ML22033A150
| ML22033A150 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 01/26/2022 |
| From: | LaCrosseSolutions |
| To: | Office of Nuclear Material Safety and Safeguards |
| Shared Package | |
| ML22033A165 | List: |
| References | |
| LS-2022-0002 | |
| Download: ML22033A150 (146) | |
Text
LA CROSSE BOILING WATER REACTOR LICENSE TERMINATION PLAN CHAPTER 6, REVISION 2 COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION
La Crosse Boiling Water Reactor License Termination Plan Revision 2 TABLE OF CONTENTS
- 6. Compliance with the Radiological Criteria for License Termination .................................. 6-1 6.1. Site Release Criteria ........................................................................................ 6-1 6.2. General Site Description ................................................................................. 6-1 6.2.1. Site Geology .................................................................................................... 6-1 6.2.2. Site Hydrogeology ........................................................................................... 6-2 6.2.3. Area Land Use ................................................................................................. 6-2 6.2.4. Area Groundwater Use .................................................................................... 6-3 6.3. Basements, Structures and Piping to Remain after License Termination (End State) .................................................................................. 6-3 6.4. Future Land Use Scenario and Average Member of the Critical Group ......... 6-5 6.5. Dose Model Overview..................................................................................... 6-6 6.5.1. Backfilled Basements ...................................................................................... 6-6 6.5.1.1. BFM Insitu Scenario ........................................................................................ 6-7 6.5.1.2. BFM Excavation Scenario ............................................................................... 6-7 6.5.2. Soil................................................................................................................... 6-8 6.5.3. Buried Piping ................................................................................................... 6-9 6.5.4. Existing Groundwater...................................................................................... 6-9 6.5.5. Remaining Above Grade Buildings .............................................................. 6-10 6.5.6. Dose Summation for Compliance ................................................................. 6-10 6.5.7. Alternate Scenarios........................................................................................ 6-11 6.6. Mixture Fractions for Initial Suite Radionuclides ......................................... 6-11 6.6.1. Potential Radionuclides of Concern and Initial Suite ................................... 6-11 6.6.2. Radionuclide Mixture Fractions .................................................................... 6-12 6.7. Soil Dose Assessment and DCGL ................................................................. 6-15 6.7.1. Soil Source Term ........................................................................................... 6-15 6.7.2. Soil Exposure Pathways and Critical Group ................................................. 6-16 6.8. Soil DCGL Computation Model - RESRAD v7.0........................................ 6-16 6.8.1. Parameter Selection Process .......................................................................... 6-16 6.8.2. RESRAD Parameter Selection for Uncertainty Analysis.............................. 6-17 6.8.3. Soil DCGL Uncertainty Analysis Results ..................................................... 6-19 6.9. Soil Deterministic Analysis and Soil DCGLs ............................................... 6-20 6.10. Basement Fill Conceptual Model .................................................................. 6-21 6.10.1. Source Term .................................................................................................. 6-22 6.10.1.1. Reactor Building ............................................................................................ 6-22 6.10.1.2. Waste Gas Tank Vault ................................................................................... 6-22 6.10.2. BFM Exposure Pathways .............................................................................. 6-23 6.11. BFM Insitu Scenario ..................................................................................... 6-23 6.11.1. BFM Insitu Groundwater Scenario ............................................................... 6-23 6.11.1.1. BFM Insitugw Computation Model ................................................................ 6-24 6.11.1.2. BFM Insitugw RESRAD Uncertainty Analysis for Initial Suite..................... 6-24 6.11.1.3. BFM Insitugw RESRAD Deterministic Analysis and DSR Results ............... 6-27 6.11.1.4. BFM Insitugw DCGL ...................................................................................... 6-28 6.11.2. BFM Insitu Drilling Spoils Scenario and DCGL Calculation ....................... 6-30 6.12. BFM Excavation Scenario............................................................................. 6-31 6-i
La Crosse Boiling Water Reactor License Termination Plan Revision 2 6.13. Basement Summation DCGL ........................................................................ 6-33 6.14. Insignificant Dose Contributors, Radionuclides of Concern and Surrogate Ratio ....................................................................................... 6-35 6.14.1. Insignificant Contributor Dose and Radionuclides of Concern .................... 6-35 6.14.2. Surrogate Radionuclide Ratio ....................................................................... 6-38 6.15. BFM Groundwater Scenario Mixing Volume Sensitivity Analysis for ROC 6-38 6.16. BFM DCGLs for ROC Adjusted for Insignificant Contributor Dose and Mixing Sensitivity ......................................................................................... 6-40 6.16.1. Soil DCGLs for ROC Adjusted for Insignificant Contributor Dose and Alternate Scenario Dose ................................................................................ 6-42 6.17. Concentrations in Excavated Fill Material .................................................... 6-42 6.18. Alternate Land Use Scenario Dose ............................................................... 6-43 6.18.1. Resident Gardener Dose: Soil ....................................................................... 6-44 6.18.2. Resident Gardener Dose: Reactor Building and WGTV............................... 6-47 6.18.2.1. Resident Gardener Dose: Basement Insitu Groundwater .............................. 6-47 6.18.2.2. Resident Gardener Dose: Basement Drilling Spoils ...................................... 6-49 6.18.2.3. Resident Gardener Dose: Basement Concrete Excavation Scenario ............. 6-51 6.18.2.4. Total Resident Gardener Alternate Scenario Dose: Reactor Building and WGTV .................................................................................................... 6-52 6.18.2.5. Resident Gardener Dose from Excavated Basement Fill ............................... 6-52 6.18.2.6.
Conclusion:
Resident Gardener Alternate Scenario Dose ............................. 6-52 6.19. Soil Area Factors ........................................................................................... 6-52 6.20. Buried Piping Dose Assessment and DCGL ................................................. 6-53 6.20.1. Source Term and Radionuclide Mixture ....................................................... 6-53 6.20.2. Exposure Scenario and Critical Group .......................................................... 6-53 6.20.3. Buried Piping Dose Assessment.................................................................... 6-54 6.20.4. Buried Pipe DCGLs Initial Suite ................................................................... 6-55 6.20.5. Buried Pipe Radionuclides of Concern and Adjusted DCGLs...................... 6-55 6.21. Existing Groundwater Dose .......................................................................... 6-57 6.22. Demonstrating Compliance with Dose Criterion .......................................... 6-58 6.23. References ..................................................................................................... 6-59 LIST OF TABLES Table 6-1 Basements and Below Ground Structures to Remain in LACBWR End State Ground Surface Elevation is 639 feet AMSL ................................ 6-4 Table 6-2 Buried Piping to Remain in LACBWR End State .......................................... 6-5 Table 6-3 Initial Suite of Potential Radionuclides and Mixture Fractions ................... 6-15 Table 6-4 Soil DCGL Uncertainty Analysis Results for Kd and Selected Deterministic Values ..................................................................................... 6-19 Table 6-5 Soil DCGL Uncertainty Analysis Results for Non-Nuclide Specific Parameters and Selected Deterministic Values ............................................. 6-20 Table 6-6 LACBWR Soil DCGLs for Initial Suite Radionuclides ............................... 6-21 Table 6-7 Deterministic Geometry RESRAD Parameters used in the Uncertainty Analysis for the Two BFM Insitugw Configurations ..................................... 6-24 6-ii
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-8 BFM Insitugw Reactor Building and WGTV Deterministic Values Selected for Distribution Coefficients (Kd) .................................................. 6-25 Table 6-9 Reactor Building: BFM Insitugw Uncertainty Analysis Results .................... 6-26 Table 6-10 WGTV: BFM Insitugw Uncertainty Analysis Results and Deterministic Values Selected for Non-Nuclide Specific Parameter Distributions............. 6-27 Table 6-11 BFM Insitugw DSRs for Reactor Building and WGTV ................................. 6-28 Table 6-12 BFM Insitugw DCGLs1 .................................................................................. 6-29 Table 6-13 BFM Insitu Drilling Spoils DCGLs for Both Basements1 ............................ 6-31 Table 6-14 BFM Concrete Excavation SA/V Ratios for Full and Partial Excavation .... 6-32 Table 6-15 BFM Excavation DCGLs for Both Basements1 ............................................ 6-33 Table 6-16 Summed Basement DCGL (DCGLB) for Initial Suite Radionuclides .......... 6-34 Table 6-17 Relative Dose Fractions, RDFi,k .................................................................... 6-35 Table 6-18 Np-237 Detection Statistics ........................................................................... 6-36 Table 6-19 ROC Mixture Fractions Renormalized to 1.00 ............................................. 6-37 Table 6-20 Sr-90 to Cs-137 Surrogate Ratios to be applied during FSS (95th Percentile values) .................................................................................. 6-38 Table 6-21 WGTV Mixing Sensitivity RESRAD Source Term Geometries .................. 6-39 Table 6-22 Reactor Building Mixing Sensitivity RESRAD Source Term Geometries ... 6-39 Table 6-23 Mixing Sensitivity Analysis Results Summary Maximum Ratio of Dose Factor Partial Mix/Full Mix ............................................................. 6-40 Table 6-24 Reactor Building BFM DCGLs for ROC Individual BFM Scenarios (DCGLBS) Adjusted for Insignificant Contributor Dose Fraction, Mixing Sensitivity and Alternate Scenario Dose .......................................... 6-41 Table 6-25 WGTV BFM DCGLs for ROC Individual BFM Scenarios (DCGLBS)
Adjusted for Insignificant Contributor Dose Fraction, Mixing Sensitivity and Alternate Scenario Dose ....................................................... 6-41 Table 6-26 BFM DCGLB Values for ROC Adjusted for Insignificant Contributor Dose Fraction and Alternate Scenario Dose.................................................. 6-42 Table 6-27 Soil DCGLs for ROC Adjusted for Insignificant Contributor Dose and Alternate Scenario Dose ................................................................................ 6-42 Table 6-28 Maximum Fill Concentration for Full and Partial Mix ................................. 6-43 Table 6-29 Additions/Revisions to Industrial Use Parameters Required for Resident Gardener Scenario .......................................................................... 6-44 Table 6-30 Soil Alternate Scenario Resident Gardener Uncertainty Analysis Results and Deterministic Values Selected for Non-Nuclide Specific Parameter Distributions .................................................................................................. 6-46 Table 6-31 Soil Alternate Scenario Resident Gardener Uncertainty Analysis Results for Distribution Coefficients (Kd) and Deterministic Values Selected ......... 6-47 Table 6-32 BFM Insitugw Alternate Scenario Resident Gardener Uncertainty Analysis Results and Deterministic Values Selected for Non-Nuclide Specific Parameter Distributions ................................................................... 6-50 Table 6-33 BFM Insitugw Alternate Scenario Resident Gardener Uncertainty Analysis Results for Distribution Coefficients (Kd) and Deterministic Values Selected ............................................................................................. 6-51 Table 6-34 Surface Soil Area Factors .............................................................................. 6-53 6-iii
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-35 RESRAD Source Term Parameters for Buried Piping DCGL Calculations . 6-55 Table 6-36 Buried Piping DCGLs ................................................................................... 6-56 Table 6-37 Buried Pipe DCGLs for ROCs Adjusted for Insignificant Radionuclide Fractions ........................................................................................................ 6-56 Table 6-38 Summed Buried Pipe DCGLs for ROCs adjusted for Insignificant Radionuclide Fractions .................................................................................. 6-57 Table 6-39 Groundwater Exposure Factors for a Water Concentration of 1 pCi/L ........ 6-58 LIST OF FIGURES Figure 6-1 Site Regional Location .................................................................................. 6-59 Figure 6-2 Site Overview ................................................................................................ 6-61 Figure 6-3 LACBWR Buildings ..................................................................................... 6-62 Figure 6-4 LACBWR End State ..................................................................................... 6-63 Figure 6-5 LACBWR End State - Backfilled Structures ................................................ 6-64 Figure 6-6 LACBWR End State - Backfilled Reactor Building Basement Elevation View .............................................................................................. 6-65 Figure 6-7 RESRAD Parameter Selection Flow Chart ................................................... 6-66 ATTACHMENTS 6-1, RESRAD Input Parameters for LACBWR Soil DCGL Uncertainty Analysis .................. 6-67 6-2, RESRAD Input Parameters for LACBWR BFM Uncertainty Analysis ............................ 6-91 6-3, RESRAD Input Parameters for LACBWR Alternate Scenario Uncertainty Analysis .... 6-115 6-iv
La Crosse Boiling Water Reactor License Termination Plan Revision 2 LIST OF ACRONYMS AND ABBREVIATIONS AF Area Factor ALARA As Low As Reasonable Achievable AMCG Average Member of the Critical Group AMSL Above Mean Sea Level ANL Argonne National Laboratory BcDCGL Base Case Derived Concentration Guideline Level BFM Basement Fill Model BFM Insitugw BFM Insitu Groundwater BFM Insituds BFM Insitu Drilling Spoils bgs Below Ground Surface DCGL Derived Concentration Guideline Level DPC Dairyland Power Cooperative FOV Field of view of ISOCS measurements FSS Final Status Survey G-3 Genoa-3 GW Groundwater HSA Historical Site Assessment HTD Hard-to-Detect ISFSI Independent Spent Fuel Storage Installation ISOCS In Situ Object Counting System LACBWR La Crosse Boiling Water Reactor LSE LACBWR Site Enclosure LTP License Termination Plan MARSSIM Multi-Agency Radiation Survey and Site Investigation Manual MDA Minimum Detectable Activity MDC Minimal Detectable Concentration NRC Nuclear Regulatory Commission PDF Probability Density Function PRCC Partial Rank Correlation Coefficient RESRAD RESidual RADioactive materials ROC Radionuclides of Concern TEDE Total Effective Dose Equivalent USACE U.S. Army Corps of Engineers WGTV Waste Gas Tank Vault WTB Waste Treatment Building 6-v
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Page Intentionally Left Blank 6-vi
La Crosse Boiling Water Reactor License Termination Plan Revision 2
- 6. Compliance with the Radiological Criteria for License Termination 6.1. Site Release Criteria The site release criteria for the La Crosse Boiling Water Reactor (LACBWR) are the radiological criteria for unrestricted release specified in Title 10, Section 20.1402, of the Code of Federal Regulations (10 CFR 20.1402):
Dose Criterion: The residual radioactivity that is distinguishable from background radiation results in a Total Effective Dose Equivalent (TEDE) to an average member of the critical group that does not exceed 25 mrem/yr, including that from groundwater sources of drinking water; and As Low As Reasonably Achievable (ALARA) Criterion: The residual radioactivity has been reduced to levels that are ALARA.
Chapter 4 describes the methods and results for demonstrating compliance with the ALARA Criterion. This Chapter describes the methods for demonstrating compliance with the Dose Criterion.
6.2. General Site Description This section provides a general description of the geology and hydrogeology at the LACBWR site. Land and groundwater use in the vicinity of site are also described. A detailed description of site geology and hydrogeology is provided in Haley & Aldrich, Inc., Hydrogeological Investigation Report, La Crosse Boiling Water Reactor, Dairyland Power Cooperative, Genoa Wisconsin (1).
The LACBWR facility is located 17 miles south of the City of La Crosse and a mile south of the Village of Genoa (population about 800) (see Figure 6-1). The nearest community (three miles to the northwest) is Reno, Minnesota, an unincorporated hamlet of about 300 people located on the west shore of the Mississippi River. The nearest community in Iowa is New Albin (pop.
522), five miles southwest of the plant. Victory, Wisconsin, five miles south of the plant on the east shore, is an unincorporated hamlet of about 80 people. The LACBWR licensed site area is shown in Figure 6-2, with a more detailed view of the facilities inside the LACBWR Site Enclosure (LSE) fence shown in Figure 6-3. See License Termination Plan (LTP) Chapter 2, section 2.1.6.1 for discussion of area classification and survey unit boundaries.
6.2.1. Site Geology LACBWR is located on the east bank of the Mississippi River in the Wisconsin Driftless section of the Central Lowland Physiographic Province. The local geology of the site is generally described as approximately 15 feet of hydraulic fill overlying 100 to 130 feet of glacial outwash and fluvial deposits. These unconsolidated deposits are underlain by flat lying sandstone and shales of the Dreshbach Group and by dense Precambrian crystalline rocks encountered at approximately 650 feet below the ground surface (bgs).
6-1
La Crosse Boiling Water Reactor License Termination Plan Revision 2 The primary soil types encountered at the site are:
0 to 20 feet bgs. Hydraulic Fill - Fill sands are encountered from approximately 0 to 20 feet bgs and described as light brown to brown, fine to medium sands with occasional fine gravel, 20 to 30 feet bgs. Brown to grey, fine to medium fine sands underlie the fill, with an average thickness of 7 to 28 feet, 30 to 100 feet bgs. Brown, fine to medium sands that also have zones of coarse sand and fine gravel below the finer sands, 100 to 115 feet bgs. Brown fine to medium sand and fine to medium gravels, 115 to 135 feet bgs. Brown fine to medium sand with trace silt, occasional zones of gravel.
6.2.2. Site Hydrogeology Regionally, groundwater flows from the bluff towards the Mississippi River. Closer to the river, it is likely that the groundwater flow direction turns downstream as groundwater discharges to the surface water. Groundwater elevation data from eight onsite wells agrees with regional groundwater flow and also shows seasonal variation on upward and downward gradients that are influenced by the river stage.
6.2.3. Area Land Use The LACBWR facility is located in the far western portion of Vernon County on the east bank of the Mississippi River. Although the site area is 163.5 acres, it is relatively isolated as it is bounded by the Mississippi River to the west, a rail line to the east, U.S. Army Corps of Engineers (USACE) property to the north, and a wildlife and fish refuge to the south.
The 163.5 acre site is comprised of the following:
the 1.5 acre LACBWR facility, the Genoa-3 (G-3) coal-fired, 379 MWe electric power station that was completed in 1969 and is owned and operated by Dairyland Power Cooperative (DPC). The G-3 station is located to the south, adjacent to LACBWR, an area south of G-3 where the LACBWR Independent Spent Fuel Storage Installation (ISFSI) is located, an approximately 36 acre area surrounding the ISFSI containing the closed coal ash landfill from past operations, the land north of the LACBWR plant was the site of the former Genoa-1 (G-1) coal (and later oil) fueled power plant (removed in 1989) which now includes the site switchyard and barge washing area (an approximately 900 m2 coal ash landfill is also present),
a parcel of land to the east of Highway 35, across from LACBWR.
The area of the Mississippi River adjacent to the site is used for recreational purposes (boating and fresh water fishing) and commercial barge and ship traffic (e.g. barges of coal are delivered to the G-3 station located south of the LACBWR plant). There is a public boat landing on the 6-2
La Crosse Boiling Water Reactor License Termination Plan Revision 2 site, located approximately 4,000 feet south of the plant. There is a portion (Pool 9) of the Upper Mississippi River National Wildlife & Fish Refuge just south of the site which has limited access for hunting, fishing and recreational activities. Further south are public land areas and the Genoa National Fish Hatchery.
Lock and Dam No. 8, located on the Mississippi River at mile 679.2, is approximately 0.6 miles north of the site. The dam is a 110 feet wide, 600 feet long lock and dam structure owned and operated by the USACE. This facility also allows public access to an observation platform, open from dawn to dusk during the months of April through November. The State of Wisconsin maintains a highway wayside off State Highway 35 approximately 1/2 mile north of the LACBWR site, across from Lock and Dam No. 8.
The closest town is Genoa, located approximately 1 mile to the northeast of the site. There are no residences within 2,000 feet of any LACBWR structure.
6.2.4. Area Groundwater Use There are two water supply wells on the site (Deep Well 3 and Deep Well 4) that were installed in 1963 to 129 feet and 116 feet bgs, respectively. Both wells are located up-gradient from the LSE and are still in use as potable water for LACBWR. Separate groundwater wells supply water to the G-3 plant. Regionally, there are five domestic wells south of the LACBWR site and east of Highway 35.
6.3. Basements, Structures and Piping to Remain after License Termination (End State)
The configuration of the remaining backfilled basements, buried piping, open land areas and above grade buildings at the time of license termination is designated as the End State.
As indicated in Chapter 3 of this LTP, the Reactor Building and WGTV will be demolished and removed to a depth of at least three feet below grade (which corresponds to an elevation of 636 foot Average Mean Sea Level (AMSL)) and backfilled. See Table 6-1 and Figures 6-4 and 6-5.
Figure 6-6 provides a cross-section of the Reactor Building basement below 636 foot elevation.
All other LACBWR buildings will be fully removed, including above and below grade portions.
All other impacted LACBWR buildings, structures and components, other than the following structures, will be demolished and removed in their entirety. The impacted above grade structures that will remain are:
- LACBWR Administration building
- G-3 Crib House
- LACBWR Crib House
- Transmission Sub-Station Switch House
- G-1 Crib House
- Barge Wash Break Room
- Back-up Control Center
- Security Station As seen in Figure 6-2, there are numerous buildings associated with the G-3 coal plant that are outside of the LSE, but within the LACBWR licensed boundary, that will remain intact and functional for G-3 power operations. However, these buildings and adjacent open land areas were not used to support LACBWR operations. The majority of open land areas and buildings outside of the LSE fence are designated as non-impacted. The G-3 Crib House is an exception in that it is classified as impacted due to its location in an impacted soil survey unit. See LTP 6-3
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Chapters 2 and 5 for additional discussion and justification of the non-impacted classification of areas outside of the LSE.
For the basements that will remain, all systems and components will be removed. The backfilled structures are generally referred to as either basements or structures in this LTP Chapter. For the Reactor Building, only the concrete exterior to the steel liner will remain; all interior concrete and the steel liner will be removed. The WGTV basement is comprised of concrete only. All remaining concrete will be decontaminated as necessary to meet the 10 CFR 20.1402 unrestricted use criteria.
Table 6-1 Basements and Below Ground Structures to Remain in LACBWR End State Ground Surface Elevation is 639 feet AMSL Material Floor and Wall Floor Elevation Basement/Structure remaining Surface Area (m2) (feet AMSL)
Reactor Building Concrete 511.54 612 Waste Gas Tank Vault Concrete 310.56 621 The End State will also include a limited number of buried pipes, with the majority not associated with contaminated operational systems. The exception is the remaining portion of the Circulating Water Discharge pipe which was used as for liquid effluent discharge as well as circulating water discharge. The buried piping to remain in the LACBWR End State is listed in Table 6-2.
For the purpose of this LTP, buried piping is defined as that contained in soil. Typical commercial power plants also contain piping that penetrates walls or is embedded in concrete.
However, the design of the LACBWR plant precludes the need for penetrations or embedded piping and none are present.
The potential for significant surface or subsurface soil contamination at LACBWR is low based on the findings of the EnergySolutions Technical Support Document (TSD) RS-TD-313196-003, La Crosse Boiling Water Reactor Historical Site Assessment (HSA) (2) and the results of extensive characterization performed in 2014 (see LTP Chapter 2). There are indications of subsurface soil contamination under the Turbine Building based on positive groundwater monitoring results down gradient of suspected broken drain lines. However, GeoProbe samples collected under the Turbine Building in the vicinity of the suspect drain lines did not identify plant-derived radionuclides above background. Additional characterization of the subsurface soil was performed after the Turbine Building foundation was removed and the underlying soil exposed. Very low levels of plant-derived radionuclides were identified.
6-4
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-2 Buried Piping to Remain in LACBWR End State Description of Piping Quantity Pipe Elevation (Bottom)
High Pressure Service Water from LACBWR 40 of 6 627 Crib House to G-31 222 of 8 Deicing Line 105' of 18" 630.5 South Storm Drain 630' of 48" 625.0 100' of 10" 635.0 North Storm Drain 435' of 24' 632 .0 250' of 32" 626.0 Remaining Portion of High Pressure Service Water Supply to LACBWR Fire Suppression 863' of 6" 632.75 System Remaining Portion of Low Pressure Service 44' of 16" 632.0 Water Pipe Remaining portion of Circulating Water 40' of 60" 630.5 Intake Pipe Well Water piping for Well #31 well water #3 pipe 438 of 3", 115 of 2", 285 of 1.5" installed vertically to depth of 129 Remaining portion of Circulating Water 525 of 60 steel pipe 630.5 Discharge Pipe2 Note 1: G-3 service water supply pipe and well water supply pipe considered non-impacted. No FSS will be performed.
Note 2: CW Discharge drops 10 to 620.5 at the outfall.
6.4. Future Land Use Scenario and Average Member of the Critical Group The Reasonably Foreseeable Scenario, is defined in NUREG-1757, Volume 2, Revision 1 Consolidated Decommissioning Guidance - Characterization, Survey, and Determination of Radiological Criteria (NUREG-1757), (3), Table 5.1 as a land use scenario that is likely within the next 100 years. The Reasonably Foreseeable Scenario for LACBWR is industrial use.
DPC first acquired land at the site in 1941 to build its first coal-fired generation station, G-1. In 1949 an additional 18.6 acres were acquired for G-1 coal ash disposal. An additional 80.8 acres were acquired in 1962. Finally, an additional 26.8 acres was filled by dredging in 1962-63. The total area owned or controlled by DPC was then, and continues to be, 163.5 acres.
The G-1 plant began operation in 1941 and was decommissioned and dismantled in 1989.
LACBWR construction began in 1963, was completed in 1967, and was permanently shut down in 1987. DPCs third plant onsite, G-3, began commercial operation in 1969 and continues operation as a major generation resource to the DPC system. The G-3 plant has been updated to meet current environmental standards and is currently performing a major turbine overhaul. The projected remaining operation life of G-3 is projected to be 20-25 years.
Residential use of the site over the next 100 years appears unlikely. The site is relatively isolated as it is bounded by the Mississippi River to the west, a rail line to the east, USACE property to the north, and a wildlife and fish refuge to the south. There are several small communities nearby that would be more suitable to additional residential development than the DPC site. In 6-5
La Crosse Boiling Water Reactor License Termination Plan Revision 2 addition, the presence of over 36 acres of closed coal ash landfills further supports the assumption of no foreseeable future residential development. Based on surrounding land use, the conversion of the property to recreational use at some point in the distant future is more likely than residential use.
In summary, the DPC site has been in continuous industrial use for 74 years and DPC has no plans to change the land use in the future. The site contains a transmission station and valuable infrastructure to support the sites future use for power supply after G-3 is decommissioned in 20-25 years. Adjacent land, and a part of the DPC site is currently used for recreational purposes.
Finally, there are large tracts of land nearby that would be preferable, and more cost-effective to develop for residential use within at least the next 100 year time period than the conversion of the DPC site from industrial use. Residential land use is therefore categorized as less likely but plausible in accordance with NUREG-1757, Table 5.1 definitions. Based on the above discussion of future land use, the Average Member of the Critical Group (AMCG) for the LACBWR dose assessment is the Industrial Worker.
6.5. Dose Model Overview Dose modeling is performed to demonstrate that residual radioactivity remaining at the time of license termination will not result in a dose to the AMCG (industrial worker) exceeding the 25 mrem/yr criterion. This section provides a general overview of the LACBWR site conceptual model and computational methods for dose assessment.
There are five potential sources of residual radioactivity in the End State that were categorized as follows for the purpose of dose modeling; backfilled basements, soil, buried piping, above grade buildings, and existing groundwater. The majority of residual radioactivity to remain at the time of license termination will be contained in basement concrete. There is no indication that significant soil contamination is currently present at the LACBWR site or will be present in the End State. With the exception of the remaining portion of the Circulating Water Discharge Pipe, the limited buried piping that will remain was not associated with contaminated systems and is expected to contain minimal, if any, contamination. Groundwater exposure factors were calculated to address existing groundwater contamination. The dose from each of the five sources will be summed to demonstrate compliance with the 25 mrem/yr dose criterion.
An overview of the dose assessment methods for the five sources is provided below. Detailed descriptions and results are provided in sections 6.7 to 6.16.
6.5.1. Backfilled Basements The dose model for backfilled basements to remain below 636 foot elevation (see Table 6-1) is designated as the Basement Fill Model (BFM). The BFM calculates the dose to the AMCG from residual radioactivity remaining in the basements.
The basement End State will be comprised of backfilled concrete structures that are physically altered to a condition that would not realistically allow the remaining structures to be occupied.
The BFM conceptual model includes two source term geometries; 1) the Insitu geometry where the concrete remains in the as-left configuration at the time of license termination and 2) the Excavation geometry where some or all of the concrete is excavated and brought to the surface.
6-6
La Crosse Boiling Water Reactor License Termination Plan Revision 2 The results of the BFM dose assessments are expressed as Derived Concentration Guideline Levels (DCGLs) in units of pCi/m2 for each of the two basements. It cannot be ruled out that a portion of the backfilled structures could be excavated and a portion remain in the ground.
Therefore, the final dose for demonstrating compliance will conservatively include the sum of the BFM Insitu and BFM Excavation dose.
6.5.1.1. BFM Insitu Scenario The BFM Insitu geometry includes two exposure scenarios; 1) exposure to well water containing residual radioactivity as a result of leaching from the backfilled concrete surfaces into the fill material and 2) exposure to drilling spoils that are brought to the surface during the assumed installation of an onsite water supply well. The well water scenario is designated as BFM Insitu Groundwater (BFM Insitugw). The drilling spoils scenario is designated as BFM Insitu Drilling Spoils (BFM Insituds).
The conceptual model for the BFM Insitugw scenario assumes that the residual radioactivity in floors and walls is released to adjacent fill through leaching into water that comes into contact with the concrete surfaces after backfill. The scenario assumes that 100% of the residual radioactivity is released instantly into the fill which is then treated as contaminated soil. The release occurs immediately after license termination taking no credit for radioactive decay. A water supply well is assumed to be installed adjacent to the down gradient edge of the building and flow to the well is conservatively modeled assuming that structures are not present. Because a clean cover of at least 3 feet is present over potentially contaminated surfaces the dose from direct exposure and soil ingestion/inhalation is negligible or zero. The BFM Insitugw exposure pathway to the industrial worker AMCG is drinking water from the onsite well.
The BFM Insituds scenario assumes that the drilling spoils from the installation of the onsite well are brought to the surface. The well installation is assumed to occur immediately after license termination, before any leaching from concrete occurs and taking no credit for radioactive decay.
The residual radioactivity in the floor concrete that is contacted by the borehole during installation is assumed to be inadvertently mixed with the fill material above the floor surface, brought to the ground surface, and spread over a 15 cm thick layer. The BFM Insituds exposure pathways are the same as those that apply to contaminated surface soil.
The BFM Insitugw dose assessment is implemented using the RESidual RADioactive materials (RESRAD) v7.0 model. The BFM Insituds dose assessment is implemented using Excel spreadsheet calculations coupled with RESRAD modeling. The results of the BFM Insitu dose assessments are expressed as DCGLs in units of pCi/m2 for each basement. (see Table 6-1). The concentrations of residual radioactivity (in units of pCi/m2) at the time of license termination, as determined by the FSS, will be divided by the BFM DCGLs for each Radionuclide of Concern (ROC), and summed as necessary, to demonstrate compliance with the 25 mrem/yr dose criterion.
6.5.1.2. BFM Excavation Scenario The BFM Excavation scenario assumes that some or all of the backfilled structure concrete is excavated and spread on the ground surface at some time after license termination. For conservatism, the excavation is assumed to occur immediately after license termination taking no 6-7
La Crosse Boiling Water Reactor License Termination Plan Revision 2 credit for radioactive decay. The residual radioactivity remaining in the backfilled basements is assumed to inadvertently mix with the mass of structural concrete removed during excavation which is generally consistent with the guidance in NUREG-1757, Appendix J, for addressing subsurface contamination. The calculation is performed using Excel spreadsheet and results in BFM Excavation DCGLs that are expressed in the same units as the DCGLs for the BFM Insitu scenario, i.e., pCi/m2. The fundamental driver of the BFM Excavation DCGL calculation is that the average concentration in the excavated concrete is limited such that the surface soil DCGLs are not exceeded.
BFM Excavation DCGLs were calculated separately for the Reactor Building and WGTV consistent with the BFM Insitu DCGL calculations.
6.5.2. Soil Derived Concentration Guideline Levels, in units of pCi/g, were developed for residual radioactivity in surface soils that correspond to the 25 mrem/yr dose criterion. RESRAD was used to perform the calculation.
Surface soil is defined as the first 15 cm layer of soil and FSS for surface soil will be performed on the first 15 cm. However, for conservatism, and to ensure efficient implantation of FSS, the surface soil dose assessment assumed a depth of 1 m from the surface. A standard surface soil contamination thickness of 15 cm would result in lower dose (i.e., higher DCGL). In the unlikely event that soil contamination is identified at LACBWR with a thickness greater than 15 cm, additional dose modeling may be required if the conceptual model assumed a 15 cm contamination thickness. Using a 1 m thickness reduces the potential for delays or unnecessary remediation if contamination with a thickness somewhat greater than 15 cm is encountered.
There is low potential for significant subsurface contamination to remain in the End State with a geometry comprised of a clean soil layer over a contaminated soil layer at depth.
In the unlikely event that geometries are encountered during continuing characterization or during FSS that are not bounded by the assumed 1 m soil contamination thickness, the discovered geometries will be addressed by additional modeling. If the modeling dictates that the new geometry results in a change to the soil DCGLs, then LS will seek approval from the NRC before implementing the change.
Survey Unit L1-SUB-TDS B consists of the sloped area boundaries of the excavation for the former RPGPA Sump, and this excavation is unusually deep. The maximum depth of the excavation in L1-SUB-TDS B is at the 618' elevation, and the area experiences groundwater intrusion due to rising Mississippi River levels. Because of the depth of the excavation is unusually deep and is in contact with the water table, Survey Unit L1-SUB-TDS B is considered outside of the geometry bounded by the 1 m surface soil thickness previously modeled.
Therefore, a sensitivity analysis (Reference 15) has been performed to address the differing condition in the survey unit and to confirm that the existing DCGLs are bounding with the exception of Sr-90 when the surrogate ratio with Cs-137 exceeds 1.11, the existing DCGLs are bounding. Additional soil samples were taken and confirmed that the surrogate ratio for Sr-90 and Cs-137 did not exceed 1.11, and thus the existing DCGLs were appropriate for use in the area.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Standard methods for RESRAD parameter selection and uncertainty analysis are used consistent with guidance in NUREG-1757. The AMCG for soil is the Industrial Worker.
6.5.3. Buried Piping With the exception of the portion of the Circulating Water Discharge Pipe, none of the buried piping to remain at LACBWR was associated with contaminated systems and therefore contamination potential is minimal (see Table 6-2). A buried piping dose assessment was conducted to develop DCGLs for pipe. There is no embedded piping present at LACBWR (i.e.,
embedded in concrete).
The conceptual model for the buried piping dose assessment is similar to the BFM and includes two scenarios: Insitu and Excavation. In the Insitu scenario the residual radioactivity on the internal surfaces of the pipe is assumed to instantaneously release and mix with a thin 2.54 cm layer of soil in an area equal to the internal surface area of the pipe. For the Excavation scenario, the soil mixing layer is 15 cm due to the extensive ground surface disturbance associated with the pipe excavation. The Insitu scenario assumes that the released radioactivity is in a below ground, 2.54 cm layer, of soil with no credit taken for the presence of the pipe to reduce environmental transport and migration. This is a conservative assumption, particularly for the Circulating Water Discharge Pipe which will be filled with a flowable fill material. The Excavation scenario assumes that the pipe is excavated followed by instant release of all radioactivity into a 15 cm layer of soil on the ground surface with no cover. The Industrial Worker is exposed to the Insitu and Excavated soil via the same pathways applicable to the BFM and soil dose assessment scenarios. RESRAD modeling is used in conjunction with Excel spreadsheet to calculate DCGLs in units of dpm/100 cm2.
