGLbs and Soil DCGLs proposed for the full initial suite of radionuclides (i.e., Soil, Reactor Building, WGTV, and Remaining Structures). The analysis should provide a technical basis for the maximum hypothetical concentrations assumed that takes into account uncertainty in the mixture percentages and applies a conservative mixture percentage. (For undetected radionuclides, the concentration assumed may be based on the MDC value for each radionuclide.)

The analysis should also incorporate any relevant parameter changes resulting from analyses related to other RAI responses (e.g., groundwater elevation). For example, for the Reactor Building, a Resident Gardener Dose should be calculated for a conceptual model that models the Reactor Building as one building, using an appropriate value for groundwater elevation. The analysis should assign appropriate values for the sensitive parameters identified.

d. Reevaluate the Resident Gardener Drilling Spoils Dose by running RESRAD using a method similar to that implemented to calculate the Drilling Spoils DSRs for the industrial scenario, with the decreased contaminated zone area of 0.457 m2, but with the appropriate sensitive parameters and pathways turned on for the Resident Gardener. Use those revised DSRs to calculate a drilling spoils dose assuming the maximum concentration, which considers a conservative mixture percentage for the radionuclides.

e. Reevaluate the Excavation Resident Gardener Dose using the revised Resident Gardener Soil DSRs and a concentration that assumes the Industrial Surface DCGLb, at a conservative mixture percentage. The analysis should consider a reasonably conservative dilution with fill volume as opposed to full mixing with the entire fill volume (See RAI PAB-7 related to check calculations for the Excavation DCGLs). The analysis should be done for each set of DCGLb (or the most limiting concentration across basements).

f. Calculate the dose attributable to the insignificant radionuclides from the revised Resident Gardener Scenarios for the basement and Soil DCGLs for the Groundwater, Drilling Spoils, and Excavation Scenarios. Compare any

differences to the insignificant radionuclides identified for the Industrial Use Scenario and assumed dose percentage contribution of insignificant radionuclides (5 percent). Discuss how the surveys during continuing characterization and FSS will ensure that the percentage assumed is adequate.

PAB-5 Additional information is needed regarding continuing characterization, specifically when samples will be analyzed for the full initial suite of HTDs.

Basis: In accordance with 10 CFR 50.82, the license termination plan must include a site characterization. NUREG-1700, Revision 1, Standard Review Plan for Evaluating Nuclear Power Reactor License Termination Plans, states that the NRC staff should review the licensee's site characterization plans and site records (as required under 10 CFR 50.75(g)). Section 5.3.4.4 of the LACBWR LTP states that Continuing Characterization Plans and Reports will be provided to the NRC for information. The licensee is not committing to submit continuing characterization plans to the NRC for review and/or approval, but only the results.

Chapter 2 of the LACBWR LTP lists the five areas at the La Crosse site that will undergo continuing characterization and states that the continuing characterization data will be used to confirm the classification of the inaccessible areas and identify areas requiring remediation. At least one sample from each of the inaccessible areas listed above will be analyzed for the full initial radionuclide suite including HTD. The five areas are as follows:

1. WGTV interior structural surfaces

2. Underlying concrete in the Reactor Building basement after liner removal

3. Soil under the Turbine Building (suspect broken drain line)

4. Soil adjacent to and beneath basement structures

5. Interior of buried pipe that may remain Given that some of these areas are broadly defined (i.e., soil adjacent to and beneath basement structures, or interior of buried pipe that may remain), more information is necessary for the NRC staff to complete its review of the adequacy of the continuing characterization for these areas, including the technical basis for why a minimum of one sample to be analyzed for the initial suite of radionuclides from each area is sufficient.

In addition, the commitment in Chapter 2 to take at least one sample from each inaccessible area and analyze it for the full suite of radionuclides is not in agreement with commitments in Chapter 5 of the LACBWR LTP. Section 5.1 of the LTP states:

If any continuing characterization sample and/or measurement is taken in soil or buried pipe, and the result indicates a [sum of fractions] (SOF) in excess of 0.5 based on gamma spectroscopy results (and inferring Sr-90), then a sample will be collected at the location of the highest accessible individual measurement and analyzed for HTD radionuclides. In this unlikely situation, if the analysis indicates HTD radionuclides (other than Sr-90) at

concentrations exceeding the MDC, then the insignificant contributor dose will be calculated using the core data.

If the insignificant contributor dose from the sample is less than the insignificant contributor dose assigned for DCGL adjustment (see LTP Chapter 6, Section 6.13.1), then the current adjustment to the soil or buried pipe DCGLs will be retained. If the insignificant contributor dose from the continuing characterization soil or buried pipe sample is greater than the insignificant contributor dose assigned for DCGL adjustment, then the soil or buried pipe DCGLs for the affected survey unit will be re-adjusted to account for the increased dose. It is possible, but not likely, that the data could indicate different ROC for the area where the additional characterization data was taken. If so, the specific ROC will be applied to the survey unit.

Similarly, Section 6.18.1 of the LTP states:

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 50 percent of the 25 mrem/yr dose criterion, then samples will be analyzed for HTD radionuclides and additional assessments performed.

It is unclear why the licensee has committed to taking a minimum of one sample for HTDs per inaccessible area without any dependency on the SOF, but also states that one sample will be analyzed for HTDs from the location of highest measurement only if the SOF > 0.5 for soil and buried pipe. In addition, three out of the five inaccessible areas listed in the LACBWR LTP are for soil or buried pipe so it is unclear which rule will apply to each inaccessible area.

Furthermore, while the LTP describes a trigger for re-evaluating the insignificant contribution dose should the one sample show HTDs exceeding MDCs, it does not describe any triggers for further investigation and sampling for HTDs. In the case that HTDs are positively identified in the one sample, the NRC staff would expect this event to trigger additional sampling and analysis of HTDs to get a better sense of the insignificant contribution and surrogate ratios. Accordingly, the LTP should describe any situations that would trigger additional sampling beyond the commitment of 10 percent or a minimum of one sample per inaccessible area.

Section 5.1 of the LTP also states that based on process knowledge, the radionuclide mixture found in the Reactor Building, WTB and Remaining Structures was applied to the WGTV as it is expected to be the most representative mixture.

Additional details on the process knowledge applied should be provided, especially in light of the fact that the WTB and Piping Tunnel will no longer remain on site.

Section 5.5.2.1.1 of the LTP states that the entire source term in the Reactor Building basement will be removed. In addition, Page 6-22 of the LTP states that cores were not collected in the Reactor Plant / Ventilation Plant area or the Turbine

Building sump and pit due to the low expectation of significant contamination being present and their very small areas. Additional information is necessary to confirm both statements in order to complete the NRC staffs review.

Path Forward:

a. Provide additional details regarding the plans for continuing characterization at the La Crosse site, including the survey units comprising the five areas listed as inaccessible areas in Chapter 2 of the LACBWR LTP.

b. Explain the technical basis for taking a minimum of one sample to be analyzed for the initial suite of radionuclides from each inaccessible area identified in Chapter 2 of the LTP. Clarify whether the commitment is to take one sample per survey unit, or one sample in each of the five areas that are listed.

c. Explain why there is a commitment to take a minimum of one sample in the five inaccessible areas listed in Chapter 2 without a dependency on SOF, but to take one sample from the highest measurement area only if the SOF > 0.5 for soil and buried pipe. Furthermore, since items 3 to 5 listed above are either soil or buried pipe areas, clarify whether there will be a minimum of one sample taken in these areas or whether one sample will only be taken when the SOF > 0.5.

d. Describe whether any continuing characterization results would trigger additional analysis of HTDs beyond 10 percent of the samples (or a minimum of one sample) from each survey area. The licensee should specify when the additional sampling of HTDs would take place (i.e., during further continuing characterization activities or during the FSS).

e. Describe the process knowledge basis for using measurement results from other buildings as the most representative radionuclide mixture for the WGTV building and/or other structures. Discuss this conclusion especially given that the WTB and Piping Tunnel will no longer remain on site. Provide the characterization plan for the WGTV interior and structural surfaces, including the basis for the number and location of samples, as well as which samples will be analyzed for the full suite of radionuclides.

f. Provide the characterization plan for the underlying concrete in the Reactor Building Basement after liner removal to confirm the statement that the entire source term in the Reactor Building basement will be removed. Describe the measurements, including the basis for the number and location of samples, that will be conducted to confirm that there is no contamination remaining below or behind the steel liner after removal.

g. Clarify why there is a low expectation of significant contamination in the Reactor Plant / Ventilation Plant area and/or the Turbine Building sump and pit. Discuss this considering the history of radiological spill incidents in the Turbine Building.

Clarify whether any additional continuing characterization will be done for these areas. Describe the number and location of samples to be taken in these areas.

PAB-6 Additional justification is needed regarding the conceptual model for the Reactor Building.