6.5.4. Existing Groundwater Groundwater monitoring has been and will continue to be conducted. The final monitoring date will depend on the evaluation of sampling results. Low concentrations of groundwater contamination have been identified adjacent to the suspected broken floor drains under the Turbine Building. Groundwater sampling in 1983 from a well located down gradient of the Turbine Building identified contamination at relatively low concentrations. In late 2012, five additional monitoring well pairs were installed to support site characterization and license termination. Results indicated lower groundwater contamination levels than found in 1983, predominantly H-3. See LTP Chapter 2, section 2.3.7 for a summary of groundwater sampling results prior to submittal of LACBWR LTP Revision 0. In December 2017, the analysis identified elevated H-3 with a maximum concentration of 24,200 pCi/L identified in February 2018 (well MW-203A). A subsequent sample from well MW-203A, collected in April 2018, contained lower H-3 concentrations of 12,100 pCi/L indicating a downward trend. An investigation is ongoing to identify the source of the contamination and the extent of contaminant migration. A report of the investigation results will be provided to NRC for evaluation.
Groundwater exposure factors in units of mrem/yr per pCi/L groundwater concentration are provided to perform dose assessment for existing groundwater contamination.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 6.5.5. Remaining Above Grade Buildings As indicated in Chapters 3 and 5 of this LTP, all impacted LACBWR buildings, structures and components, other than the following structures, will be demolished and removed to a depth of at least 3 feet below grade or removed in their entirety. The impacted above grade structures that will remain are; LACBWR Administration building G-3 Crib House LACBWR Crib House Transmission Sub-Station Switch House G-1 Crib House Barge Wash Break Room Back-up Control Center Security Station None of the buildings and structures associated with the Genoa 3 Fossil Station (G-3) are expected to be radiologically impacted. Therefore, the structures associated with G-3 will remain intact and functional for G-3 power operations. The G-3 Crib House is classified as impacted due to its location in an impacted soil survey unit. The impacted above grade structures that will remain are not expected to contain residual radioactivity and will be subjected to FSS.
The Screening Values in NUREG-1757, Table H-1 will be applied to the FSS of above grade buildings.
6.5.6. Dose Summation for Compliance The DCGLs for each ROC, in each medium, represent the concentration that would result in a dose of 25 mrem/yr for that ROC and medium independently after adjusting for the dose from insignificant contributor radionuclides. Compliance is demonstrated through the summation of the dose from each of the ROCs in each of the five media (basement concrete, soil, buried pipe above grade buildings and existing groundwater). The LACBWR basements do not contain embedded pipes or penetrations.
To ensure that the dose summation from the residual radioactivity in each of the five media is 25 mrem/yr or less after all FSS is completed, the DCGLs for each medium are reduced based on an assigned, or a priori, fraction of the 25 mrem/yr dose criterion. The reduced DCGLs are designated as Operational DCGLs. The summation of the dose from all ROC, in all five media will equal 25 mrem/yr if residual radioactivity is present at the Operational DCGL concentrations.
A detailed description of the methods and equations used to calculate the Operational DCGLs and perform the compliance dose summation is provided in Energy Solutions TSD LC-FS-TSD-002, Operational Derived Concentration Guideline Levels for Final Status Survey, Revision 0 (4). Additional discussion of the compliance dose summation is provided in section 6.22.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 6.5.7. Alternate Scenarios As discussed in section 6.4, the industrial use scenario was selected as a reasonably foreseeable scenario based on past and projected use of the LACBWR site. Two alternate land use scenarios were considered including recreational use, and residential use with a water supply well and onsite garden.
A qualitative evaluation of a recreational land use scenario concluded that the dose would be lower than that calculated for the industrial use scenario because occupancy time would be less than that assigned to an industrial worker. In addition, if a water supply well were installed in the recreational land use scenario, the recreational users intake rate from the well would be less than assumed for the industrial worker.
The resident gardener land use was considered to be a less likely but plausible land use scenario as defined in NUREG-1757, Table 5-1. RESRAD was used, in conjunction with Excel spreadsheet, to calculate the dose from the Resident Gardener alternate scenario. NUREG-1757 states that if the peak dose from a less likely put plausible scenario is significant, then greater assurance that the scenario is unlikely would be necessary. The alternate scenario is addressed in section 6.15.
6.6. Mixture Fractions for Initial Suite Radionuclides A comprehensive initial suite of 22 radionuclides were identified that could potentially be present at the LACBWR site. As discussed below, the initial suite includes many radionuclides that have not been positively identified in LACBWR 10 CFR Part 61 analysis or during characterization but could theoretically be present, albeit at very low mixture fractions, relative the primary radionuclide which is Cs-137.
The initial suite mixture fractions were used in conjunction with the soil DCGLs and BFM DCGLs to identify the insignificant dose contributors and final list of ROC that will be considered during the FSS and final dose summation for compliance as described in section 6.5.6.
6.6.1. Potential Radionuclides of Concern and Initial Suite EnergySolutions TSD RS-TD-313196-001, Radionuclides of Concern During LACBWR Decommissioning, (5) established an initial suite of potential ROC. Two industry guidance documents were reviewed including NUREG/CR-3474, Long-Lived Activation Products in Reactor Materials, (6), and NUREG/CR-4289, Residual Radionuclide Concentration Within and Around Commercial Nuclear Power Plants; Origin, Distribution, Inventory, and Decommissioning Assessment (7). The review also included an evaluation of a LACBWR spent fuel inventory assessment conducted in 1988 that was decay corrected to January 2015. Several 10 CFR Part 61 waste stream analyses were also reviewed.
The list of activation product radionuclides for consideration in the initial suite was developed after eliminating noble gases, radionuclides with half-lives less than two years, and radionuclides with theoretical neutron activation products with abundances less than 0.01 percent relative to Co-60 and Ni-63 (the prominent activation products identified in LACBWR 10 CFR Part 61 samples). The review of Reference (7), which includes both activation and fission products, coupled with review of the LACBWR fuel inventory and 10 CFR Part 61 analyses resulted in 6-11
La Crosse Boiling Water Reactor License Termination Plan Revision 2 several additional radionuclides being included in the list of those that could be potentially present during the decommissioning of LACBWR.
Finally, the results of concrete core samples collected from the Waste Treatment Building (WTB), Reactor Building and Piping and Ventilation Tunnels were reviewed. No radionuclides were positively identified that were not already accounted for by the assessments described above.
The resulting list of potentially present radionuclides is called the Initial Suite and is provided in Table 6-3. The process for determining the mixture fractions is described below.
6.6.2. Radionuclide Mixture Fractions As described in Reference (5), the mixture fractions for the initial suite radionuclides were developed from the 38 concrete cores collected during initial and continuing characterization of the Reactor Building, WTB, WGTV, and Piping/Ventilation Tunnels. The core data was decayed to the scheduled license termination date of March 2020. The vast majority of the activity was located in the first 1.27 cm slices from the cores which were therefore used for the mixture calculation. The use of cores with higher concentrations provides the most accurate estimate of mixture fractions by minimizing the bias caused by non-detect results.
Three different methods were used in Reference (5) to calculate the mixture fractions which are summarized as follows:
- 1. The first was to calculate the radionuclide activity fraction, fAi,j,k, for each sample, j, each radionuclide, i, within each population, k, from the reported radionuclide activity concentrations, Ci,j,k, using Equation 6-1 and then calculating the average activity fraction, fAi,j,k, for each radionuclide, i, and population, k, of N samples using Equation 6-2.
Equations 6-1 and 6-2
,, =
(),,
(),,
, =
- 2. The second was to calculate the 75th percentile of the population of samples from Equations 6-1 above. Once the 75th percentile fraction were calculated for each radionuclide, fi,k,.75, data set was re-normalized to determine the percentile-based activity fractions, fAi,k,.75 using Equation 6-3.
Equation 6-3
,,,.75
,,.75 =
(),,,.75
- 3. The third was to calculate the individual radionuclide ratios to Cs-137 for each sample, Ri,Cs-137,j, calculate the 75th percentile for the sample group, Ri,Cs-137,k, .75 then renormalize to determine the activity fractions, fRAi,k,.75 using Equation 6-4.
Equation 6-4 6-12
La Crosse Boiling Water Reactor License Termination Plan Revision 2
,137,,.75
,,.75 =
(),137,,.75 The analyses give equal statistical weight to each of the sample results and are not affected by activity weighted averaging. Also, the analyses apply the actual reported activity concentrations in the calculation of activity fractions regardless of whether the reported activity was less than the detectability criteria.
The 75th percentile values were calculated to conservatively account for uncertainty in the mixture fractions. It is recognized that a higher percentile value will provide a different mixture profile, but such an approach would give undo weight to outliers of the samples population. In addition, there is precedence in using the 75th percentile, particularly in the parameter selection process for dose modeling to support DCGL calculations. The 75th percentile is therefore considered sufficient to provide overall conservatism in the development of the radionuclide mixtures.
The results of the three methods were evaluated in section 6.3 of Reference (5). The two methods involving the use of the 75th percentiles (Equations 6-3 and 6-4) provide very similar results, particularly the dose fraction attributed to the insignificant contributor (IC) radionuclides (see section 6.13 for discussion of IC dose fractions), and both are more conservative than using averages (Equations 6-1 and 6-2). It was initially expected that the method using the 75th percentile of the Cs-137 fractions (Equation 6-4) would yield higher IC dose fractions in all cases. However, the data shows that the IC dose fractions are maximized slightly higher for soil and the Reactor Building using Equation 6-4 but slightly lower (by 0.2%) for the WGTV as compared to the 75th percentile of the activity averages (Equation 6-3). But the differences appear to be within the overall uncertainties and variance of the data sets leading to a conclusion that the methods are effectively very similar. For consistency and simplicity, the 75th percentile of the Cs-137 fractions was chosen to conservatively represent the overall nuclide mixtures for soil, the Reactor Building, and the WGTV. The soil mixture fractions are also applied to buried pipe and above grade buildings.
The mixture fractions for the Reactor Building and WGTV are based on cores collected from the respective buildings. For soil, there were only a few positive soil sample results identified during characterization, predominantly Cs-137, and the concentrations were insufficient to provide a meaningful evaluation of the relative mixture fractions for the HTD radionuclides which were all reported as less than the MDC. Therefore, the soil radionuclide mixture fractions were determined from an analysis of all 38 concrete cores under the premise that the source of soil contamination could be from any of the LACBWR buildings during operation as well during remediation and demolition.
The radionuclide mixtures for soil, Reactor Building and WGTV are listed in Table 6-3.
The analysis of the Np-237 activity fraction in the WGTV is different than the soil and Reactor Building. Np-237 was not detected in any of the 38 samples (see Attachment 2 of Reference (5)). The average reported activity fraction and the normalized 75th percentile fraction are 6.07E-5 and 4.12E-4 respectively (5). These clearly suggest the absence of Np-237. However, even at these low fractions, the values artificially skew the IC dose analysis provided in section since 6-13
La Crosse Boiling Water Reactor License Termination Plan Revision 2 the Np-237 DCGLs are much lower than the other radionuclides. Therefore, the Np-237 IC dose was calculated using an alternate approach to eliminate this artifact as further explained in section 6.13. As such, the Np-237 fraction for the WGTV in Table 6-3 has been set to zero.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-3 Initial Suite of Potential Radionuclides and Mixture Fractions Soil Mix Rx Bldg Mix WGTV Mix Nuclide Fraction Fraction Fraction H-3 1.51E-01 2.36E-02 2.52E-01 C-14 1.72E-03 1.27E-03 9.37E-03 Fe-55 2.36E-02 1.40E-02 -8.13E-03 Ni-59 7.40E-04 2.48E-04 4.74E-02 Co-60 3.43E-02 4.58E-02 4.76E-03 Ni-63 2.64E-01 2.77E-01 1.89E-01 Sr-90 5.22E-02 7.59E-02 9.12E-03 Nb-94 1.68E-04 1.07E-04 1.01E-03 Tc-99 2.06E-03 2.16E-03 6.91E-03 Cs-137 4.41E-01 4.92E-01 4.49E-01 Eu-152 2.93E-03 1.84E-03 4.49E-03 Eu-154 1.50E-03 2.49E-03 1.60E-03 Eu-155 2.08E-03 6.61E-04 4.56E-03 Np-237 2.15E-06 2.17E-06 0.00E+00 Pu-238 1.16E-03 2.27E-03 7.95E-04 Pu-239/240 7.80E-04 3.17E-03 1.90E-04 Pu-241 1.56E-02 4.58E-02 2.35E-02 Am-241 3.56E-03 1.03E-02 3.25E-03 Am-243 5.85E-04 6.18E-04 4.55E-04 Cm-243/244 1.65E-04 1.58E-04 1.78E-04 Total 1.00 1.00 1.00 6.7. Soil Dose Assessment and DCGL Site-specific DCGLs were developed for residual radioactivity in surface soil that represent the 10 CFR 20.1402 dose criterion of 25 mrem/yr. A DCGL was calculated for each initial suite radionuclide in order to provide an input to the determination of IC dose percentage and the final list of ROC.
The surface soil conceptual model assumes that the soil contamination is contained in a uniformly contaminated 1 m layer of soil from the ground surface downward. There are no expectations that at the time of license termination, residual radioactivity will remain with a geometry consisting of a clean surface layer of soil over a contaminated subsurface soil layer with concentrations exceeding the surface soil DCGL.
6.7.1. Soil Source Term There is limited potential for significant surface or subsurface soil contamination at LACBWR based on the results of soil characterization performed in 2014. There are indications that subsurface soil contamination may be present under the Turbine Building based on positive groundwater monitoring results down gradient of suspected broken drain lines. However, 6-15
La Crosse Boiling Water Reactor License Termination Plan Revision 2 GeoProbe samples collected under the Turbine Building in the vicinity of the suspect drain lines did not identify plant-derived radionuclides above background. Additional subsurface soil samples were collected during continuing characterization after the Turbine Building foundation was removed and the underlying soil exposed. Very low levels of plant-derived radionuclides were identified.
Soil characterization results are provided in LTP Chapter 2 and summarized here. A total of 22 biased surface soil samples and 79 subsurface soil samples (at varying depths to six meters) were collected inside the LSE fence. Thirteen of the surface soil sample analyses results indicated Cs-137 above the MDC with a maximum of 1.07 pCi/g. Two of the surface soil sample results indicated Co-60 above the MDC with a maximum of 0.287 pCi/g. Fifteen of the subsurface soil samples indicated Cs-137 above MDC with a maximum of 0.161 pCi/g with no subsurface Co-60 results greater than MDC. The Cs-137 results are all in the range of natural background.
6.7.2. Soil Exposure Pathways and Critical Group The AMCG at the LACBWR site is the Industrial Worker. The following exposure pathways apply:
Direct exposure to external radiation Inhalation of airborne radioactivity Ingestion of soil Ingestion of water from onsite well The agricultural and gardening pathways are not applicable to industrial land use. The meat, milk, grain and vegetable ingestion pathways are therefore not included.
6.8. Soil DCGL Computation Model - RESRAD v7.0 The RESRAD model was used to calculate DCGLs for surface soil. The RESRAD output reports for the modeling discussed below are provided electronically in conjunction with EnergySolutions TSD RS-TD-313196-004, LACBWR Soil DCGL and Concrete DCGL (8).
6.8.1. Parameter Selection Process RESRAD parameters are classified as behavioral, metabolic or physical. Some parameters may belong to more than one category. Physical parameters are determined by the geometry and location of the source term and the geological characteristics of the site (i.e., source-specific and site-specific) including the geohydrologic, geochemical, and meteorologic characteristics. The characteristics of atmospheric and biospheric transport up to, but not including, uptake by, or exposure to the dose receptor, are also considered physical input parameters.
Behavioral parameters define the receptors behavior considering the conceptual model selected for the site. For the same group of receptors, a parameter value could change if the scenario changed (e.g., parameters for industrial use could be different from residential use). For LACBWR, the behavioral parameters are based on an industrial use scenario and are the same for both the BFM and soil dose assessments.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Metabolic parameters define certain physiological characteristics of the potential receptor. One set of metabolic parameters applies to both the BFM and soil dose assessments.
Physical, behavioral and metabolic parameters are treated as deterministic parameters in the final dose modeling to calculate soil DCGLs (and BFM DCGLs). The deterministic module of the code uses single values for input parameters and generates a single value for dose. The parameter selection process is described below.
Argonne National Laboratory (ANL) ranked physical parameters by priority as 1, 2, or 3.
Priority 1 parameters have the highest potential impact on dose and Priority 3 the least. This ranking is documented in Attachment B to the ANL report, NUREG/CR-6697, Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0 Computer Codes, (NUREG/CR-6697) (9).
Priority 3 physical parameters were assigned either a site-specific value or the median values from the Probability Density Functions (PDF) defined in NUREG/CR-6697. Priority 1 and 2 parameters were either assigned site-specific deterministic values, evaluated by uncertainty analysis using the PDFs defined in NUREG/CR-6697, or evaluated by uncertainty analysis using site-specific PDFs. The Partial Rank Correlation Coefficient (PRCC) values from the RESRAD uncertainty analysis were used to evaluate the relative sensitivity of the Priority 1 and 2 parameters. A PRCC value less than -0.25 was considered sensitive and negatively correlated to dose. The 25th percentile of the parameter distribution was assigned to negatively correlated parameters. A PRCC value greater than +0.25 was considered sensitive and positively correlated to dose. The 75th percentile of the parameter distribution was assigned to positively correlated parameters. Priority 1 and 2 parameters with a PRCC absolute value less than 0.25 were assigned the median value of the PDF.
Consistent with the guidance in NUREG-1757, section I.6.4.2, metabolic and behavioral parameters were assigned the mean values from NUREG/CR-5512 Vol. 3, Residual Radioactive Contamination From Decommissioning Parameter Analysis (NUREG/CR-5512) (10)
Table 6.87.
Figure 6-7 provides a flow chart of the parameter selection process. The set of selected deterministic parameters and PDFs is used in a RESRAD Uncertainty Analysis to determine the final deterministic parameter set used to calculate soil DCGLs. The uncertainty analysis was performed for each radionuclide individually. This conservatively disregards the reduced influence of low abundance radionuclides on the total dose and eliminates the potential impact of uncertainty in mixture fractions. The number of sample runs in the uncertainty analysis ranged from 300 to 500 to manage run time.
6.8.2. RESRAD Parameter Selection for Uncertainty Analysis -1 provides a table with the deterministic values and PDFs used for the uncertainty analysis. The references or justifications for the parameter selections are listed in Attachment 6-
- 1. The basis for the behavioral and metabolic parameters (NUREG/CR-5512) and the generic PDFs (NUREG/CR-6697) are straightforward and consistent with the process flow chart in Figure 6-7. The Kd PDFs for the contaminated zone, unsaturated zone and saturated zone were correlated in the uncertainty analysis with Rank Correlation Coefficients of 0.99. The basis for the site-specific deterministic parameters and PDFs are discussed in more detail below.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 The contaminated area was assumed to be the full 7500 m2 area inside the LSE fence. The site-specific soil type is sand. The depth of soil contamination was conservatively assumed to be 1
- m. This allows for a more efficient remediation (if necessary) and FSS process because a single DCGL applies to the soil depths from 0 to 1 m as opposed to requiring separate DCGLs for 0.0 to 0.15 m and 0.15 m to 1.0 m. Site-specific deterministic values from Reference (1) were applied to the following hydrogeological parameters:
Contaminated Zone Hydraulic Conductivity, Soil Density, Soil Porosity, Soil Effective Porosity, Saturated Zone Hydraulic Gradient.
A site-specific deterministic value was also selected for the Saturated Zone Field Capacity parameter based on a calculation performed for a sand soil type in ZionSolutions Technical Support Document 14-006, Conestoga Rovers & Associates (CRA) Report, Evaluation of Hydrological Parameters in Support of Dose Modeling for the Zion Restoration Project (11).
The Inhalation Rate parameter for the industrial worker AMCG was derived from NUREG/CR-5512, Section 5.3.4 which recommends an inhalation rate for workers in light industry of 1.4 m3/hr. The annual inhalation rate the industrial worker was then calculated as follows:
Inhalation Rate (m3/yr) = 1.4 m3/hr*2190 hr/yr = 3066 m3/yr.
A similar process was followed to determine the Drinking Water Intake Rate parameter for the industrial worker. NUREG/CR-5512, Table 6.87, provides a water intake rate of 478 L/yr for a residential user which corresponds to 1.31 L/d. This rate was conservatively applied as the intake rate for a worker as follows: 1.31 L/d
- 250 work days/yr = 327 L/yr.
The RESRAD parameters Indoor and Outdoor Time Fractions were derived from NUREG/CR-6697 Att. C, Table 7.6-1, which recommends a median indoor work day of 8.76 hour8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br />s/day.
Assuming 5 days a week and 50 weeks per year, this equates to 2190 hours0.0253 days <br />0.608 hours <br />0.00362 weeks <br />8.33295e-4 months <br /> per year. The majority of industrial work is expected to be indoors. Consistent with Table 2-3 of the ANL report Users Manual for RESRAD Version 6 (12), 75% of work time is assumed to be indoors and 25% outdoors. The corresponding RESRAD Indoor Fraction parameter =
(2190*0.75)/(24*365) = 0.1875. The Outdoor Time Fraction is then calculated as (2190*0.25)/(24*365) = 0.0625.
Site-specific PDFs were developed for the Well Pump Intake Depth and Well Pumping Rate Parameters. There are two existing onsite industrial water supply wells supporting LACBWR.
The well depths are 116 feet and 129 feet below the ground surface (bgs) (1). The 129 foot depth is 36.3 m below the maximum water table elevation 629 feet. The 36.3 m depth is assumed to be maximum well depth. The minimum well depth is assumed to be represented by a nominal 20 foot screen depth (6.1 m) starting at the maximum water table elevation. The mode of the recommended triangular distribution is assumed to be mid-point between 6.1 m and 36.3 m which is 21.2 m. Note that the site-specific distribution is reasonably similar to the NUREG/CR-6697 distribution values of 6, 10, and 30 for the triangular distribution.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 NUREG/CR-6697 does not provide a recommended value for well pumping rate due to high variability. For an industrial use scenario, the pump rate depends on industry. To ensure that well pumping rate is included in the uncertainty analysis, a nominal uniform distribution was developed. NUREG-6697, Table 3.10-1 applies a sanitary and potable water usage rate for four persons of 328.7 m3/yr. This value is assumed to be the minimum industrial well pumping rate assuming four workers. A nominal maximum rate is assumed based on supply to 20 workers which equates to 1643.5 m3/yr. These minimum and maximum values are not intended to predict actual water use at an unknown future industrial facility on the site after license termination but to provide a range that can be used to determine if the dose is sensitive to well pumping rate.
The remaining parameters in Attachment 6-1 not discussed above were selected in accordance with the process flow chart in Figure 6-7.
6.8.3. Soil DCGL Uncertainty Analysis Results The full RESRAD Uncertainty Reports for Soil DCGL parameters are provided electronically with Reference (8). The uncertainty analysis results are provided in Table 6-4 and Table 6-5.
Table 6-4 Soil DCGL Uncertainty Analysis Results for Kd and Selected Deterministic Values Nuclide Correlation Basis of Selected Nuclide Correlation Basis of Selected to Dose Deterministic Deterministic to Dose Deterministic Deterministic Parameter Value Parameter Value Selection1 (cm3/g) Selection1 (cm3/g) 2 3 H-3 NS mean 0.06 Eu-154 NS mean 825 C-14 NS mean 5 Eu-1553 NS mean 825 Fe-55 NS mean 220 Np-237 Negative 25th 1 Ni-59 Positive 75th 1110 Pu-238 NS mean 550 Co-60 NS mean 60 Pu-239 NS mean 550 Ni-63 Positive 75th 1110 Pu-240 NS mean 550 Sr-90 NS mean 15 Pu-241 NS mean 550 Nb-94 NS mean 160 Am-241 NS mean 1900 Tc-99 Negative 25th 0.04 Am-243 NS mean 1900 Cs-137 NS mean 280 Cm-243 NS mean 4000 Eu-1523 NS mean 825 Cm-244 NS mean 4000 Note 1: Mean values for sand from NUREG/CR-6697 Table 3.9.2. The 75th and 25th values for sand from Reference (13)
Note 2: NS indicates non-sensitive parameter.
Note 3: Sand Kds not listed in NUREG-6697 Table 3.9-2 for this radionuclide. The mean value from NUREG-6697 was used.
The deterministic Kd values assigned to radionuclides that are sensitive to Kd are the 75th or 25th percentile values from the sand Kd distributions in Sheppard and Thibault, Default Soil/Solid
/Liquid Partition Coefficients, Kds, for Four Major Soil Types: A Compendium, (13) which are listed in Reference (8). The Kd values assigned to radionuclides that were not sensitive to Kd were assigned the mean deterministic values for sand from Reference (13) as listed in NUREG-6697, Table 3.9-2.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-5 Soil DCGL Uncertainty Analysis Results for Non-Nuclide Specific Parameters and Selected Deterministic Values Basis of Selected Correlation Deterministic Parameter Radionuclide Deterministic to Dose Parameter Value Selection Contaminated Nb-94 75th Positive 0.0029 zone erosion rate Contaminated NA median NS 0.97 zone b parameter th Evapotranspiration H-3, Tc-99 25 Negative 0.56 coefficient Am-241, C-14, Cm-243, Cm-244, Minimum Site-Fe-55, Ni-59, Ni-63, Pu-238, Pu-239, Specific Wind Speed1 Negative 3.7 Pu-240, Pu-241, Monthly Average1 Negative H-3, Np-237, Tc-99 25th 0.27 Runoff coefficient Saturated zone b NA median NS 0.97 parameter Well pump intake H-3, Np-237, Tc-99 minimum site-Negative 6.1 depth specific value b Parameter of NA median NS 0.97 Unsaturated zone Mass loading for Cm-243, Cm-244, Fe-55, Ni-59, Ni-Positive 75th 2.87E-05 inhalation 63, Pu-238, Pu-239, Pu-240, Pu-241, Indoor dust Cm-243, Cm-244, Fe-55, Ni-59, Ni- 75th Positive 0.75 filtration factor 63, Pu-238, Pu-239, Pu-240, Pu-241 Am-241, Am-243, Cm-243, Co-60, External gamma Positive Cs-137, Eu-152, Eu-154, Eu-155, 75th 0.40 shielding factor Nb-94, Np-237, Pu-241, Sr-90 Well Pumping NA median NS 986.1 Rate Depth of Soil NS NA median 0.23 Mixing Layer (1) Site-specific average wind speed data from Wisconsin State Climatology Office converted to m/s (Web Address:
http://www.aos.wisc.edu/~sco/clim-history/7cities/la_crosse.html) 6.9. Soil Deterministic Analysis and Soil DCGLs The soil DCGLs were calculated using the parameter set provided in Attachment 6-1 with the PDFs replaced by the deterministic values listed in Table 6-4 and Table 6-5. The full RESRAD Summary Report for Soil DCGLs is provided electronically with Reference (8). The Soil DCGLs for the Initial Suite are listed in Table 6-6.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-6 LACBWR Soil DCGLs for Initial Suite Radionuclides Radionuclide Soil DCGL (pCi/g)
H-3 1.746E+04 C-14 2.448E+05 Fe-55 1.018E+07 Ni-59 2.594E+07 Co-60 1.281E+01 Ni-63 9.478E+06 Sr-90 6.586E+03 Nb-94 2.018E+01 Tc-99 3.563E+02 Cs-137 5.812E+01 Eu-152 2.844E+01 Eu-154 2.636E+01 Eu-155 1.122E+03 Np-237 7.991E-01 Pu-238 1.660E+03 Pu-239 1.494E+03 Pu-240 1.496E+03 Pu-241 3.637E+04 Am-241 1.089E+03 Am-243 1.868E+02 Cm-243 2.884E+02 Cm-244 2.668E+03 6.10. Basement Fill Conceptual Model The BFM is used to calculate dose to the industrial worker AMCG from residual radioactivity in the backfilled basements to remain after license termination. A general description of the BFM conceptual model was provided in section 6.5.1. This section describes the conceptual model in more detail including assumed physical configuration, geohydrology, source-term and exposure pathways. The computational model is described in sections 6.11 and 6.12.
The BFM conceptual model assumes that all structures are removed to a depth of three feet bgs (639 foot), i.e., to an elevation of 636 foot, and then backfilled. The site groundwater is in direct communication with the Mississippi river resulting in river driven seasonal groundwater fluctuation. The maximum groundwater elevation is 629 foot. Although the maximum elevation is seasonal, the water table is assumed to be constantly present at 629 foot.
The BFM includes two scenarios; in-situ (designated as Insitu in this Chapter) and Excavation.
The BFM Insitu scenario includes two exposure pathways; ingestion of drinking water from an onsite well and direct exposure to drilling spoils that are assumed to be brought to the surface during the installation of the onsite well. The BFM Excavation scenario assumes large scale industrial excavation of some or all of the backfilled concrete and spreading the concrete over a 1 m layer on the ground surface. The dose from the Insitu and Excavation scenarios are summed in the DCGL calculation.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 6.10.1. Source Term The source term for the BFM is the residual radioactivity remaining in the backfilled basement End State at the time of license termination. LTP Chapter 2 provides the characterization data for the basements that will remain. The dimensions and surface areas are provided in RS-TD-313196-002 Final LACBWR End State Basement Concrete Surface Areas, Volumes, and Void Spaces (14). The expected source term configurations and activity levels projected to remain in each basement are summarized below.
6.10.1.1. Reactor Building The Reactor Building is a right circular cylinder with a hemispherical dome and semi-ellipsoidal bottom. It has an overall internal height of 144 feet and an inside diameter of 60 feet, and it extends 26 feet 6 inches below grade level. The steel shell thickness is 1.16 inch, except for the upper hemispherical dome, which is 0.60 inch thick. The lowest floor elevation is at 612 foot elevation.
The total wall/floor surface area in the portion of the building to be backfilled, i.e., below 636 foot elevation, is 512 m2. T Remediation plans call for all below grade concrete interior to the steel liner to be removed exposing the steel liner. Subsequent to interior concrete removal, the remaining portion of the steel liner will be removed. The remaining structural concrete outside the liner and below the 636 foot elevation will remain. The remaining concrete bowl sits on an external support structure comprised of a concrete pile cap and piles. The pile cap and piles were isolated from reactor operations by the interior concrete, the steel liner and the exterior concrete bowl. There is no evidence of contamination leakage beyond the steel liner or the concrete bowl exterior to the liner, therefore the pile cap and piles are considered to be non-impacted areas.
Six 1.27 cm thick core slices were collected from the surface downward and shipped to an offsite laboratory for analysis. The cores were collected from biased locations as indicated by elevated survey measurements and represent the areas expected to contain the highest levels of contamination. As shown in Table 6-3, Cs-137 is the predominate radionuclide at 88% of the radionuclide mixture. The Cs-137 results from the six cores ranged from 66 pCi/g to 7,500 pCi/g with an average of 1,903 pCi/g.
There is no indication that contamination is present in the concrete to remain after the steel lineris removed, therefore no cores were collected from the concrete outside the steel liner. In addition, general cleanup of loose contamination on the steel liner (concrete dust) after demolition and removal of the internal concrete is expected for operational radiation protection purposes before the steel liner is removed. This reduces the already low potential for transfer of activity in contaminated dust from the steel liner to the underlying concrete during removal of the liner. Therefore, minimal source term is expected in the Reactor Building End State.
6.10.1.2. Waste Gas Tank Vault The WGTV is a 29 foot by 31 foot underground concrete structure with 14 feet high walls and 2 feet thick floors, walls, and ceiling located 3 feet bgs just outside of the WTB. The gas decay system routed main condenser gases through various components for drying, filtering, recombining, monitoring and holdup for decay in the WGTV. The vault floor is at 621 foot elevation. A small sump is present that extends to 618 foot elevation.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Eight concrete cores were collected from the WGTV in September 2017 during continuing characterization. Cs-137 was positively detected in seven of the eight cores with concentrations ranging from 4.4 pCi/g to 240 pCi/g. Very low concentrations of Co-60, Ni-63, Pu-238, Pu-239/240, Pu-241, and Am-241 were sporadically identified in three cores with a range of 0.021 pCi/g to 10.5 pCi/g.
6.10.2. BFM Exposure Pathways The BFM includes two exposure scenarios, Insitu and Excavation. As discussed in section 6.4 the reasonably foreseeable future land use at the LACBWR site is industrial and the AMCG is therefore the industrial worker.
The following exposure pathways are applicable to the BFM scenarios:
Direct exposure to external radiation in as left End State geometry (negligible),
Inhalation of airborne radioactivity in as left End State geometry (negligible),
Ingestion of concrete or fill material in as left End State geometry (negligible),
Ingestion of water from onsite well, Direct exposure, inhalation dose and ingestion dose from contaminated drilling spoils brought to the surface during installation of the onsite well into the fill material, and Direct exposure, inhalation dose and ingestion dose from concrete that is brought to the surface by excavation.
The agricultural and gardening pathways are not applicable to industrial land use. The meat, milk, grain and vegetable ingestion pathways are therefore not included.
6.11. BFM Insitu Scenario 6.11.1. BFM Insitu Groundwater Scenario The BFM Insitu groundwater (BFM Insitugw) conceptual model is based on a conservative screening approach. One hundred percent of the inventory in the backfilled basement concrete is assumed to instantly release and mix with the fill material at the time of license termination with no credit for decay. RESRAD is then used to perform the dose modeling assuming that the source term is in the fill and that the structures provide no resistance to groundwater flow, i.e.,
are not present.
The two remaining basements (Reactor Building and WGTV) have different geometries and contamination potential and are therefore modeled as separate contaminated zones with portions above and below the water table after backfill. This configuration is addressed in the RESRAD model using the contaminated fraction below the water table parameter. The conceptual model assumes full mixing over the fill volume. For the portions of the Reactor Building and WGTV that are below the maximum water table elevation of 629, full mixing over the entire fill volume below 629 is a reasonable assumption given that the fill is saturated and the conceptual model assumes unrestricted groundwater flow. Mixing within the fill that is above the 629 water table elevation is more uncertain because the source of the water is vertical rainwater infiltration as 6-23
La Crosse Boiling Water Reactor License Termination Plan Revision 2 opposed to horizontal flowing groundwater. To evaluate the effect of mixing volume, a sensitivity analysis was conducted to determine the dependence of dose on the mixing distance into the fill (see section 6.15).
6.11.1.1. BFM Insitugw Computation Model RESRAD was used to calculate Dose to Source Ratios (DSR), with units of mrem/yr per pCi/g, for each radionuclide in the initial suite. Separate DSRs were calculated for the Reactor Building and WGTV. The DSRs were then used in conjunction with unitized fill concentrations (i.e., the pCi/g concentration in fill resulting from the release of 1 pCi/m2 from the structure concrete) to calculate DCGLs for the BFM Insitugw scenario in units of pCi/m2. The BFM Insitugw DCGL calculations are provided in Reference (8).
6.11.1.2. BFM Insitugw RESRAD Uncertainty Analysis for Initial Suite An uncertainty analysis was performed to select deterministic RESRAD parameters for the calculation of DSRs for the BFM Insitugw model. The process for determining the input parameters for the BFM Insitugw RESRAD uncertainty analysis was the same as that used for the soil DCGL uncertainty analysis (see process flowchart in Figure 6-7).
The parameter set developed to perform the soil DCGL uncertainty analysis is applicable to the BFM Insitugw analysis with changes to account for the geometries of the backfilled structures.
The affected RESRAD geometry parameters are Cover Depth, Area of Contaminated Zone, Thickness of Contaminated Zone, Length Parallel to Aquifer Flow, Unsaturated Zone Thickness, and Fraction of Contaminated Zone below the Water Table.
The uncertainty analysis was performed separately for the two structures to remain at license termination; i.e., Reactor Building and WTGV. The parameters used for the five BFM Insitugw uncertainty analyses are listed in Attachment 6-2. The parameters listed as Variable in -2 were assigned the values shown in Table 6-7 for the respective structures.
Table 6-7 Deterministic Geometry RESRAD Parameters used in the Uncertainty Analysis for the Two BFM Insitugw Configurations Parameter Rx Building WGTV Above 619 Cover Depth (m) 0.91 0.91 Area of Contaminated Zone (m2) 262.68 86.33 Thickness of Contaminated zone (m) 7.32 4.57 Length Parallel to Aquifer Flow (m) 18.29 9.6 Unsaturated Zone Thickness (m) 0 0 Contaminated Fraction Below the Water Table 0.71 0.53 The uncertainty analysis was performed for each radionuclide individually. This conservatively disregards the reduced influence of low abundance radionuclides on the total dose and eliminates the potential impact of uncertainty in mixture fractions. The RESRAD Uncertainty Reports are provided electronically with Reference (8).
The Kd values are radionuclide-specific. The uncertainty analyses for all radionuclides, except Nb-94 which had a positive PRCC value near zero, showed a negative correlation between dose 6-24
La Crosse Boiling Water Reactor License Termination Plan Revision 2 and Kd. With a few exceptions, the negative correlation exceeded the PRCC threshold of l0.25l.
The predominance of negative correlation with Kd was expected because the majority of the contamination is below the water table and the primary dose pathway in the BFM Insitugw scenario is through the ingestion of well water. To ensure conservatism, the deterministic Kd values selected for all negatively correlated radionuclides, in both the Reactor Building and WGTV, were the 25th percentile values from Reference (13). The positive correlation to the Nb-94 Kd was investigated further primarily because a cross-check of the Standardized Rank Regression Coefficient (SRRC) indicated more significant positive correlation than the PRCC.
The cause of the positive correlation was found to be direct dose from the water independent pathway after long term cover erosion. The maximum Nb-94 dose occurs when the 75th percentile Kd is applied, as opposed to the 25th percentile, for both the Reactor Building and WGTV. The time of maximum dose is year 312. This result is due to the unique decay characteristics of Nb-94 which include gamma emission with a very long half-life. The 75th percentile Kd was applied to ensure conservatism.
The parameter distributions for the site-specific soil type of sand from Reference (13) were used to generate the 25th and 75th percentile deterministic values. Note that Reference (13) does not contain values for Europium; the 25th percentile from the NUREG-6697 Kd parameter distribution was used. The assigned Kd values for all radionuclides are listed in Table 6-8.
Table 6-8 BFM Insitugw Reactor Building and WGTV Deterministic Values Selected for Distribution Coefficients (Kd)
Kd in Contaminated Zone and Saturated Zone (No Unsaturated Zone Present)
Radionuclide Correlation Value (cm3/g)
H-3 Negative 0.05 C-14 Negative 1.8 Fe-55 Negative 38 Ni-59 Negative 147 Co-60 Negative 9 Ni-63 Negative 147 Sr-90 Negative 5 Nb-94 Positive 611 Tc-99 Negative 0.04 Cs-137 Negative 50 Eu-1521 Negative 95 Eu-1541 Negative 95 Eu-1551 Negative 95 Np-237 Negative 1 Pu-238 Negative 173 Pu-239/240 Negative 173 Pu-241 Negative 173 Am-241 Negative 329 Am-243 Negative 329 Cm-243/244 Negative 881 6-25
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-9 and Table 6-10 provide the uncertainty analysis results for the parameters that are not radionuclide-specific (i.e., all parameters other than Kd) and the selected deterministic values. If a parameter is sensitive for any radionuclide the corresponding deterministic value (75th or 25th percentile depending on the correlation) is assigned to all radionuclides. This approach can be used since there are no inconsistencies in the results of the uncertainty analysis, i.e., no parameters were positively correlated for one radionuclide and negatively correlated for another.
Table 6-9 Reactor Building: BFM Insitugw Uncertainty Analysis Results and Deterministic Values Selected for Non-Nuclide Specific Parameter Distributions Correlation Radionuclide Basis of Selected to Dose1 Deterministic Parameter Deterministic Parameter Value Selection Contaminated zone erosion rate NS NA median 0.0015 Contaminated zone b parameter NS NA median 0.97 Evapotranspiration coefficient NS NA median 0.62 Wind Speed 2 NS NA median 4.5 Runoff coefficient NS NA median 0.45 Saturated zone b parameter NS NA inactive NA Negative H-3, C-14, Fe-55, Ni-59, Co-60, Ni-63, Sr-90, Tc-99, Cs-137, Eu-152, Eu-25th Percentile 154, Eu-155, Np-Well pump intake depth 6.1 237, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, Am-243, Pu-241, Cm-243, Cm-244 Mass loading for inhalation NS NA median 2.35E-05 Indoor dust filtration factor NS NA median 0.55 External gamma shielding factor NS NA median 0.27 Well Pumping Rate NS NA median 986.1 Depth of Soil Mixing Layer NS NA median 0.15 Positive Nb-94, Cs-137, Eu- th 75 Percentile Cover Erosion Rate 152, Eu-154, Am- 0.0029 243 (1) NS = Not Sensitive (2) Site-specific annual average wind speed from Wisconsin State Climatology Office converted to m/s ( Web Address http://www.aos.wisc.edu/~sco/clim-history/7cities/la_crosse.html) 6-26
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-10 WGTV: BFM Insitugw Uncertainty Analysis Results and Deterministic Values Selected for Non-Nuclide Specific Parameter Distributions Correlation Radionuclide Basis of Selected to Dose1 Deterministic Parameter Deterministic Parameter Value Selection Contaminated zone erosion rate NS NA Median 0.0015 Contaminated zone b parameter NS NA Median 0.97 Evapotranspiration coefficient NS NA Median 0.62 Wind Speed 2 NS NA Median 4.5 Runoff coefficient NS NA Median 0.45 Saturated zone b parameter NS NA Inactive NA Negative H-3, C-14, Fe-55, Ni-59, Co-60, Ni-63, Sr-90, Tc-99, Cs-137, Eu-152, Eu-Minimum 154, Eu-155, Np-Well pump intake depth Depth 6.1 237, Pu-238, Pu-239, Pu-240, Pu-241, Am-241, Am-243, Cm-243, Cm-244 Mass loading for inhalation NS NA Median 2.35E-05 Indoor dust filtration factor NS NA Median 0.55 External gamma shielding factor NS NA Median 0.27 Well Pumping Rate NS NA Median 986.1 Depth of Soil Mixing Layer NS NA Median 0.15 Positive Nb-94, Cs-137, Eu-75th Percentile Cover Erosion Rate 152, Eu-154, Am- 0.0029 241, Am-243 (1) NS = Not Sensitive (2) Site-specific annual average wind speed data from Wisconsin State Climatology Office converted to m/s (Web Address:
http://www.aos.wisc.edu/~sco/clim-history/7cities/la_crosse.html) 6.11.1.3. BFM Insitugw RESRAD Deterministic Analysis and DSR Results As discussed above, the BFM Insitugw RESRAD dose assessments were performed separately for the Reactor Building and WGTV. The parameters provided in Attachment 6-2 were applied to both analyses with the two exceptions; 1) the deterministic values in Table 6-7 were used to replace the parameters listed as variable in Attachment 6-2 and 2) the values in Table 6-8, Table 6-9 and Table 6-10 provide the uncertainty analysis results for the parameters that are not radionuclide-specific (i.e., all parameters other than Kd) and the selected deterministic values. If a parameter is sensitive for any radionuclide the corresponding deterministic value (75th or 25th percentile depending on the correlation) is assigned to all radionuclides. This approach can be used since there are no inconsistencies in the results of the uncertainty analysis, i.e., no parameters were positively correlated for one radionuclide and negatively correlated for another.
Table 6-9, and Table 6-10 were used to replace the PDFs listed in Attachment 6-2.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 The RESRAD Uncertainty Reports are provided electronically with Reference (8). The resulting DSRs are provided in Table 6-11.
Table 6-11 BFM Insitugw DSRs for Reactor Building and WGTV Radionuclide Rx Building WGTV (mrem/yr (mrem/yr per pCi/g) per pCi/g)
H-3 5.916E-03 2.703E-03 C-14 1.581E-01 8.069E-02 Fe-55 3.787E-03 1.852E-03 Ni-59 3.940E-04 2.032E-04 Co-60 6.691E-01 3.759E-01 Ni-63 1.079E-03 5.003E-04 Sr-90 6.440E+00 3.661E+00 Nb-94 1.546E-01 6.640E-02 Tc-99 1.355E-01 6.195E-02 Cs-137 2.674E-01 1.295E-01 Eu-152 1.824E-02 8.675E-03 Eu-154 2.649E-02 1.260E-02 Eu-155 4.112E-03 1.955E-03 Np-237 3.887E+02 1.854E+02 Pu-238 5.089E+00 2.398E+00 Pu-239 5.652E+00 2.911E+00 Pu-240 5.651E+00 2.907E+00 Pu-241 1.090E-01 5.311E-02 Am-241 3.066E+00 1.499E+00 Am-243 3.054E+00 1.567E+00 Cm-243 7.831E-01 3.658E-01 Cm-244 6.260E-01 2.924E-01 6.11.1.4. BFM Insitugw DCGL The BFM Insitugw DCGLs were calculated in Reference (8) using Equation 6-5.
Equation 6-5 25
()() =
()
Where:
BFM Insitu (gw) DCGL = DCGL for radionuclide (i) (pCi/m2)
DSR (i) = RESRAD Dose to Source Ratio for radionuclide i (mrem/yr per pCi/g)
Unit Fill Concentration = Basement-specific concentration in fill (pCi/g) assuming release of unit inventory of 1 pCi/m2 from concrete (pCi/g per pCi/m2) 6-28
La Crosse Boiling Water Reactor License Termination Plan Revision 2 The BFM Insitugw DCGLs are listed in Table 6-12. Note that the DCGLs for the initial suite radionuclides provided in Table 6-12 (and in Table 6-13 and Table 6-15) are calculated values only and do not indicate the concentrations relative to Cs-137 that are present or could be present at license termination. The site-specific radionuclide mixture fractions are provided in Table 6-3.
Table 6-12 BFM Insitugw DCGLs1 Rx Building WGTV BFM Insitugw DCGL BFM Insitugw DCGL Radionuclide (pCi/m2) (pCi/m2)
H-3 2.16E+10 1.60E+10 C-14 8.08E+08 5.35E+08 Fe-55 3.37E+10 2.33E+10 Ni-59 3.24E+11 2.13E+11 Co-60 1.91E+08 1.15E+08 Ni-63 1.18E+11 8.63E+10 Sr-90 1.98E+07 1.18E+07 Nb-94 8.26E+08 6.50E+08 Tc-99 9.43E+08 6.97E+08 Cs-137 4.78E+08 3.34E+08 Eu-152 7.00E+09 4.98E+09 Eu-154 4.82E+09 3.43E+09 Eu-155 3.11E+10 2.21E+10 Np-237 3.29E+05 2.33E+05 Pu-238 2.51E+07 1.80E+07 Pu-239 2.26E+07 1.48E+07 Pu-240 2.26E+07 1.49E+07 Pu-241 1.17E+09 8.13E+08 Am-241 4.17E+07 2.88E+07 Am-243 4.18E+07 2.76E+07 Cm-243 1.63E+08 1.18E+08 Cm-244 2.04E+08 1.48E+08 Note 1: The DCGLs for the initial suite radionuclides are calculated values only and do not indicate the concentrations relative to Cs-137 that are present or could be present at license termination. The site-specific radionuclide mixture fractions are provided in Table 6-3.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 6.11.2. BFM Insitu Drilling Spoils Scenario and DCGL Calculation The BFM Insitu Drilling Spoils (BFM Insituds) scenario addresses one of the BFM exposure pathways listed in section 6.10.2. The dose from residual radioactivity in the concrete is assumed to be brought to the surface during the installation of a well that randomly hits backfilled structural concrete. The driller is assumed to be unaware that the backfilled structure is present. The residual radioactivity in the concrete surfaces is brought to the surface with the drilling spoils which includes the fill material above the structure floor. 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 the fill. The BFM Insituds DCGLs are calculated with units of pCi/m2.
There are a number of ways that installers handle and dispose of drilling spoils, including the use of slurry pits, tanks, and dumping the drilling spoils on the existing surface soils. The use of pits would likely involve additional dilution by refilling the pit with the material excavated during its construction. As a conservative assumption, no dilution of the spoil material is assumed after being brought to the surface.
The well is assumed to be drilled into the basement fill down to the concrete floor where refusal is met and drilling stopped. The extent of drilling into concrete is assumed to be sufficient to capture 100 percent of the remaining residual radioactivity in the concrete, at all depths, within the borehole area. The volume of spoil material brought to the surface is calculated based on the borehole diameter and depth of drilling which is conservatively assumed to be the minimum fill depth of three feet which minimizes the mixing volume for both basements. The concrete and fill are uniformly mixed and spread over a circular area on the ground surface to a depth of 0.15
- m. The three-foot drilling depth represents the minimum backfill depth after demolition to three feet below grade.
Both basement floors are greater than three feet deep and therefore assuming a three foot well depth is conservative for all basements. Because the same fill depth was assumed for both basements the BFM Insituds DCGLs are the same for both basements.
The dose from the circular area at the surface was calculated by RESRAD using the surface soil DCGLs and Area Factors (AF). The AFs were calculated using the deterministic parameters applied for soil DCGLs and the spoils spread area which was determined to be 0.457 m 2. The RESRAD Summary Reports are provided electronically with Reference (8).
The AFs and BFM Insituds DCGLs were calculated in Reference (8) using the RESRAD results.
The BFM Insituds DCGLs are listed in Table 6-13.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-13 BFM Insitu Drilling Spoils DCGLs for Both Basements1 BFM Insituds DCGLs for Both Basements (Reactor Building and WGTV)
Radionuclide (pCi/m2)
H-3 5.09E+13 C-14 4.40E+15 Fe-55 4.53E+16 Ni-59 6.15E+16 Co-60 6.00E+08 Ni-63 2.52E+16 Sr-90 3.41E+11 Nb-94 8.72E+08 Tc-99 1.37E+12 Cs-137 2.45E+09 Eu-152 1.27E+09 Eu-154 1.19E+09 Eu-155 3.44E+10 Pu-238 9.83E+09 Np-237 5.59E+11 Pu-239 5.07E+11 Pu-240 5.11E+11 Pu-241 4.95E+12 Am-241 9.68E+10 Am-243 7.21E+09 Cm-243 1.17E+10 Cm-244 8.90E+11 Note 1: The DCGLs for the initial suite radionuclides are calculated values only and do not indicate the concentrations relative to Cs-137 that are present or could be present at license termination. The site-specific radionuclide mixture fractions are provided in Table 6-3.
6.12. BFM Excavation Scenario The BFM Excavation scenario assumes that some or all of the backfilled structures are excavated and the concrete spread on the surface immediately after license termination taking no credit for decay. A typical excavation process for a backfilled structure would entail using a medium sized excavator with a 1.0 to 1.5 cubic yard bucket to excavate and stockpile fill. After removing the fill to the planned excavation depth, a hoe-ram would be used to pound out the concrete walls and floor (if the excavation reaches the floor). The concrete would be segregated, the rebar removed, and remaining concrete size reduced. The excavation scenario assumes that the size reduced concrete is used as onsite fill. Large-scale industrial excavation of the entire basement may require different methods but the result would be the same, i.e., a volume of sized concrete to be used as onsite fill.
The DCGLs are calculated to ensure that the average radionuclide concentrations in the excavated, mixed, and sized concrete do not exceed the soil DCGLs which is a conservative 6-31
La Crosse Boiling Water Reactor License Termination Plan Revision 2 approach given that the surface area of excavated concrete, assuming 1 m spread depth, would be less than the 7500 m2 area assumed in the calculation of the soil DCGL. The assessment provides BFM Excavation DCGLs in units of pCi/m2. Due to differences in configuration and contamination potential, the BFM Excavation DCGLs were calculated separately for the Reactor Building and WGTV in the same manner as was done for the BFM Insitu assessments.
The radionuclide concentrations (pCi/g) in the inadvertently mixed, excavated concrete that is assumed to be spread on the site surface is a linear function of the ratio of concrete surface area (SA) to concrete volume (V). The SA/V ratio was calculated in two ways; 1) assuming full excavation of the entire basement, and 2) assuming partial excavation that includes only the walls with the minimum thickness (0.75 feet for both the Reactor Building and WGTV). The walls with minimum thickness will have the maximum SA/V ratio and will result in the maximum concentration in the excavated concrete.
As seen in Table 6-14Error! Reference source not found., the SA/V ratios for the partial excavation of the minimum thickness wall is greater than the SA/V ratio assuming full excavation. To ensure conservatism, the maximum SA/V ratio was used in the DCGL calculation which results in the maximum radionuclide concentrations in the excavated concrete.
Table 6-14 BFM Concrete Excavation SA/V Ratios for Full and Partial Excavation Partial Excavation SA/V Full Excavation SA/V Structure (Minimum Wall Thickness) Partial SA/V Full SA/V (m2/m3)
(m2/m3)
Waste Gas 2.55 4.37 1.72 Tank Vault Reactor 0.95 4.37 4.61 Building The BFM Excavation DCGLs are calculated in Reference (8). The results are listed in Table 6-15 and are the same for the Reactor Building and WGTV because the concentrations in the excavated concrete are both based on the same partial (i.e., maximum) SA/V ratio of 4.37.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-15 BFM Excavation DCGLs for Both Basements1 BFM Excavation DCGLs for Both Basements Radionuclide (Reactor Building and WGTV)
(pCi/m2)
H-3 9.38E+09 C-14 1.32E+11 Fe-55 5.47E+12 Ni-59 1.39E+13 Co-60 6.88E+06 Ni-63 5.09E+12 Sr-90 3.54E+09 Nb-94 1.08E+07 Tc-99 1.91E+08 Cs-137 3.12E+07 Eu-152 1.53E+07 Eu-154 1.42E+07 Eu-155 6.03E+08 Np-237 4.29E+05 Pu-238 8.92E+08 Pu-239 8.03E+08 Pu-240 8.04E+08 Pu-241 1.95E+10 Am-241 5.85E+08 Am-243 1.00E+08 Cm-243 1.55E+08 Cm-244 1.43E+09 Note 1: The DCGLs for the initial suite radionuclides are calculated values only and do not indicate the concentrations relative to Cs-137 that are present or could be present at license termination. The site-specific radionuclide mixture fractions are provided in Table 6-3.
6.13. Basement Summation DCGL The DCGLs to be used for compliance during FSS of basements (the MARSSIM DCGLw value) are designated as DCGL Basement (DCGLB) and represent the summation of the dose from the three scenarios. The DCGLB is calculated in Reference (8) using Equation 6-6. The DCGLB values for the initial suite are provided in Table 6-16.
6-33
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Equation 6-6 1
,, =
31[1 ]
Where:
DCGLB,i,j = DCGLB for radionuclide (i) and basement (j)
DCGLBS =DCGL Basement Scenarios for (i) (Groundwater, Drilling Spoils and Excavation from Table 6-12, Table 6-13 and Table 6-15) and basement (j)
Table 6-16 Summed Basement DCGL (DCGLB) for Initial Suite Radionuclides Rx Building WGTV Radionuclide DCGLB DCGLB (pCi/m2) (pCi/m2)
H-3 6.54E+09 5.91E+09 C-14 8.03E+08 5.33E+08 Fe-55 3.35E+10 2.32E+10 Ni-59 3.17E+11 2.09E+11 Co-60 6.57E+06 6.42E+06 Ni-63 1.16E+11 8.49E+10 Sr-90 1.97E+07 1.18E+07 Nb-94 1.06E+07 1.05E+07 Tc-99 1.59E+08 1.50E+08 Cs-137 2.90E+07 2.82E+07 Eu-152 1.51E+07 1.51E+07 Eu-154 1.40E+07 1.39E+07 Eu-155 5.81E+08 5.77E+08 Np-237 1.86E+05 1.51E+05 Pu-238 2.44E+07 1.77E+07 Pu-239 2.20E+07 1.46E+07 Pu-240 2.20E+07 1.46E+07 Pu-241 1.11E+09 7.81E+08 Am-241 3.89E+07 2.75E+07 Am-243 2.94E+07 2.16E+07 Cm-243 7.89E+07 6.66E+07 Cm-244 1.79E+08 1.34E+08 The DCGLs for the individual Basement Scenarios (DCGLBS) in Table 6-12, Table 6-13 and Table 6-15 are all larger than the DCGLB values for the same basement in Table 6-16. This is 6-34
La Crosse Boiling Water Reactor License Termination Plan Revision 2 because the DCGLBS represent the concentration that results in a dose of 25 mrem/yr for each scenario individually. For conservatism, the DCGLB is used for FSS which is the concentration that leads to a dose of 25 mrem/yr under the assumption that all three exposure scenarios occur simultaneously, which is not physically possible.
6.14. Insignificant Dose Contributors, Radionuclides of Concern and Surrogate Ratio NUREG-1757, section 3.3 states that radionuclides contributing no greater than 10% of the dose criterion (i.e., 2.5 mrem/yr) are considered to be insignificant contributors (IC). The 10%
criterion applies to the sum of the dose contributions from the aggregate of IC radionuclides.
This section provides the assessment of dose contributions from the initial suite and identifies the radionuclides that can be designated as insignificant dose contributors.
The radionuclides remaining after the removal of IC radionuclides were designated as the ROC.
The ROC list is determined for concrete using the BFM DCGLB values and for soil using the soil DCGLs. The dose from the removed ICs was accounted for by adjusting the soil DCGLs and BFM DCGLs for each ROC. The IC radionuclides are excluded from further detailed evaluations during FSS and demonstration of compliance with the 25 mrem/yr dose criteria.
6.14.1. Insignificant Contributor Dose and Radionuclides of Concern The soil DCGLs in Table 6-6 and BFM DCGLB values in Table 6-16 were used in Reference (5) to calculate the dose contributions from the initial suite radionuclides, to identify the IC dose percentage, and to select the ROCs. The IC percentages represent the percentage of the 25 mrem/yr dose limit. The IC percentages were calculated using the 75th percentile mixture fractions as listed in Table 6-3.
The Relative Dose Fraction, RDFi,k, for radionuclide i and media k is calculated using the applicable DCGLs, the radionuclide activity fractions from Table 6-3, and Equation 6-7. The results of the calculations are provided in Table 6-17.
Equation 6-7
, 1
, =
()
[ , ]
Where:
RDFi,k = Relative dose fraction for radionuclide I and media k (soil, Reactor Building and WGTV) fAi,k = Activity fraction of radionuclide I and media k DCGLi,k = DCGL for radionuclide I and media k Table 6-17 Relative Dose Fractions, RDFi,k Radionuclide ROC ? Combined Relative Doses, RDFi,k 6-35
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Soil Rx Bldg WGTV H-3 8.27E-04 1.26E-04 2.33E-03 C-14 6.69E-07 5.51E-05 9.61E-04 Fe-55 2.22E-07 1.46E-05 -1.91E-05 Ni-59 2.72E-09 2.73E-08 1.24E-05 Co-60 Y 2.55E-01 2.43E-01 4.05E-02 Ni-63 2.66E-06 8.35E-05 1.22E-04 Sr-90 Y 7.57E-04 1.34E-01 4.24E-02 Nb-94 7.96E-04 3.51E-04 5.22E-03 Tc-99 5.52E-04 4.72E-04 2.51E-03 Cs-137 Y 7.25E-01 5.92E-01 8.70E-01 Eu-152 Y 9.82E-03 4.25E-03 1.63E-02 Eu-154 Y 5.42E-03 6.22E-03 6.29E-03 Eu-155 1.77E-04 3.96E-05 4.32E-04 Np-237 2.56E-04 4.06E-04 0.00E+00 Pu-238 6.65E-05 3.24E-03 2.46E-03 Pu-239/240 4.98E-05 5.01E-03 7.15E-04 Pu-241 4.08E-05 1.44E-03 1.64E-03 Am-241 3.12E-04 9.27E-03 6.46E-03 Am-243 2.99E-04 7.31E-04 1.15E-03 Cm-243/244 5.48E-05 6.99E-05 1.46E-04 Sum 1.00E+00 1.00E+00 1.00E+00 ROC Dose Fraction 9.97E-01 9.79E-01 9.76E-01 IC Dose Fraction 3.44E-03 2.13E-02 2.42E-02 In addition to the relative dose fraction for each radionuclide, Table 6-17 shows the relative doses for the ROC and IC radionuclides as ROC Dose Fraction and IC Dose Fraction respectively. This shows that the IC dose fraction ranges from approximately 0.3% to 2.42%
(not including negative values) of the total dose. As noted above, the activity fraction for Np-237 was set to zero for the WGTV in Table 6-3 and must be included in the final calculation of the IC dose fraction.
As discussed above, Np-237 was not detected in any of the 38 samples. Table 6-18 provides summary detection statistics for all samples and for the WGTV.
Table 6-18 Np-237 Detection Statistics Parameter Value pCi/g Min MDC All Samples 0.015 Min MDC WGTV 0.078 Max MDC All Samples 0.239 Max MDC WGTV 0.239 Min Result -0.062 Max Result 0.102 Average Result -0.002 The maximum MDC among all samples is 0.239 pCi/g and was used to calculate the Np-237 IC dose fraction for the WGTV (see Reference (5)). The mass-based core concentration (pCi/g) is 6-36
La Crosse Boiling Water Reactor License Termination Plan Revision 2 converted to areal-based concentration (pCi/m2) by assuming a contamination depth of 1.27 cm (0.5 inch) corresponding to the thickness of each concrete puck and a concrete density of 2.35 g/cm3. The 0.239 pCi/g MDC corresponds to 7.13E+3 pCi/m2. From Table 6-16, the WGTV DCGLB is 1.51E+5 pCi/m2. Using Equations 6-8 and 6-9, the dose fraction from the IC nuclides, DFIC,WGTV, is calculated by adding the dose contribution of Np-237, TEDENp-237 (1.18 mrem/yr), to the remaining IC dose fractions, DFIC/ scaled to 25 mrem (25*DFIC/). The resulting DFIC,WGTV is 0.0729.
Equation 6-8 237()(),
, =
237(),
Equation 6-9 25 + 237
, = = 0.0729 25 Where:
DFIC,WGTV = Dose fraction of IC radionuclides in the WGTV excluding Np-237 RDFi,WGTV = Relative dose fraction for radionuclide i in WGTV DFIC,WGTV = IC dose fraction including Np-237 dose (0.0729)
TEDE,Np-237 = Total Effective Dose Equivalent for Np-237 in WGTV (1.18 mrem/yr)
As shown in Table 6-17 and the results of IC dose fraction calculation for the WGTV using Equation 6-9, the IC dose percentage for soil, Reactor Building and WGTV ranged from 0.3% to 7.3%. To provide additional margin beyond the application of the 75th percentile mixture and the conservative approach used to calculate Np-237 dose for the WGTV, the IC dose percentage assigned to adjust the DCGLs for the ROC in all media (soil, basements, buried pipe and above grade building) is increased to 10%.
The ROC are designated as Cs-137, Co-60, Sr-90, Eu-154, and Eu-152. Table 6-19 provides the radionuclide mixture for the ROC renormalized to 1.00.
Table 6-19 ROC Mixture Fractions Renormalized to 1.00 Soil Mix Rx Bldg Mix WGTV Mix Radionuclide Fraction Fraction Fraction Co-60 6.44E-02 7.41E-02 1.01E-02 Sr-90 9.81E-02 1.23E-01 1.94E-02 Cs-137 8.29E-01 7.96E-01 9.57E-01 Eu-152 5.49E-03 2.97E-03 9.56E-03 Eu-154 2.81E-03 4.04E-03 3.42E-03 Sum 1.0 1.0 1.0 6-37
La Crosse Boiling Water Reactor License Termination Plan Revision 2 6.14.2. Surrogate Radionuclide Ratio The FSS will be performed using gamma spectroscopy. The ROC include one radionuclide that is not a gamma emitter, i.e., Sr-90. As discussed in LTP Chapter 5, the Sr-90 concentration will be accounted for using a surrogate approach during FSS. The ratio of Sr-90/Cs-137 from the site radionuclide mixture is required to implement the surrogate approach.
The Sr-90/Cs-137 surrogate is calculated separately for the Reactor Building, WGTV, and soil in section 6.4 of Reference (5) using applicable core data. When Sr-90 was not detected, the MDC was assigned. The result of one core sample was non-detect for both Cs-137 and Sr-90. The ratio of the MDCs is not used since this is merely a ratio of the detectability of the two nuclides for that specific sample and has no relationship to the activity ratio. The 95th percentile of the data sets are assigned as the surrogate ratios for use during FSS as listed in Table 6-20.
Table 6-20 Sr-90 to Cs-137 Surrogate Ratios to be applied during FSS (95th Percentile values)
Sr-90/Cs-137 Building or Area Surrogate Activity Ratio WGTV 6.75E-02 Reactor Building 5.00E-01 Soil1 5.02E-01 Note 1: The soil designation represents all concrete core bores and can also be used for other miscellaneous structures to remain if needed.
6.15. BFM Groundwater Scenario Mixing Volume Sensitivity Analysis for ROC The BFM DCGLgw are based on a conceptual model that assumes instant release of all activity and uniform mixing throughout the entire fill volume. The concentration in fill is a function of the mixing volume which is proportional to the distance away from the wall that the released activity is assumed to mix. The fill concentration increases with decreasing mix distance but the source term geometries, and dose per pCi/g in fill, also change with mixing distance. A sensitivity analysis was performed for the ROC to evaluate the impact of mixing distance on the dose from the BFM Groundwater scenario. Mixing distance does not impact the dose from the BFM Drilling Spoils scenario which assumes all activity remains in the concrete.
The mixing distance sensitivity analyses for the BFM Groundwater scenario were performed assuming that released activity mixes over distances of 1 m, 2 m, and 3 m from the floor and wall surfaces, as opposed to full mixing. A mixing distance of 1 m was assumed to conservatively represent a minimum width from which a well drawdown zone would not include dilution with adjacent, uncontaminated, groundwater.
Separate analyses were performed for the Reactor Building and the WGTV. For the WGTV, three fill mixing geometries were evaluated including; 1) mixing in fill adjacent to the two walls perpendicular to groundwater flow, 2) mixing in fill adjacent to a wall parallel to groundwater flow, and 3) mixing in fill adjacent to the floor. RESRAD modeling was performed with the deterministic parameters used to calculate the BFM Insitugw DCGLs (see section 6.11.1) but 6-38
La Crosse Boiling Water Reactor License Termination Plan Revision 2 changing the source term parameters to represent the three mixing geometries, for 1, 2, and 3 m distances as shown in Table 6-21.
Table 6-21 WGTV Mixing Sensitivity RESRAD Source Term Geometries Parameter Mixing Perpendicular Parallel Wall Floor Distance Walls Area of 1 18 9.6 86.4 Contaminated 2 36 19.2 86.4 Zone (m2) 3 54 28.8 86.4 Thickness of 1 4.57 4.57 1 Contaminated 2 4.57 4.57 2 zone (m) 3 4.57 4.57 3 Length Parallel to 1 2 9.6 9.6 Aquifer Flow (m) 2 4 9.6 9.6 3 6 9.6 9.6 The values in Table 6-21 are directly related to the WGTV structure geometry with the exception of the values listed for the perpendicular walls. There are two walls perpendicular to groundwater flow and the activity in each wall is assumed to mix in the layer of fill immediately adjacent to the wall. The fill layers from each of the two perpendicular walls are assumed to be contiguous resulting in a length parallel to flow of 2 times the mixing distance.
The Reactor Building is circular as opposed to rectangular. Therefore, there is no geometry equivalent to the Parallel Wall geometry in the WGTV. The Perpendicular Walls and Floor geometries were evaluated. The Reactor Building mix sensitivity source term geometries for 1, 2, and 3 m distances are provided in Table 6-22.
Table 6-22 Reactor Building Mixing Sensitivity RESRAD Source Term Geometries Parameter Mixing Perpendicular Floor Distance Walls Area of 1 102.34 262.68 Contaminated 2 179.55 262.68 Zone (m2) 3 231.62 262.68 Thickness of 1 7.32 1 Contaminated 2 7.32 2 zone (m) 3 7.32 3 Length Parallel to 1 2 18.29 Aquifer Flow (m) 2 4 18.29 3 6 18.29 The RESRAD Summary Reports for the analyses of the 15 mixing sensitivity geometries in Table 6-21 and Table 6-22 are provided electronically in Reference (8).
The impact of mixing distance on dose was evaluated by calculating the ratio of the dose factor (mrem/year per pCi/m2) for each partial mixing distance (1, 2, and 3 m) to the dose factor under the full mixing assumption. The maximum values of the ratios of Partial/Full mixing dose, for all ROC, and all 15 geometries listed in Table 6-21 and Table 6-22 are provided in Table 6-23. The dose from the perpendicular wall and floor geometries were summed, for each partial mixing distance, to determine the maximum ratio because both the floor and perpendicular wall fill 6-39
La Crosse Boiling Water Reactor License Termination Plan Revision 2 source terms can contribute to a well simultaneously under partial mixing. The parallel wall source term (WGTV only) includes the fill from the top of the wall down to the floor. The residual radioactivity in the floor is therefore included in the calculation of fill concentration.
The well is assumed to be placed at the downstream edge of the parallel wall source term.
Table 6-23 Mixing Sensitivity Analysis Results Summary Maximum Ratio of Dose Factor Partial Mix/Full Mix ROC Rx Building WGTV Wall + Floor Wall + Floor Co-60 1.25 1.19 Cs-137 1.91 1.42 Sr-90 1.08 1.18 Eu-152 2.03 1.41 Eu-154 2.03 1.40 To ensure a conservative and bounding DCGLgw for ROCs in backfilled basement surfaces, the DCGLgw values calculated for each ROC under the full mixing assumption (Table 6-12) were reduced by the ratios in Table 6-23 (see section 6.16).
6.16. BFM DCGLs for ROC Adjusted for Insignificant Contributor Dose and Mixing Sensitivity The DCGLBS for each ROC, each basement, and each dose scenario, are adjusted for the IC dose percentage of 10% and mixing sensitivity. The adjusted DCGLBS are calculated using Equation 6-10.