Basis: Section 6.11.1 of the LACBWR LTP states, with regard to the Reactor Building Conceptual Model:

The mixing distance into the fill after leaching from floor and wall surfaces above 619 foot is more uncertain than leaching into the Reactor Building fill volume below 619 foot. Above 619 foot, contact with water is the result of periodic flushing from seasonal water level rise, or rainwater infiltration, as opposed to continual contact with water in the saturated zone.

A different approach to defining the vertical flow rate was required for LACBWR due to the site groundwater being hydraulically connected to the Mississippi River and the resulting seasonal water level fluctuation into and out of the unsaturated zone. The vertical water flow rate is conservatively modeled in RESRAD by assuming that the annual rate is defined by the distance that water recedes from the seasonal high elevation of 629 foot to the 619 foot average water table elevation (10 feet (ft) or 3.05 meter (m)). This seasonal 3.05 m water elevation change is used to conservatively bound the net flow of water to the saturated zone in RESRAD by forcing the infiltration rate to be 3.05 m/yr.

Given the uncertainty associated with the water table level (see RAI HYDRO-1 on determining the groundwater level), and therefore the extent to which the Reactor Building walls may be in contact with the saturated zone, the use of an infiltration /

precipitation rate to mimic water elevation changes is not appropriate.

The NRC staff notes that the sensitivity report associated with the LACBWR LTP showed that precipitation is not a sensitive parameter for the Reactor Building conceptual models above or below the 619 foot level. However, groundwater elevation changes, and the fraction of the contaminated zone in the saturated zone, would be expected to significantly impact dose. Accordingly, if the precipitation /

infiltration rate was influencing dose in a meaningful way, it would have been identified as a sensitive parameter. The fact that precipitation rate is not a sensitive parameter is a clear indication that the licensees manipulation of precipitation rate in their attempt to model the groundwater elevation changes is not adequate.

The justification for this approach is explained by the licensee as follows:

The use of 3.05 m/yr is conservative because the seasonal water level fluctuation includes periods of increase and decrease, as well as a horizontal component, that are not accounted for. Full resolution of the actual water flow pattern during the seasonal fluctuations into the unsaturated zone would require detailed groundwater transport and dispersion modeling which is not justified

given the bounding / screening approach used to develop the BFM conceptual model.

The NRC staff agrees with the licensee that full resolution of the actual water pattern is not justified. However, the value for groundwater elevation should reflect a conservative value given the changes in water table level. Therefore, an alternative conceptual modeling of the Reactor Building as a single structure, which assumes that a fraction of the contamination is in contact with the water table, is warranted.

Path Forward:

a. Instead of modeling the Reactor Building as two separate conceptual models (above the 619 foot elevation, and below the 619 foot elevation), analyze the impact on the DCGLs should the Reactor Building be modeled as one conceptual model with an appropriate groundwater elevation (see RAI HYDRO-1 on Groundwater Level and Unsaturated Zone Thickness). The parameter value for the contaminated fraction below the water table should reflect the appropriate value for groundwater elevation based on this new conceptual model.

b. Analyze the impact on the insignificant radionuclide dose when using the alternative conceptual model for the Reactor Building.

PAB-7 Additional information is necessary to determine the potential dose from excavated concrete or fill material.

Basis: Section 6.12 of the LACBWR LTP describes the conceptual model for the Excavation Scenario, stating that the calculation is driven by the baseline limitation in the conceptual model that the average radionuclide concentrations in the concrete after inadvertent mixing during excavation will not exceed the surface soil DCGLs.

To derive the Excavation DCGLs, the licensee starts with the Surface Soil DCGL concentrations in pCi/g and finds the total Curies associated with the total volume of concrete in the basement. This activity is homogeneously distributed over the entire surface area of the basement. So, in essence, the intruder is assumed to excavate all the surfaces and total volume of concrete in a basement, and homogeneously mix it to obtain a concentration in pCi/g. The concrete is assumed to act like soil, which is a conservative assumption. However, this analysis dilutes, to an unrealistic extent, the source term which is allowed on the building surfaces.

Section 6.5.1.2 of the LACBWR LTP states that the residual radioactivity remaining in the backfilled basements is assumed to inadvertently mix with the mass of structural concrete removed during excavation which is consistent with the guidance in NUREG-1757, Appendix J, for addressing subsurface contamination. The NRC staff would argue that the degree of mixing assumed by the licensee in this scenario is not consistent with the guidance in NUREG-1757. Specifically, while some degree of mixing is expected, assuming that the intruder excavates the entire volume of structural concrete for any given basement is unrealistic and non-conservative. This is especially true for a larger basement like the Reactor Building.

NUREG-1757, Volume 2, Appendix J, Assessment Strategy for Buried Material, describes three acceptable approaches for determining source concentration (see Sections J.3.1, J.3.2, and J.3.3). In the first approach in Section J.3.1 (Mass Balance) the total buried source term (in pCi) is assumed to be brought to the surface and spread evenly across 360 cubic meters (m3), and then spread out to an area of 2400 m2 at a thickness of 0.15 m. For example, the DCGLb for Cs-137 in the Reactor Building is 1.05 x 108 pCi/m2 and the surface area of the Reactor Building is 511.54 m2. Therefore, the total source term would be 5.37 x 1010 pCi. If you divide this source term by the mass in 360 m3, at an assumed fill density of 1.76 g/cm3, the resulting concentration is 85 pCi/g. This value is four times greater than the maximum fill concentration calculated by the licensee of 21 pCi/g, and about 1.5 times greater than the Adjusted Soil DCGL of 55.2 pCi/g.

NUREG-1757, Volume 2, Page J-6 states that this approach should be used if the thickness of the residual radioactivity is unknown and it can be safely assumed that the volume of residual radioactivity is greater than or equal to 360 m3. The total volume of fill in the Reactor Building Basement is 1485.16 m3. However, note that since the WTB, WGTV, and Reactor Building have fill volumes of 60 m3, 305 m3, and 357 m3, the approach described in Section J.3.1 may not be appropriate.

The approach in Section J.3.2 describes a simulation assuming that the contaminants are distributed uniformly within the volume of contaminated soil while interspersing clean soil, and assuming that the soil is distributed over a surface to a depth of 0.15 m. This second approach requires that the depth of residual radioactivity be known. The approach in Section J.3.3 describes a dual simulation with two contaminated zones (one which has been excavated, and one which remains buried), each with a separate concentration. In order to find the concentration of the excavated material, the initial concentration of the buried residual radioactivity must be known. The depth (and volume) of fill is known, but the extent of mixing between the surface contamination and the fill volume is not known.

In addition, the depth of the contamination on the wall and floor surfaces is not known. Therefore, these two methods, as they are specifically described in Appendix J of NUREG-1757, Volume 2, Revision 1, will not adequately guide the licensee in calculating a source term. Accordingly, the licensee may need to evaluate different values for depth of contamination and mixing in order to determine an appropriate value for the thickness of contamination and/or initial concentration.

Section 6.14 of the LTP describes a check calculation on the final DCGLb values:

A check calculation was performed in Reference 8 to determine the maximum hypothetical concentrations of ROC in fill material after excavation. The calculation assumed that 100 percent of the residual radioactivity in the concrete was instantly released to the fill and uniformly mixed in the fill during basement concrete excavation.

Therefore, the source term would be in the fill and not in the concrete.

The calculation assumed that the residual radioactivity is uniformly distributed over basement surfaces at the BFM DCGLB

concentrations. For all ROC and all basements, the hypothetical maximum fill concentrations were less than the surface soil DCGLs.

The check calculation in the LACBWR LTP was not appropriate, in part, because the maximum fill concentrations are determined for Cs-137, Sr-90, and Co-60 assuming they are at the adjusted DCGLb values. Since the adjusted DCGLb values are based on a 5 percent Insignificant Radionuclide contribution, this calculation may not adequately take into account uncertainty in the radionuclide mixture should the insignificant contribution be higher (see RAI PAB-1 on the dose contribution from Insignificant Radionuclides). This evaluation also does not discuss potential hotspots that could be excavated. Finally, assuming complete mixing with the total volume of fill is optimistic and not appropriate for an excavation scenario in this case.

Path Forward:

a. Evaluate the impact on the excavation dose due to uncertainty in the radionuclide mixture percentages, and therefore the percent insignificant contribution.

b. Evaluate the excavation of the DCGLb concentrations with more defensible assumptions regarding the depth of contamination on the building surfaces and dilution with fill or concrete material.

c. Evaluate the impact on potential dose based on excavation of the maximum hotspot which could be allowed based on the area factor equations proposed.

PAB-8 Additional information is necessary to determine the potential dose from the Drilling Spoils Scenario.

Basis: Section 6.11.2 of the LACBWR LTP describes the Drilling Spoils Scenario. In this scenario, concrete is assumed to be brought to the surface during the installation of a well that randomly hits backfilled structural concrete. Section 6.11.2 states that the source term for the BFM Insituds scenario is the residual radioactivity remaining in concrete at the time of license termination assuming no decay or release to fill.