Equation 6-10
, =
, = ,
, =
Where:
DCGLBS,i GW = BFM Insitu Groundwater, for ROC (i), adjusted for IC dose and mixing sensitivity BFM Insitugw,i DCGL = BFM Insitugw DCGL, for ROC (i), from Table 6-12 IC Dose Adjust = Insignificant contributor adjustment factor of 0.9 (equal to 1.0 minus the assigned IC dose of 0.1)
MS Ratioi = Maximum Mixing Sensitivity ratio, for ROC (i), from Table 6-23 DCGLBS,i DS = DCGLBS, for ROC (i), adjusted for IC dose.
BFM Insituds,i DCGL = BFM Insituds DCGL, for ROC (i), from Table 6-13 DCGLBS,i Excavation = BFM Excavation DCGL, for ROC (i), adjusted for IC dose BFM Excavation DCGLi = BFM Excavation DCGL, for ROC (i), from Table 6-15 The ROC DCGLs were also adjusted to account for the maximum Resident Gardener Alternate Scenario dose of 28.4 mrem/yr and 34.9 mrem/yr for the Reactor Building and WGTV, 6-40
La Crosse Boiling Water Reactor License Termination Plan Revision 2 respectively (see section 6.18.2). The Alternate Scenario adjustment was made by multiplying the DCGLs by additional factors of 0.88 (=25/28.4) and 0.72 (=25/34.9) for the Reactor Building and WGTV, respectively. The results are provided in Table 6-24 and Table 6-25.
Table 6-24 Reactor Building BFM DCGLs for ROC Individual BFM Scenarios (DCGLBS) Adjusted for Insignificant Contributor Dose Fraction, Mixing Sensitivity and Alternate Scenario Dose Reactor Building Adjusted BFM DCGLBS ROC (pCi/m2)
Insitu GW Insitu Drilling Spoils Excavation Co-60 1.21E+08 4.75E+08 5.45E+06 Sr-90 1.46E+07 2.70E+11 2.80E+09 Cs-137 1.98E+08 1.94E+09 2.47E+07 Eu-152 2.73E+09 1.00E+09 1.21E+07 Eu-154 1.88E+09 9.43E+08 1.12E+07 Table 6-25 WGTV BFM DCGLs for ROC Individual BFM Scenarios (DCGLBS)
Adjusted for Insignificant Contributor Dose Fraction, Mixing Sensitivity and Alternate Scenario Dose WGTV Adjusted BFM DCGLBS ROC (pCi/m2)
Insitu GW Insitu Drilling Spoils Excavation Co-60 6.23E+07 3.86E+08 4.43E+06 Sr-90 6.42E+06 2.20E+11 2.28E+09 Cs-137 1.52E+08 1.58E+09 2.01E+07 Eu-152 2.28E+09 8.16E+08 9.84E+06 Eu-154 1.57E+09 7.67E+08 9.12E+06 The DCGLs to be used for compliance during the FSS of basements (the MARSSIM DCGLw value) are the DCGLB values which represent the summation of the dose from the three scenarios. The adjusted DCGLB values for each ROC and each basement are calculated using Equation 6-6 with the adjusted DCGLBS values in Table 6-24 and Table 6-25. The results are provided in Table 6-26.
Applying the DCGLB values for compliance is conservative since the summation includes dose from the BFM Insitu scenarios and the BFM Excavation scenario which are mutually exclusive.
The concrete can either be in the as-left backfilled configuration or excavated but not both simultaneously.
6-41
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-26 BFM DCGLB Values for ROC Adjusted for Insignificant Contributor Dose Fraction and Alternate Scenario Dose Rx Bldg DCGLB WGTV DCGLB (pCi/m2) (pCi/m2)
Co-60 5.16E+06 4.10E+06 Sr-90 1.45E+07 6.40E+06 Cs-137 2.17E+07 1.76E+07 Eu-152 1.19E+07 9.69E+06 Eu-154 1.10E+07 8.97E+06 6.16.1. Soil DCGLs for ROC Adjusted for Insignificant Contributor Dose and Alternate Scenario Dose As discussed in 6.14.1, the IC dose percentage assigned to soil is 10%. The soil DCGLs for the ROC are adjusted for the IC dose by multiplying the DCGLs in Table 6-6 by a factor of 0.9 which is calculated as 1.0-0.10 = 0.90. The ROC DCGLs were also adjusted to account for the maximum Resident Gardener Alternate Scenario soil dose of 27.07 mrem/yr (see section 6.18.1) by multiplying the DCGLs by an additional factor of 0.92 (25/27.07).
The adjusted soil DCGLs for each ROC are provided in Table 6-27.
Table 6-27 Soil DCGLs for ROC Adjusted for Insignificant Contributor Dose and Alternate Scenario Dose Adjusted Soil DCGLs ROC (pCi/g)
Co-60 1.06E+01 Sr-90 5.47E+03 Cs-137 4.83E+01 Eu-152 2.36E+01 2.19E+01 Eu-154 6.17. Concentrations in Excavated Fill Material A check calculation is performed in Reference (8) to determine the maximum hypothetical concentrations of the ROC in fill material after excavation. The calculation assumes that 100%
of the residual radioactivity in the concrete is instantly released to the fill and uniformly mixed in the fill during basement excavation. Therefore, the source term would be in the fill and not in the concrete. The calculation is performed for each ROC separately and assumes that the residual radioactivity is uniformly distributed over basement surfaces at the BFM DCGLB.
Consistent with the discussion in section 6.12 regarding the excavation of concrete, the concentrations in the fill are calculated for two excavation and mixing scenarios; 1) full excavation of the fill and mixing of all activity in concrete (100% release) with all of the fill in the basement, and 2) partial excavation of fill and mixing of all activity in concrete (100%
release) with the fill adjacent to the concrete that is excavated. The concentration in fill subject to partial excavation is calculated assuming that the minimum mixing volume is 1 m3 based on a 6-42
La Crosse Boiling Water Reactor License Termination Plan Revision 2 typical fill excavation process which entails using a 1.0 to 1.5 cubic yard bucket (see discussion in section 6.12). The partial excavation fill concentration is calculated assuming that all of the activity in a 1 m2 concrete surface area is captured and mixed in a single 1 m3 bucket load (1 m distance from a 1 m2 surface). To ensure conservatism, several worst-case bucket loads are assumed to be stockpiled together such that area over which the excavated fill is spread results in a soil AF of 1. A single bucket load would have an AF >1.
The fill concentrations for the Full and Partial excavation assumptions are provided in Table 6-28. The partial mix fill concentration for the WGTV are slightly less than the full mix concentrations. This is due to the relatively small size of the WGTV and the presence of a center wall and concrete tank support structures which result in a slightly greater ratio of surface area to fill volume for full mix as compared to the 1/1 ratio assumed for partial mix.
Table 6-28 Maximum Fill Concentration for Full and Partial Mix Radionuclide Soil Reactor Building WGTV DCGL (pCi/g) Full Mix Partial Full Mix Partial (pCi/g) Mix (pCi/g) Mix (pCi/g) (pCi/g)
Co-60 1.06E+01 1.15E+00 3.33E+00 3.31E+00 3.25E+00 Sr-90 5.47E+03 3.23E+00 9.37E+00 5.18E+00 5.08E+00 Cs-137 4.83E+01 4.83E+00 1.40E+01 1.42E+01 1.39E+01 Eu-152 2.36E+01 2.65E+00 7.69E+00 7.83E+00 7.69E+00 Eu-154 2.19E+01 2.45E+00 7.11E+00 7.25E+00 7.12E+00 The fill concentrations from the two excavation scenarios are compared to the adjusted soil DCGLs from Table 6-27. For all ROC and both basements, the hypothetical maximum fill concentrations are less than the respective soil DCGLs. Therefore, if all activity in basement surfaces is instantly released, mixed with the fill, and excavated, the dose from the fill material would always be less than the dose assigned using BFM DCGLB values.
6.18. Alternate Land Use Scenario Dose Two alternate less likely but plausible land use scenarios were considered, Resident Gardener with onsite well and Recreational Use with onsite well. In accordance with NUREG-1757, these less likely but plausible scenarios were not analyzed for compliance, but were used to risk-inform the decision of Industrial Use as the reasonably foreseeable land use. NUREG-1757 states that if the peak dose from a less likely put plausible scenario is significant then greater assurance that the scenario is unlikely would be necessary.
A quantitative evaluation of the dose from the Recreational Use scenario was not required. A simple qualitative evaluation concluded that the dose than the Recreational Use scenario will be less than the dose from the Industrial Use scenario because the occupancy time and well water intake rate would be less.
A dose assessment of the Resident Gardener scenario (with onsite well for drinking water and irrigation) was performed. The Resident Gardener pathways included direct dose, inhalation, soil ingestion and fruit and vegetables from an onsite garden. It was considered highly unlikely that livestock would be raised on the site so the meat and milk pathways were inactive.
6-43
La Crosse Boiling Water Reactor License Termination Plan Revision 2 The alternate scenario dose assessment was performed for soil, for the in situ geometry of two backfilled basements to remain (Reactor Building and WGTV), and for the excavation of concrete (and fill) from the basements. The doses were calculated assuming that a resident gardener could not plausibly occupy the LACBWR site until after the G-3 plant ceased operation and was decommissioned which was conservatively assumed to be 30 years after license termination.
The full initial suite of radionuclides was evaluated to determine the dose from the IC radionuclides specifically for the Resident Gardener scenario. The Resident Gardener dose was calculated using the ROC (which are the same as selected for the Industrial Use scenario), after adjusting the ROC for the Resident Gardener-specific IC dose. The dose was calculated in Reference (5) using a source term based on the conservative radionuclide mixture in Table 6-3.
6.18.1. Resident Gardener Dose: Soil The RESRAD assessment of Resident Gardener dose from soil applied the Industrial Use deterministic parameters in Attachment 6-1 with nine parameter changes or additions as listed in Table 6-29. The contaminated zone thickness was changed to 0.15 m to more closely represent actual site conditions as opposed to the 1 m depth assumed in the screening approach applied in the Industrial Use scenario. The remaining parameters in Table 6-29 are metabolic and behavioral. The reference or basis for the selected parameters are also listed in Table 6-29.
Table 6-29 Additions/Revisions to Industrial Use Parameters Required for Resident Gardener Scenario Parameter Value Basis Contaminated Zone Thickness 0.15 m Standard surface soil contamination depth assumption and consistent with expected site conditions Inhalation Rate 8400 m3/yr NUREG/CR-5512, Vol. 3 Table 6.29 (23 m3/d x 365 d)
Fraction of Time Spent Indoors 0.649 NUREG/CR-5512, Vol. 3 Table 6.87 Fraction of Time Spent Outdoors 0.124 NUREG/CR-5512, Vol. 3 Table 6.87 (outdoors +
gardening)
Fruit, Vegetable and Grain 112 kg/yr NUREG/CR-5512, Vol. 3 Table 6.87 (other Consumption vegetables + fruits + grain)
Leafy vegetable Consumption 21.4 Kg/yr NUREG/CR-5512, Vol. 3 Table 6.87 Drinking Water Intake 478 L/yr NUREG/CR-5512, Vol. 3 Table 6.87 Well Pumping Rate 530 m3/yr NUREG/CR-6697, Att. C Section 3.10 method assuming 7500 m2 land area and Wisconsin irrigation rate.
Plant Food Contaminated Fraction 1 All plant food assumed to be grown onsite An uncertainty analysis was performed specifically for the soil Resident Gardener Scenario including all radionuclides in the initial suite. The process outlined in Figure 6-7 was used to perform the uncertainty analysis and to select deterministic parameters for each radionuclide in 6-44
La Crosse Boiling Water Reactor License Termination Plan Revision 2 the initial suite. Attachment 6-3 lists the deterministic parameters and PDFs used for the uncertainty analysis.
A separate uncertainty analysis was performed for each radionuclide, ignoring the effect of radionuclide mixture fractions. This conservatively ensures that all sensitive parameters are identified for low abundance radionuclides regardless of actual impact on total dose given the mixture. The results of the uncertainty analyses are provided in Table 6-30 and Table 6-31.
Resident Gardener Alternate Scenario DCGLs for soil (pCi/g per 25 mrem/yr) were calculated for each initial suite radionuclide using the deterministic parameters in Attachment 6-3 with the PDFs replaced by the deterministic values from Table 6-30 and Table 6-31. The Plant Transfer Factors for all initial suite radionuclides were conservatively assigned the 75th percentile of the NUREG-6697 distribution.
The IC dose was determined specifically for the soil Resident Gardener scenario. The 75th percentile probabilistic radionuclide mixture in Table 6-3 was used in the analysis. The IC dose calculation is provided in Reference (8). The Resident Gardener ROC DCGLs were adjusted to account for the Resident Gardener-specific soil IC dose which was 0.31% of the 25 mrem/yr dose limit.
Two soil source terms were used to calculate the Resident Gardener dose. The first source term is comprised of the maximum ROC soil concentrations identified during characterization and was used to provide the most accurate estimate of dose from soil in the Resident Gardener scenario. The second source term is comprised of the ROC soil concentrations corresponding to 25 mrem/yr (i.e., the hypothetical maximum concentrations) that are calculating using the IC adjusted Industrial Use soil DCGLs, the 75th percentile radionuclide mixture (Table 6-3), and the unity rule.
The dose calculations are provided in Reference (8). The soil dose from the Resident Gardener alternate scenario using the maximum concentrations identified during characterization as the source term is 1.13 mrem/yr. When the source term was the hypothetical maximum soil concentrations that could remain given the Industrial Scenario DCGLs, the dose is 27.07 mrem/yr.
The calculation of the soil DCGLs reported in Table 6-27 includes an adjustment to ensure that the Alternate Scenario soil dose does not exceed 25 mrem/yr.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-30 Soil Alternate Scenario Resident Gardener Uncertainty Analysis Results and Deterministic Values Selected for Non-Nuclide Specific Parameter Distributions Correlation Radionuclide Basis of Selected to Dose1 Deterministic Parameter Deterministic Parameter Value Selection Contaminated zone erosion Negative Ni-59, Ni-63, Pu-25th 7.6E-04 rate 241 Contaminated zone b NS NA median 0.97 parameter Evapotranspiration Positive Tc-99 75th 0.69 coefficient Wind Speed Negative C-14 25th 3.6 Runoff coefficient Positive H-3, Tc-99 75 th 0.62 Well pump intake depth NS NA mode 21.2 b Parameter of Unsaturated NS NA median 0.97 zone Mass loading for inhalation NS NA median 2.35E-05 Indoor dust filtration factor NS NA median 0.55 Positive Co-60, Nb-94, Cs-External gamma shielding 137, Eu-152, Eu-75th 0.4 factor 154, Eu-155, Am-243, Cm-243 Negative Fe-55, Pu-238, Pu-239, Pu-240, Pu-Depth of Soil Mixing Layer 25th 0.15 241, Am-241, Cm-243, Cm-244 Negative H-3, Fe-55, Ni-59, Ni-63, Sr-90, Tc-99, Cs-137, Np-237, 25th 1.22 Depth of roots Pu-238, Pu-239.Pu-240, Pu-241, Am-241, Am-243, Cm-243, Cm-244 Wet weight crop yield of NS NA median fruit, grain, and non-leafy 1.75 vegetables Weathering removal NS NA median 33 constant for all vegetation Wet foliar interceptions NS NA median 0.58 fraction of leafy vegetables (1) NS = Not Sensitive 6-46
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-31 Soil Alternate Scenario Resident Gardener Uncertainty Analysis Results for Distribution Coefficients (Kd) and Deterministic Values Selected Kd in Contaminated Zone, Radionuclide Unsaturated Zone and Saturated Zone (cm3/mg)
H-3 NS 0.06 C-14 NS 5 Fe-55 NS 220 Ni-59 NS 400 Co-60 NS 60 Ni-63 NS 400 Sr-90 NS 15 Nb-94 NS 160 Tc-99 NS 0.1 Cs-137 NS 280 1
Eu-152 NS 825 Eu-1541 NS 825 Eu-1551 NS 825 Np-237 NS 5 Pu-238 NS 550 Pu-239/240 NS 550 Pu-241 NS 550 Am-241 NS 1900 Am-243 NS 1900 Cm-243/244 NS 4000 Note 1: Sand Kds not listed in NUREG-6607 Table 3.9-2 for Europium.
The mean value from NUREG-6697, Table 3.9-1 was used.
6.18.2. Resident Gardener Dose: Reactor Building and WGTV The Resident Gardener dose was also calculated using the backfilled Reactor Building and WGTV basement concrete as the source terms. As for soil, the Resident Gardener scenario is assumed to occur after 30 years. The dose assessment includes the three BFM dose scenarios, Insitu Groundwater, Insitu Drilling Spoils, and Excavation.
6.18.2.1. Resident Gardener Dose: Basement Insitu Groundwater This section provides the Resident Gardener dose from the basement Insitu Groundwater scenario and maximum hypothetical radionuclide concentrations.
An uncertainty analysis was performed specifically for the Reactor Building and WGTV as source terms using the Resident Gardener Scenario parameter set. All radionuclides in the initial suite were included. The uncertainty analysis process outlined in Figure 6-7 was followed. -3 provides the deterministic parameters and PDFs used for the uncertainty analysis. Certain parameters in Attachment 6-3, as listed below, relating to the geometry of each 6-47
La Crosse Boiling Water Reactor License Termination Plan Revision 2 basement were changed to the deterministic values listed in Table 6-7. The affected RESRAD geometry parameters are:
- cover depth,
- area of contaminated zone,
- thickness of contaminated zone,
- length parallel to aquifer flow, unsaturated zone thickness, and
- contaminated fraction below the water table.
A separate uncertainty analysis was performed for each initial suite radionuclide, ignoring the effect of radionuclide mixture fractions. This conservatively ensures that all sensitive parameters are identified for low abundance radionuclides regardless of actual impact on total dose given the mixture.
The RESRAD Uncertainty Reports for the Reactor Building and WGTV Resident Gardener scenario were submitted electronically with Reference (8). The results of the uncertainty analyses are provided in Table 6-32 and Table 6-33.
The Kds were negatively correlated for all but two radionuclides, Tc-99 and Nb-94. The absolute values of the PRCC were less than the l0.25l threshold for a number of radionuclides. However, A cross-check with the SRRC in the uncertainty analysis report indicates that the correlation may be more significant than indicated by the PRCC. A sensitivity analysis was performed using the 25th percentile Kds and the median Kds for the radionuclides with PRCC < l0.25l. The 25th percentile Kds resulted in higher dose for the negatively correlated radionuclides and were therefore assigned to all negatively correlated radionuclides. The cause of the positive correlation to the Tc-99 Kd was the effect on the plant dose, primarily the water independent pathway. The 75th percentile Kd resulted in the highest dose and was assigned to Tc-99. The cause of the positive correlation to the Nb-94 Kd was found to be direct dose from the water independent pathway after long term cover erosion. The maximum Nb-94 dose occurs when the 75th percentile Kd is applied but the time of maximum dose is year 312. This result is due to the unique decay characteristics of Nb-94 which include gamma emission with a very long half-life.
The 75th percentile Kd was applied for Nb-94 to ensure conservatism.
Resident Gardener Alternate Scenario Basement Groundwater DCGLs (pCi/g per 25 mrem/yr) were calculated for each initial suite radionuclide using the deterministic parameters in -3, with the PDFs replaced by the deterministic values from Table 6-32 and Table 6-33 and the basement geometry parameters replaced by the values in Table 6-7. The Plant Transfer Factors for all initial suite radionuclides were conservatively assigned the 75th percentile of the NUREG-6697 PDF.
The RESRAD Summary Reports for the Reactor Building and WGTV alternate scenario dose assessment were submitted electronically with Reference (8).
The IC dose is calculated specifically for the Insitu Groundwater Reactor Building and WGTV using the Alternate Scenario Groundwater DCGLs. The Resident Gardener scenario-specific IC dose calculations and inputs are provided in Reference (5). The Resident Gardener-specific IC doses for the Reactor Building and WGTV are 7.0% and 14.1%, respectively, and the Alternate Scenario Groundwater DCGLs were adjusted using these values.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 The Resident Gardener dose is calculated separately for the Reactor Building and WGTV using the Alternate Scenario Groundwater DCGLs for ROC (adjusted for the Resident Gardener-specific IC dose). The radionuclide concentrations used for the calculation were the maximum values in concrete that would be allowed to remain in each basement under the BFM Industrial Use scenario. The maximum concentrations are those corresponding to 25 mrem/yr using the BFM DCGLB values for ROC in Table 6-16 (adjusted for IC dose percentage of 10%), the 75th percentile mixture in Table 6-3, and the unity rule to account for all ROC contributions. The calculation of the maximum concentrations and Resident Gardener dose are provided in Reference (8).
The Resident Gardener Insitu Groundwater doses, after 30-year decay, for the Reactor Building and WGTV are 7.51 mrem/yr and 8.72 mrem/yr, respectively.
6.18.2.2. Resident Gardener Dose: Basement Drilling Spoils This section provides the Resident Gardener dose from the Reactor Building and WGTV basements Insitu Drilling Spoils scenario.
The Resident Gardener dose from the drilling spoils scenario is calculated using the methods described in 6.11.2 but replacing the Industrial Use soil DCGLs, and the 0.457 m2 soil area factors, with corresponding Resident Gardener Alternate Scenario Soil DCGLs and area factors. The RESRAD Summary report for the area factors is provided electronically with Reference (8). The Alternate Scenario Basement Drilling Spoils DCGLs for the ROC were adjusted for IC dose using a conservative assumption that the IC dose percentage was 10% as was used for the BFM Drilling spoils adjustment. The actual IC dose is expected to be the same as that calculated for the Resident Gardener soil scenario (i.e., 0.35%). The radionuclide concentrations applied in the Insitu Drilling Spoils Resident Gardener dose assessment were the same maximum allowable ROC concentrations that were calculated in section 6.18.2.1. The details of the alternate scenario drilling spoils dose calculation are provided in Reference (8).
The Resident Gardener Insitu Drilling Spoils doses (after 30-year decay) for the Reactor Building and WGTV are 0.25 mrem/yr and 0.35 mrem/yr, respectively.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-32 BFM Insitugw Alternate Scenario Resident Gardener Uncertainty Analysis Results and Deterministic Values Selected for Non-Nuclide Specific Parameter Distributions Correlation Radionuclide Basis of Selected to Dose1 Deterministic Parameter Rx Building WGTV Deterministic Parameter Value Selection Positive Ni-59, Nb-94, Pu-241, Pu-239, Am-241 Ni-59, Ni-63, Nb-94. Pu-239, Pu-241, Am-Cover erosion rate 75th 2.9E-03 241 Contaminated zone b parameter NS NA NA median 0.97 Contaminated Zone Erosion NS NA NA median 0.00015 Rate th Positive H-3 75 0.69 (Rx Building) (Rx Building) (Rx Building)
Evapotranspiration coefficient NS Median 0.62 (WGTV) (WGTV) (WGTV)
Wind Speed NS NA NA median 3.6 th Positive H-3 75 0.62 (Rx Building) (Rx Building) (Rx Building)
Runoff coefficient NS Median 0.45 (WGTV) (WGTV) (WGTV)
H-3, C-14, Fe-55, Sr-90, Np-237, Pu- H-3, C-14, Fe-55, Sr-90, Np-237, Pu-239 Minimum Well pump intake depth Negative 6.1 241, Pu-239 Depth b Parameter of Unsaturated NS NA NA median 0.97 zone Mass loading for inhalation NS NA NA median 2.35E-05 Indoor dust filtration factor NS NA NA median 0.55 External gamma shielding factor NS NA NA median 0.4 Depth of Soil Mixing Layer NS NA NA median 0.15 C-14, Fe-55, Co-60, Ni-59, Ni-63, Sr- H-3, C-14, Fe-55, Ni-59, Co-60, Ni-63, Sr-90, 90, Tc-99, Cs-137, Eu-152, Eu-154, Tc-99, Cs-137, Eu-152, Eu-154, Eu-155, Np-Depth of roots Positive 75th 3.07 Eu-155, Np-237, Pu-241, Pu-239, Am- 237, Pu-239, Pu-241, Am-241 241 Wet weight crop yield of fruit, NS NA NA median 1.75 grain, and non-leafy vegetables Weathering removal constant NS NA NA median 33 for all vegetation Wet foliar interceptions fraction NS NA NA median 0.58 of leafy vegetables Note 1: NS= Not Sensitive 6-50
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-33 BFM Insitugw Alternate Scenario Resident Gardener Uncertainty Analysis Results for Distribution Coefficients (Kd) and Deterministic Values Selected Rx Building WGTV Kd in Contaminated Zone Kd in Contaminated Zone Radionuclide and Saturated Zone and Saturated Zone (cm3/mg) (cm3/mg)
H-3 Negative1 0.05 Negative1 0.05 C-14 Negative 1.8 Negative 1.8 Fe-55 Negative 38 Negative 38 Ni-59 Negative 147 Negative 147 Co-60 Negative 9 Negative 9 Ni-63 Negative 147 Negative 147 Sr-90 Negative 5 Negative 5 1
Nb-94 Positive 611 Negative 611 Tc-99 Positive 0.46 Positive 1 0.46 Cs-137 Negative 50 Negative 50 2
Eu-152 Negative 95 Negative 95 2
Eu-154 Negative 95 Negative 95 Eu-155 Negative 95 Negative 95 Np-237 Negative 1 Negative 1 Pu-238 Negative 173 Negative 173 Pu-239/240 Negative 173 Negative 173 Pu-241 Negative 173 Negative 173 Am-241 Negative 329 Negative 329 Am-243 Negative 329 Negative 329 Cm-243/244 Negative 881 Negative 881 Note 1: 25th and 75th percentile values for sand from Reference (8) (see Attachment 3)
Note 2: Sand Kds not listed in NUREG-6697 Table 3.9-2 for Europium. The mean 6.18.2.3. Resident Gardener Dose: Basement Concrete Excavation Scenario This section provides the Resident Gardener dose from the Reactor Building and WGTV basement Excavation scenarios.
The BFM Industrial Use Excavation DCGLs are derived by limiting the concentrations in the excavated concrete to the concentrations of the Industrial Use soil DCGLs (see section 6.12). To calculate the Resident Gardener dose from excavated concrete, the BFM Excavation DCGLs in Table 6-15 are multiplied by the ratio of the Industrial Use Soil DSR to the Resident Gardener Alternate Scenario Soil DSR. Note that the Alternate Scenario Soil DSR are adjusted for Resident Gardener-specific IC dose as discussed in section 6.18.1. The Alternate Scenario Excavation DCGLs are used to calculate the Resident Gardener dose from excavated basement concrete using the maximum concrete ROC concentrations hypothetically allowed by the Industrial Use scenario as discussed in section 6.18.2.1. The concrete excavation dose calculations are provided in Reference (8).
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 The Resident Gardener Concrete Excavation dose (after 30-year decay) for the Reactor Building and WGTV are 20.7 mrem/yr and 25.9 mrem/yr, respectively.
6.18.2.4. Total Resident Gardener Alternate Scenario Dose: Reactor Building and WGTV The total Resident Gardener doses from the Reactor Building and WGTV are calculated by summing the dose from the Insitu Groundwater, Insitu Drilling Spoils and Excavation scenarios.
The Total Resident Gardener doses from the Reactor Building and WGTV, assuming maximum allowable concentrations in the basement concrete, are 28.4 mrem/yr and 34.9 mrem/yr, respectively. To ensure conservatism, the BFM DCGLBS and DCGLB values reported in Table 6-24, Table 6-25, and Table 6-26 include an adjustment to ensure that the Alternate Scenario dose does not exceed 25 mrem/yr.
6.18.2.5. Resident Gardener Dose from Excavated Basement Fill The Resident Gardener dose was also calculated using the maximum hypothetical fill concentrations in the Reactor Building and WGTV (see section 6.17). The doses from excavated fill (after 30-year decay) were 6.3 mrem/yr and 8.0 mrem/yr for the Reactor Building and WGTV, respectively (8).
6.18.2.6.
Conclusion:
Resident Gardener Alternate Scenario Dose The Resident Gardener dose, after 30 years decay, was evaluated for soil and basements. The maximum soil dose is 27.07 mrem/yr. The maximum basement doses are 28.4 mrem/yr and 34.9 mrem/yr, for the Reactor Building and WGTV, respectively. These doses are not considered significant and therefore greater assurance that these scenarios will not occur is not necessary.
However, adjustments were made to the soil and basement DCGLs to ensure that no Alternate Scenario dose exceeds 25 mrem/yr.
6.19. Soil Area Factors The RESRAD modeling for soil assumes that the entire area within the LSE, 7500 m2, is contaminated. Isolated areas of contamination that are smaller than 7500 m2 will have a lower dose for a given concentration. The ratio of the dose from the full source term area to the dose from a smaller area is defined as the AF.
Reference (8) calculates AFs for each ROC using RESRAD with the deterministic parameter set used to calculate soil DCGLs. The Area of Contaminated Zone parameter was varied from 1.0 m2 to 100 m2. The need to apply AFs to contaminated areas greater than 100 m2 is unlikely.
The AFs were calculated by dividing the pCi/g per 25 mrem/yr value from RESRAD for each smaller area by the soil DCGLs in Table 6-7. The full RESRAD Summary Reports are provided electronically with Reference (8). The AFs are provided in Table 6-34.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-34 Surface Soil Area Factors Area Factors Radionuclide 2 2 1m 2m 5m2 10m2 100m2 Cs-137 9.44 5.56 3.07 2.04 1.19 Co-60 9.11 5.42 3.01 2.00 1.18 Sr-90 11.21 6.66 3.69 2.45 1.41 Eu-152 9.30 5.50 3.04 2.02 1.18 Eu-154 9.38 5.54 3.06 2.03 1.18 6.20. Buried Piping Dose Assessment and DCGL Buried piping is defined as below ground pipe located outside of structures and basements. This section describes the buried pipe dose assessment methods and provides the resulting DCGLs.
The calculations are performed by RESRAD and Excel spreadsheet as detailed in Reference (8).
6.20.1. Source Term and Radionuclide Mixture With the exception of a portion of the Circulating Water System Pipe, none of the buried piping to remain at LACBWR was associated with contaminated systems and therefore contamination potential is minimal. See Table 6-2 for list of buried pipe to remain. The High Pressure Service Water from LACBWR Crib House to G-3 and Well water piping for Well #3 are considered non-impacted because they only contacted clean river water or groundwater with no potential for contamination and will continue operation after license termination.
DCGLs for buried pipe were calculated for the initial suite radionuclides. To date, no characterization has been performed in buried piping due to the very low contamination potential. The radionuclide mixture is assumed to be the same as listed in Table 6-3. As discussed in LTP Chapter 5, if continuing characterization is performed for buried pipe and the results indicate that the buried piping dose could exceed 10% of the 25 mrem/yr dose criterion, then samples will be analyzed for HTD radionuclides and additional assessments performed.
6.20.2. Exposure Scenario and Critical Group The dose assessment approach was generally consistent with the guidance in NUREG-1757, Appendix J in that two exposure scenarios were considered; 1) assuming that the buried pipe is excavated and spread across the surface (Excavation scenario), and 2) assuming that the buried pipe remains in situ (Insitu scenario).
NUREG-1757, Appendix J states that it should be appropriate to use the arithmetic average of the radionuclide concentration in the analysis, including any interspersing clean soil. The buried piping at LACBWR is a minimum of 1 m below grade. The LACBWR buried pipe excavation conceptual model is more conservative than the NUREG-1757, Appendix J conceptual model in that no mixing is assumed to occur with the soil in the 1 m cover or the interspersing clean soil between pipes during excavation.
The conceptual models for the buried pipe Insitu and Excavation scenarios are similar to those developed for the BFM. In the Insitu scenario, the residual radioactivity on the internal surfaces 6-53
La Crosse Boiling Water Reactor License Termination Plan Revision 2 of the pipe is assumed to instantaneously release and mix with a thin 2.54 cm layer of soil in an area equal to the internal surface area of the pipe. The Insitu scenario model assumes that the released radioactivity as a below ground 2.54 cm layer of soil with no credit taken for the presence of the pipe to reduce environmental transport and migration. This is a conservative assumption, particularly for the Circulating Water Discharge Pipe which will be filled with a flowable fill material. The Excavation scenario model assumes that the released radioactivity is mixed in a 15 cm layer of soil on the ground surface after excavation. A 15 cm mixing layer is assumed due to the extensive ground surface disturbance caused by the large scale excavation required to remove pipe. The Industrial Worker is exposed to the Insitu and Excavated soil via the same pathways applicable to the BFM and soil models.
6.20.3. Buried Piping Dose Assessment RESRAD modeling was performed to calculate DSRs which are the basis for determining DCGLs for the internal surfaces of the pipes after converting units to dpm/100 cm2.
The buried piping was separated into two categories. The first category included the summation and grouping of all impacted buried pipe other than the Circulating Water Discharge Piping and is designated as the Group. The second category consisted of the Circulating Water Discharge Pipe only. The separation of the Circulating Water pipe was necessary because the geometry was significantly different from the other pipe and the pipes are located in distinctly different parts of the site.
The Insitu dose calculation for the buried piping Group (which as stated above does not include the Circulating Water Discharge pipe) was performed by RESRAD modeling using the input parameters applied to the BFM Insitu Groundwater scenario with adjustments to the source term geometry as listed in Table 6-35. The lowest elevation at the bottom of the Group piping is 625 (excluding water well #3 piping which is considered non-impacted). Using the lowest elevation maximizes the Insitu dose, which is driven by the groundwater pathway, by minimizing the distance to the water table. The assumed RESRAD parameters Area of Contaminated Zone and Length Parallel to Flow were calculated assuming that the all of the pipe in the Group was located in one circular area equal in size to the summed internal surface area of all Group pipes. The internal surface area of the High-Pressure Service Water from LACBWR Crib House to G-3 Well and water piping for Well #3 are conservatively included in the calculation of total area for the Group piping notwithstanding their classification as non-impacted (i.e., no FSS to be performed).
The Insitu dose for the Circulating Water Discharge Pipe was also calculated using the BFM Insitu Groundwater parameters with the elevation of the thin contaminated layer being set at the elevation of the bottom of the pipe (630.5 foot). The Circulating Water Discharge pipe drops 10 to 620.5 elevation at the outfall but this 10-foot length is trivial compared to the total 525 length at 630.5 elevation. In addition, it is not plausible to locate a well between the location where the pipe drops and the outfall. The contaminated area was set equal to the internal surface area of the pipe.
The dose from the Excavation scenarios (and corresponding DCGLs) for both the Buried Pipe Group and the Circulating Water Discharge Pipe were calculated using the RESRAD parameters used to calculate surface soil DCGLs with source term adjustments as listed in Table 6-35.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-35 RESRAD Source Term Parameters for Buried Piping DCGL Calculations Parameter Buried Pipe Buried Pipe Circulating Water Circulating Water Group Group Discharge Pipe Discharge Pipe Insitu Excavation Insitu Excavation Cover Depth (m) 3.02 0 2.59 0 Area of Contaminated 1552.70 1552.70 766.14 766.14 Zone (m2)
Thickness of 0.0254 0.15 0.0254 0.15 Contaminated zone (m)
Length Parallel to Aquifer 44.46 160.02 160.02 Flow (m)
Unsaturated Zone 0 2.90 0.43 2.90 Thickness (m)
Depth of Soil Mixing 0.15 0.15 0.15 0.15 Layer 6.20.4. Buried Pipe DCGLs Initial Suite The RESRAD Summary Reports are provided electronically with Reference (8). The detailed inputs to the DCGL calculation, including RESRAD source term parameter calculations, a list of the resulting RESRAD DSRs generated by modeling and all unit conversions are provided in Reference (8). The buried pipe DCGLs for the initial suite are provided in Table 6-36 .
6.20.5. Buried Pipe Radionuclides of Concern and Adjusted DCGLs The Buried Pipe DCGLs in Table 6-36 were used in Reference (5) to calculate the relative dose contributions from the initial suite radionuclides, identify the IC radionuclides, select the final ROCs and adjust the ROCs for the dose fraction attributable to the removed IC radionuclides.
The dose percentages for the initial suite were calculated using the mixture fractions in Table 6-3.
The dose percentages for the initial suite radionuclides were calculated using the summed DCGLs, which include the dose from the Insitu and Excavation scenarios, and the 75th percentile mixture for soil in Table 6-3. The final list of IC radionuclides was the same for all buried pipe scenarios and the same as identified for soil and basements. The ROCs are Co-60, Sr-90, Cs-137, Eu-152 and Eu-154.
The calculated IC dose percentages was 0.5% of the 25 mrem/yr dose limit for both the Group and Circulating Water Discharge Pipe. However, consistent with approach used for soil and basements, an IC dose percentage of 10% was applied to provide additional margin to account for mixture variability. Table 6-37 provides the Buried Pipe DCGLs for the ROC adjusted for the IC dose percentage.
The final DCGLs to be used during FSS account for the fact that the dose from the Insitu and Excavation scenarios must be summed in the conceptual model for buried pipe dose assessment since the insitu and excavation scenarios may occur concurrently to some extent. The summed Buried Pipe DCGLs are provided in Table 6-38.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-36 Buried Piping DCGLs Buried Pipe Group Buried Pipe Group Circulating Water Circulating Water Discharge Pipe Discharge Pipe Radionuclide Insitu Excavation Insitu Excavation (dpm/100 cm2) (dpm/100 cm2) (dpm/100 cm2) (dpm/100 cm2)
H-3 4.59E+08 9.65E+08 1.61E+08 1.34E+09 C-14 4.65E+07 4.81E+09 2.60E+07 4.81E+09 Fe-55 1.27E+09 5.98E+10 2.03E+26 7.81E+10 Ni-59 1.15E+10 1.52E+11 2.03E+26 1.99E+11 Co-60 6.37E+06 8.44E+04 2.51E+08 8.62E+04 Ni-63 4.61E+09 5.56E+10 1.11E+11 7.26E+10 Sr-90 5.81E+05 4.03E+07 8.55E+05 4.29E+07 Nb-94 1.02E+08 1.27E+05 1.02E+08 1.29E+05 Tc-99 2.00E+07 2.41E+07 6.50E+06 3.43E+07 Cs-137 1.82E+07 3.60E+05 6.53E+08 3.67E+05 Eu-152 2.71E+08 1.82E+05 5.08E+14 1.86E+05 Eu-154 1.87E+08 1.70E+05 9.89E+17 1.73E+05 Eu-155 1.20E+09 6.62E+06 2.03E+26 6.71E+06 Np-237 6.52E+03 1.52E+04 4.34E+03 2.09E+04 Pu-238 9.79E+05 9.79E+06 6.82E+07 1.27E+07 Pu-239 8.06E+05 8.81E+06 8.15E+05 1.14E+07 Pu-240 8.10E+05 8.82E+06 8.57E+05 1.14E+07 Pu-241 4.56E+07 3.58E+08 2.45E+08 4.37E+08 Am-241 1.62E+06 6.46E+06 8.88E+06 7.84E+06 Am-243 1.51E+06 1.11E+06 1.90E+06 1.16E+06 Cm-243 6.41E+06 1.71E+06 6.78E+08 1.79E+06 Cm-244 8.02E+06 1.57E+07 3.11E+08 2.04E+07 Table 6-37 Buried Pipe DCGLs for ROCs Adjusted for Insignificant Radionuclide Fractions Buried Pipe Group Buried Pipe Group Circulating Water Circulating Water Discharge Pipe Discharge Pipe Radionuclide Insitu Excavation Insitu Excavation (dpm/100 cm2) (dpm/100 cm2) (dpm/100 cm2) (dpm/100 cm2)
Co-60 5.73E+06 7.60E+04 2.26E+08 7.76E+04 Sr-90 5.23E+05 3.63E+07 7.70E+05 3.86E+07 Cs-137 1.64E+07 3.24E+05 5.88E+08 3.30E+05 Eu-152 2.44E+08 1.64E+05 4.57E+14 1.67E+05 Eu-154 1.68E+08 1.53E+05 8.90E+17 1.56E+05 6-56
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-38 Summed Buried Pipe DCGLs for ROCs adjusted for Insignificant Radionuclide Fractions Buried Pipe Group Circulating Water Discharge Pipe Radionuclide (dpm/100 cm2) (dpm/100 cm2)
Co-60 7.50E+04 7.75E+04 Sr-90 5.16E+05 7.55E+05 Cs-137 3.18E+05 3.30E+05 Eu-152 1.64E+05 1.67E+05 Eu-154 1.52E+05 1.56E+05 6.21. Existing Groundwater Dose There is low potential for significant groundwater contamination to be present although low concentrations have been identified in groundwater adjacent to suspected broken floor drains under the Turbine Building. Sampling in 1983 from a well located down gradient of the Turbine Building indicated positive groundwater contamination at relatively low concentrations.
In late 2012, five additional monitoring well pairs were installed to support site characterization and license termination. Results indicated lower groundwater contamination levels than found in 1983, predominantly H-3. See LTP Chapter 2 for a summary of characterization and HSA results prior to submittal of LACBWR LTP Revision 0. In December 2017, the groundwater sampling program identified elevated H-3 with a maximum concentration of 24,200 pCi/L identified in February 2018 (well MW-203A). A subsequent sample from well MW-203A, collected in April 2018, contained lower H-3 concentrations of 12,100 pCi/L indicating a downward trend.
As a result an investigation was initiated to identify the source of the contamination and the extent of contaminant migration from the source. The investigation identified the source as being a reactor building ventilation exhaust that was directed toward the ground surface where H-3 condensed. This investigation included the performance of a dye tracer test and the development of a numerical groundwater transport model that concluded that the maximum groundwater concentration of H-3 would have been approximately 60,000 pCi/L for a brief period of time (less than a few months) and that the concentrations continue to decline. Prior to and during this investigation period, final status surveys were conducted using operational DCGLs with a dose contribution from groundwater as 3.25 mrem as assigned in LC-FS-TSD-002 Rev 01. To ensure that this dose is bounding, a dose calculation using a maximum H-3 concentration of 60,000 pCi/L and dose from the maximum well using the IC radionuclides from Table 2-19 (June 2014 sample data) is performed.
To support this dose calculation, Groundwater Exposure Factors from the initial suite of radionuclides were calculated using Ingestion Dose Conversion Factors from Federal Guidance Report 11 Reference (8) directly with an assumed industrial worker AMCG drinking water intake rate of 327 L/yr. See Table 6-39.
6-57
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-39 Ingestion Dose Conversion Factors and Groundwater Exposure Groundwater FGR 11 ING DCF, Nuclide Exposure Factors, mrem/pCi mrem/y per pCi/L H-3 6.40E-08 2.09E-05 C-14 2.09E-06 6.83E-04 Fe-55 6.07E-07 1.98E-04 Ni-59 2.10E-07 6.87E-05 Co-60 2.69E-05 8.80E-03 Ni-63 5.77E-07 1.89E-04 Sr-90 1.42E-04 4.64E-02 Nb-94 7.14E-06 2.33E-03 Tc-99 1.46E-06 4.77E-04 Cs-137 5.00E-05 1.64E-02 Eu-152 6.48E-06 2.12E-03 Eu-154 9.55E-06 3.12E-03 Eu-155 1.53E-06 5.00E-04 Pu-238 3.20E-03 1.05E+00 Pu-239 3.54E-03 1.16E+00 Pu-240 3.54E-03 1.16E+00 Pu-241 6.85E-05 2.24E-02 Am-241 3.64E-03 1.19E+00 The concentrations for samples that were identified as positive in each sampling event from 2014 is shown in the following tables (the bolded values from LTP Table 2-19). Also, the corresponding annual dose is shown at the bottom of each table without considering radioactive decay.
As shown in these tables, the maximum dose from the identified positive detections in 2014 was 0.471 mrem/y from well MW-DW7 from the June 2014 sampling event from Pu-239 and no identified H-3. The dose from the more recent H-3 concentration of 60,000 pCi/L is 1.26 mrem for a total groundwater dose of 1.73 mrem. Given that the assigned dose from groundwater for selection of the operational DCGL is 3.25 mrem, this leaves an additional dose margin of 1.52 mrem.
This analysis demonstrates that the dose assigned dose for groundwater is well bounded using the maximum plume concentrations including the potential contributions of the IC radionuclides from 2014 results.
6-58
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-40 June 2014 Well Concentrations (pCi/L) for Radionuclides Identified as Positive and the Corresponding Dose (mrem/yr)
MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW-Nuclide DW3 DW4 DW5 DW7 B11R B11AR MW-B2 MW-B3 200-A 200-B 201-A 201-B 202-A 202-B 203-A 203-B H-3 2.45E+02 3.36E+02 2.79E+02 2.79E+02 C-14 Fe-55 Ni-59 Co-60 Ni-63 Sr-90 1.12E+00 1.17E+00 Nb-94 Tc-99 5.08E+00 Cs-137 Eu-152 9.48E+00 1.42E+01 9.71E+00 1.12E+01 1.08E+01 Eu-154 Eu-155 Pu-238 Pu-239 4.07E-01 Pu-240 Pu-241 1.72E-01 9.71E-01 1.16E-01 Am-241 Dose, 2.01E-02 0.00E+00 0.00E+00 5.13E-03 0.00E+00 0.00E+00 3.01E-02 2.44E-02 2.42E-02 0.00E+00 2.37E-02 6.18E-02 2.29E-02 6.03E-02 5.84E-03 4.71E-01 mrem/yr 6-57a
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Table 6-40 June 2014 Well Concentrations (pCi/L) for Radionuclides Identified as Positive and the Corresponding Dose (mrem/yr)
(Continued)
Radionuclid MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW- MW-e 204-A 204-B DW5 B11R B11AR 200-A 200-B 201-A 201-B 202-A 202-B 203-A 203-B 204-A 204-B H-3 C-14 1.30E+01 Fe-55 Ni-59 Co-60 3.56E+00 3.67E+00 Ni-63 5.00E+00 Sr-90 2.01E+00 1.14E+00 Nb-94 Tc-99 6.95E+00 6.31E+00 Cs-137 3.97E+00 2.17E+01 Eu-152 9.40E+00 Eu-154 4.18E+00 Eu-155 Pu-238 Pu-239 1.41E-01 Pu-240 Pu-241 1.56E-01 2.29E-01 1.40E-01 2.51E-01 2.69E-01 Am-241 Dose, 1.00E-01 1.19E-02 6.48E-02 0.00E+00 0.00E+00 4.39E-01 5.13E-03 4.08E-03 1.63E-01 1.99E-02 1.30E-02 3.79E-02 0.00E+00 0.00E+00 6.02E-03 mrem/yr 6-57b
La Crosse Boiling Water Reactor License Termination Plan Revision 2 6.22. Demonstrating Compliance with Dose Criterion As discussed in section 6.5.6, the final demonstration of compliance with the dose criterion will be made through the summation of dose from each of the five media. The compliance dose will be calculated using Equation 6-11 after FSS has been completed in all survey units. Note that the acronym BcDCGL in Equation 6-11 is defined in Reference (4) as Base Case DCGL which is equivalent to the DCGLs developed in this Chapter. Different terminology was required in Reference (4) to distinguish the full DCGLs (which represent 25 mrem/yr) from the Operational DCGLs which represent a fraction of 25 mrem/yr (see Reference (4)). The acronym SOF in Equation 6-11 is defined as Sum of Fractions.
The Release Record for each FSS unit will be reviewed to determine the maximum mean dose from each of the five source terms (e.g. basement, soil, buried pipe, above grade buildings, and existing GW). The compliance dose must be less than or equal to 25 mrem/yr. The calculation of the compliance dose will be documented in the final FSS Report for the site.
A detailed description of the terms in Equation 6-11 and the method for calculating the dose for each term is provided in Reference (4).
Equation 6-11 Compliance Dose = (Max BcSOFBASEMENT + Max BcSOFSOIL + Max BcSOFBURIED PIPE +
BcSOFAG BUILDING + GW BcSOFBS OB + GW BcSOFBPS OBP + Max SOFEGW) x 25 mrem/yr where:
Compliance Dose = must be less than or equal to 25 mrem/yr, Max BcSOFBASEMENT = Maximum BcSOF (mean of FSS systematic results plus the dose from any identified elevated areas) for backfilled Basements, Max BcSOFSOIL = Maximum BcSOF (mean of FSS systematic results plus the dose from any identified elevated areas) for open land survey units, Max BcSOFBURIED PIPE = Maximum BcSOF (mean of FSS systematic results plus the dose from any identified elevated areas) from buried piping survey units, Max BcSOFAG BUILDING = Maximum BcSOF (mean of FSS systematic results plus the dose from any identified elevated areas) from above grade standing building survey units, GW BcSOFBS OB = Groundwater scenario dose from the Other Basement (OB) which is defined as the basement not used to generate the Max BcSOFBASEMENT term in Equation 1 GW BcSOFBPS OBP = Groundwater scenario dose from the Other Buried Pipe (OBP) which is defined as the buried pipe survey unit not used to generate the Max BcSOFBURIED PIPE term in Equation 1 Max SOFEGW = Maximum SOF from existing groundwater (EGW) 6-58
La Crosse Boiling Water Reactor License Termination Plan Revision 2 6.23. References
- 1. Haley & Aldrich Inc., Hydrogeological Investigation Report, La Crosse Boiling Water Reactor, Dairyland Power Cooperative, Genoa Wisconsin, File No. 38705-008, January 2015.
- 2. EnergySolutions Technical Support Document RS-TD-313196-003, Revision 0, La Crosse Boiling Water Reactor Historical Site Assessment (HSA).
- 3. U.S. Nuclear Regulatory Commission, NUREG-1757, Volume 2, Revision 1, Consolidated Decommissioning Guidance - Characterization, Survey, and Determination of Radiological Criteria, Final Report - September 2006.
- 4. EnergySolutions LC-FS-TSD-002, Revision 2, Operational Derived Concentration Guideline Levels for Final Status Survey.
- 5. EnergySolutions Technical Support Document RS-TD-313196-001, Revision 5, Radionuclides of Concern During LACBWR Decommissioning.
- 6. Pacific Northwest Laboratory, NUREG/CR-3474, Long-Lived Activation Products in Reactor Materials, Pacific Northwest Laboratory - 1984.
- 7. Pacific Northwest Laboratory, NUREG/CR-4289, Residual Radionuclide Concentration Within and Around Commercial Nuclear Power Plants; Origin, Distribution, Inventory, and Decommissioning Assessment - 1985.
- 8. EnergySolutions Technical Support Document RS-TD-313196-004, Revision 4, LACBWR Soil DCGL, Basement Concrete DCGL and Buried Pipe DCGL.
- 9. Argonne National Laboratory, NUREG/CR-6697, Development of Probabilistic RESRAD 6.0 and RESRAD-BUILD 3.0 Computer Codes - December 2000.
- 10. Sandia National Laboratory, NUREG/CR-5512, Volume 3, Residual Radioactive Contamination From Decommissioning Parameter Analysis - October 1999.
- 11. ZionSolutions Technical Support Document 14-006, Revision 5, Conestoga Rovers &
Associates (CRA) Report, Evaluation of Hydrological Parameters in Support of Dose Modeling for the Zion Restoration Project.
- 12. Argonne National Laboratory, Users Manual for RESRAD Version 6, ANL/EAD-4, July 2001.
- 13. Sheppard and Thibault, Default Soil/Solid /Liquid Partition Coefficients, Kds, for Four Major Soil Types: A Compendium, Health Physics, Vol. 59 No 4, October 1990.
- 14. EnergySolutions Technical Support Document RS-TD-313196-002, Revision 0, Final LACBWR End State Basement Concrete Surface Areas, Volumes, and Void Spaces.
- 15. EnergySolutions, LC-FS-TSD-004, Revision 0, Dose Assessment of Post-Remediation Subsurface Geometry of Survey Unit L1-SUB-TDS B (including the RPGPA Sump).
6-59
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Figure 6-1 Site Regional Location 6-60
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Figure 6-2 Site Overview 6-61
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Figure 6-3 LACBWR Buildings 6-62
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Figure 6-4 LACBWR End State 6-63
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Figure 6-5 LACBWR End State - Backfilled Structures 6-64
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Figure 6-6 LACBWR End State - Backfilled Reactor Building Basement Elevation View 6-65
La Crosse Boiling Water Reactor License Termination Plan Revision 2 Figure 6-7 RESRAD Parameter Selection Flow Chart Select model parameter Classify as Behavioral, Metabolic, or Physical Behavioral parameter Metabolic parameter Physical parameter Assign default value Assign default value from NUREG/CR 5512 from NUREG/CR 5512 Site data Vol. 3 Vol. 3 available?
Yes Input parameter Input parameter value value No Input parameter value Classify as Priority 1, 2 or 3 Priority 1 or 2 Priority 3 Assign distribution from Assign Median value NUREG/CR-6697, Att C from RESRAD v7.0 Complete sensitivity analysis using RESRAD v7.0 Input parameter value Classify parameter as Sensitive or Non-sensitive Sensitive, lPRCCl > 0.25 Non-sensitive, lPRCCl < 0.25 Assign 25% percentile value if Assign 75% percentile value if TEDE is negatively correlated Assign 50% value TEDE is positively correlated with the parameter distribution with the parameter Input parameter Input parameter Input parameter value value value 6-66
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ATTACHMENT 6-1 RESRAD Input Parameters for LACBWR Soil DCGL Uncertainty Analysis 6-67
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Soil Concentrations Basic radiation dose limit (mrem/y) 3 D 25 10 CFR 20.1402 NR NR NR NR Initial principal radionuclide (pCi/g) P 2 D 1 Unit Value NR NR NR NR Distribution coefficients (contaminated, unsaturated, and saturated zones) (cm 3/g)
Mean Kd Value for sand Ac-227 (daughter of Cm-243 and NUREG/CR-6697, Table P 1 D 450 NA NA NA NA Pu-239) 3.9-2, Sheppard and Thibault Am-241 (also daughter of Cm-245 S Lognormal-N NUREG/CR-6697 Att. C 7.28 3.15 1445 P 1 and Pu-241)
S Lognormal-N NUREG/CR-6697 Att. C 7.28 3.15 NA NA 1445 Am-243 P 1 C-14 P 1 S Lognormal-N NUREG/CR-6697 Att. C 2.4 3.22 NA NA 11 Lognormal-N NUREG/CR-6697 Att. C 8.82 1.82 NA NA 6761 Cm-243 P 1 S Lognormal-N NUREG/CR-6697 Att. C 8.82 1.82 NA NA 6761 Cm-244 P 1 S Co-60 P 1 S Lognormal-N NUREG/CR-6697 Att. C 5.46 2.53 NA NA 235 Cs-137 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.1 2.33 NA NA 446 Eu-152 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.72 3.22 NA NA 825 Eu-154 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.72 3.22 NA NA 825 Eu-155 P 1 S Lognormal-N NUREG/CR-6697 Att. C 5.94 3.22 NA NA 380 Fe-155 P 1 S Lognormal-N NUREG/CR-6697 Att. C 5.34 2.67 NA NA 209 Mean Value NUREG/CR-6697, Att. C Gd-152 (daughter for Eu-152) P 1 D 825 NA NA NA NA NA (No sand value listed in Table 3.9-2)
H-3 P 1 S Lognormal-N NUREG/CR-6697 Att. C -2.81 0.5 NA NA 0.06 Nb-94 P 1 S Lognormal NUREG/CR-6697 Att. C 5.94 3.22 NA NA 380 RESRADv.7.0 Default Nd-144 (daughter for Eu-152) P 1 D 158 Nd not listed in NA NA NA NA NA NUREG/CR-6697 Ni-59 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.05 1.46 NA NA 424 Ni-63 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.05 1.46 NA NA 424 6-68
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Np-237 (also daughter for Am-241, P 1 S Lognormal-N NUREG/CR-6697 Att. C 2.84 2.25 NA NA 17 Cm-245, and Pu-241)
Mean Kd Value for sand Pa-231 (daughter for Cm-243 and NUREG/CR-6697, Table P 1 D 550 NA NA NA NA NA Pu-239) 3.9-2, Sheppard and Thibault Mean Kd Value for sand NUREG/CR-6697, Table Pb-210 (daughter for Pu-238) P 1 D 270 NA NA NA NA NA 3.9-2, Sheppard and Thibault Mean Kd Value for sand NUREG/CR-6697, Table Po-210 (daughter Pu-238) P 1 D 150 NA NA NA NA NA 3.9-2, Sheppard and Thibault Pu-238 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.86 1.89 NA NA 953 Pu-239 (also daughter for Cm-243) P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.86 1.89 NA NA 953 Pu-240 (also daughter for Cm-244) P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.86 1.89 NA NA 953 Pu-241 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.86 1.89 NA NA 953 Mean Kd Value for sand NUREG/CR-6697, Table Ra-226 (daughter Pu-238) P 1 D 500 NA NA NA NA NA 3.9-2, Sheppard and Thibault Mean Kd Value for sand Ra-228 (daughter Cm-244 and Pu- NUREG/CR-6697, Table P 1 D 500 NA NA NA NA NA 240) 3.9-2, Sheppard and Thibault Mean Kd Value for sand NUREG/CR-6697, Table Sm-148 (daughter Eu-152) P 1 D 245 6.72 3.22 NA NA 825 3.9-2, Sheppard and Thibault Sr-90 P 1 S Lognormal-N NUREG/CR-6697 Att. C 3.45 2.12 NA NA 32 Tc-99 P 1 S Lognormal-N NUREG/CR-6697 Att. C 0.67 3.16 NA NA 0.51 Mean Kd Value for sand Th-228 (daughter Cm-244 and Pu- NUREG/CR-6697, Table P 1 D 3200 NA NA NA NA NA 240) 3.9-2, Sheppard and Thibault 6-69
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Mean Kd Value for sand Th-229 (daughter Am-241, Cm-245, NUREG/CR-6697, Table P 1 D 3200 NA8.68 NA3.62 NA NA NA Np-237, and Pu-241) 3.9-2, Sheppard and Thibault Mean Kd Value for sand Th-230 (daughter Cm-246 and Pu- NUREG/CR-6697, Table P 1 D 3200 NA NA NA NA NA 238) 3.9-2, Sheppard and Thibault Mean Kd Value for sand Th-232 (daughter Cm-244 and Pu- NUREG/CR-6697, Table P 1 D 3200 NA NA NA NA NA 240) 3.9-2, Sheppard and Thibault Mean Kd Value for sand U-233 (daughter Am-241, Cm-245, NUREG/CR-6697, Table P 1 D 35 NA NA NA NA NA Np-237, and Pu-241) 3.9-2, Sheppard and Thibault C Mean Kd Value for sand NUREG/CR-6697, Table U-234 (daughter Pu-238) P 1 D 35 3.9-2, Sheppard and NA NA NA NA NA Thibault parent < 0.1% of radionuclide mixture Mean Kd Value for sand U-235 (daughter Cm-243 and Pu- NUREG/CR-6697, Table P 1 D 35 NA NA NA NA NA 239) 3.9-2, Sheppard and Thibault Mean Kd Value for sand U-236 (daughter Cm-244 and Pu- NUREG/CR-6697, Table P 1 D 35 NA NA NA NA NA 240) 3.9-2, Sheppard and Thibault Initial concentration of radionuclides No existing groundwater P 3 D 0 NR NR NR NR present in groundwater (pCi/l) contamination Calculation Times Start of dose calculation Time since placement of material (y) P 3 D 0 immediately after license NR NR NR NR termination 0, 1, 3, 10, 30, 100, 300, Time for calculations (y) P 3 D RESRAD Default NR NR NR NR 1000 Contaminated Zone Size of LACBWR Licensed Area of contaminated zone (m2) P 2 D 7500 NR NR NR NR Site Exclusion (LSE) area 6-70
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Surface Soil contamination Thickness of contaminated zone (m) P 2 D 1 thickness not expected to NR NR NR NR exceed 1 m.
Diameter of 7500 m2 Length parallel to aquifer flow (m) P 2 D 98 NR NR NR NR contaminated zone Does the initial contamination Contaminated zone at NA NA NA No NA NA NA NA penetrate the water table? surface Contaminated fraction below water Contaminated zone at P 3 D 0 NR NR NR NR table surface Cover and Contaminated Zone Hydrological Data Cover depth (m) P 2 D 0 No cover NR NR NR NR Density of cover material P 2 D NA No cover NR NR NR NR Cover erosion rate P,B 2 D NA No cover NR NR NR NR Density of contaminated zone e P 1 D 1.76 Site specific NR NR NR NR (g/cm3)
Contaminated zone erosion rate NUREG/CR-6697 Att. C m/y) P,B 2 S Continuous Logarithmic 5E-08 0.0007 0.005 0.2 0.0015 Table 3.8-1 Contaminated zone total porosity P 2 D 0.31 Site specifice NR NR NR NR Calculated value for sand Contaminated zone field capacity P 3 D 0.066 NR NR NR NR soil typef Contaminated zone hydraulic Site specifice P 2 D 34822 NR NR NR NR conductivity (m/y) 313 feet/day = 34822 m/y Site specific soil type sand Contaminated zone b parameter P 2 S Lognormal-N NUREG/CR-6697 Att. C -.0253 0.216 NA NA 0.97 Table 3.5-1 Median Humidity in air (g/m3) P 3 D 7.2 1.98 0.334 0.001 0.999 7.2 NUREG/CR-6697 Att. C NUREG/CR-6697 Att. C Evapotranspiration coefficient P 2 S Uniform 0.5 0.75 NR NR 0.625 Figure 4.3-1 NUREG/CR-6697 Att. C Average annual wind speed (m/s) P 2 S Bounded Lognormal - N 1.445 0.2419 1.4 13 4.2 Figure 4.5-1 NUREG/CR-6697 Att. C Precipitation (m/y) P 2 D 0.78 La Crosse, WI NR NR NR NR Table 4.1-2 Irrigation (m/y) B 3 D NA Industrial Scenario NR NR NR NR 6-71
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Irrigation mode B 3 D NA Industrial Scenario NR NR NR NR NUREG/CR-6697 Att. C Runoff coefficient P 2 S Uniform 0.1 0.8 NR NR 0.45 Figure 4.2-1 Watershed area for nearby stream P 3 D 1.00E+06 RESRAD Default NR NR NR NR or pond (m2)
Accuracy for water/soil
- 3 D 1.00E-03 RESRAD Default NR NR NR NR computations Saturated Zone Hydrological Data Density of saturated zone (g/cm3) P 2 D 1.76 Site-specifice NR NR NR NR e
Saturated zone total porosity P 1 D 0.31 Site-specific NR NR NR NR Saturated zone effective porosity P 1 D 0.28 Site-specifice NR NR NR NR Calculated values for sand Saturated zone field capacity P 3 D 0.066 NR NR NR NR soil typef Site-specific valuee Saturated zone hydraulic P 1 D 34822 NR NR NR NR conductivity (m/y) 313 feet/day = 34822 m/y Saturated zone hydraulic gradient P 2 D 0.0045 Site-specifice NR NR NR NR Site specific soil type sand 0.97 Saturated zone b parameter P 2 S Lognormal-N NUREG/CR-6697 Att. C -.0253 0.216 NA NA Table 3.5-1 Assumed zero due to Water table drop rate (m/y) P 3 D 0 hydraulic connectivity with NR NR NR NR Mississippi river.
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La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Site-specific distribution Existing industrial water supply wells onsite at depth of 116 and 129 below ground surface (the 129 depth equals 36.3 m below the water table). 36.3 m assumed to be maximum well depth.
Minimum well depth assumed to be represented by a nominal 20 screen depth (6.1 m) starting at the Well pump intake depth (m below P 2 S Triangular maximum seasonal water 618 21.2 36.3 NR 21.2 water table) table elevation of 629 and extending to 10 below 619 elevation where water table continuously present.
Mode is assumed to be mid-point between 6.1 m and 36.3 m which is 21.2 m.
Note that the site-specific distribution is reasonably similar to the NUREG-6697 distribution values of 6, 10, and 30 for the triangular distribution.
Model: Nondispersion (ND) or Mass- Applicable to flowing P 3 D ND NR NR NR NR Balance (MB) groundwater 6-73
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median NUREG/CR-6697, Att. C provides no recommended value due to high variability.
Industrial Scenario pump rate depends on industry.
NUREG-6697, Table 3.10-1 Well pumping rate (m3/y) P 2 S Uniform applies a sanitary and 328.7 1643.5 NR NR 986.1 potable water usage rate for four persons of 328.7 m3/yr.
This value is assumed to be the minimum industrial rate.
Maximum industrial rate assumed to supply 20 workers which equates to1643.5 m3/yr.
Unsaturated Zone Hydrological Data Number of unsaturated zone strata P 3 D 1 Site-specifice NR NR NR NR Site Specifice Calculated assuming ground surface 639 elevation, Unsat. zone thickness (m) P 1 D 2.05 m Contaminated Zone NR NR NR NR Thickness 1 m, and maximum water table elevation of 629.
1.76 Unsat. zone soil density (g/cm 3) P 2 D Site-specifice NR NR NR NR e
Unsat. zone total porosity P 2 D 0.31 Site-specific NR NR NR NR e
Unsat. zone effective porosity P 2 D 0.28 Site-specific NR NR NR NR 0.066 Calculated values for sand Unsat. zone field capacity P 3 D NR NR NR NR soil typef Unsat. zone hydraulic conductivity Site-specifice P 2 D 34822 NR NR NR NR (m/y)
Site specific soil type sand Unsat. zone soil-specific b P 2 S Lognormal-N NUREG/CR-6697 Att. C -.0253 0.216 NA NA 0.97 parameter Table 3.5-1 Occupancy 6-74
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median NUREG-6697 Att. C, Table 7.6-1 recommends a median indoor work day as 8.76 hour8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br />s/day. Assuming 5 days a week and 50 weeks per years this equates to 2190 hours0.0253 days <br />0.608 hours <br />0.00362 weeks <br />8.33295e-4 months <br /> per year.
NUREG/CR-5512, Vol. 3, Inhalation rate (m3/y) M,B 3 D 3066 Section 5.3.4 recommends NR NR NR NR an inhalation rate for workers in light industry of 1.4 m3/hr.
Industrial Scenario m3/yr =
1.4 m3/hr
- 2190 hr/yr =
3066 m3/yr.
See See See See NUREG- NUREG-NUREG- NUREG-Mass loading for inhalation (g/m3) P,B 2 S Continuous Linear NUREG/CR-6697, Att. C 6697 6697 2.35E-05 6697 Table 6697 Table Table Table 4.6-1 4.6-1 4.6-1 4.6-1 RESRAD Users Manual Exposure duration B 3 D 30 parameter value not used in NR NR NR NR dose calculation NUREG/CR-6697, Att. C Indoor dust filtration factor P,B 2 S Uniform 0.15 0.95 NR NR 0.55 Figure 7.1-1 NUREG/CR-6697, Att. C Shielding factor, external gamma P 2 S Bounded Lognormal-N -1.3 0.59 0.044 1 0.2725 Table 7.10-1 6-75
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median NUREG-6697 Att. C, Table 7.6-1 recommends a median indoor work day as 8.76 hour8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br />s/day. Assuming 5 days a week and 50 weeks per years this equates to 2190 hours0.0253 days <br />0.608 hours <br />0.00362 weeks <br />8.33295e-4 months <br /> per year.
Majority of industrial work is expected to be indoors.
Fraction of time spent indoors B 3 D 0.1875 NR NR NR NR Consistent with Table 2-3 of the Users Manual for RESRAD Version 6g 75%
of work time is indoors and 25% outdoors.
The corresponding RESRAD indoor Fraction parameter =
(2190*.75)/(24*365) = .1875 As explained in the basis for the Indoor Fraction parameter, the indoor time Fraction of time spent outdoors (on fraction was set at 75% and B 3 D 0.0625 NR NR NR NR site) outdoor time fraction at 25%.
(2190*.25)/(24*365) =
0.0625 Circular contaminated zone Shape factor flag, external gamma P 3 D Circular assumed for modeling NR NR NR NR purposes Ingestion, Dietary Fruits, non-leafy vegetables, grain M,B 2 D NA Industrial Scenario NR NR NR NR consumption (kg/y)
Leafy vegetable consumption (kg/y) M,B 3 D NA Industrial Scenario NR NR NR NR Milk consumption (L/y) M,B 2 D NA Industrial Scenario NR NR NR NR M,B Meat and poultry consumption (kg/y) 3 D NA Industrial Scenario NR NR NR NR M,B Fish consumption (kg/y) 3 D NA Industrial Scenario NR NR NR NR 6-76
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median M,B Other seafood consumption (kg/y) 3 D NA Industrial Scenario NR NR NR NR M,B NUREG/CR-5512, Vol. 3 Soil ingestion rate (g/y) 2 D 18.3 NR NR NR NR Table 6.87 NUREG/CR-5512, Vol. 3 Table 6.87 Industrial Scenario water supply assumed to be from an onsite well.
M,B Drinking water intake (L/y) 2 D 327 478 L/y from NUREG/CR- NR NR NR NR 5512 corresponds to 1.31 L/d which is considered a conservative value for 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> work day.