In order to derive the Drilling Spoils DCGL for all basements, the licensee ran the Soil Conceptual Model, which was used to derive the Industrial Use soil DCGLs, in RESRAD with a Contaminated Zone area of 0.457 m2 and a thickness of 15 cm.

The Drilling Spoils concentrations in pCi/g corresponding to 25 mrem/yr from RESRAD were multiplied by a conversion factor of pCi/m2 per pCi/g. The conversion factor finds the amount of contamination (in pCi) that is contained in the total grams of drilling spoils assuming a concentration of 1pCi/g in the drilling spoils, and divides by the surface area of the borehole. The drill is assumed to be a 12-inch borehole, and the contamination is assumed to be in the first 2.54 cm of concrete. A minimum of 3 feet of backfill is assumed to minimize mixing dilution with fill material.

In other words, the contamination source is assumed to be the activity found in the concrete diluted by the mass of the drilling spoils. However, this derivation does not

consider a scenario where a driller might drill into a hotspot existing on the basement concrete, buried piping, or equipment drains.

Path Forward:

a. Analyze the potential dose from drilling through hotspots that may be left in the basement concrete, buried piping, or equipment drains (see RAI PAB-10 on Area Factors). Include the inventory or concentration assumed, the amount of material assumed to be excavated, and the amount of dilution from mixing with overburden material that was assumed. In addition, provide details on how the assumed inventory compares to the maximum activity that will remain (i.e., the area factors proposed). Provide a basis for how it is known that the assumed inventory bounds the potential dose from the drilling spoils scenario.

PAB-9 More information is needed on the Sensitivity Analysis of the Kd Factors for some radionuclides.

Basis: Table 6-13, BFM Insitugw WGTV Uncertainty Analysis Results for Distribution Coefficients (Kd), of the LACBWR LTP lists the WGTV uncertainty analysis results for the associated distribution coefficients (Kd). Page 6-27 of the LTP states:

The predominance of negative correlation with Kd was expected since the primary dose pathway in the BFM Insitugw scenario is through the ingestion of well water and lower Kd values result in a greater percentage of radioactivity in the water phase at equilibrium. Therefore, the deterministic Kd values selected for the non-sensitive radionuclides that were included in the uncertainty analysis were also conservatively assigned the 25th percentile values from Reference 14. The 75th percentiles were assigned as indicated by the [partial rank correlation coefficient] (PRCC) results in order to follow the parameter selection process in Figure 6-7 [RESRAD Parameter Selection Flow Chart] but this will have a very minor, if any, effect on dose since the parameters were shown to be only slightly sensitive.

The implemented deterministic model for the WGTV needs clarification because the distribution coefficients for Iron-55 (Fe-55), Nickel-39 (Ni-59), Nickel-63 (Ni-63), and H-3 are set to the non-sensitive values (or 25th percentile) for the contaminated zone, but the values for the unsaturated zone (or saturated zone for Ni-63) are set to the 75th percentile values because the PRCC values were greater than 0.25 in one of the repetitions. Erosion of the cover and contaminated zone may be one explanation for positive correlations of distribution coefficients with dose, but that explanation is inconsistent with the use of higher Kd values in the deeper layers than the shallow layers. The cause of the conflicting PRCC values in the sensitivity analysis (strong negative correlation in some repetitions versus strong positive correlations in other repetitions) is not adequately explained in the LACBWR LTP.

As implemented for the WGTV, lower distribution coefficients in the contaminated zone lead to higher leach rates from the source area, thereby pushing more

radionuclides to the unsaturated and saturated zones where the larger distribution coefficients cause radionuclides to sorb to the solids rather than stay in the water phase. Therefore, the selected set of distribution coefficients leads to non-conservative assumptions for both the soil inhalation / ingestion and groundwater pathways. In addition, the LTP does not specify the treatment of parameters whose lPRCCl = 0.25 exactly.

Path Forward:

a. Provide justification that the uncertainty analysis was accurate given that it did not produce consistent results. Specifically, rerun the sensitivity analysis with more realizations to reassess the consistency of the correlations.

b. Provide the rationale for the values used in the three zones as distribution coefficients for Fe-55, Ni-59, Ni-63, and H-3 for the WGTV BFM Insitu Groundwater Model (Table 6-13) using the updated uncertainty analysis.

c. Analyze the impact on the DCGLs should the Kd values for Fe-55, Ni-59, Ni-63, and H-3 beset to the 25th percentile consistently across the contaminated zone, unsaturated zone, and saturated zone.

d. Clarify the treatment of a parameter whose lPRCCl = 0.25.

PAB-10 More information is needed regarding Area Factors and the Elevated Measurement Comparison.

Basis: Section 6.16 of the LACBWR LTP presents the Area Factors and Elevated Measurement Comparison (EMC) test to be applied in basement structures during FSS. Given that the WTB was the only Class 1 basement proposed, the discussion in Section 6.16 focuses on the WTB. The NRC staff acknowledge that the WTB is now going to be removed. However, if other basement structures are to be classified as Class 1, much of this request for information remains relevant to the elevated measurement comparison for those basement structures. Section 6.16 states that the EMC test will be performed for the WTB basement concrete survey unit (or any other Class 1 basement survey unit if there is a classification change) using Equation A-1 from NUREG-1757 which is shown in Equation 6-5 with the required site-specific modifications. Equation 6-5 from the LTP for the EMC test and corresponding Equations 6-3 and 6-4 for the Area Factors are reproduced below.

The Area Factors for the Insitugw DCGLbs are calculated using Equation 6-3 from the LACBWR LTP and captured in Table 6-27, Area Factors for WTB Insitugw Scenario Assuming FSS [In Situ Object Counting System] (ISOCS) [Field of View] (FOV) of 28m2. The title and column headings of this table indicate an area of 28 m2 for the FOV, but the text indicates an area of 7.3 m2. Section 6.16.1 of the LTP states that the DSRs and Area Factors for the assumed 7.3 m2 ISOCS FOV in the WTB are provided in Table 6-28. The reference to Table 6-28 is a typo as the reference should be to Table 6-27; in addition, the appropriate FOV area should be clarified.

The Area Factors for Excavation DCGLbs are calculated using Equation 6-4. The Excavation DCGLbs Area Factor is determined by taking the ratio of the Survey Unit Area for the WTB over the area of the ISOCS FOV.

The differences in approach between the Area Factor for the excavation scenario and the drilling spoils scenario need additional explanation given the similarities in how the DCGLbss are derived for each scenario.

With regard to the Excavation Scenario, Section 6.16.2 of the LACBWR LTP states:

The Excavation scenario is a mixing model with dose pathways that are dependent only on the total source term in the concrete and are independent of the activity distribution. The total source term for each ROC that represents 25 mrem/yr for the Excavation scenario can be calculated by multiplying the DCGLbs for the excavation scenario in Table 6-27 (pCi/m2) by the survey unit area (m2).

The reference to Table 6-27 should be corrected since this table does not reference DCGLbs. In addition, since the Excavation DCGLbs are derived by assuming the Soil DCGL concentration is in the total volume of concrete, and then dividing that activity amount by the surface area of the basement (including floor and walls), one would expect this sentence to read: The total source term for each ROC that represents 25 mrem/yr for the Excavation scenario can be calculated by multiplying the DCGLbs for the excavation scenario (pCi/m2) by the surface area of the WTB (m2).

With regard to the Drilling Spoils Area Factor, Section 6.16.3 of the LTP states:

Unlike the Insitugw and Excavation scenarios, the dose from the Insituds scenario does not systematically decrease with decreasing

size of elevated areas. Theoretically, an area on a basement floor as small as the assumed eight-inch diameter of the well borehole could be the source term in the Insituds scenario. Therefore, an EMC test will be applied to each contiguous elevated area on the WTB floor that contains activity exceeding the DCGLb (or a SOF of one considering all ROC) to account for the Insituds scenario dose from the elevated areas. The Insituds scenario does not apply to contamination on walls and therefore the drilling spoils EMC test does not apply to walls. A typical [Area Factor] (AF) calculation is not applicable to the drilling spoils scenario because the dose from drilling spoils does not decrease as a function of decreasing elevated area size. Therefore, the AF for the EMC test for the Insituds scenario will have an AF of one, which is the most conservative value that can be applied.

Similar to the Drilling Spoils Scenario, the Excavation Scenario DCGLbs are derived based on the total activity brought to the surface since the assumed clean mixing volume is kept constant within a particular basement. Specifically, the activity is assumed to be in the volume / mass of concrete, which is constant for a particular basement in the derivation of the DCGLbs. Similarly, the volume / mass of fill material is kept constant per basement in the check calculation performed in Section 6.14 of the LACBWR LTP. This is similar to the concept used for the Drilling Spoils Scenario, whereby the concentration of the drilling spoils once brought to the surface is limited to be the Soil DCGL, and the pCi/m2 is derived from that concentration and the mass of the drilling spoils.