1.31 L/d
- 250 work days =
327 L/y Contamination fraction of drinking All water assumed B,P 3 D 1 NR NR NR NR water contaminated Contamination fraction of household B,P 3 1 All water from well water (if used)
Contamination fraction of livestock B,P 3 D NA Industrial Scenario NR NR NR NR water Contamination fraction of irrigation B,P 3 D NA Industrial Scenario NR NR NR NR water Contamination fraction of aquatic B,P 2 D NA Industrial Scenario NR NR NR NR food Contamination fraction of plant food B,P 3 D NA Industrial Scenario NR NR NR NR Contamination fraction of meat B,P 3 D NA Industrial Scenario NR NR NR NR Contamination fraction of milk B,P 3 D NA Industrial Scenario NR NR NR NR Ingestion, Non-Dietary Livestock fodder intake for meat M 3 D NA Industrial Scenario NR NR NR NR (kg/day)
Livestock fodder intake for milk M 3 D NA Industrial Scenario NR NR NR NR (kg/day)
Livestock water intake for meat M 3 D NA Industrial Scenario NR NR NR NR (L/day) 6-77
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Livestock water intake for milk M 3 D NA Industrial Scenario NR NR NR NR (L/day)
Livestock soil intake (kg/day) M 3 D NA Industrial Scenario NR NR NR NR Mass loading for foliar deposition P 3 D NA Industrial Scenario NR NR NR NR (g/m3)
NUREG/CR-6697, Att. C Depth of soil mixing layer (m) P 2 S Triangular 0 0.15 0.6 NR 0.15 Figure 3.12-1 Depth of roots (m) P 1 D NA Industrial Scenario NR NR NR NR Drinking water fraction from ground B,P 3 D 1 Industrial Scenario NR NR NR NR water Household water fraction from B,P 3 1 Industrial Scenario NR NR NR NR ground water (if used)
Livestock water fraction from ground B,P 3 D NA Industrial Scenario NR NR NR NR water Irrigation fraction from ground water B,P 3 D NA Industrial Scenario NR NR NR NR Wet weight crop yield for Non-Leafy P 2 D NA Industrial Scenario NR NR NR NR (kg/m2)
Wet weight crop yield for Leafy P 3 D NA Industrial Scenario NR NR NR NR (kg/m2)
Wet weight crop yield for Fodder P 3 D NA Industrial Scenario NR NR NR NR (kg/m2)
Growing Season for Non-Leafy (y) P 3 D NA Industrial Scenario NR NR NR NR Growing Season for Leafy (y) P 3 D NA Industrial Scenario NR NR NR NR Growing Season for Fodder (y) P 3 D NA Industrial Scenario NR NR NR NR Translocation Factor for Non-Leafy P 3 D NA Industrial Scenario NR NR NR NR Translocation Factor for Leafy P 3 D NA Industrial Scenario NR NR NR NR Translocation Factor for Fodder P 3 D NA Industrial Scenario NR NR NR NR Weathering Removal Constant for P 2 D NA Industrial Scenario NR NR NR NR Vegetation (1/y)
Wet Foliar Interception Fraction for P 3 D NA Industrial Scenario NR NR NR NR Non-Leafy Wet Foliar Interception Fraction for P 2 D NA Industrial Scenario NR NR NR NR Leafy Wet Foliar Interception Fraction for P 3 D NA Industrial Scenario NR NR NR NR Fodder 6-78
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Dry Foliar Interception Fraction for P 3 D NA Industrial Scenario NR NR NR NR Non-Leafy Dry Foliar Interception Fraction for P 3 D NA Industrial Scenario NR NR NR NR Leafy Dry Foliar Interception Fraction for P 3 D NA Industrial Scenario NR NR NR NR Fodder Storage times of contaminated foodstuffs (days):
Fruits, non-leafy vegetables, and B 3 D NA Industrial Scenario NR NR NR NR grain Leafy vegetables B 3 D NA Industrial Scenario NR NR NR NR Milk B 3 D NA Industrial Scenario NR NR NR NR Meat and poultry B 3 D NA Industrial Scenario NR NR NR NR Fish B 3 D NA Industrial Scenario NR NR NR NR Crustacea and mollusks B 3 D NA Industrial Scenario NR NR NR NR Well water B 3 D NA Industrial Scenario NR NR NR NR Surface water B 3 D NA Industrial Scenario NR NR NR NR Livestock fodder B 3 D NA Industrial Scenario NR NR NR NR Special Radionuclides (C-14)
C-12 concentration in water (g/cm3) P 3 D NA Industrial Scenario NR NR NR NR C-12 concentration in contaminated P 3 D NA Industrial Scenario NR NR NR NR soil (g/g)
Fraction of vegetation carbon from P 3 D NA Industrial Scenario NR NR NR NR soil Fraction of vegetation carbon from P 3 D NA Industrial Scenario NR NR NR NR air C-14 evasion layer thickness in soil P 2 D NA Industrial Scenario NR NR NR NR (m)
C-14 evasion flux rate from soil P 3 D NA Industrial Scenario NR NR NR NR (1/sec)
C-12 evasion flux rate from soil P 3 D NA Industrial Scenario NR NR NR NR (1/sec)
Fraction of grain in beef cattle feed B 3 D NA Industrial Scenario NR NR NR NR Fraction of grain in milk cow feed B 3 D NA Industrial Scenario NR NR NR NR Dose Conversion Factors (Inhalation mrem/pCi) 6-79
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Ac-227 M 3 D 6.70E+00 FGR11 NR NR NR NR Am-241 M 3 D 4.44E-01 FGR11 NR NR NR NR Am-243 M 3 D 4.40E-01 FGR11 NR NR NR NR C-14 M 3 D 2.09E-06 FGR11 NR NR NR NR Cm-243 M 3 D 3.07E-01 FGR11 NR NR NR NR Cm-244 M 3 D 2.48E-01 FGR11 NR NR NR NR Cm-245 M 3 D 4.55E-01 FGR11 NR NR NR NR Cm-246 M 3 D 4.51E-01 FGR11 NR NR NR NR Co-60 M 3 D 2.19E-04 FGR11 NR NR NR NR Cs-134 M 3 D 4.62E-05 FGR11 NR NR NR NR Cs-137 M 3 D 3.19E-05 FGR11 NR NR NR NR Eu-152 M 3 D 2.21E-04 FGR11 NR NR NR NR Eu-154 M 3 D 2.86E-04 FGR11 NR NR NR NR Gd-152 M 3 D 2.43E-01 FGR11 NR NR NR NR H-3 M 3 D 6.40E-08 FGR11 NR NR NR NR I-129 M 3 D 1.74E-04 FGR11 NR NR NR NR Nb-94 M 3 D 4.14E-04 FGR11 NR NR NR NR e
Nd-144 M 3 D 7.04E-02 ICRP60 NR NR NR NR Ni-59 M 3 D 2.70E-06 FGR11 NR NR NR NR Ni-63 M 3 D 6.29E-06 FGR11 NR NR NR NR Np-237 M 3 D 5.40E-01 FGR11 NR NR NR NR Pa-231 M 3 D 1.28E+00 FGR11 NR NR NR NR Pb-210 M 3 D 1.36E-02 FGR11 NR NR NR NR Po-210 M 3 D 9.40E-03 FGR11 NR NR NR NR Pu-238 M 3 D 3.92E-01 FGR11 NR NR NR NR Pu-239 M 3 D 4.29E-01 FGR11 NR NR NR NR Pu-240 M 3 D 4.29E-01 FGR11 NR NR NR NR Pu-241 M 3 D 8.25E-03 FGR11 NR NR NR NR Pu-242 M 3 D 4.11E-01 FGR11 NR NR NR NR Ra-226 M 3 D 8.58E-03 FGR11 NR NR NR NR 6-80
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Ra-228 M 3 D 4.77E-03 FGR11 NR NR NR NR e
Sm-148 M 3 D 7.34E-02 ICRP60 NR NR NR NR Sr-90 M 3 D 1.30E-03 FGR11 NR NR NR NR Tc-99 M 3 D 8.32E-06 FGR11 NR NR NR NR Th-228 M 3 D 3.42E-01 FGR11 NR NR NR NR Th-229 M 3 D 2.15E+00 FGR11 NR NR NR NR Th-230 M 3 D 3.26E-01 FGR11 NR NR NR NR Th232 M 3 D 1.64e+00 FGR11 NR NR NR NR U-233 M 3 D 1.35E-01 FGR11 NR NR NR NR U-234 M 3 D 1.32E-01 FGR11 NR NR NR NR U-235 M 3 D 1.23E-01 FGR11 NR NR NR NR U-236 M 3 D 1.25E-01 FGR11 NR NR NR NR U-238 M 3 D 1.18E-01 FGR11 NR NR NR NR Dose Conversion Factors (Ingestion mrem/pCi)
Ac-227 M 3 D 1.41E-02 FGR11 NR NR NR NR Am-241 M 3 D 3.64E-03 FGR11 NR NR NR NR Am-243 M 3 D 3.62E-03 FGR11 NR NR NR NR C-14 M 3 D 2.09E-06 FGR11 NR NR NR NR Cm-243 M 3 D 2.51E-03 FGR11 NR NR NR NR Cm-244 M 3 D 2.02E-03 FGR11 NR NR NR NR Cm-245 M 3 D 3.74E-03 FGR11 NR NR NR NR Cm-246 M 3 D 3.70E-03 FGR11 NR NR NR NR Co-60 M 3 D 2.69E-05 FGR11 NR NR NR NR Cs-134 M 3 D 7.33E-05 FGR11 NR NR NR NR Cs-137 M 3 D 5.00E-05 FGR11 NR NR NR NR Eu-152 M 3 D 6.48E-06 FGR11 NR NR NR NR Eu-154 M 3 D 9.55E-06 FGR11 NR NR NR NR Gd-152 M 3 D 1.61E-04 FGR11 NR NR NR NR H-3 M 3 D 6.40E-08 FGR11 NR NR NR NR 6-81
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median I-129 M 3 D 2.76E-04 FGR11 NR NR NR NR Nb-94 M 3 D 7.14E-06 FGR11 NR NR NR NR Nd-144e M 3 D 1.51E-04 ICRP60 NR NR NR NR Ni-59 M 3 D 2.10E-07 FGR11 NR NR NR NR Ni-63 M 3 D 5.77E-07 FGR11 NR NR NR NR Np-237 M 3 D 4.44E-03 FGR11 NR NR NR NR Pa-231 M 3 D 1.06E-02 FGR11 NR NR NR NR Pb-210 M 3 D 5.37E-03 FGR11 NR NR NR NR Po-210 M 3 D 1.90E-03 FGR11 NR NR NR NR Pu-238 M 3 D 3.20E-03 FGR11 NR NR NR NR Pu-239 M 3 D 3.54E-03 FGR11 NR NR NR NR Pu-240 M 3 D 3.54E-03 FGR11 NR NR NR NR Pu-241 M 3 D 6.84E-05 FGR11 NR NR NR NR Pu-242 M 3 D 3.36E-03 FGR11 NR NR NR NR Ra-226 M 3 D 1.32E-03 FGR11 NR NR NR NR Ra-228 M 3 D 1.44E-03 FGR11 NR NR NR NR Sm-148e M 3 D 1.58E-04 ICRP60 NR NR NR NR Sr-90 M 3 D 1.42E-04 FGR11 NR NR NR NR Tc-99 M 3 D 1.46E-06 FGR11 NR NR NR NR Th-228 M 3 D 3.96E-04 FGR11 NR NR NR NR Th-229 M 3 D 3.53E-03 FGR11 NR NR NR NR Th-230 M 3 D 5.48E-04 FGR11 NR NR NR NR Th-232 M 3 D 2.73E-03 FGR11 NR NR NR NR U-233 M 3 D 2.89E-04 FGR11 NR NR NR NR U-234 M 3 D 2.83E-04 FGR11 NR NR NR NR U-235 M 3 D 2.66E-04 FGR11 NR NR NR NR U-236 M 3 D 2.69E-04 FGR11 NR NR NR NR U-238 M 3 D 2.55E-04 FGR11 NR NR NR NR Plant Transfer Factors (pCi/g plant)/(pCi/g soil)
Ac-227 P 1 D NA Industrial Scenario NR NR NR NR 6-82
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Am-241 P 1 D NA Industrial Scenario NR NR NR NR Am-243 P 1 D NA Industrial Scenario NR NR NR NR C-14 P 1 D NA Industrial Scenario NR NR NR NR Cm-243 P 1 D NA Industrial Scenario NR NR NR NR Cm-244 P 1 D NA Industrial Scenario NR NR NR NR Co-60 P 1 D NA Industrial Scenario NR NR NR NR Cs-134 P 1 D NA Industrial Scenario NR NR NR NR Cs-137 P 1 D NA Industrial Scenario NR NR NR NR Eu-152 P 1 D NA Industrial Scenario NR NR NR NR Eu-154 P 1 D NA Industrial Scenario NR NR NR NR Fe-55 P 1 D NA Industrial Scenario NR NR NR NR Gd-152 P 1 D NA Industrial Scenario NR NR NR NR H-3 P 1 D NA Industrial Scenario NR NR NR NR Nb-94 P 1 D NA Industrial Scenario NR NR NR NR Nd-144 P 1 D NA Industrial Scenario NR NR NR NR Ni-59 P 1 D NA Industrial Scenario NR NR NR NR Ni-63 P 1 D NA Industrial Scenario NR NR NR NR Np-237 P 1 D NA Industrial Scenario NR NR NR NR Pa-231 P 1 D NA Industrial Scenario NR NR NR NR Pb-210 P 1 D NA Industrial Scenario NR NR NR NR Pm-147 P 1 D NA Industrial Scenario NR NR NR NR Po-210 P 1 D NA Industrial Scenario NR NR NR NR Pu-238 P 1 D NA Industrial Scenario NR NR NR NR Pu-239 P 1 D NA Industrial Scenario NR NR NR NR Pu-240 P 1 D NA Industrial Scenario NR NR NR NR Pu-241 P 1 D NA Industrial Scenario NR NR NR NR Ra-226 P 1 D NA Industrial Scenario NR NR NR NR Ra-228 P 1 D NA Industrial Scenario NR NR NR NR Sb-125 P 1 D NA Industrial Scenario NR NR NR NR Sm-148 P 1 D NA Industrial Scenario NR NR NR NR 6-83
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Sr-90 P 1 D NA Industrial Scenario NR NR NR NR Tc-99 P 1 D NA Industrial Scenario NR NR NR NR Th-228 P 1 D NA Industrial Scenario NR NR NR NR Th-229 P 1 D NA Industrial Scenario NR NR NR NR Th-230 P 1 D NA Industrial Scenario NR NR NR NR Th-232 P 1 D NA Industrial Scenario NR NR NR NR U-233 P 1 D NA Industrial Scenario NR NR NR NR U-234 P 1 D NA Industrial Scenario NR NR NR NR U-235 P 1 D NA Industrial Scenario NR NR NR NR U-236 P 1 D NA Industrial Scenario NR NR NR NR Meat Transfer Factors (pCi/kg)/(pCi/d)
Ac-227 P 2 D NA Industrial Scenario NR NR NR NR Ag-108m P 2 D NA Industrial Scenario NR NR NR NR Am-241 P 2 D NA Industrial Scenario NR NR NR NR Am-243 P 2 D NA Industrial Scenario NR NR NR NR C-14 P 2 D NA Industrial Scenario NR NR NR NR Cm-243 P 2 D NA Industrial Scenario NR NR NR NR Cm-244 P 2 D NA Industrial Scenario NR NR NR NR Co-60 P 2 D NA Industrial Scenario NR NR NR NR Cs-134 P 2 D NA Industrial Scenario NR NR NR NR Cs-137 P 2 D NA Industrial Scenario NR NR NR NR Eu-152 P 2 D NA Industrial Scenario NR NR NR NR Eu-154 P 2 D NA Industrial Scenario NR NR NR NR Fe-55 P 2 D NA Industrial Scenario NR NR NR NR Gd-152 P 2 D NA Industrial Scenario NR NR NR NR H-3 P 2 D NA Industrial Scenario NR NR NR NR Nb-94 P 2 D NA Industrial Scenario NR NR NR NR Nd-144 P 2 D NA Industrial Scenario NR NR NR NR Ni-59 P 2 D NA Industrial Scenario NR NR NR NR Ni-63 P 2 D NA Industrial Scenario NR NR NR NR 6-84
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Np-237 P 2 D NA Industrial Scenario NR NR NR NR Pa-231 P 2 D NA Industrial Scenario NR NR NR NR Pb-210 P 2 D NA Industrial Scenario NR NR NR NR Po-210 P 2 D NA Industrial Scenario NR NR NR NR Pu-238 P 2 D NA Industrial Scenario NR NR NR NR Pu-239 P 2 D NA Industrial Scenario NR NR NR NR Pu-240 P 2 D NA Industrial Scenario NR NR NR NR Pu-241 P 2 D NA Industrial Scenario NR NR NR NR Ra-226 P 2 D NA Industrial Scenario NR NR NR NR Ra-228 P 2 D NA Industrial Scenario NR NR NR NR Sb-125 P 2 D NA Industrial Scenario NR NR NR NR Sm-148 P 1 D NA Industrial Scenario NR NR NR NR Sr-90 P 2 D NA Industrial Scenario NR NR NR NR Tc-99 P 2 D NA Industrial Scenario NR NR NR NR Th-228 P 2 D NA Industrial Scenario NR NR NR NR Th-229 P 2 D NA Industrial Scenario NR NR NR NR Th-230 P 2 D NA Industrial Scenario NR NR NR NR Th-232 P 2 D NA Industrial Scenario NR NR NR NR U-233 P 2 D NA Industrial Scenario NR NR NR NR U-234 P 2 D NA Industrial Scenario NR NR NR NR U-235 P 2 D NA Industrial Scenario NR NR NR NR U-236 P 2 D NA Industrial Scenario NR NR NR NR Milk Transfer Factors (pCi/L)/(pCi/d)
Ac-227 P 2 D NA Industrial Scenario NR NR NR NR Am-241 P 2 D NA Industrial Scenario NR NR NR NR Am-243 P 2 D NA Industrial Scenario NR NR NR NR C-14 P 2 D NA Industrial Scenario NR NR NR NR Cm-243 P 2 D NA Industrial Scenario NR NR NR NR Cm-244 P 2 D NA Industrial Scenario NR NR NR NR Co-60 P 2 D NA Industrial Scenario NR NR NR NR 6-85
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Cs-134 P 2 D NA Industrial Scenario NR NR NR NR Cs-137 P 2 D NA Industrial Scenario NR NR NR NR Eu-152 P 2 D NA Industrial Scenario NR NR NR NR Eu-154 P 2 D NA Industrial Scenario NR NR NR NR Fe-55 P 2 D NA Industrial Scenario NR NR NR NR Gd-152 P 2 D NA Industrial Scenario NR NR NR NR H-3 P 2 D NA Industrial Scenario NR NR NR NR Nb-94 P 2 D NA Industrial Scenario NR NR NR NR Nd-144 P 2 D NA Industrial Scenario NR NR NR NR Ni-59 P 2 D NA Industrial Scenario NR NR NR NR Ni-63 P 2 D NA Industrial Scenario NR NR NR NR Np-237 P 2 D NA Industrial Scenario NR NR NR NR Pa-231 P 2 D NA Industrial Scenario NR NR NR NR Pb-210 P 2 D NA Industrial Scenario NR NR NR NR Po-210 P 2 D NA Industrial Scenario NR NR NR NR Pu-238 P 2 D NA Industrial Scenario NR NR NR NR Pu-239 P 2 D NA Industrial Scenario NR NR NR NR Pu-240 P 2 D NA Industrial Scenario NR NR NR NR Pu-241 P 2 D NA Industrial Scenario NR NR NR NR Ra-226 P 2 D NA Industrial Scenario NR NR NR NR Ra-228 P 2 D NA Industrial Scenario NR NR NR NR Sm-148 P 2 D NA Industrial Scenario NR NR NR NR Sr-90 P 2 D NA Industrial Scenario NR NR NR NR Tc-99 P 2 D NA Industrial Scenario NR NR NR NR Th-228 P 2 D NA Industrial Scenario NR NR NR NR Th-229 P 2 D NA Industrial Scenario NR NR NR NR Th-230 P 2 D NA Industrial Scenario NR NR NR NR Th-232 P 2 D NA Industrial Scenario NR NR NR NR U-233 P 2 D NA Industrial Scenario NR NR NR NR U-234 P 2 D NA Industrial Scenario NR NR NR NR 6-86
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median U-235 P 2 D NA Industrial Scenario NR NR NR NR U-236 P 2 D NA Industrial Scenario NR NR NR NR Bioaccumulation Factors for Fish ((pCi/kg)/(pCi/L))
Ac-227 P 2 D NA Industrial Scenario NR NR NR NR Am-241 P 2 D NA Industrial Scenario NR NR NR NR Am-243 P 2 D NA Industrial Scenario NR NR NR NR C-14 P 2 D NA Industrial Scenario NR NR NR NR Cm-243 P 2 D NA Industrial Scenario NR NR NR NR Cm-244 P 2 D NA Industrial Scenario NR NR NR NR Cm-245 P 2 D NA Industrial Scenario NR NR NR NR Cm-246 P 2 D NA Industrial Scenario NR NR NR NR Co-60 P 2 D NA Industrial Scenario NR NR NR NR Cs-137 P 2 D NA Industrial Scenario NR NR NR NR Eu-152 P 2 D NA Industrial Scenario NR NR NR NR Eu-154 P 2 D NA Industrial Scenario NR NR NR NR Gd-152 P 2 D NA Industrial Scenario NR NR NR NR H-3 P 2 D NA Industrial Scenario NR NR NR NR I-129 P 2 D NA Industrial Scenario NR NR NR NR Nb-94 P 2 D NA Industrial Scenario NR NR NR NR Ni-59 P 2 D NA Industrial Scenario NR NR NR NR Ni-63 P 2 D NA Industrial Scenario NR NR NR NR Np-237 P 2 D NA Industrial Scenario NR NR NR NR Pa-231 P 2 D NA Industrial Scenario NR NR NR NR Po-210 P 2 D NA Industrial Scenario NR NR NR NR Pb-210 P 2 D NA Industrial Scenario NR NR NR NR Pu-238 P 2 D NA Industrial Scenario NR NR NR NR Pu-239 P 2 D NA Industrial Scenario NR NR NR NR Pu-240 P 2 D NA Industrial Scenario NR NR NR NR Pu-241 P 2 D NA Industrial Scenario NR NR NR NR 6-87
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Pu-242 P 2 D NA Industrial Scenario NR NR NR NR Ra-226 P 2 D NA Industrial Scenario NR NR NR NR Ra-228 P 2 D NA Industrial Scenario NR NR NR NR Sr-90 P 2 D NA Industrial Scenario NR NR NR NR Th-228 P 2 D NA Industrial Scenario NR NR NR NR Th-229 P 2 D NA Industrial Scenario NR NR NR NR Th-230 P 2 D NA Industrial Scenario NR NR NR NR Th-232 P 2 D NA Industrial Scenario NR NR NR NR U-233 P 2 D NA Industrial Scenario NR NR NR NR U-234 P 2 D NA Industrial Scenario NR NR NR NR U-235 P 2 D NA Industrial Scenario NR NR NR NR U-236 P 2 D NA Industrial Scenario NR NR NR NR U-238 P 2 D NA Industrial Scenario NR NR NR NR Bioaccumulation Factors for Crustacea/ Mollusks ((pCi/kg)/(pCi/L))
Ac-227 P 3 D NA Industrial Scenario NR NR NR NR Am-241 P 3 D NA Industrial Scenario NR NR NR NR Am-243 P 3 D NA Industrial Scenario NR NR NR NR C-14 P 3 D NA Industrial Scenario NR NR NR NR Cm-243 P 3 D NA Industrial Scenario NR NR NR NR Cm-244 P 3 D NA Industrial Scenario NR NR NR NR Cm-245 P 3 D NA Industrial Scenario NR NR NR NR Cm-246 P 3 D NA Industrial Scenario NR NR NR NR Co-60 P 3 D NA Industrial Scenario NR NR NR NR Cs-137 P 3 D NA Industrial Scenario NR NR NR NR Eu-152 P 3 D NA Industrial Scenario NR NR NR NR Eu-154 P 3 D NA Industrial Scenario NR NR NR NR Gd-152 P 3 D NA Industrial Scenario NR NR NR NR H-3 P 3 D NA Industrial Scenario NR NR NR NR I-129 P 3 D NA Industrial Scenario NR NR NR NR 6-88
La Crosse Boiling Water Reactor License Termination Plan Revision 2 SOIL DCGL: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Nb-94 P 3 D NA Industrial Scenario NR NR NR NR Ni-59 P 3 D NA Industrial Scenario NR NR NR NR Ni-63 P 3 D NA Industrial Scenario NR NR NR NR Np-237 P 3 D NA Industrial Scenario NR NR NR NR Pa-231 P 3 D NA Industrial Scenario NR NR NR NR Pb-210 P 3 D NA Industrial Scenario NR NR NR NR Po-210 P S D NA Industrial Scenario NR NR NR NR Pu-238 P 3 D NA Industrial Scenario NR NR NR NR Pu-239 P 3 D NA Industrial Scenario NR NR NR NR Pu-240 P 3 D NA Industrial Scenario NR NR NR NR Pu-241 P 3 D NA Industrial Scenario NR NR NR NR Pu-242 P 3 D NA Industrial Scenario NR NR NR NR Ra-226 P 3 D NA Industrial Scenario NR NR NR NR Ra-228 P 3 D NA Industrial Scenario NR NR NR NR Sr-90 P 3 D NA Industrial Scenario NR NR NR NR Th-228 P 3 D NA Industrial Scenario NR NR NR NR Th-229 P 3 D NA Industrial Scenario NR NR NR NR Th-230 P 3 D NA Industrial Scenario NR NR NR NR Th-232 P 3 D NA Industrial Scenario NR NR NR NR U-233 P 3 D NA Industrial Scenario NR NR NR NR U-234 P 3 D NA Industrial Scenario NR NR NR NR U-235 P 3 D NA Industrial Scenario NR NR NR NR U-236 P 3 D NA Industrial Scenario NR NR NR NR U-238 P 3 D NA Industrial Scenario NR NR NR NR Graphics Parameters Number of points 32 RESRAD Default NR NR NR NR Spacing log RESRAD Default NR NR NR NR Time integration parameters Maximum number of points for dose 17 RESRAD Default NR NR NR NR Notes:
6-89
La Crosse Boiling Water Reactor License Termination Plan Revision 2 a P = physical, B = behavioral, M = metabolic; (see NUREG/CR-6697, Attachment B, Table 4.)
b 1 = high-priority parameter, 2 = medium-priority parameter, 3 = low-priority parameter (see NUREG/CR-6697, Attachment B, Table 4.1) c D = deterministic, S = stochastic d Distributions Statistical Parameters:
Lognormal-n: 1= mean, 2 = standard deviation Bounded lognormal-n: 1= mean, 2 = standard deviation, 3 = minimum, 4 = maximum Truncated lognormal-n: 1= mean, 2 = standard deviation, 3 = lower quantile, 4 = upper quantile Bounded normal: 1 = mean, 2 = standard deviation, 3 = minimum, 4 = maximum Beta: 1 = minimum, 2 = maximum, 3 = P-value, 4 = Q-value Triangular: 1 = minimum, 2 = mode, 3 = maximum Uniform: 1 = minimum, 2 = maximum e Sm-148 an ND-144 not listed in RESRAD FGR 11 DCF file e
Reference:
Haley and Aldrich, Inc., "Hydrogeological Investigation Report La Crosse Boiling Water Reactor, Dairyland Power Cooperative, Genoa, WI January 2015 f ZionSolutions Technical Support Document 14-003, Conestoga Rovers & Associates (CRA) Report, Zion Hydrogeologic Investigation Report g Argonne National Laboratory, Users Manual for RESRAD Version 6, ANL/EAD 4, July 2001 6-90
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ATTACHMENT 6-2 RESRAD Input Parameters for LACBWR BFM Uncertainty Analysis 6-91
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Soil Concentrations Basic radiation dose limit (mrem/y) 3 D 25 10 CFR 20.1402 NR NR NR NR Initial principal radionuclide (pCi/g) P 2 D 1 Unit Value NR NR NR NR Distribution coefficients (contaminated, unsaturated, and saturated zones) (cm 3/g)
Mean Kd Value for sand Ac-227 (daughter of Cm-243 and NUREG/CR-6697, Table P 1 D 450 6.72 3.22 NA NA 825 Pu-239) 3.9-2, Sheppard and Thibault Am-241 (also daughter of Cm-245 Not Included in < 0.1% of radionuclide P 1 S NA NA NA NA NA and Pu-241) Uncertainty Analysis mixture Not Included in < 0.1% of radionuclide Am-243 P 1 S NA NA NA NA NA Uncertainty Analysis mixture C-14 P 1 S Lognormal-N NUREG/CR-6697 Att. C 2.4 3.22 NA NA 11 Not Included in < 0.1% of radionuclide Cm-243 P 1 S NA NA NA NA NA Uncertainty Analysis mixture Not Included in Uncertainty < 0.1% of radionuclide Cm-244 P 1 S NA NA NA NA NA Analysis mixture Co-60 P 1 S Lognormal-N NUREG/CR-6697 Att. C 5.46 2.53 NA NA 235 Cs-137 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.1 2.33 NA NA 446 Eu-152 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.72 3.22 NA NA 825 Eu-154 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.72 3.22 NA NA 825 Not Included in < 0.1% of radionuclide Eu-155 P 1 S NA NA NA NA NA Uncertainty Analysis mixture Fe-55 P 1 S Lognormal-N NUREG/CR-6697 Att. C 5.34 2.67 NA NA 209 Median Value NUREG/CR-6697, Att. C Gd-152 (daughter for Eu-152) P 1 D 825 6.72 3.22 NA NA 825 (No sand value listed in Table 3.9-2)
H-3 P 1 S Lognormal-N NUREG/CR-6697 Att. C -2.81 0.5 NA NA 0.06 Not Included in < 0.1% of radionuclide Nb-94 P 1 S NA NA NA NA NA Uncertainty Analysis mixture RESRADv.7.0 Default Nd-144 (daughter for Eu-152) P 1 D 158 Nd not listed in NA NA NA NA NA NUREG/CR-6697 Ni-59 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.05 1.46 NA NA 424 Ni-63 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.05 1.46 NA NA 424 6-92
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Np-237 (also daughter for Am-241, Not Included in < 0.1% of radionuclide P 1 S NA NA NA NA NA Cm-245, and Pu-241) Uncertainty Analysis mixture Pa-231 (daughter for Cm-243 and Not Included in parent < 0.1% of P 1 D NA NA NA NA NA Pu-239) Uncertainty Analysis radionuclide mixture Not Included in parent < 0.1% of Pb-210 (daughter for Pu-238) P 1 D NA NA NA NA NA Uncertainty Analysis radionuclide mixture Not Included in parent < 0.1% of Po-210 (daughter Pu-238) P 1 D NA NA NA NA NA Uncertainty Analysis radionuclide mixture Not Included in < 0.1% of radionuclide Pu-238 P 1 S NA NA NA NA NA Uncertainty Analysis mixture Not Included in < 0.1% of radionuclide Pu-239 (also daughter for Cm-243) P 1 S NA NA NA NA NA Uncertainty Analysis mixture Not Included in < 0.1% of radionuclide Pu-240 (also daughter for Cm-244) P 1 S NA NA NA NA NA Uncertainty Analysis mixture Pu-241 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.86 1.89 NA NA 953 Not Included in parent < 0.1% of Ra-226 (daughter Pu-238) P 1 D NA NA NA NA NA Uncertainty Analysis radionuclide mixture Ra-228 (daughter Cm-244 and Pu- Not Included in parent < 0.1% of P 1 D NA NA NA NA NA 240) Uncertainty Analysis radionuclide mixture Mean Kd Value for sand NUREG/CR-6697, Table Sm-148 (daughter Eu-152) P 1 D 245 6.72 3.22 NA NA 825 3.9-2, Sheppard and Thibault Sr-90 P 1 S Lognormal-N NUREG/CR-6697 Att. C 3.45 2.12 NA NA 32 Not Included in < 0.1% of radionuclide Tc-99 P 1 S NA NA NA NA NA Uncertainty Analysis mixture Th-228 (daughter Cm-244 and Pu- Not Included in parent < 0.1% of P 1 D NA NA NA NA NA 240) Uncertainty Analysis radionuclide mixture Mean Kd Value for sand Th-229 (daughter Am-241, Cm-245, NUREG/CR-6697, Table P 1 D 3200 8.68 3.62 NA NA 5884 Np-237, and Pu-241) 3.9-2, Sheppard and Thibault Th-230 (daughter Cm-246 and Pu- Not Included in parent < 0.1% of P 1 D NA NA NA NA NA 238) Uncertainty Analysis radionuclide mixture Th-232 (daughter Cm-244 and Pu- Not Included in parent < 0.1% of P 1 D NA NA NA NA NA 240) Uncertainty Analysis radionuclide mixture Mean Kd Value for sand U-233 (daughter Am-241, Cm-245, NUREG/CR-6697, Table P 1 D 35 4.84 3.13 NA NA 126 Np-237, and Pu-241) 3.9-2, Sheppard and Thibault C 6-93
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Not Included in parent < 0.1% of U-234 (daughter Pu-238) P 1 D NA NA NA NA NA Uncertainty Analysis radionuclide mixture U-235 (daughter Cm-243 and Pu- Not Included in parent < 0.1% of P 1 D NA NA NA NA NA 239) Uncertainty Analysis radionuclide mixture U-236 (daughter Cm-244 and Pu- Not Included in parent < 0.1% of P 1 D NA NA NA NA NA 240) Uncertainty Analysis radionuclide mixture Initial concentration of radionuclides No existing groundwater P 3 D 0 NR NR NR NR present in groundwater (pCi/l) contamination Calculation Times Start of dose calculation Time since placement of material (y) P 3 D 0 immediately after license NR NR NR NR termination 0, 1, 3, 10, 30, 100, 300, Time for calculations (y) P 3 D RESRAD Default NR NR NR NR 1000 Contaminated Zone Source term and physical geometries vary for the five structures evaluated for Area of contaminated zone (m2) P 2 D Variable NR NR NR NR BFM Insitugw Uncertainty Analyses (see Table 6-8 in text)
Source term and physical geometries vary for the five structures evaluated for Thickness of contaminated zone (m) P 2 D Variable NR NR NR NR BFM Insitugw Uncertainty Analyses (see Table 6-8 in text)
Source term and physical geometries vary for the five structures evaluated for Length parallel to aquifer flow (m) P 2 D Variable NR NR NR NR BFM Insitugw Uncertainty Analyses (see Table 6-8 in text)
Contaminated fraction below water table Source term and physical geometries vary for the five structures evaluated for P 3 D Variable NR NR NR NR BFM Insitugw Uncertainty Analyses (see Table 6-8 in text) 6-94
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Cover and Contaminated Zone Hydrological Data Source term and physical geometries vary for the five structures evaluated for Cover depth (m) P 2 D Variable NR NR NR NR BFM Insitugw Uncertainty Analyses (see Table 6-8 in text)
Density of cover material P 2 D 1.76 Site-specifice NR NR NR NR NUREG/CR-6697 Att. C Cover erosion rate P,B 2 S Continuous Logarithmic 5E-08 0.0007 0.005 0.2 0.0015 Table 3.8-1 Density of contaminated zone P 1 D 1.76 Site-specifice NR NR NR NR (g/cm3)
Contaminated zone erosion rate NUREG/CR-6697 Att. C P,B 2 S Continuous Logarithmic 5E-08 0.0007 0.005 0.2 0.0015 m/y) Table 3.8-1 e
Contaminated zone total porosity P 2 D 0.31 Site-specific NR NR NR NR 0.43 RESRAD default. No distribution or median value Contaminated zone field capacity P 3 D 0.2 NR NR NR NR provided in NURE/CR-6697 Att. C Site-specifice Contaminated zone hydraulic P 2 D 34822 NR NR NR NR conductivity (m/y) 313 feet/day = 34822 m/y Site specific soil type sand 0.97 Contaminated zone b parameter P 2 S Lognormal-N NUREG/CR-6697 Att. C -.0253 0.216 NA NA Table 3.5-1 Median Humidity in air (g/m3) P 3 D 7.2 1.98 0.334 0.001 0.999 7.2 NUREG/CR-6697 Att. C Site-specific value to force the Precipitation parameter Evapotranspiration coefficient P 2 S 0 to equal to the infiltration NR NR NR NR NR rate (see text section 6.11.1.2)
NUREG/CR-6697 Att. C Average annual wind speed (m/s) P 2 S Bounded Lognormal - N 1.445 0.2419 1.4 13 4.2 Figure 4.5-1 6-95
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median For the BFM analysis the Evapotranspiration coefficient and Runoff coefficient were set to zero to force the Precipitation parameter to equal the infiltration rate.
This was necessary to incorporate the effect of seasonal ground water elevation rise associated with the Mississippi river stage.