The dose from the Excavation Scenario may not systematically decrease with decreasing size of the elevated area. This will depend on assumptions regarding the size of the hotspot(s), the volume of material excavated, and dilution with clean fill. It is the NRC staffs position that the Area Factor for the Excavation Scenario should be one instead of having it be based on Equation 6-4 from the LACBWR LTP.

In addition, the NRC staff does not agree that a driller cannot conceivably drill into the remaining basement walls at the La Crosse site. Drilling technology would conceivably allow an intruder to drill into these walls at an angle. Therefore, the drilling spoils EMC test should also apply to the remaining basement walls (see RAI PAB-13 on classification and survey coverage for basement structures).

The EMC test will be analyzed using Equation 6-5 from the LTP, which is:

With regard to the Maximum Elevated Floor Core term in Equation 6-5, Section 6.16.3 of the LACBWR LTP states that a concrete core sample will be collected at the location within the bounded area that exhibits the maximum reading and the activity quantified. If the total activity in the core, including all core slices with depth, exceeds the DCGLb, the EMC test will be performed using the DCGLbs from Table 6-27 applicable to the WTB BFM Insituds scenario. The reference to Table 6-27 should be corrected since this table does not reference DCGLbs.

In addition, the way in which the total activity in the concrete core will be calculated from survey data needs further explanation. Does the licensee intend to compare concentration values or total activity values? If the licensee intends to average over the entire depth of the core, that is not appropriate for the drilling spoils scenario since only a portion of concrete, 2.54 cm thick with an area of 730 cm2 at a certain concentration, needs to be brought to the surface. Furthermore, it is unclear from the LACBWR LTP what, if any, Area Factors are to be used for Buried Piping. The LTP should stipulate how hotspots will be handled during FSS of buried piping.

Page 5-23 of the LACBWR LTP states that if more than one elevated area is identified during FSS, then the same calculation will be performed for eachThe fractions associated with each elevated area are summed and the result must be less than unity for the survey unit to pass. It unclear whether the licensee intends to apply the unity rule to the sum of all hotspots within each survey unit.

Section 8.5.2 of NUREG-1575, Revision 1, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), states that if there is more than one elevated area, a separate term should be included for each. It is unclear from the LACBWR LTP whether the licensee considers all elevated areas together and assumes that the receptor is affected by them all. If the licensee intends to comply with MARSSIM Section 8.5.2, the sum of all the hotspots (i.e., elevated areas) needs to comply with the unity rule. Deviations from this method would need to be justified.

For comparison, the discussion in Equation 5-15 (see Page 5-62) of the LACBWR LTP (which is intended for areas other than the basements) and also Equation 5-13 from the Zion Nuclear Power Station LTP, clearly includes an n factor and indicates that if the individual elevated areas pass, then they are combined and evaluated under the unity rule. As such, Equation 6-5 from the LACBWR LTP needs clarification as it is not readily evident that the MARSSIM guidance is being followed.

Path Forward:

a. Choose one of the following (i or ii):

i. Provide a more detailed technical basis for using Equation 6-3 through Equation 6-5,, addressing the issues raised in the Basis of this RAI.

ii. As an alternative to using Equation 6-3 through Equation 6-5, provide a table of Area Factors to be used with the DCGLb (similar to what was supplied in Table 5-7, Area Factors for Soils, for Soil) for each of the basement

categories to remain (i.e., Reactor Basement, WGTV, and Remaining Structures). The table should indicate either the Area Factor or the hotspot concentration that will be considered acceptable for various area sizes of contamination. Revise Equation 5-5 and Equation 6-5 accordingly to incorporate the Area Factors in the requested table and the DCGLb, similar to what is presented in Equation 5-4 for Soil.

b. Provide a more detailed description of how the Maximum Elevated Floor Core term in Equation 6-5 is determined from survey and sampling data, assuming it would still be used after responding to this RAI.

c. Provide a table of Area Factors to be used with Piping DCGLs.

d. Provide a detailed example of how survey units with more than one hotspot will be evaluated for the Elevated Measurement Comparison.

PAB-11 More information is needed for Equation 5-3 on Compliance.

Basis: Page 5-10 and 5-11 of the LACBWR LTP states that the mean values from FSS will include the results of judgmental samples based on an area weighted average approach. Detailed instructions for calculating the compliance dose using Equation 5-3 will be provided in a procedure which will be submitted to NRC for information. The level of detail provided in the LTP for Equation 5-3 is not sufficient.

For example, the area weighted average approach is not described. In addition, the maximum basement term is described as a survey unit dose, but the scenario in which there is more than one survey unit per basement is not described.

Path Forward:

a. The details regarding Equation 5-3 should be provided to the NRC for approval, not just information. In addition, provide a detailed example showing how the factors for Equation 5-3 will be determined from survey measurements.

PAB-12 More information is needed on the conceptual model for the Waste Gas Tank Vault and Remaining Structures.

Basis: Table 6-8, Deterministic Geometry RESRAD Parameters Used in the Uncertainty Analysis for the Five BFM Insitugw Configurations, of the LACBWR LTP shows that the thickness of the contaminated zone is set to 2.54 cm for the WGTV, WTB, and Remaining Structures. The area of the contaminated zone is set to be the total surface area of the basements including floors, walls, and ceiling (where applicable). The 2.54 cm thick contaminated zone is assumed to be entirely at an elevation equivalent to the floor elevation. While this conceptual model does artificially assume a lower elevation for some of the wall and/or ceiling contamination for the WGTV and WTB, it also artificially alters the geometry and extent of the contaminated zone for these structures in a way that makes it difficult to tell if it is an overall conservative assumption.

In reality, there will be some contamination mixing with fill near the floor, wall and/or ceiling surfaces at various elevations. The extent of mixing into the fill is uncertain.

The licensee performed a sensitivity analysis on the extent of contamination mixing with the fill in which the mixing distance ranged from 2.54 cm to the full height or length of the fill for the both the Reactor Building Above 619 foot and the WGTV.

This sensitivity analysis assumed all the surface area was in a horizontal position for the WGTV, whereas for the Reactor Building Above 619 foot the wall contamination was modeled vertically. The sensitivity analysis showed that for the Reactor Building Above 619 foot, the full mixing distance out from the wall of 9 meters yielded higher doses, while for the WGTV, the smaller mixing distance of 2.54 cm yielded higher doses. The NRC staff notes that contamination on the walls of other structures may be more likely to yield sensitivity results similar to the Reactor Building above 619 foot. Therefore, the assumptions for the orientation of the contamination in the sensitivity analysis should be reviewed to ensure that the related assumptions regarding extent of contamination are appropriate.

For the Remaining Structures Conceptual Model, one conceptual model is applied to represent a variety of the Remaining Structures at the LACBWR site (i.e., Piping and Ventilation Tunnels, Reactor / Generator Plant, Chimney Slab, Turbine Sump 1C, and Turbine Pit at G5). The elevation of the floor of the WGTV is 621 foot and the WTB is 630 foot, while the Remaining Structures elevation varies from 639 foot to 618 foot. The licensee applies an Area Weighted Average Floor Elevation of the Remaining Structures for the Remaining Structures Conceptual Model. This approach conceptually moves some of the contamination further away from the Saturated Zone, especially any contamination remaining in the Turbine Sump and Turbine Pit, which are at the 618 foot elevation. The adequacy of these assumptions should be reviewed, especially given the uncertainty in the groundwater elevation (see RAI HYDRO-1 on Groundwater Level and Unsaturated Zone Thickness).

Path Forward:

a. Provide additional justification regarding the geometry assumptions for the contaminated zone for the WGTV and Remaining Structures. Include additional discussion on the related assumptions for the contaminated zone thickness and contaminated zone area. Analyze the impact on dose and sensitivity of mixing depth when modeling the contamination on the walls vertically, as opposed to assuming all contamination is at the floor elevation.

b. Provide an additional basis for why the single BFM conceptual model (and an area weighted average floor elevation) appropriately represents the variety of Remaining Structures. Analyze the impact on the DCGLs should these structures be modeled separately with the contamination at separate elevations.

c. Analyze the impact on the DCGLs should a higher groundwater elevation be applied in the conceptual models for the WGTV and Remaining Structures.

PAB-13 Additional information is needed regarding classification and survey coverage of basement structures.

Basis: Table 2-2, LACBWR Structural Survey Units, of the LACBWR LTP lists several survey areas which were initially classified as Class 1: Reactor Building, Waste Treatment Building, Ventilation Stack, WGTV, Turbine Building, Turbine Office Building, Piping Tunnel, etc. However, Table 5-8, Number of ISOCS Measurements per FSS Unit Based on Areal Coverage, of the LACBWR LTP lists the WTB Basement as the only remaining Class 1 survey unit for FSS purposes.

MARSSIM Section 2.2 provides one example of a MARSSIM Class 1 area as site areas previously subjected to remedial actions and goes on to add a footnote that remediated areas are identified as Class 1 areas because the remediation process often results in less than 100 percent removal of the contamination, even though the goal of remediation is to comply with regulatory standards and protect human health and the environment. The contamination that remains on the site after remediation is often associated with relatively small areas with elevated levels of residual radioactivity. This results in a non-uniform distribution of the radionuclide and a Class 1 classification. If an area is expected to have no potential to exceed the DCGLw and was remediated to demonstrate the residual radioactivity is as low as reasonably achievable (ALARA), the remediated area might be classified as Class 2 for the final status survey.

Page 2-5 of MARSSIM further states that to justify changing the classification from Class 1 to 2, there should be measurement data that provides a high degree of confidence that no individual measurement would exceed the DCGLw. To date, the licensee has not yet provided characterization data showing with a high degree of confidence that no individual measurement would exceed the DCGLb.

Due to the initial classification of the Reactor Building, WTB, Ventilation Stack, WGTV, and Turbine Building as Class 1 impacted areas, and the fact that these areas are being subjected to remedial actions, according to MARSSIM Section 2.2, the survey units listed in Table 5-8 of the LACBWR LTP should remain as Class 1 until additional characterization results are provided for review and approval to sufficiently justify changing the classification of these areas from Class 1 to Class 2.

The NRC staff acknowledge the licensees recent decision to remove the WTB in its entirety. However, given that other structures will now likely be Class 1 for the purposes of FSS, the following discussion is still relevant. Regarding the WTB surveys, Section 5.5.2 of the LACBWR LTP states that the basement structure FSS units will be comprised of the combined wall and floor surfaces of each remaining building basement. The WTB is a single survey unit, and Table 5-8 of the LTP specifies that the survey unit will receive a 100 percent scan. The surface area of the floors of the WTB are about 37 m2, while the surface area of the walls are about 65 m2. Therefore, the entire surface area of the WTB in Table 5-8 is listed as 101.9 m2, which includes both the floors and the walls.

However, it is unclear whether the WTB walls will receive a 100 percent scan due to other statements made in the LACBWR LTP. For example, Page 5-2 of the LTP states that in the WTB, the only portion of the structure that will remain is the 630 and 635 foot floor, sump and concrete footers below the 636 foot elevation. It is unclear which parts of this description of the WTB correspond to walls and which parts correspond to floors. Furthermore, Page 5-22 of the LTP states that for the FSS of basement structures, the EMC is only applicable to Class 1 survey units when an elevated area is identified by systematic or biased / judgmental ISOCS measurements. As stated in Section 5.5.2.1, this is expected to apply only to the 630 and 635 foot elevations of the WTB.

Page 5-23 states that elevated areas will only be investigated on the floor surfaces:

As discussed in Section 5.4.3, small elevated areas (smaller area than the ISOCS FOV), if present, are identified and investigated during the post-demolition survey of the floor only. The areal extent of identified elevated areas on the floor will be bounded and a concrete core sample collected at the location within the bounded area that exhibits the highest activity. If the activity in the core (summing activity in all core slices) exceeds the DCGLb values in Table 5-3 [Basement DCGLs (pCi/m2)], then an EMC test will be performed using the DCGLbs for the Drilling Spoils scenario from Table 5-4 [DCGLs for Individual BFM Scenarios (DCGLbs)] with an Area Factor of one. The Drilling Spoils scenario, and corresponding EMC test, does not apply to walls.

The reference to Section 5.4.3 of the LTP in this paragraph seems to be a typo.

Instead, this should refer to Section 5.4.4 of the LTP, which states:

Following demolition, after all debris is removed and the floors cleaned, an additional scan survey will be performed on 100 percent of the Class 1 basement floors, again using hand-held beta-gamma instrumentation as presented in Table 5-13 [Typical FSS Survey Instrumentation] in typical scanning and measurement modes. Any identified elevated areas that could potentially exceed the DCGLb will be identified and bounded. A concrete core sample will be collected at the location that exhibits the highest activity within any bounded area. If the activity in the core (summing activity in all core slices) exceeds the DCGLb values in Table 5-3, an EMC test will be performed using the DCGLbs for the Drilling Spoils scenario from Table 5-4 (see Section 5.5.3).

Given that Section 5.4.4 of the LACBWR LTP uses the phrase 100 percent of Class 1 basement floors, it is unclear whether the wall surface areas of Class 1 survey units will receive 100 percent scan coverage as well. In addition, the licensee has not adequately justified why the EMC test will only apply to the floors and not also the walls of Class 1 areas (see RAI PAB-10 on Area Factors and the Elevated Measurement Comparison). It is also unclear whether the licensee intends to apply

the EMC test as it is described in Equation 5-5 of the LTP, or only the Drilling Spoils term from Equation 5-5. The statement an EMC test will be performed using the DCGLbs for the Drilling Spoils scenario seems to disagree with the statement on Page 5-23 of the LTP that for the FSS of Class 1 basement structures, the EMC test will be performed using Equation A-1 from NUREG-1757, as shown in Equation 5-5.

Path Forward:

a. List the areas identified in Table 5-8 of the LACBWR LTP as Class 1 survey areas until additional characterization results are provided to the NRC staff for review and approval to sufficiently justify changing the classification of these areas to Class 2 survey areas. Adjust the minimum areal coverage and number of ISOCS measurements for these survey units accordingly. If additional characterization results are available showing a high degree of confidence that no individual measurement would exceed the DCGLb, provide these to the NRC.

b. Clarify LACBWRs current FSS plan for Class 1 basement structures, which areas are walls versus floors, and their various elevations. In addition, clarify whether or not there are multiple survey units within each basement structure. If possible, provide a diagram for each basement structure or identify where this information may be available in the materials which have already been submitted to the NRC. Finally, clarify which surfaces will be receive 100 percent scan coverage and which surfaces will be subject to the EMC test.

c. Commit to 100 percent scan coverage and systematic sampling of all Class 1 surface areas, and to appropriately investigate and/or remediate hotspots according to the MARSSIM guidance for Class 1 surfaces (regardless of whether the surfaces are walls, floors, or ceilings).

d. With regard to the EMC test for Class 1 structures, clarify what is meant by an EMC test will be performed using the DCGLbs for the Drilling Spoils scenario.

Reconcile this with the later statement that for the FSS of Class 1 basement structures, the EMC test will be performed using Equation A-1 from NUREG-1757, as shown in Equation 5-5.

e. Clarify whether operational DCGLs will be applied in the design of Final Status Surveys for basement structures.

PAB-14 Clarification is needed on whether the underground tanks should be categorized as Buried Pipes.

Basis: Page 5-44 of the LACBWR LTP states that designated sections of buried piping will be remediated in place and undergo FSS. The tanks associated with the Administration Building Sanitation system and leach field will also be addressed during FSS as buried pipe. Compliance with the DCGL values, as presented in Table 5-6, DCGLs for Buried Piping (dpm/100cm2), will be primarily demonstrated by measurements of total surface contamination and by the collection of sediment samples when available. Page 6-5 of the LTP states that the Administrative

Building Sanitary System includes three small tanks that are also designated as buried pipe for the purpose of dose assessment and FSS. Table 6-2, Buried Piping to Remain in LACBWR End State, of the LTP lists the three tanks as a 2500 gallon septic tank, a 750 gallon dosing tank, and a 94 gallon distribution tank. Additional information is needed to determine the acceptability of treating the underground tanks as buried piping for the purposes of performing FSS.

Path Forward:

a. Provide additional justification that the conceptual model for developing the buried piping DCGLs is appropriate for these tanks. Include a description of the location (survey unit), geometry (diameter and height), and elevation of each tank, the potential for contamination given historical use, and the end state (e.g., whether they will be backfilled or grouted).

b. Provide a technical basis for why the mixture fractions assumed in deriving the buried piping DCGLs apply to these underground tanks.

c. Provide additional information on the surveys that will be conducted on the tanks; clarify the classification of the tanks; and include a description of what is meant by compliancewill be primarily demonstrated by measurements of the total surface. In addition, include a specific description of how it is determined whether sediment samples are available. For example, is availability determined by accessibility, the mass of sediment existing, or some combination of both?

Health Physics Topics (Identified by HP)

HP-1 ISOCS FOV Uncertainties for Remedial Action Support Surveys (RASS) and FSS Measurements Basis: Section 5.4 and Section 5.5 of the LACBWR LTP address the use of ISOCS measurements in a discussion of RASS and FSS. NUREG-1700, Section 5, Final Radiation Survey Plan, calls for a demonstration that the insitu sample measurements with field instruments, and the associated survey methods, have adequate sensitivity as part of the FSS design. Accordingly, the LACBWR LTP should address the effects of ISOCS FOVs that are smaller than the source area.