Precipitation (m/y) P 2 S Uniform 0.25 3.05 NR NR Minimum value of 0.25 is traditional infiltration rate based on Evapotranspiration coefficient, Runoff coefficient and site precipitation rate parameters listed in the parameter set for Soil DCGL.
The maximum value of 3.05 is the seasonal high groundwater elevation at the site as driven by river stage.
Irrigation (m/y) B 3 D NA Industrial Scenario NR NR NR NR Irrigation mode B 3 D NA Industrial Scenario NR NR NR NR Site-specific value to force the Precipitation parameter Runoff coefficient P 2 D 0 NR NR NR NR NR to equal to the infiltration rate. See section 6.11.1.2.
Watershed area for nearby stream P 3 D 1.00E+06 RESRAD Default NR NR NR NR or pond (m2)
Accuracy for water/soil
- 3 D 1.00E-03 RESRAD Default NR NR NR NR computations 6-96
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Saturated Zone Hydrological Data Density of saturated zone (g/cm3) P 2 D 1.76 Site-specifice NR NR NR NR Saturated zone total porosity P 1 D 0.31 Site-specifice NR NR NR NR Saturated zone effective porosity P 1 D 0.28 Site-specifice NR NR NR NR Calculated values for sand Saturated zone field capacity P 3 D 0.066 NR NR NR NR soil typef Site-specifice Saturated zone hydraulic P 1 D 34822 NR NR NR NR conductivity (m/y) 313 feet/day = 34822 m/y Saturated zone hydraulic gradient P 2 D 0.0045 Site-specific valuee NR NR NR NR Site specific soil type sand Saturated zone b parameter P 2 S Lognormal-N NUREG/CR-6697 Att. C -.0253 0.216 NA NA 0.97 Table 3.5-1 Assumed zero due to Water table drop rate (m/y) P 3 D 0 hydraulic connectivity with NR NR NR NR Mississippi river.
6-97
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Site-specific distribution Existing industrial water supply wells onsite at depth of 116 and 129 below ground surface (the 129 depth equals 33.1 m below the water table). 33.1 m assumed to be maximum well depth.
Minimum well depth assumed to be represented Well pump intake depth (m below P 2 S Triangular by nominal 20 screen depth 6.08 19.6 33.1 NR 10 water table)
(6.08 m) from top of water table.
Mode is mid-point between 6.08m and 33.1 m which is 19.6 m.
Note that the site-specific distribution is reasonably similar to the NUREG-6697 distribution values of 6, 10, and 30 for the triangular distribution.
Model: Nondispersion (ND) or Mass- P 3 D ND Applicable to moving NR NR NR NR Balance (MB) groundwater 6-98
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median NUREG/CR-6697, Att. C provides no recommended value due to high variability.
Industrial Scenario pump rate depends on industry.
General water usage rate Well pumping rate (m3/y) P 2 S Uniform 328.7 1643.5 NR NR 986.1 for four persons is 328.7 m3/yr (NUREG-6697, Table 3.10-1) which is assumed to be minimum industrial rate.
Maximum industrial rate assumed to supply 20 workers which equals 1643.5 m3/yr.
Unsaturated Zone Hydrological Data Number of unsaturated zone strata P 3 D 1 Site-specific NR NR NR NR Unsat. zone thickness (m) P 1 D Variable Structure specific NR NR NR NR Unsat. zone soil density (g/cm 3) P 2 D 1.76 Site specific e NR NR NR NR Unsat. zone total porosity P 2 D 0.31 Site specific e NR NR NR NR e
Unsat. zone effective porosity P 2 D 0.28 Site specific NR NR NR NR Calculated values for sand Unsat. zone field capacity P 3 D 0.066 NR NR NR NR soil typef Unsat. zone hydraulic conductivity P 2 D 34822 Site-specific valuee NR NR NR NR (m/y)
Site specific soil type sand Unsat. zone soil-specific b 0.97 P 2 S Lognormal-N NUREG/CR-6697 Att. C -.0253 0.216 NA NA parameter Table 3.5-1 Occupancy NUREG/CR-5512, Vol. 3 Table 5.1.1 mean value is 8400/y which equates to 23 Inhalation rate (m3/y) M,B 3 D 1917 m3/d NR NR NR NR Industrial Scenario m3/yr
=23 m3/d ÷ 24h/d
- 2000 h/y) 6-99
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median See See See See NUREG- NUREG-3 NUREG- NUREG-Mass loading for inhalation (g/m ) P,B 2 S Continuous Linear NUREG/CR-6697, Att. C 6697 6697 2.35E-05 6697 Table 6697 Table Table Table 4.6-1 4.6-1 4.6-1 4.6-1 RESRAD Users Manual Exposure duration B 3 D 30 parameter value not used in NR NR NR NR dose calculation NUREG/CR-6697, Att. C Indoor dust filtration factor P,B 2 S Uniform 0.15 0.95 NR NR 0.55 Figure 7.1-1 NUREG/CR-6697, Att. C Shielding factor, external gamma P 2 S Bounded Lognormal-N Table 7.10-1 -1.3 0.59 0.044 1 0.2725 NUREG-6697 Att. C, Table 7.6-1 recommends a median indoor work day as 8.76 hour8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br />s/day. Assuming 5 days a week and 50 weeks per years this equates to 2190 hours0.0253 days <br />0.608 hours <br />0.00362 weeks <br />8.33295e-4 months <br /> per year.
Majority of industrial work is expected to be indoors.
Fraction of time spent indoors B 3 D 0.1875 NR NR NR NR Consistent with Table 2-3 of the Users Manual for RESRAD Version 6g 75%
of work time is indoors and 25% outdoors.
The corresponding RESRAD indoor Fraction parameter =
(2190*.75)/(24*365) = .1875 As explained in the basis for the Indoor Fraction parameter, the indoor time Fraction of time spent outdoors (on fraction was set at 75% and B 3 D 0.0625 NR NR NR NR site) outdoor time fraction at 25%.
(2190*.25)/(24*365) =
0.0625 Circular contaminated zone Shape factor flag, external gamma P 3 D Circular assumed for modeling NR NR NR NR purposes 6-100
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Ingestion, Dietary Fruits, non-leafy vegetables, grain M,B 2 D NA Industrial Scenario NR NR NR NR consumption (kg/y)
Leafy vegetable consumption (kg/y) M,B 3 D NA Industrial Scenario NR NR NR NR Milk consumption (L/y) M,B 2 D NA Industrial Scenario NR NR NR NR M,B Meat and poultry consumption (kg/y) 3 D NA Industrial Scenario NR NR NR NR M,B Fish consumption (kg/y) 3 D NA Industrial Scenario NR NR NR NR M,B Other seafood consumption (kg/y) 3 D NA Industrial Scenario NR NR NR NR M,B NUREG/CR-5512, Vol. 3 Soil ingestion rate (g/y) 2 D 18.3 NR NR NR NR Table 6.87 NUREG/CR-5512, Vol. 3 Table 6.87 Industrial Scenario water supply assumed to be from an onsite well.
M,B Drinking water intake (L/y) 2 D 327 478 L/y from NUREG/CR- NR NR NR NR 5512 corresponds to 1.31 L/d which is considered a conservative value for 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> work day.
1.31 L/d
- 250 work days =
327 L/y Contamination fraction of drinking All water assumed B,P 3 D 1 NR NR NR NR water contaminated Contamination fraction of household B,P 3 1 All water from well water (if used)
Contamination fraction of livestock B,P 3 D NA Industrial Scenario NR NR NR NR water Contamination fraction of irrigation B,P 3 D NA Industrial Scenario NR NR NR NR water Contamination fraction of aquatic B,P 2 D NA Industrial Scenario NR NR NR NR food Contamination fraction of plant food B,P 3 D NA Industrial Scenario NR NR NR NR Contamination fraction of meat B,P 3 D NA Industrial Scenario NR NR NR NR 6-101
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Contamination fraction of milk B,P 3 D NA Industrial Scenario NR NR NR NR Ingestion, Non-Dietary Livestock fodder intake for meat M 3 D NA Industrial Scenario NR NR NR NR (kg/day)
Livestock fodder intake for milk M 3 D NA Industrial Scenario NR NR NR NR (kg/day)
Livestock water intake for meat M 3 D NA Industrial Scenario NR NR NR NR (L/day)
Livestock water intake for milk M 3 D NA Industrial Scenario NR NR NR NR (L/day)
Livestock soil intake (kg/day) M 3 D NA Industrial Scenario NR NR NR NR Mass loading for foliar deposition P 3 D NA Industrial Scenario NR NR NR NR (g/m3)
NUREG/CR-6697, Att. C Depth of soil mixing layer (m) P 2 D Triangular 0 0.15 0.6 NR 0.15 Figure 3.12-1 Depth of roots (m) P 1 D NA Industrial Scenario NR NR NR NR Drinking water fraction from ground B,P 3 D 1 Industrial Scenario NR NR NR NR water Household water fraction from B,P 3 1 Industrial Scenario NR NR NR NR ground water (if used)
Livestock water fraction from ground B,P 3 D NA Industrial Scenario NR NR NR NR water Irrigation fraction from ground water B,P 3 D NA Industrial Scenario NR NR NR NR Wet weight crop yield for Non-Leafy P 2 D NA Industrial Scenario NR NR NR NR (kg/m2)
Wet weight crop yield for Leafy P 3 D NA Industrial Scenario NR NR NR NR (kg/m2)
Wet weight crop yield for Fodder P 3 D NA Industrial Scenario NR NR NR NR (kg/m2)
Growing Season for Non-Leafy (y) P 3 D NA Industrial Scenario NR NR NR NR Growing Season for Leafy (y) P 3 D NA Industrial Scenario NR NR NR NR Growing Season for Fodder (y) P 3 D NA Industrial Scenario NR NR NR NR Translocation Factor for Non-Leafy P 3 D NA Industrial Scenario NR NR NR NR Translocation Factor for Leafy P 3 D NA Industrial Scenario NR NR NR NR Translocation Factor for Fodder P 3 D NA Industrial Scenario NR NR NR NR Weathering Removal Constant for P 2 D NA Industrial Scenario NR NR NR NR Vegetation (1/y) 6-102
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Wet Foliar Interception Fraction for P 3 D NA Industrial Scenario NR NR NR NR Non-Leafy Wet Foliar Interception Fraction for P 2 D NA Industrial Scenario NR NR NR NR Leafy Wet Foliar Interception Fraction for P 3 D NA Industrial Scenario NR NR NR NR Fodder Dry Foliar Interception Fraction for P 3 D NA Industrial Scenario NR NR NR NR Non-Leafy Dry Foliar Interception Fraction for P 3 D NA Industrial Scenario NR NR NR NR Leafy Dry Foliar Interception Fraction for P 3 D NA Industrial Scenario NR NR NR NR Fodder Storage times of contaminated foodstuffs (days):
Fruits, non-leafy vegetables, and B 3 D NA Industrial Scenario NR NR NR NR grain Leafy vegetables B 3 D NA Industrial Scenario NR NR NR NR Milk B 3 D NA Industrial Scenario NR NR NR NR Meat and poultry B 3 D NA Industrial Scenario NR NR NR NR Fish B 3 D NA Industrial Scenario NR NR NR NR Crustacea and mollusks B 3 D NA Industrial Scenario NR NR NR NR Well water B 3 D NA Industrial Scenario NR NR NR NR Surface water B 3 D NA Industrial Scenario NR NR NR NR Livestock fodder B 3 D NA Industrial Scenario NR NR NR NR Special Radionuclides (C-14)
C-12 concentration in water (g/cm3) P 3 D NA Industrial Scenario NR NR NR NR C-12 concentration in contaminated P 3 D NA Industrial Scenario NR NR NR NR soil (g/g)
Fraction of vegetation carbon from P 3 D NA Industrial Scenario NR NR NR NR soil Fraction of vegetation carbon from P 3 D NA Industrial Scenario NR NR NR NR air C-14 evasion layer thickness in soil P 2 D NA Industrial Scenario NR NR NR NR (m)
C-14 evasion flux rate from soil P 3 D NA Industrial Scenario NR NR NR NR (1/sec)
C-12 evasion flux rate from soil P 3 D NA Industrial Scenario NR NR NR NR (1/sec) 6-103
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Fraction of grain in beef cattle feed B 3 D NA Industrial Scenario NR NR NR NR Fraction of grain in milk cow feed B 3 D NA Industrial Scenario NR NR NR NR Dose Conversion Factors (Inhalation mrem/pCi)
Ac-227 M 3 D 6.70E+00 FGR11 NR NR NR NR Am-241 M 3 D 4.44E-01 FGR11 NR NR NR NR Am-243 M 3 D 4.40E-01 FGR11 NR NR NR NR C-14 M 3 D 2.09E-06 FGR11 NR NR NR NR Cm-243 M 3 D 3.07E-01 FGR11 NR NR NR NR Cm-244 M 3 D 2.48E-01 FGR11 NR NR NR NR Cm-245 M 3 D 4.55E-01 FGR11 NR NR NR NR Cm-246 M 3 D 4.51E-01 FGR11 NR NR NR NR Co-60 M 3 D 2.19E-04 FGR11 NR NR NR NR Cs-134 M 3 D 4.62E-05 FGR11 NR NR NR NR Cs-137 M 3 D 3.19E-05 FGR11 NR NR NR NR Eu-152 M 3 D 2.21E-04 FGR11 NR NR NR NR Eu-154 M 3 D 2.86E-04 FGR11 NR NR NR NR Eu-155 M 3 D 4.14E-05 FGR11 NR NR NR NR Gd-152 M 3 D 2.43E-01 FGR11 NR NR NR NR H-3 M 3 D 6.40E-08 FGR11 NR NR NR NR I-129 M 3 D 1.74E-04 FGR11 NR NR NR NR Nb-94 M 3 D 4.14E-04 FGR11 NR NR NR NR Nd-144e M 3 D 7.04E-02 ICRP60 NR NR NR NR Ni-59 M 3 D 2.70E-06 FGR11 NR NR NR NR Ni-63 M 3 D 6.29E-06 FGR11 NR NR NR NR Np-237 M 3 D 5.40E-01 FGR11 NR NR NR NR Pa-231 M 3 D 1.28E+00 FGR11 NR NR NR NR Pb-210 M 3 D 1.36E-02 FGR11 NR NR NR NR Po-210 M 3 D 9.40E-03 FGR11 NR NR NR NR Pu-238 M 3 D 3.92E-01 FGR11 NR NR NR NR 6-104
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Pu-239 M 3 D 4.29E-01 FGR11 NR NR NR NR Pu-240 M 3 D 4.29E-01 FGR11 NR NR NR NR Pu-241 M 3 D 8.25E-03 FGR11 NR NR NR NR Pu-242 M 3 D 4.11E-01 FGR11 NR NR NR NR Ra-226 M 3 D 8.58E-03 FGR11 NR NR NR NR Ra-228 M 3 D 4.77E-03 FGR11 NR NR NR NR Sm-148e M 3 D 7.34E-02 ICRP60 NR NR NR NR Sr-90 M 3 D 1.30E-03 FGR11 NR NR NR NR Tc-99 M 3 D 8.32E-06 FGR11 NR NR NR NR Th-228 M 3 D 3.42E-01 FGR11 NR NR NR NR Th-229 M 3 D 2.15E+00 FGR11 NR NR NR NR Th-230 M 3 D 3.26E-01 FGR11 NR NR NR NR Th232 M 3 D 1.64e+00 FGR11 NR NR NR NR U-233 M 3 D 1.35E-01 FGR11 NR NR NR NR U-234 M 3 D 1.32E-01 FGR11 NR NR NR NR U-235 M 3 D 1.23E-01 FGR11 NR NR NR NR U-236 M 3 D 1.25E-01 FGR11 NR NR NR NR U-238 M 3 D 1.18E-01 FGR11 NR NR NR NR Dose Conversion Factors (Ingestion mrem/pCi)
Ac-227 M 3 D 1.41E-02 FGR11 NR NR NR NR Am-241 M 3 D 3.64E-03 FGR11 NR NR NR NR Am-243 M 3 D 3.62E-03 FGR11 NR NR NR NR C-14 M 3 D 2.09E-06 FGR11 NR NR NR NR Cm-243 M 3 D 2.51E-03 FGR11 NR NR NR NR Cm-244 M 3 D 2.02E-03 FGR11 NR NR NR NR Cm-245 M 3 D 3.74E-03 FGR11 NR NR NR NR Cm-246 M 3 D 3.70E-03 FGR11 NR NR NR NR Co-60 M 3 D 2.69E-05 FGR11 NR NR NR NR Cs-134 M 3 D 7.33E-05 FGR11 NR NR NR NR Cs-137 M 3 D 5.00E-05 FGR11 NR NR NR NR 6-105
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Eu-152 M 3 D 6.48E-06 FGR11 NR NR NR NR Eu-154 M 3 D 9.55E-06 FGR11 NR NR NR NR Eu-155 M 3 D 1.53E-06 FGR11 NR NR NR NR Gd-152 M 3 D 1.61E-04 FGR11 NR NR NR NR H-3 M 3 D 6.40E-08 FGR11 NR NR NR NR I-129 M 3 D 2.76E-04 FGR11 NR NR NR NR Nb-94 M 3 D 7.14E-06 FGR11 NR NR NR NR Nd-144e M 3 D 1.51E-04 ICRP60 NR NR NR NR Ni-59 M 3 D 2.10E-07 FGR11 NR NR NR NR Ni-63 M 3 D 5.77E-07 FGR11 NR NR NR NR Np-237 M 3 D 4.44E-03 FGR11 NR NR NR NR Pa-231 M 3 D 1.06E-02 FGR11 NR NR NR NR Pb-210 M 3 D 5.37E-03 FGR11 NR NR NR NR Po-210 M 3 D 1.90E-03 FGR11 NR NR NR NR Pu-238 M 3 D 3.20E-03 FGR11 NR NR NR NR Pu-239 M 3 D 3.54E-03 FGR11 NR NR NR NR Pu-240 M 3 D 3.54E-03 FGR11 NR NR NR NR Pu-241 M 3 D 6.84E-05 FGR11 NR NR NR NR Pu-242 M 3 D 3.36E-03 FGR11 NR NR NR NR Ra-226 M 3 D 1.32E-03 FGR11 NR NR NR NR Ra-228 M 3 D 1.44E-03 FGR11 NR NR NR NR e
Sm-148 M 3 D 1.58E-04 ICRP60 NR NR NR NR Sr-90 M 3 D 1.42E-04 FGR11 NR NR NR NR Tc-99 M 3 D 1.46E-06 FGR11 NR NR NR NR Th-228 M 3 D 3.96E-04 FGR11 NR NR NR NR Th-229 M 3 D 3.53E-03 FGR11 NR NR NR NR Th-230 M 3 D 5.48E-04 FGR11 NR NR NR NR Th-232 M 3 D 2.73E-03 FGR11 NR NR NR NR U-233 M 3 D 2.89E-04 FGR11 NR NR NR NR 6-106
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median U-234 M 3 D 2.83E-04 FGR11 NR NR NR NR U-235 M 3 D 2.66E-04 FGR11 NR NR NR NR U-236 M 3 D 2.69E-04 FGR11 NR NR NR NR U-238 M 3 D 2.55E-04 FGR11 NR NR NR NR Plant Transfer Factors (pCi/g plant)/(pCi/g soil)
Ac-227 P 1 D NA Industrial Scenario NR NR NR NR Am-241 P 1 D NA Industrial Scenario NR NR NR NR Am-243 P 1 D NA Industrial Scenario NR NR NR NR C-14 P 1 D NA Industrial Scenario NR NR NR NR Cm-243 P 1 D NA Industrial Scenario NR NR NR NR Cm-244 P 1 D NA Industrial Scenario NR NR NR NR Co-60 P 1 D NA Industrial Scenario NR NR NR NR Cs-134 P 1 D NA Industrial Scenario NR NR NR NR Cs-137 P 1 D NA Industrial Scenario NR NR NR NR Eu-152 P 1 D NA Industrial Scenario NR NR NR NR Eu-154 P 1 D NA Industrial Scenario NR NR NR NR Fe-55 P 1 D NA Industrial Scenario NR NR NR NR Gd-152 P 1 D NA Industrial Scenario NR NR NR NR H-3 P 1 D NA Industrial Scenario NR NR NR NR Nb-94 P 1 D NA Industrial Scenario NR NR NR NR Nd-144 P 1 D NA Industrial Scenario NR NR NR NR Ni-59 P 1 D NA Industrial Scenario NR NR NR NR Ni-63 P 1 D NA Industrial Scenario NR NR NR NR Np-237 P 1 D NA Industrial Scenario NR NR NR NR Pa-231 P 1 D NA Industrial Scenario NR NR NR NR Pb-210 P 1 D NA Industrial Scenario NR NR NR NR Pm-147 P 1 D NA Industrial Scenario NR NR NR NR Po-210 P 1 D NA Industrial Scenario NR NR NR NR Pu-238 P 1 D NA Industrial Scenario NR NR NR NR Pu-239 P 1 D NA Industrial Scenario NR NR NR NR 6-107
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Pu-240 P 1 D NA Industrial Scenario NR NR NR NR Pu-241 P 1 D NA Industrial Scenario NR NR NR NR Ra-226 P 1 D NA Industrial Scenario NR NR NR NR Ra-228 P 1 D NA Industrial Scenario NR NR NR NR Sb-125 P 1 D NA Industrial Scenario NR NR NR NR Sm-148 P 1 D NA Industrial Scenario NR NR NR NR Sr-90 P 1 D NA Industrial Scenario NR NR NR NR Tc-99 P 1 D NA Industrial Scenario NR NR NR NR Th-228 P 1 D NA Industrial Scenario NR NR NR NR Th-229 P 1 D NA Industrial Scenario NR NR NR NR Th-230 P 1 D NA Industrial Scenario NR NR NR NR Th-232 P 1 D NA Industrial Scenario NR NR NR NR U-233 P 1 D NA Industrial Scenario NR NR NR NR U-234 P 1 D NA Industrial Scenario NR NR NR NR U-235 P 1 D NA Industrial Scenario NR NR NR NR U-236 P 1 D NA Industrial Scenario NR NR NR NR Meat Transfer Factors (pCi/kg)/(pCi/d)
Ac-227 P 2 D NA Industrial Scenario NR NR NR NR Ag-108m P 2 D NA Industrial Scenario NR NR NR NR Am-241 P 2 D NA Industrial Scenario NR NR NR NR Am-243 P 2 D NA Industrial Scenario NR NR NR NR C-14 P 2 D NA Industrial Scenario NR NR NR NR Cm-243 P 2 D NA Industrial Scenario NR NR NR NR Cm-244 P 2 D NA Industrial Scenario NR NR NR NR Co-60 P 2 D NA Industrial Scenario NR NR NR NR Cs-134 P 2 D NA Industrial Scenario NR NR NR NR Cs-137 P 2 D NA Industrial Scenario NR NR NR NR Eu-152 P 2 D NA Industrial Scenario NR NR NR NR Eu-154 P 2 D NA Industrial Scenario NR NR NR NR Fe-55 P 2 D NA Industrial Scenario NR NR NR NR 6-108
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Gd-152 P 2 D NA Industrial Scenario NR NR NR NR H-3 P 2 D NA Industrial Scenario NR NR NR NR Nb-94 P 2 D NA Industrial Scenario NR NR NR NR Nd-144 P 2 D NA Industrial Scenario NR NR NR NR Ni-59 P 2 D NA Industrial Scenario NR NR NR NR Ni-63 P 2 D NA Industrial Scenario NR NR NR NR Np-237 P 2 D NA Industrial Scenario NR NR NR NR Pa-231 P 2 D NA Industrial Scenario NR NR NR NR Pb-210 P 2 D NA Industrial Scenario NR NR NR NR Po-210 P 2 D NA Industrial Scenario NR NR NR NR Pu-238 P 2 D NA Industrial Scenario NR NR NR NR Pu-239 P 2 D NA Industrial Scenario NR NR NR NR Pu-240 P 2 D NA Industrial Scenario NR NR NR NR Pu-241 P 2 D NA Industrial Scenario NR NR NR NR Ra-226 P 2 D NA Industrial Scenario NR NR NR NR Ra-228 P 2 D NA Industrial Scenario NR NR NR NR Sb-125 P 2 D NA Industrial Scenario NR NR NR NR Sm-148 P 1 D NA Industrial Scenario NR NR NR NR Sr-90 P 2 D NA Industrial Scenario NR NR NR NR Tc-99 P 2 D NA Industrial Scenario NR NR NR NR Th-228 P 2 D NA Industrial Scenario NR NR NR NR Th-229 P 2 D NA Industrial Scenario NR NR NR NR Th-230 P 2 D NA Industrial Scenario NR NR NR NR Th-232 P 2 D NA Industrial Scenario NR NR NR NR U-233 P 2 D NA Industrial Scenario NR NR NR NR U-234 P 2 D NA Industrial Scenario NR NR NR NR U-235 P 2 D NA Industrial Scenario NR NR NR NR U-236 P 2 D NA Industrial Scenario NR NR NR NR Milk Transfer Factors (pCi/L)/(pCi/d)
Ac-227 P 2 D NA Industrial Scenario NR NR NR NR Am-241 P 2 D NA Industrial Scenario NR NR NR NR 6-109
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Am-243 P 2 D NA Industrial Scenario NR NR NR NR C-14 P 2 D NA Industrial Scenario NR NR NR NR Cm-243 P 2 D NA Industrial Scenario NR NR NR NR Cm-244 P 2 D NA Industrial Scenario NR NR NR NR Co-60 P 2 D NA Industrial Scenario NR NR NR NR Cs-134 P 2 D NA Industrial Scenario NR NR NR NR Cs-137 P 2 D NA Industrial Scenario NR NR NR NR Eu-152 P 2 D NA Industrial Scenario NR NR NR NR Eu-154 P 2 D NA Industrial Scenario NR NR NR NR Fe-55 P 2 D NA Industrial Scenario NR NR NR NR Gd-152 P 2 D NA Industrial Scenario NR NR NR NR H-3 P 2 D NA Industrial Scenario NR NR NR NR Nb-94 P 2 D NA Industrial Scenario NR NR NR NR Nd-144 P 2 D NA Industrial Scenario NR NR NR NR Ni-59 P 2 D NA Industrial Scenario NR NR NR NR Ni-63 P 2 D NA Industrial Scenario NR NR NR NR Np-237 P 2 D NA Industrial Scenario NR NR NR NR Pa-231 P 2 D NA Industrial Scenario NR NR NR NR Pb-210 P 2 D NA Industrial Scenario NR NR NR NR Po-210 P 2 D NA Industrial Scenario NR NR NR NR Pu-238 P 2 D NA Industrial Scenario NR NR NR NR Pu-239 P 2 D NA Industrial Scenario NR NR NR NR Pu-240 P 2 D NA Industrial Scenario NR NR NR NR Pu-241 P 2 D NA Industrial Scenario NR NR NR NR Ra-226 P 2 D NA Industrial Scenario NR NR NR NR Ra-228 P 2 D NA Industrial Scenario NR NR NR NR Sm-148 P 2 D NA Industrial Scenario NR NR NR NR Sr-90 P 2 D NA Industrial Scenario NR NR NR NR Tc-99 P 2 D NA Industrial Scenario NR NR NR NR Th-228 P 2 D NA Industrial Scenario NR NR NR NR 6-110
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Th-229 P 2 D NA Industrial Scenario NR NR NR NR Th-230 P 2 D NA Industrial Scenario NR NR NR NR Th-232 P 2 D NA Industrial Scenario NR NR NR NR U-233 P 2 D NA Industrial Scenario NR NR NR NR U-234 P 2 D NA Industrial Scenario NR NR NR NR U-235 P 2 D NA Industrial Scenario NR NR NR NR U-236 P 2 D NA Industrial Scenario NR NR NR NR Bioaccumulation Factors for Fish ((pCi/kg)/(pCi/L))
Ac-227 P 2 D NA Industrial Scenario NR NR NR NR Am-241 P 2 D NA Industrial Scenario NR NR NR NR Am-243 P 2 D NA Industrial Scenario NR NR NR NR C-14 P 2 D NA Industrial Scenario NR NR NR NR Cm-243 P 2 D NA Industrial Scenario NR NR NR NR Cm-244 P 2 D NA Industrial Scenario NR NR NR NR Cm-245 P 2 D NA Industrial Scenario NR NR NR NR Cm-246 P 2 D NA Industrial Scenario NR NR NR NR Co-60 P 2 D NA Industrial Scenario NR NR NR NR Cs-137 P 2 D NA Industrial Scenario NR NR NR NR Eu-152 P 2 D NA Industrial Scenario NR NR NR NR Eu-154 P 2 D NA Industrial Scenario NR NR NR NR Gd-152 P 2 D NA Industrial Scenario NR NR NR NR H-3 P 2 D NA Industrial Scenario NR NR NR NR I-129 P 2 D NA Industrial Scenario NR NR NR NR Nb-94 P 2 D NA Industrial Scenario NR NR NR NR Ni-59 P 2 D NA Industrial Scenario NR NR NR NR Ni-63 P 2 D NA Industrial Scenario NR NR NR NR Np-237 P 2 D NA Industrial Scenario NR NR NR NR Pa-231 P 2 D NA Industrial Scenario NR NR NR NR Po-210 P 2 D NA Industrial Scenario NR NR NR NR Pb-210 P 2 D NA Industrial Scenario NR NR NR NR 6-111
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Pu-238 P 2 D NA Industrial Scenario NR NR NR NR Pu-239 P 2 D NA Industrial Scenario NR NR NR NR Pu-240 P 2 D NA Industrial Scenario NR NR NR NR Pu-241 P 2 D NA Industrial Scenario NR NR NR NR Pu-242 P 2 D NA Industrial Scenario NR NR NR NR Ra-226 P 2 D NA Industrial Scenario NR NR NR NR Ra-228 P 2 D NA Industrial Scenario NR NR NR NR Sr-90 P 2 D NA Industrial Scenario NR NR NR NR Th-228 P 2 D NA Industrial Scenario NR NR NR NR Th-229 P 2 D NA Industrial Scenario NR NR NR NR Th-230 P 2 D NA Industrial Scenario NR NR NR NR Th-232 P 2 D NA Industrial Scenario NR NR NR NR U-233 P 2 D NA Industrial Scenario NR NR NR NR U-234 P 2 D NA Industrial Scenario NR NR NR NR U-235 P 2 D NA Industrial Scenario NR NR NR NR U-236 P 2 D NA Industrial Scenario NR NR NR NR U-238 P 2 D NA Industrial Scenario NR NR NR NR Bioaccumulation Factors for Crustacea/ Mollusks ((pCi/kg)/(pCi/L))
Ac-227 P 3 D NA Industrial Scenario NR NR NR NR Am-241 P 3 D NA Industrial Scenario NR NR NR NR Am-243 P 3 D NA Industrial Scenario NR NR NR NR C-14 P 3 D NA Industrial Scenario NR NR NR NR Cm-243 P 3 D NA Industrial Scenario NR NR NR NR Cm-244 P 3 D NA Industrial Scenario NR NR NR NR Cm-245 P 3 D NA Industrial Scenario NR NR NR NR Cm-246 P 3 D NA Industrial Scenario NR NR NR NR Co-60 P 3 D NA Industrial Scenario NR NR NR NR Cs-137 P 3 D NA Industrial Scenario NR NR NR NR Eu-152 P 3 D NA Industrial Scenario NR NR NR NR Eu-154 P 3 D NA Industrial Scenario NR NR NR NR 6-112
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Gd-152 P 3 D NA Industrial Scenario NR NR NR NR H-3 P 3 D NA Industrial Scenario NR NR NR NR I-129 P 3 D NA Industrial Scenario NR NR NR NR Nb-94 P 3 D NA Industrial Scenario NR NR NR NR Ni-59 P 3 D NA Industrial Scenario NR NR NR NR Ni-63 P 3 D NA Industrial Scenario NR NR NR NR Np-237 P 3 D NA Industrial Scenario NR NR NR NR Pa-231 P 3 D NA Industrial Scenario NR NR NR NR Pb-210 P 3 D NA Industrial Scenario NR NR NR NR Po-210 P S D NA Industrial Scenario NR NR NR NR Pu-238 P 3 D NA Industrial Scenario NR NR NR NR Pu-239 P 3 D NA Industrial Scenario NR NR NR NR Pu-240 P 3 D NA Industrial Scenario NR NR NR NR Pu-241 P 3 D NA Industrial Scenario NR NR NR NR Pu-242 P 3 D NA Industrial Scenario NR NR NR NR Ra-226 P 3 D NA Industrial Scenario NR NR NR NR Ra-228 P 3 D NA Industrial Scenario NR NR NR NR Sr-90 P 3 D NA Industrial Scenario NR NR NR NR Th-228 P 3 D NA Industrial Scenario NR NR NR NR Th-229 P 3 D NA Industrial Scenario NR NR NR NR Th-230 P 3 D NA Industrial Scenario NR NR NR NR Th-232 P 3 D NA Industrial Scenario NR NR NR NR U-233 P 3 D NA Industrial Scenario NR NR NR NR U-234 P 3 D NA Industrial Scenario NR NR NR NR U-235 P 3 D NA Industrial Scenario NR NR NR NR U-236 P 3 D NA Industrial Scenario NR NR NR NR U-238 P 3 D NA Industrial Scenario NR NR NR NR Graphics Parameters Number of points 32 RESRAD Default NR NR NR NR Spacing log RESRAD Default NR NR NR NR 6-113
La Crosse Boiling Water Reactor License Termination Plan Revision 2 BFM INSITUgw: RESRAD PARAMETERS FOR UNCERTAINTY ANALYSIS a b Parameter (unit) Type Priority Treatmentc Value/Distribution Basis Distribution's Statistical Parametersd 1 2 3 4 Mean/
Median Time integration parameters Maximum number of points for dose 17 RESRAD Default NR NR NR NR Notes:
a P = physical, B = behavioral, M = metabolic; (see NUREG/CR-6697, Attachment B, Table 4.)