The additional information should specifically address the measurement of inaccessible areas and the licensees method for demonstrating that they are capable of 100 percent scan coverage for the Class 1 survey units.

Path Forward:

a. Provide additional information on how ISOCS measurements compare to conventional measurements. Identify how ISOCS FOVs provide for 100 percent scan coverage of Class 1 survey units and how inaccessible areas are addressed. Discuss the possibility of utilizing combinations of insitu and conventional measurements as part of the overall FSS design.

b. Consider reevaluating characterization data and survey unit classifications before conducting FSS measurements. MARSSIM Section 8.5.2, Interpretation of Statistical Test Results, states that a separate term should be included for each elevated area when determining whether the total dose is within the release criterion of a survey unit. Accordingly, clarify whether all elevated areas are considered together or independently in a Class 1 survey unit.

HP-2 Judgmental Sampling, Laboratory Analyses, and Surveyor / Instrument Efficiencies Basis: Section 5.5 of the LACBWR LTP includes a statement on judgmental sampling and basing judgmental measurements on a higher potential for elevated concentrations of residual concentration based on characterization, contamination verification survey (CVS) results, or professional judgment.

Section 5.3 of the LTP discusses laboratory instrument methods and sensitivities, including the use of offsite vendor laboratories. Instrument calibrations are also dependent on source geometry, specifically 2 or 4 geometry. In addition, surveyor efficiency is an important component of calculating scan MDCs as noted in NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, and NUREG/CR-6364, Human Performance in Radiological Survey Scanning. Scan MDC values based on a surveyor efficiency of between 0.5 and 0.75 are generally acceptable; use of a surveyor efficiency above 0.7 should have a documented basis.

Path Forward:

a. Provide more detail on specific triggers for judgmental sampling. Identify the minimum criteria, if any, that surveyors use to select judgmental sample locations. Provide details on how judgmental samples are based on an area weighted average and do not dilute dose by taking judgmental samples from clean areas rather than areas of higher potential.

b. Clarify whether the onsite and offsite laboratories used by LACBWR participate in an independent inter-laboratory comparison program.

c. Identify the basis for false positive analytical results and any follow-up measures used to investigate those results. Provide a basis for eliminating any data points identified as a false positive result. Identify whether there are minimum trigger levels based on analytical results that warrant sample recount.

d. Provide additional information on survey instrument calibration calculations and specify whether 2 or 4 source geometry is utilized.

e. Describe the technical basis for surveyor efficiencies which exceed 0.7.

HP-3 Background Reference Areas and Materials Basis: Section 2.3.1 of the LACBWR LTP discusses a background study which concludes that no suitable offsite background radiation reference area was identified for the La Crosse site, resulting in the selection of an onsite area to assess background radioactivity. This section also references the use of Zion Nuclear Power Station data for the global fallout of Cs-137 in soils at LACBWR. Section 5.2 of the LTP identifies the DCGLs as the level of radioactivity above background levels that could result in a dose of 25 mrem/yr to the average member of the critical group.

Therefore, an accurate assessment of background radioactivity is important for determining compliance with the DCGLs. MARSSIM Section 4.5 discusses background reference areas, and states that reference areas and materials should not be part of the survey unit being evaluated. In addition, background radiation levels can vary by a factor of five or more within a few acres.

Path Forward:

a. Provide a basis for not selecting an offsite background radiation reference area at the La Crosse site and not using LACBWR specific historical Cs-137 global fallout data. Assess whether existing historical data from the annual radiological environmental monitoring reports or semiannual effluent release reports can be used to more appropriately document background and Cs-137 fallout.

b. With the exception of basement structures, which will not subtract background radiation from measurements, provide additional information regarding the basis for the background levels used with the applicable DCGLs to demonstrate compliance with the release criteria of each survey unit.

HP-4 Sub-Foundation and Inaccessible Area Sampling Basis: Section 5.3.4.4 of the LACBWR LTP defines areas where surveys are deferred for soils under structures, soils under concrete or asphalt coverings, inaccessible concrete basements surfaces, and interiors of buried pipe (i.e.,

inaccessible areas). This section also states that continuing characterization plans and reports for these areas will be provided to the NRC for information.

NUREG-1757, Volume 2, Appendix A, Implementing the MARSSIM Approach for Conducting Final Radiological Surveys," provides guidance on characterization of subsurface soil. Section 5.7.1.8 of the LACBWR LTP addresses buried piping, but does not discuss grouting of pipes, how underground tanks will be assessed, or continuing assessment of the areas around and under structures.

Path Forward:

a. Additional information is needed to address the scope and adequacy of deferred surveys for currently inaccessible areas. Specifically, provide additional information on the GeoProbe technology that will be used for soil surveys and

sampling in these areas. In addition, identify the methods and criteria that will be used to survey the interior surfaces of buried piping that will remain onsite.

b. Provide additional information on grouting of pipes and subsequent dose assessment of any grouted piping.

c. Provide a technical support document for the use of ISOCS measurements for buried piping and underground tanks at LACBWR.

d. Submittal for NRC review and approval of continuing characterization plans and reports is expected in lieu of being provided to the NRC for information.

HP-5 Confirmatory Surveys, Split Sampling, and Side by Side Measurements Basis: Section 5.9.6 of the LACBWR LTP briefly discusses confirmatory measurements, but does not address split sampling and/or analysis of side by side instrument comparisons with the NRC. NUREG-1757, Volume 2, Chapter 4, Facility Radiation Surveys, describes the performance of confirmatory surveys, split sampling, and side by side measurements as part of FSS design. NUREG-1575, Chapter 7, Sampling and Preparation for Laboratory Measurements, discusses collocated samples, field duplicate samples, and laboratory duplicate samples.

Path Forward:

a. Provide additional information that addresses split sampling and analysis, as well as the performance of side by side instrument comparisons with the NRC, both during and after FSS. Split sampling and side by side measurement locations need to be identified prior to the performance of decommissioning activities that may compromise the ability to replicate the samples or measurements. The number of samples and measurements will be directed by the NRC.

b. In accordance with NUREG-1575, Chapter 7, provide additional information on the criteria used to identify the location and number of collocated samples, field duplicate samples, and laboratory duplicate samples.

Hydrogeology Topics (Identified by HYDRO)

HYDRO-1 Groundwater Level and Unsaturated Zone Thickness Provide a basis for selecting a water table at 619 foot mean sea level (MSL) for calculations in RESRAD for the LACBWR site. The water table elevation is used to determine the unsaturated zone thickness, which in turn is used to set the distance between contaminated structures and the saturated zone.

River stage levels, which are highly correlated with water table levels across the La Crosse site, exhibit a high season from April through September and a low season from October through March. Measurements from onsite wells during the low

season appear to be in the 620 to 624 foot MSL range, whereas those in the high season are in the 624 to 630 foot MSL range.

Alternatively, demonstrate that the use of a water table at 619 foot MSL for RESRAD calculations is conservative compared to use of a water table that more closely reflects onsite measurements of water table elevation. RESRAD inputs for unsaturated zone thickness, mixing distances used for setting concentration distributions in the building infill, and contaminated fraction below the water table would be expected to change depending on the water table elevation. Accordingly, analyze the sensitivity of the BFM and soil conceptual models to a higher groundwater elevation assumption.

HYDRO-2 Embankment Design and Design Criteria Provide information on the design and design criteria for the embankment and rip-rap that protects the hydraulic fill used to elevate the site to 639 foot MSL.

The site elevation is above the recorded river stage levels; however, the 1965 recorded flood stage and the estimated 100-year flood stage are within 1.8 feet of the site elevation. In addition, the upstream and adjacent configuration of the Mississippi River indicates that the possibility of cut-bank erosion along the site cannot be discounted, especially during flood events near the site. Finally, there is no information provided on the capability of the embankment and rip-rap to withstand cut-bank erosion during high river flow periods over the 1000-year performance period of the embankment, or the 200-year design period.

HYDRO-3 Contamination of Groundwater from a 1983 Spill Provide additional information on the potential spill incident in 1983 that is briefly described on Page 2-36 of the LACBWR LTP. Any additional information will enable the staff to assess the LTP with regard to the acceptance criteria in Section B.2 of NUREG-1700 [10 CFR 50.82(a)(9)(ii)(A)], which recommends sampling strategies for collecting hydrogeologic information in accordance with the guidance in Section F.4 and Section F.5 of NUREG-1757. Specifically, the licensee should address:

a. Are there indications that other radionuclides, besides those listed in Table 2-17, 1983 Groundwater Analysis from Temporary Well-Point South of Turbine Building, of the LACBWR LTP, may have been involved in the spill and subsequently entered the groundwater?

b. What indications are there that the spatial extent and magnitude of contamination of the groundwater related to the spill is small, and thus not important for site characterization?

c. Is there any information to suggest that the well-point was located in the zone of peak concentration of the plume?

A temporary well-point was sampled once and then later abandoned. Table 2-17 lists analytical results for Manganese-54 (Mn-54), Cobalt-57 (Co-57), Co-60, Niobium-95 (Nb-95), Cs-137, and Cerium-141 (Ce-141). When considered individually, only the Co-60 result was above (by a factor of five) the derived concentration corresponding to the primary drinking water standard for beta emitters.

After approximately fifteen years, the Co-60 levels would have dropped below the standard. In terms of monitoring, there were no other wells downgradient of this location, including the new well pairs that were later installed. The MW-203A/B monitoring well pair were later installed upgradient of the temporary well-point location. Please provide additional information to describe how the current wells are adequate to monitor the potential contamination from the 1983 spill.

HYDRO-4 Basement Backfill Description and Surveys Clarify the description of basement backfill to be used at the LACBWR site.

Properties are used in the RESRAD modeling that infer the basement backfill will be identical to the onsite sediments. For the description of backfill, constraints are needed in terms of textural and hydrologic properties and geochemistry. Is there a supporting document that provided a description of the basement backfill?

Section 5.1.7.6 of the LACBWR LTP discusses excavating buried structures and using that soil to backfill the excavated feature. But the backfill for basements is only described in the LTP as a radiological assessment is performed prior to introducing offsite material to ZSRP for use as backfill in a basement, or for any other use from a barrow pit, landfill, or other location. The radiological assessment consists of gamma scans and material sampling. Gamma scans are performed insitu, or by package (using a hand-held instrument or through the use of a truck monitor). Material samples are analyzed by gamma spectroscopy. Please clarify whether offsite backfill materials, including soil and sand, that will be used to fill the basements are subject to a survey plan, and describe that plan.

HYDRO-5 River Sediment Contamination in Outfall Area Provide a description of the contamination in the river sediments in the outfall area of the Circulating Water Discharge Pipe, clarify the areas categorization as impacted or non-impacted, and provide the basis for that treatment. If categorized as an impacted area, provide a discussion of how the outfall area will be treated in the process used to meet the requirements for terminating the LACBWR license.

The outfall area is outside the site boundary and is not considered an impacted area in the LACBWR LTP. The available information from the Initial Site Characterization Survey (see Reference LAC-TR-138, dated 1995, updated 2009; Page 19) and the annual Radiological Environmental Monitoring Reports indicate the possibility of contamination remaining in the river sediments in the outfall area. Results from the 1994 scuba diver sampling and the 2014 biannual sampling are on par with the biannual sampling results from the operating period, all of which are in the 2 to 4

pCi/g range for Cs-137 (except for the 21 pCi/g result from 1985); the annual reports reviewed only covered the 1985-1989, and 1999-2016 periods.

Based on the description of the sampling track by scuba divers in 1994 and the sampling location at the outfall in the annual reports, it is not clear that the peak sediment contamination was found. The NRC staff notes that the plume pathway, sediment deposition and erosion, and possible past dredging in the area may have imparted a level of uncertainty in the pattern and presence of contamination.

Specifically, decay alone is not consistent with the temporal pattern of radionuclide contamination in the sediments in the outfall area, thus possibly supporting the influence of other processes affecting the levels of contamination found. For example, the uptick in Cs-137 results in 2014 and 2015 may have followed a period of erosion uncovering previously buried contamination.

Depending on how the outfall area is treated in the FSS, additional information may be needed. Provide clarification of the biannual sediment sampling location in relation to the sample track of the 1994 scuba divers. Does the biannual sample location at the outfall correspond to the 1994 sample track that is described as starting 60 feet from the edge of the Mississippi River? The ambiguity may be related to the diffuser ramp that extends into the river, and the likelihood that sediments are not deposited on the ramp. Further clarification is also needed for the 1994 sample track location extending into the river, especially as compared to the general pattern of a plume discharging from the pipe.

Environmental Topics (Identified by ENV)

The purpose of these RAIs is to obtain information and data for the NRC to fulfill its responsibilities under the National Environmental Policy Act of 1969 (NEPA), as implemented by 10 CFR Part 51, Environmental Protection Regulations for Domestic Licensing and Related Regulatory Functions. The requirements for the contents of an Environmental Assessment are found in 10 CFR 51.53, Postconstruction Environmental Reports.

ENV-1 Status of Decommissioning Activities Update Chapter 3 of the LACBWR LTP, including Table 3-1, Status of Major LACBWR Systems and Components, to provide the plans for and status of decommissioning efforts as of August 2017.

ENV-2 Summary of Environmental Impacts In Chapter 8 of the LACBWR LTP, Table 8-1, Summary of the Environmental Impacts from Decommissioning Nuclear Power Facilities, includes references to subsections on independent spent fuel storage installation (ISFSI) construction (Section 8.6.3.18), Vertical Concrete Cask (VCC) construction (Section 8.6.3.19),

and rail line upgrades (Section 8.6.3.20). Those sections do not appear in Chapter 8 of the LTP, Supplement to the Environmental Report. Accordingly, update Table 8-1 or provide the missing sections for Chapter 8 of the LTP. In addition,

please update Table 8-1 to reflect any changes to the overall decommissioning plans and activities as of August 2017.

ENV-3 Land Use Section 8.6.3.2 of the LACBWR LTP states that appropriate isolation and control measures will be instituted to prevent the spread of contamination from the LSE

[LACBWR Site Enclosure] area to the adjacent LACBWR property as well as offsite areas. Explain what isolation and control measures will be used to prevent the spread of contamination. Provide enough information to understand how the measures will accomplish the required function.

ENV-4 Surface Water Section 8.6.3.4 of the LACBWR LTP states that storm water runoff from active impacted areas undergoing remediation or demolition will be monitored and controlled, if necessary. Explain how this runoff will be monitored and controlled and how the licensee will determine if it is necessary.

In addition, provide the surface water monitoring requirements established by the Wisconsin Department of Natural Resources (WDNR) permit. Explain what parameters need to be monitored as required by the permit.

Finally, describe the best management practices (BMPs) that will be implemented to reduce impacts to the Mississippi River and the Thief Slough tributary.

ENV-5 Air Quality Please provide the Clean Air Act, Section 107, National Ambient Air Quality Standards attainment status of Vernon County per 40 CFR Part 81, Designation of Areas for Air Quality Planning Purposes.

ENV-6 In-Water Activities Section 8.6.3.2 of the LACBWR LTP states that public access to the boat launch at the southern end of the site will be minimally impacted during decommissioning by rail and truck traffic entering or leaving the site. However, Section 8.6.3.2 does not specify what minimal impacts would occur.

Accordingly, describe any activities that would occur at or near the boat launch, the Thief Slough tributary, or the Mississippi River that would directly or indirectly support decommissioning or license termination of the LACBWR site. Explain whether the associated rail and truck traffic would impact the boat launch on the days it is most often used (e.g., if the boat launch is most often used on weekends, would rail and truck traffic impact it or would rail and truck traffic be less on weekends?).

ENV-7 Runoff and Erosion Containment In addition to the isolation and control measures mentioned in Section 8.6.3.2 of the LACBWR LTP, Section 8.6.3.6 of the LTP states that protection of the nearby Refuge wetlands and habitats is, and will continue to be a priority when planning for site dismantling or waste management operations. Clarify what isolation, control, and other protective measures will be used to minimize erosion and degradation of the nearby ecological habitats, such as the Mississippi River and the Upper Mississippi River National Wildlife and Fish Refuge.

ENV-8 Impingement and Entrainment Section 8.6.3.3 of the LACBWR LTP states that the facility withdraws Mississippi River water to supply the intact portion of the Low Pressure and High Pressure service water systems. However, Section 8.6.3.3 of the LTP does not state the withdrawal rates, which can lead to impingement and entrainment of aquatic organisms. Describe the withdrawal rates to supply the Low Pressure and High Pressure service water systems, as well as any operational or engineering features that minimize impacts from impingement and entrainment.

ENV-9 Impingement and Entrainment Reference Provide a copy of Reference 19 in Chapter 8 of the LTP titled, Impingement and Entrainment of Fishes at Dairyland Power Cooperatives Genoa Site.

ENV-10 Bird and Bat Collisions Section 8.6.3.8 of the LACBWR LTP describes potential impacts to State and Federally protected species, including birds, as a result of activities associated with the license termination plan and decommissioning. However, Section 8.6.3.8 of the LTP does not describe the likelihood of bird collisions with decommissioning equipment and intact structures and buildings. Bird collisions may result in injury or mortality. In addition, artificial night lighting can increase the likelihood of such collisions. For example, migratory songbirds are most likely to collide with artificially lighted structures or cranes because of their propensity to migrate at night, their low flight altitudes, and their tendency to be trapped and disoriented by artificial light.

Although not mentioned in Chapter 8 of the LACBWR LTP, U.S. Fish and Wildlifes Information for Planning and Conservation database indicates that bats are known to occur in the decommissioning project area. Therefore, in order for the NRC staff to evaluate potential impacts to birds and bats as a result of the proposed license termination plan, please provide the following:

a. Any data, recorded observations, or studies related to bird and bat collisions at the LACBWR site. If available, please provide the date, time, number of individuals, species, and impact to each individual (e.g., death or injury) for each recorded bird and bat collision.

b. Describe whether artificial lighting would be used at night during decommissioning activities at the LACBWR site.

c. Describe whether any best management practices, such as light source shielding and appropriate directional lighting, would be used to mitigate impacts associated with artificial nighttime illumination and potential bird and bat collisions.

ENV-11 Peregrine Falcon Restoration Program The Peregrine Falcon is listed as a State-Endangered Species in Wisconsin.

According to the Dairyland Power Cooperative (DPC) website (http://www.dairylandpower.com/environment/falcon_program.php), DPC has a Peregrine Falcon Restoration Program at its facilities. Please provide details on that program as it relates to the LACBWR / Genoa No. 3 site. Describe efforts taken, if any, to ensure that decommissioning activities at LACBWR do not adversely affect onsite peregrine falcons, if applicable.

ENV-12 General Characterization of Species and Habitats on Site In Sections 8.6.3.6 - 8.6.3.8 of the LACBWR LTP, descriptions of the aquatic, terrestrial, and threatened and endangered species that may occur at the LACBWR site are given. In addition, Section 8.6.3.7 and Section 8.6.3.8 of the LTP state that no protected species occur at the LACBWR site. Chapter 8 of the LTP, however, does not describe the studies that provide a basis for determining whether State or Federally protected species occur or do not occur on site.

Provide the basis for the licensees determination that no protected species occur at the LACBWR site. In addition, please provide a copy of any field studies that have been conducted to characterize State and Federally protected species and habitats at or within the vicinity of LACBWR.

ENV-13 Impacts to Terrestrial and Aquatic Ecology Beyond the Site Boundary In Section 8.6.3.6 and Section 8.6.3.7 of the LACBWR LTP, the licensee states that the potential impacts to aquatic and terrestrial ecology beyond the site boundary have been evaluated and are considered to be Small. Please describe the activities that could cause direct or indirect offsite impacts to ecological resources.

ENV-14 Radiological Health - Worker Dose Section 3.4.2 of the LACBWR LTP provides projections for occupational radiation exposure from activities associated with decommissioning. The projections are based on RS-TD-313196-007, Radiation Exposure Projections for LACBWR Decommissioning, which was submitted to the NRC in November 2015. Provide any updates to these radiation exposure projections; if available, provide data from 2016 for the decommissioning activities and spent fuel management activities.

ENV-15 Socioeconomics Approximately how many permanent workers were at the LACBWR facility during operation of the nuclear power reactor (until approximately 1987)?

Approximately how many workers were at the LACBWR facility during the time it was being transitioned from permanent shutdown to SAFSTOR? How many workers were at the LACBWR facility while it was in SAFSTOR?

Provide the number of workers that will be required to complete dismantlement and decommissioning activities at LACBWR. Section 8.6.3.11 of the LACBWR LTP states that this increase in workers will occur in the 2016-2017 timeframe.

Section 8.6.3.11 of the LTP also states that some of these decommissioning workers will be from the local and regional area. Approximately how many of the workers will be from the local and regional areas?

After decommissioning of the LACBWR facility is complete in approximately 2018, approximately how many workers will remain on the former nuclear plant site?

ENV-16 Historic and Cultural Resources Section 8.6.3.13 of the LACBWR LTP states that land disturbance for the removal of large components will be minimized as the waste will be primarily shipped via rail.

Explain how shipping by rail will minimize land disturbances.

ENV-17 Noise Section 8.6.3.15 of the LACBWR LTP states that the licensee agrees to comply with any noise limitations imposed by the Village of Genoa. What noise limitations, if any, has the Village of Genoa imposed?

ENV-18 Transportation Barge Section 8.6.3.17 of the LACBWR LTP states that there will be no shipments of radioactive waste via barge, and that existing barge traffic is associated with the fossil station. Please confirm that no material, equipment, or supplies associated with decommissioning will be brought onsite or removed from the site by barge.

ENV-19 Transportation Rail Provide an update on the plans for shipment of radioactive waste from the LACBWR facility by rail. Specifically, provide details on the transportation of waste material by rail, such as the name and location of the rail line, number of shipments / cars anticipated / completed to date, and the general frequency of rail shipments. In addition, provide information on the number of truck shipments, frequency of shipments, and route used to transport the waste material by truck to the rail line.

ENV-20 Transportation Truck Section 8.6.3.17 of the LACBWR LTP states that the shipments of radioactive waste from the LACBWR facility via truck would be greater during decommissioning than during operation. Please provide the frequency of truck shipments (e.g., how many trucks per day, week, or month would be shipping waste material) and describe how this compares to similar shipments during the operation of LACBWR.

ENV-21 Traffic Section 8.6.3.17 of the LACBWR LTP states that the non-radiological impacts of transportation include increased traffic and wear and tear on roadways. It is anticipated that there will be no significant effect on traffic flow or road wear. Please explain why increased traffic and wear and tear on roadways is not expected.

ENV-22 Projected Waste Quantities Section 3.3.4 of the LACBWR LTP, Table 3-3, Projected Waste Quantities, provides the projected waste quantities that will be generated as a result of decommissioning activities at LACBWR. Please clarify if Table 3-3 includes the 75 cubic yards of legacy waste mentioned in Section 3.3.7 of the LTP. In addition, if the waste projections have changed since original submittal of the LACBWR LTP, please provide an updated projection of waste quantities.

ENV-23 Characterization of Hazardous Waste NUREG-0586, Supplement 1, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, does not evaluate the impacts of nonradioactive waste (e.g., solid waste, hazardous waste). The LACBWR LTP does not account for nonradioactive waste management and the potential impacts of its disposal on the capacity of receiving waste management facilities.

Provide summary information about the nonradioactive wastes that were or will be generated by the LACBWR decommissioning activities, and the plans for its onsite and offsite management and disposal. If a document already exists that describes these activities, please provide this document to the NRC staff for reference.

Information can be generally descriptive and does not need to be highly detailed.

ENV-24 Disposal of Contaminated Soils Provide summary information regarding the quantities and types of contaminated soils that were or will be generated as a result of the decommissioning and license termination activities at the LACBWR site, and indicate where such soils will ultimately be disposed of (e.g., local landfill).

ENV-25 Backfill Material Provide additional information on (1) where the backfill material to be used at the LACBWR site will come from, (2) how many shipments are expected, and (3) will the backfill material be stored onsite until its use. This information will allow the NRC staff to determine if there are additional impacts on land use and transportation.

Miscellaneous Topics (Identified by MISC)

MISC-1 Typos in the La Crosse LTP

a. Section 5.1.7.6: Appears to be cut-and-paste from the Zion LTP text; need to edit ZSRP and spell check the section in general.

b. Page 5-63: Correct the reference in the last sentence of the third paragraph of Section 5.10.5. The existing reference is to a spreadsheet on the Wilcoxon Rank Sum (WRS) test.

c. Page 6-18: Reference 13 is cited twice; it should instead point to Reference 12.

d. Page 6-19: Table 6-5 footnote: NUREG-6607 should be NUREG/CR-6697.

e. Page 6-27: Reference 14 is cited several places; however, there are only 13 references in the list on Page 6-56. Reference 14 likely refers to Sheppard and Thibault (1990), which is listed as Reference 12.

f. Table 6-6: The Entry for Runoff Coefficient has Uncertainty Result = 75th, but the Selected Deterministic Value = 0.45 is the median value of the distribution in NUREG/CR-6697, uniform [0.1, 0.8].

MISC-2 Clarifying Comments on the La Crosse LTP

a. List the correct revisions to the references in the reference section of each Chapter of the LACBWR LTP.

b. Clarify whether ZSRP as used in the first sentence of the last paragraph of Section 5.7.1.6 of the LACBWR LTP is a typo or is intended as written.

c. Inhalation Rate for Industrial Workers: NUREG-5512, Volume 3, Table 5.7 lists inhalation rates for Construction Workers ranging from 1.26 m3/hr to 1.68 m3/hr.

The value chosen in the soil conceptual model for the LACBWR LTP is 1917 m3/hr. This value is based on the NUREG/CR-5512, Volume 3, Table 5.1.1 mean value of 8400 m3/yr, which equates to 23 m3/day, or 0.96 m3/hr. Assuming 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year at this rate gives you 1917 m3/hr; however, the basis for multiplying by 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> per year is not explained.