b 1 = high-priority parameter, 2 = medium-priority parameter, 3 = low-priority parameter (see NUREG/CR-6697, Attachment B, Table 4.1) c D = deterministic, S = stochastic d Distributions Statistical Parameters:
Lognormal-n: 1= mean, 2 = standard deviation Bounded lognormal-n: 1= mean, 2 = standard deviation, 3 = minimum, 4 = maximum Truncated lognormal-n: 1= mean, 2 = standard deviation, 3 = lower quantile, 4 = upper quantile Bounded normal: 1 = mean, 2 = standard deviation, 3 = minimum, 4 = maximum Beta: 1 = minimum, 2 = maximum, 3 = P-value, 4 = Q-value Triangular: 1 = minimum, 2 = mode, 3 = maximum Uniform: 1 = minimum, 2 = maximum e Sm-148 an ND-144 not listed in RESRAD FGR 11 DCF file e
Reference:
Haley and Aldrich, Inc., "Hydrogeological Investigation Report La Crosse Boiling Water Reactor, Dairyland Power Cooperative, Genoa, WI January 2015 f ZionSolutions Technical Support Document 14-003, Conestoga Rovers & Associates (CRA) Report, Zion Hydrogeologic Investigation Report g Argonne National Laboratory, Users Manual for RESRAD Version 6, ANL/EAD 4, July 2001 6-114
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ATTACHMENT 6-3 RESRAD Input Parameters for LACBWR Alternate Scenario Uncertainty Analysis 6-115
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Soil Concentrations Basic radiation dose limit (mrem/y) 3 D 25 10 CFR 20.1402 NR NR NR NR Initial principal radionuclide (pCi/g) P 2 D Varies based on Radionuclide Unit Value NR NR NR NR Mixture Distribution coefficients (contaminated, unsaturated, and saturated zones) (cm3/g)
Ac-227 (daughter of Cm-243 and P 1 D 450 Mean Kd Value for sand 6.72 3.22 NA NA 825 Pu-239) NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault Am-241 (also daughter of Cm-245 P 1 S Lognormal-N NUREG/CR-6697 Att. C 7.28 3.15 NA NA 1445 and Pu-241)
Am-243 P 1 S Lognormal-N NUREG/CR-6697 Att. C 7.28 3.15 NA NA 1445 C-14 P 1 S Lognormal-N NUREG/CR-6697 Att. C 2.4 3.22 NA NA 11 Cm-243 P 1 S Lognormal-N NUREG/CR-6697 Att. C 8.82 1.82 NA NA 6761 Cm-244 P 1 S Lognormal-N NUREG/CR-6697 Att. C 8.82 1.82 NA NA 6761 Co-60 P 1 S Lognormal-N NUREG/CR-6697 Att. C 5.46 2.53 NA NA 235 Cs-137 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.1 2.33 NA NA 446 Eu-152 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.72 3.22 NA NA 825 Eu-154 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.72 3.22 NA NA 825 Eu-155 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.72 3.22 NA NA 825 Fe-55 P 1 S Lognormal-N NUREG/CR-6697 Att. C 5.34 2.67 NA NA 209 Gd-152 (daughter for Eu-152) P 1 D 825 Median Value 6.72 3.22 NA NA 825 NUREG/CR-6697, Att. C (No sand value listed in Table 3.9-2)
H-3 P 1 S Lognormal-N NUREG/CR-6697 Att. C -2.81 0.5 NA NA 0.06 Nb-94 P 1 S Lognormal-N NUREG/CR-6697 Att. C 5.94 3.22 NA NA 380 Nd-144 (daughter for Eu-152) P 1 D 158 RESRADv.7.0 Default NA NA NA NA NA Nd not listed in NUREG/CR-6697 Ni-59 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.05 1.46 NA NA 424 6-116
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Ni-63 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.05 1.46 NA NA 424 Np-237 (also daughter for Am-241, P 1 S Lognormal-N NUREG/CR-6697 Att. C 2.84 2.25 NA NA 17 Cm-245, and Pu-241)
Pa-231 (daughter for Cm-243 and P 1 D 550 Mean Kd Value for sand 5.94 3.22 NA NA 380 Pu-239) NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault Pb-210 (daughter for Pu-238) P 1 D 270 Mean Kd Value for sand 7.78 2.76 NA NA 181 NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault Po-210 (daughter Pu-238) P 1 D 150 Mean Kd Value for sand 5.20 1.68 NA NA 181 NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault Pu-238 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.86 1.89 NA NA 953 Pu-239 (also daughter for Cm-243) P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.86 1.89 NA NA 953 Pu-240 (also daughter for Cm-244) P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.86 1.89 NA NA 953 Pu-241 P 1 S Lognormal-N NUREG/CR-6697 Att. C 6.86 1.89 NA NA 953 Ra-226 (daughter Pu-238) P 1 D 500 Mean Kd Value for sand 8.17 1.70 NA NA 3533 NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault Ra-228 (daughter Cm-244 and Pu- P 1 D 500 Mean Kd Value for sand 8.17 1.70 NA NA 3533 240) NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault Sm-148 (daughter Eu-152) P 1 D 245 Mean Kd Value for sand 6.72 3.22 NA NA 825 NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault Sr-90 P 1 S Lognormal-N NUREG/CR-6697 Att. C 3.45 2.12 NA NA 32 Tc-99 P 1 S Lognormal-N NUREG/CR-6697 Att. C -0.67 3.16 NA NA 0.51 Th-228 (daughter Cm-244 and Pu- P 1 D 3200 Mean Kd Value for sand 8.68 3.62 NA NA 5884 240) NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault 6-117
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Th-229 (daughter Am-241, Cm-245, P 1 D 3200 Mean Kd Value for sand 8.68 3.62 NA NA 5884 Np-237, and Pu-241) NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault Th-230 (daughter Cm-246 and Pu- P 1 D 3200 Mean Kd Value for sand 8.68 3.62 NA NA 5884 238) NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault Th-232 (daughter Cm-244 and Pu- P 1 D 3200 Mean Kd Value for sand 8.68 3.62 NA NA 5884 240) NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault U-233 (daughter Am-241, Cm-245, P 1 D 35 Mean Kd Value for sand 4.84 3.13 NA NA 126 Np-237, and Pu-241) NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault C U-234 (daughter Pu-238) P 1 D 35 Mean Kd Value for sand 4.84 3.13 NA NA 126 NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault C U-235 (daughter Cm-243 and Pu- P 1 D 35 Mean Kd Value for sand 4.84 3.13 NA NA 126 239) NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault C U-236 (daughter Cm-244 and Pu- P 1 D 35 Mean Kd Value for sand 4.84 3.13 NA NA 126 240) NUREG/CR-6697, Table 3.9-2, Sheppard and Thibault C Initial concentration of radionuclides P 3 D 0 No existing groundwater NR NR NR NR present in groundwater (pCi/l) contamination Calculation Times Time since placement of material (y) P 3 D 0 Start of dose calculation NR NR NR NR immediately after license termination Time for calculations (y) P 3 D 0, 1, 3, 10, 30, 100, 300, 1000 RESRAD Default NR NR NR NR Contaminated Zone Area of contaminated zone (m2) P 2 D 7500 Size of LACBWR NR NR NR NR Licensed Site Exclusion (LSE) area 6-118
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Thickness of contaminated zone (m) P 2 D 0.15 Assumed depth of NR NR NR NR surface soil Length parallel to aquifer flow (m) P 2 D 98 Diameter of 7500 m2 NR NR NR NR contaminated zone Does the initial contamination NA NA NA No Contaminated zone at NA NA NA NA penetrate the water table? surface Contaminated fraction below water P 3 D 0 Contaminated zone at NR NR NR NR table surface Cover and Contaminated Zone Hydrological Data Cover depth (m) P 2 D 0 No cover NR NR NR NR Density of cover material P 2 D NA No cover NR NR NR NR Cover erosion rate P,B 2 S Continuous Logarithmic NUREG/CR-6697 Att. C 5E-08 0.0007 0.005 0.2 0.0015 Table 3.8-1 Density of contaminated zone P 1 D 1.76 Site specifice NR NR NR NR (g/cm3)
Contaminated zone erosion rate P,B 2 S Continuous Logarithmic NUREG/CR-6697 Att. C 5E-08 0.0007 0.005 0.2 0.0015 m/y) Table 3.8-1 Contaminated zone total porosity P 2 D 0.31 Site specifice NR NR NR NR Contaminated zone field capacity P 3 D 0.2 RESRAD default. No NR NR NR NR distribution or median value provided in NURE/CR-6697 Att. C Contaminated zone hydraulic P 2 D 34822 Site specifice NR NR NR NR conductivity (m/y) 313 feet/day = 34822 m/y Contaminated zone b parameter P 2 S Lognormal-N Site specific soil type -.0253 0.216 NA NA 0.97 sand NUREG/CR-6697 Att. C Table 3.5-1 Humidity in air (g/m3) P 3 D 7.2 Median 1.98 0.334 0.001 0.999 7.2 NUREG/CR-6697 Att. C Evapotranspiration coefficient P 2 S Uniform NUREG/CR-6697 Att. C 0.5 0.75 NR NR 0.625 Figure 4.3-1 Average annual wind speed (m/s) P 2 S Bounded Lognormal - N NUREG/CR-6697 Att. C 1.445 0.2419 1.4 13 4.2 Figure 4.5-1 6-119
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Precipitation (m/y) P 2 D 0.78 NUREG/CR-6697 Att. C NR NR NR NR La Crosse, WI Table 4.1-2 Irrigation (m/y) B 3 D 0.20 NUREG-5512, Vol. 3, NR NR NR NR Table 6-18 (Wisconsin Average)
Converted 0.56 L/m2/d to m/y Irrigation mode B 3 D NA Overhead irrigation is NR NR NR NR common practice in U. S.
Runoff coefficient P 2 S Uniform NUREG/CR-6697 Att. C 0.1 0.8 NR NR 0.45 Figure 4.2-1 Watershed area for nearby stream P 3 D 1.00E+06 RESRAD Default NR NR NR NR or pond (m2)
Accuracy for water/soil - 3 D 1.00E-03 RESRAD Default NR NR NR NR computations Saturated Zone Hydrological Data Density of saturated zone (g/cm3) P 2 D 1.76 Site-specifice NR NR NR NR e
Saturated zone total porosity P 1 D 0.31 Site-specific NR NR NR NR e
Saturated zone effective porosity P 1 D 0.28 Site-specific NR NR NR NR Saturated zone field capacity P 3 D 0.066 Calculated values for NR NR NR NR sand soil typef Saturated zone hydraulic P 1 D 34822 Site-specific valuee NR NR NR NR conductivity (m/y) 313 feet/day = 34822 m/y Saturated zone hydraulic gradient P 2 D 0.0045 Site-specifice NR NR NR NR Saturated zone b parameter P 2 S Lognormal-N Site specific soil type -.0253 0.216 NA NA 0.97 sand NUREG/CR-6697 Att. C Table 3.5-1 Water table drop rate (m/y) P 3 D 0 Assumed zero due to NR NR NR NR hydraulic connectivity with Mississippi river.
6-120
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Well pump intake depth (m below P 2 S Triangular Site-specific distribution 6.1 21.2 36.3 NR water table)
Existing industrial water supply wells onsite at depth of 116 and 129 below ground surface (the 129 depth equals 36.3 m below the water table).
36.3 m assumed to be maximum well depth.
Minimum well depth assumed to be represented by a nominal 20 screen depth (6.1 m) starting at the maximum seasonal water table elevation of 629 and extending to 10 below 619 elevation where water table continuously present.
Mode is assumed to be mid-point between 6.1 m and 36.3 m which is 21.2 m.
Note that the site-specific distribution is reasonably similar to the NUREG-6697 distribution values of 6, 10, and 30 for the triangular distribution.
Model: Nondispersion (ND) or Mass- P 3 D ND Applicable to flowing NR NR NR NR Balance (MB) groundwater 6-121
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Well pumping rate (m3/y) P 2 S 530 Calculated according to NR NR NR NR method described in NUREG/CR-6697, Att. C Section 3.10 assuming 7,500 m2 land area (LACBWR site area) and Wisconsin specific irrigation rate. Resident gardener with no livestock.
Unsaturated Zone Hydrological Data Number of unsaturated zone strata P 3 D 1 Site-specifice NR NR NR NR Unsat. zone thickness (m) P 1 D 2.90 m Site Specific NR NR NR NR Assumed ground surface 639 elevation, contaminated Zone thickness 0.15 m, and maximum water table elevation of 629.
Unsat. zone soil density (g/cm 3) P 2 D 1.76 Site-specifice NR NR NR NR Unsat. zone total porosity P 2 D 0.31 Site-specifice NR NR NR NR e
Unsat. zone effective porosity P 2 D 0.28 Site-specific NR NR NR NR Unsat. zone field capacity P 3 D 0.066 Calculated values for NR NR NR NR sand soil typef Unsat. zone hydraulic conductivity P 2 D 34822 Site-specifice NR NR NR NR (m/y)
Unsat. zone soil-specific b P 2 S Lognormal-N Site specific soil type -.0253 0.216 NA NA 0.97 parameter sand NUREG/CR-6697 Att. C Table 3.5-1 Occupancy Inhalation rate (m3/y) M,B 3 D 8400 NUREG/CR-5512, Vol. 3 NR NR NR NR Table 6.29 (23 m3/d x 365 d) 6-122
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Mass loading for inhalation (g/m3) P,B 2 S Continuous Linear NUREG/CR-6697, Att. C See See See See 2.35E-05 NUREG- NUREG- NUREG- NUREG-6697 6697 Table 6697 6697 Table Table 4.6-1 Table 4.6-1 4.6-1 4.6-1 Exposure duration B 3 D 30 RESRAD Users Manual NR NR NR NR parameter value not used in dose calculation Indoor dust filtration factor P,B 2 S Uniform NUREG/CR-6697, Att. C 0.15 0.95 NR NR 0.55 Figure 7.1-1 Shielding factor, external gamma P 2 S Bounded Lognormal-N NUREG/CR-6697, Att. C Table 7.10-1 -1.3 0.59 0.044 1 0.2725 Fraction of time spent indoors B 3 D 0.649 NUREG/CR-5512, Vol. 3 NR NR NR NR Table 6.87 Fraction of time spent outdoors (on B 3 D 0.124 NUREG/CR-5512, Vol. 3 NR NR NR NR site) Table 6.87 (outdoors +
gardening)
Shape factor flag, external gamma P 3 D Circular Circular contaminated NR NR NR NR zone assumed for modeling purposes Ingestion, Dietary Fruits, non-leafy vegetables, grain M,B 2 D 112 NUREG/CR-5512, Vol. 3 NR NR NR NR consumption (kg/y) (other vegetables + fruits
+ grain) Table 6.87 Leafy vegetable consumption (kg/y) M,B 3 D 21.4 NUREG/CR-5512, Vol. 3 NR NR NR NR Table 6.87 Milk consumption (L/y) M,B 2 D NA No Livestock NR NR NR NR Meat and poultry consumption (kg/y) M,B 3 D NA No Livestock NR NR NR NR Fish consumption (kg/y) M,B 3 D NA No Fish consumption NR NR NR NR Other seafood consumption (kg/y) M,B 3 D NA No Seafood consumption NR NR NR NR Soil ingestion rate (g/y) M,B 2 D 18.3 NUREG/CR-5512, Vol. 3 NR NR NR NR Table 6.87 Drinking water intake (L/y) M,B 2 D 478 NUREG/CR-5512, Vol. 3 NR NR NR NR Table 6.87 Contamination fraction of drinking B,P 3 D 1 All water assumed NR NR NR NR water contaminated 6-123
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Contamination fraction of household B,P 3 NA Only applicable to radon water (if used) pathway Contamination fraction of livestock B,P 3 D NA No Livestock NR NR NR NR water Contamination fraction of irrigation B,P 3 D 1 All water assumed NR NR NR NR water contaminated Contamination fraction of aquatic B,P 2 D NA No aquatic food ingestion NR NR NR NR food Contamination fraction of plant food B,P 3 D 1 NUREG/CR-5512, Table NR NR NR NR 6.87 ingestion rates assumes source is residential garden Contamination fraction of meat B,P 3 D NA No Livestock NR NR NR NR Contamination fraction of milk B,P 3 D NA No Livestock NR NR NR NR Ingestion, Non-Dietary Livestock fodder intake for meat M 3 D NA No Livestock NR NR NR NR (kg/day)
Livestock fodder intake for milk M 3 D NA No Livestock NR NR NR NR (kg/day)
Livestock water intake for meat M 3 D NA No Livestock NR NR NR NR (L/day)
Livestock water intake for milk M 3 D NA No Livestock NR NR NR NR (L/day)
Livestock soil intake (kg/day) M 3 D NA No Livestock NR NR NR NR Mass loading for foliar deposition P 3 D 4.00E-04 NUREG/CR-5512, Vol. 3 NR NR NR NR (g/m3) Table 6.87, gardening Depth of soil mixing layer (m) P 2 S Triangular NUREG/CR-6697, Att. C 0 0.15 0.6 NR 0.15 Figure 3.12-1 Depth of roots (m) P 1 S Uniform NUREG/CR-6697, Att. C 0.3 4.0 2.15 Drinking water fraction from ground B,P 3 D 1 All water assumed to be NR NR NR NR water supplied from groundwater Household water fraction from B,P 3 D NA Only applicable to radon NR NR NR NR ground water (if used) pathway Livestock water fraction from ground B,P 3 D NA No Livestock NR NR NR NR water 6-124
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Irrigation fraction from ground water B,P 3 D 1 All water assumed to be NR NR NR NR supplied from groundwater Wet weight crop yield for Non-Leafy P 2 S Truncated Lognormal - N NUREG/CR-6697, Att. C 0.56 0.48 0.001 0.999 1.75 (kg/m2)
Wet weight crop yield for Leafy P 3 S 2.89 NUREG/CR-5512, Vol. 3 NR NR NR NR (kg/m2) Table 6.87 Wet weight crop yield for Fodder P 3 D NA Industrial Scenario NR NR NR NR (kg/m2)
Growing Season for Non-Leafy (y) P 3 D 0.246 NUREG/CR-5512, Vol. 3 NR NR NR NR Table 6.87 Growing Season for Leafy (y) P 3 D 0.123 NUREG/CR-5512, Vol. 3 NR NR NR NR Table 6.87 Growing Season for Fodder (y) P 3 D NA No Livestock NR NR NR NR Translocation Factor for Non-Leafy P 3 D 0.1 NUREG/CR-5512, Vol. 3 NR NR NR NR Table 6.87 Translocation Factor for Leafy P 3 D 1 NUREG/CR-5512, Vol. 3 NR NR NR NR Table 6.87 Translocation Factor for Fodder P 3 D NA No Livestock NR NR NR NR Weathering Removal Constant for P 2 S Triangular NUREG/CR-6697, Att. C 5.1 18 84 33 Vegetation (1/y)
Wet Foliar Interception Fraction for P 3 D 0.35 NUREG/CR-5512, Vol. 3 NR NR NR NR Non-Leafy Table 6.87 Wet Foliar Interception Fraction for P 2 S Triangular NUREG/CR-6697, Att. C 0.06 0.67 0.95 0.58 Leafy Vegetables Wet Foliar Interception Fraction for P 3 D NA No Livestock NR NR NR NR Fodder Dry Foliar Interception Fraction for P 3 D 0.35 NUREG/CR-5512, Vol. 3 NR NR NR NR Non-Leafy Table 6.87 Dry Foliar Interception Fraction for P 3 D 0.35 NUREG/CR-5512, Vol. 3 NR NR NR NR Leafy Table 6.87 Dry Foliar Interception Fraction for P 3 D NA No Livestock NR NR NR NR Fodder Storage times of contaminated foodstuffs (days):
Fruits, non-leafy vegetables, and B 3 D 14 NUREG/CR-5512, Vol. 3 NR NR NR NR grain Table 6.87 Leafy vegetables B 3 D 1 NUREG/CR-5512, Vol. 3 NR NR NR NR Table 6.87 6-125
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Milk B 3 D NA No Livestock NR NR NR NR Meat and poultry B 3 D NA No Livestock NR NR NR NR Fish B 3 D NA No Aquatic Food NR NR NR NR Consumption Crustacea and mollusks B 3 D NA No Aquatic Food NR NR NR NR Consumption Well water B 3 D 1 RESRAD Users Manual NR NR NR NR Table D.6 Surface water B 3 D 1 RESRAD Users Manual NR NR NR NR Table D.6 Livestock fodder B 3 D NA No Livestock NR NR NR NR Special Radionuclides (C-14)
C-12 concentration in water (g/cm 3) P 3 D NA NA NR NR NR NR C-12 concentration in P 3 D NA NA NR NR NR NR contaminated soil (g/g)
Fraction of vegetation carbon from P 3 D NA NA NR NR NR NR soil Fraction of vegetation carbon from P 3 D NA NA NR NR NR NR air C-14 evasion layer thickness in soil P 2 D NA NA NR NR NR NR (m)
C-14 evasion flux rate from soil P 3 D NA NA NR NR NR NR (1/sec)
C-12 evasion flux rate from soil P 3 D NA NA NR NR NR NR (1/sec)
Fraction of grain in beef cattle feed B 3 D NA NA NR NR NR NR Fraction of grain in milk cow feed B 3 D NA NA NR NR NR NR Dose Conversion Factors (Inhalation mrem/pCi)
All Nuclides (except two listed M 3 D Values in RESRAD FGR FGR11 NR NR NR NR below) 11 DCF File Nd-144e M 3 D 7.04E-02 ICRP 107 NR NR NR NR e
Sm-148 M 3 D 7.34E-02 ICRP 107 NR NR NR NR Dose Conversion Factors (Ingestion mrem/pCi)
All Nuclides (except two listed M 3 D Values in RESRAD FGR FGR11 NR NR NR NR below) 11 DCF File 6-126
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Nd-144e M 3 D 1.51E-04 ICRP107 NR NR NR NR Sm-148e M 3 D 1.58E-04 ICRP107 NR NR NR NR Plant Transfer Factors (pCi/g plant)/(pCi/g soil)
Ac-227 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 1.1 NR NR 9.98E-04 Am-241 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 Am-243 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 C-14 P 1 D Lognormal - N NUREG/CR-6697, Att. C -0.36 0.9 NR NR 6.98E-01 Cm-243 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 Cm-244 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 Co-60 P 1 S Lognormal - N NUREG/CR-6697, Att. C -2.53 0.9 NR NR 7.97E-02 Cs-134 P 1 S Lognormal - N NUREG/CR-6697, Att. C -3.22 1.0 NR NR 4.00E-02 Cs-137 P 1 S Lognormal - N NUREG/CR-6697, Att. C -3.22 1.0 NR NR 4.00E-02 Eu-152 P 1 S Lognormal - N NUREG/CR-6697, Att. C -6.21 1.1 NR NR 2.01E-03 Eu-154 P 1 S Lognormal - N NUREG/CR-6697, Att. C -6.21 1.1 NR NR 2.01E-03 Eu-155 Lognormal - N NUREG/CR-6697, Att. C -6.21 1.1 NR NR 2.01E-03 Fe-55 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 Gd-152 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.21 1.1 NR NR 2.01E-03 H-3 P 1 D Lognormal - N NUREG/CR-6697, Att. C 1.57 1.1 NR NR 4.81E+00 Nb-94 P 1 D Lognormal - N NUREG/CR-6697, Att. C -4.61 1.1 NR NR 9.95E-03 Nd-144 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.21 1.0 NR NR 2.01E-03 Ni-59 P 1 D Lognormal - N NUREG/CR-6697, Att. C -3.00 0.9 NR NR 4.98E-02 Ni-63 P 1 D Lognormal - N NUREG/CR-6697, Att. C -3.00 0.9 NR NR 4.98E-02 Np-237 P 1 D Lognormal - N NUREG/CR-6697, Att. C -3.91 0.9 NR NR 2.00E-02 Pa-231 P 1 D Lognormal - N NUREG/CR-6697, Att. C -4.61 1.1 NR NR 9.95E-03 Pb-210 P 1 D Lognormal - N NUREG/CR-6697, Att. C -5.52 0.9 NR NR 4.01E-03 Pm-147 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.21 1.1 NR NR 2.01E-03 Po-210 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.9 0.9 NR NR 1.01E-03 Pu-238 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 Pu-239 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 6-127
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Pu-240 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 Pu-241 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 Ra-226 P 1 D Lognormal - N NUREG/CR-6697, Att. C -3.22 0.9 NR NR 4.00E-02 Ra-228 P 1 D Lognormal - N NUREG/CR-6697, Att. C -3.22 0.9 NR NR 4.00E-02 Sb-125 P 1 D Lognormal - N NUREG/CR-6697, Att. C -4.61 1.0 NR NR 9.95E-03 Sm-148 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.21 1.1 NR NR 2.01E-03 Sr-90 P 1 D Lognormal - N NUREG/CR-6697, Att. C -1.20 1.0 NR NR 3.01E-01 Tc-99 P 1 D Lognormal - N NUREG/CR-6697, Att. C 1.61 0.9 NR NR 5.00E+00 Th-228 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 Th-229 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 Th-230 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 Th-232 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.91 0.9 NR NR 9.98E-04 U-233 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.21 0.9 NR NR 2.01E-03 U-234 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.21 0.9 NR NR 2.01E-03 U-235 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.21 0.9 NR NR 2.01E-03 U-236 P 1 D Lognormal - N NUREG/CR-6697, Att. C -6.21 0.9 NR NR 2.01E-03 Meat Transfer Factors (pCi/kg)/(pCi/d)
Ac-227 P 2 D NA Resident Gardener No NR NR NR NR Livestock Ag-108m P 2 D NA Resident Gardener No NR NR NR NR Livestock Am-241 P 2 D NA Resident Gardener No NR NR NR NR Livestock Am-243 P 2 D NA Resident Gardener No NR NR NR NR Livestock C-14 P 2 D NA Resident Gardener No NR NR NR NR Livestock Cm-243 P 2 D NA Resident Gardener No NR NR NR NR Livestock Cm-244 P 2 D NA Resident Gardener No NR NR NR NR Livestock Co-60 P 2 D NA Resident Gardener No NR NR NR NR Livestock 6-128
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Cs-134 P 2 D NA Resident Gardener No NR NR NR NR Livestock Cs-137 P 2 D NA Resident Gardener No NR NR NR NR Livestock Eu-152 P 2 D NA Resident Gardener No NR NR NR NR Livestock Eu-154 P 2 D NA Resident Gardener No NR NR NR NR Livestock Fe-55 P 2 D NA Resident Gardener No NR NR NR NR Livestock Gd-152 P 2 D NA Resident Gardener No NR NR NR NR Livestock H-3 P 2 D NA Resident Gardener No NR NR NR NR Livestock Nb-94 P 2 D NA Resident Gardener No NR NR NR NR Livestock Nd-144 P 2 D NA Resident Gardener No NR NR NR NR Livestock Ni-59 P 2 D NA Resident Gardener No NR NR NR NR Livestock Ni-63 P 2 D NA Resident Gardener No NR NR NR NR Livestock Np-237 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pa-231 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pb-210 P 2 D NA Resident Gardener No NR NR NR NR Livestock Po-210 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pu-238 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pu-239 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pu-240 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pu-241 P 2 D NA Resident Gardener No NR NR NR NR Livestock Ra-226 P 2 D NA Resident Gardener No NR NR NR NR Livestock 6-129
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Ra-228 P 2 D NA Resident Gardener No NR NR NR NR Livestock Sb-125 P 2 D NA Resident Gardener No NR NR NR NR Livestock Sm-148 P 1 D NA Resident Gardener No NR NR NR NR Livestock Sr-90 P 2 D NA Resident Gardener No NR NR NR NR Livestock Tc-99 P 2 D NA Resident Gardener No NR NR NR NR Livestock Th-228 P 2 D NA Resident Gardener No NR NR NR NR Livestock Th-229 P 2 D NA Resident Gardener No NR NR NR NR Livestock Th-230 P 2 D NA Resident Gardener No NR NR NR NR Livestock Th-232 P 2 D NA Resident Gardener No NR NR NR NR Livestock U-233 P 2 D NA Resident Gardener No NR NR NR NR Livestock U-234 P 2 D NA Resident Gardener No NR NR NR NR Livestock U-235 P 2 D NA Resident Gardener No NR NR NR NR Livestock U-236 P 2 D NA Resident Gardener No NR NR NR NR Livestock Milk Transfer Factors (pCi/L)/(pCi/d)
Ac-227 P 2 D NA Resident Gardener No NR NR NR NR Livestock Am-241 P 2 D NA Resident Gardener No NR NR NR NR Livestock Am-243 P 2 D NA Resident Gardener No NR NR NR NR Livestock C-14 P 2 D NA Resident Gardener No NR NR NR NR Livestock Cm-243 P 2 D NA Resident Gardener No NR NR NR NR Livestock Cm-244 P 2 D NA Resident Gardener No NR NR NR NR Livestock 6-130
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Co-60 P 2 D NA Resident Gardener No NR NR NR NR Livestock Cs-134 P 2 D NA Resident Gardener No NR NR NR NR Livestock Cs-137 P 2 D NA Resident Gardener No NR NR NR NR Livestock Eu-152 P 2 D NA Resident Gardener No NR NR NR NR Livestock Eu-154 P 2 D NA Resident Gardener No NR NR NR NR Livestock Fe-55 P 2 D NA Resident Gardener No NR NR NR NR Livestock Gd-152 P 2 D NA Resident Gardener No NR NR NR NR Livestock H-3 P 2 D NA Resident Gardener No NR NR NR NR Livestock Nb-94 P 2 D NA Resident Gardener No NR NR NR NR Livestock Nd-144 P 2 D NA Resident Gardener No NR NR NR NR Livestock Ni-59 P 2 D NA Resident Gardener No NR NR NR NR Livestock Ni-63 P 2 D NA Resident Gardener No NR NR NR NR Livestock Np-237 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pa-231 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pb-210 P 2 D NA Resident Gardener No NR NR NR NR Livestock Po-210 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pu-238 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pu-239 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pu-240 P 2 D NA Resident Gardener No NR NR NR NR Livestock Pu-241 P 2 D NA Resident Gardener No NR NR NR NR Livestock 6-131
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Ra-226 P 2 D NA Resident Gardener No NR NR NR NR Livestock Ra-228 P 2 D NA Resident Gardener No NR NR NR NR Livestock Sm-148 P 2 D NA Resident Gardener No NR NR NR NR Livestock Sr-90 P 2 D NA Resident Gardener No NR NR NR NR Livestock Tc-99 P 2 D NA Resident Gardener No NR NR NR NR Livestock Th-228 P 2 D NA Resident Gardener No NR NR NR NR Livestock Th-229 P 2 D NA Resident Gardener No NR NR NR NR Livestock Th-230 P 2 D NA Resident Gardener No NR NR NR NR Livestock Th-232 P 2 D NA Resident Gardener No NR NR NR NR Livestock U-233 P 2 D NA Resident Gardener No NR NR NR NR Livestock U-234 P 2 D NA Resident Gardener No NR NR NR NR Livestock U-235 P 2 D NA Resident Gardener No NR NR NR NR Livestock U-236 P 2 D NA Resident Gardener No NR NR NR NR Livestock Bioaccumulation Factors for Fish ((pCi/kg)/(pCi/L))
Ac-227 (daughter of Cm-243 and P 2 D NA No Fish Consumption NR NR NR NR Pu-239)
Am-241 (also daughter of Pu-241) P 2 D NA No Fish Consumption NR NR NR NR Am-243 P 2 D NA No Fish Consumption NR NR NR NR C-14 P 2 D NA No Fish Consumption NR NR NR NR Cm-243 P 2 D NA No Fish Consumption NR NR NR NR Cm-244 P 2 D NA No Fish Consumption NR NR NR NR Co-60 P 2 D NA No Fish Consumption NR NR NR NR Cs-137 P 2 D NA No Fish Consumption NR NR NR NR Eu-152 P 2 D NA No Fish Consumption NR NR NR NR 6-132
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Eu-154 P 2 D NA No Fish Consumption NR NR NR NR Eu-155 P 2 D NA No Fish Consumption NR NR NR NR Fe-155 P 2 D NA No Fish Consumption NR NR NR NR Gd-152 (daughter of Eu-152) P 2 D NA No Fish Consumption NR NR NR NR H-3 P 2 D NA No Fish Consumption NR NR NR NR Nb-94 P 2 D NA No Fish Consumption NR NR NR NR Nd-144 (daughter of Eu-152) P 2 D NA No Fish Consumption NR NR NR NR Ni-59 P 2 D NA No Fish Consumption NR NR NR NR Ni-63 P 2 D NA No Fish Consumption NR NR NR NR Np-237 (also daughter of Am-241, P 2 D NA No Fish Consumption NR NR NR NR Cm-245, and Pu-241)
Pa-231 (daughter of Cm-243 and P 2 D NA No Fish Consumption NR NR NR NR Pu-239)
Pb-210 (daughter of Pu-238) P 2 D NA No Fish Consumption NR NR NR NR Po-210 (daughter of Pu-238) P 2 D NA No Fish Consumption NR NR NR NR Pu-238 P 2 D NA No Fish Consumption NR NR NR NR Pu-239 (also daughter of Cm-243) P 2 D NA No Fish Consumption NR NR NR NR Pu-240 (also daughter of Cm-244) P 2 D NA No Fish Consumption NR NR NR NR Pu-241 P 2 D NA No Fish Consumption NR NR NR NR Ra-226 (daughter of Pu-238) P 2 D NA No Fish Consumption NR NR NR NR Ra-228 (daughter of Cm-244 and P 2 D NA No Fish Consumption NR NR NR NR Pu-240)
Sm-148 (daughter Eu-152) P 2 D NA No Fish Consumption NR NR NR NR Sr-90 P 2 D NA No Fish Consumption NR NR NR NR Tc-99 P 2 D NA No Fish Consumption NR NR NR NR Th-228 (daughter Cm-244 and Pu- P 2 D NA No Fish Consumption NR NR NR NR 240)
Th-229 (daughter Am-241, Cm-245, P 2 D NA No Fish Consumption NR NR NR NR Np-237, and Pu-241)
Th-230 (daughter Cm-246 and Pu- P 2 D NA No Fish Consumption NR NR NR NR 238)
Th-232 (daughter Cm-244 and Pu- P 2 D NA No Fish Consumption NR NR NR NR 240) 6-133
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median U-233 (daughter Am-241, Cm-245, P 2 D NA No Fish Consumption NR NR NR NR Np-237, and Pu-241)
U-234 (daughter Pu-238) P 2 D NA No Fish Consumption NR NR NR NR U-235 (daughter Cm-243 and Pu- P 2 D NA No Fish Consumption NR NR NR NR 239)
U-236 (daughter Cm-244 and Pu- P 2 D NA No Fish Consumption NR NR NR NR 240)
Bioaccumulation Factors for Crustacea/ Mollusks ((pCi/kg)/(pCi/L))
Ac-227 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Am-241 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Am-243 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption C-14 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Cm-243 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Cm-244 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Cm-245 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Cm-246 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Co-60 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Cs-137 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Eu-152 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Eu-154 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Gd-152 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption H-3 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption I-129 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Nb-94 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption 6-134
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median Ni-59 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Ni-63 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Np-237 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Pa-231 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Pb-210 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Po-210 P S D NA No Crustacea/Mollusks NR NR NR NR Consumption Pu-238 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Pu-239 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Pu-240 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Pu-241 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Pu-242 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Ra-226 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Ra-228 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Sr-90 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Th-228 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Th-229 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Th-230 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Th-232 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption U-233 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption U-234 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption 6-135
La Crosse Boiling Water Reactor License Termination Plan Revision 2 ALTERNATE SCENARIO RESIDENT GARDENER RESRAD INPUT PARAMETERS INITIAL SUITE RADIONUCLIDES UNCERTAINTY ANALYSIS Parameter (unit)
Typea Priorityb Treatmentc Value/Distribution Basis Distribution's Statistical Parameters d 1 2 3 4 Mean/
Median U-235 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption U-236 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption U-238 P 3 D NA No Crustacea/Mollusks NR NR NR NR Consumption Graphics Parameters Number of points 32 RESRAD Default NR NR NR NR Spacing log RESRAD Default NR NR NR NR Time integration parameters Maximum number of points for dose 17 RESRAD Default NR NR NR NR Notes:
a P = physical, B = behavioral, M = metabolic; (see NUREG/CR-6697, Attachment B, Table 4.1) b 1 = high-priority parameter, 2 = medium-priority parameter, 3 = low-priority parameter (see NUREG/CR-6697, Attachment B, Table 4.1) c D = deterministic, S = stochastic d NUREG/CR-6697 Distributions Statistical Parameters:
Lognormal-n: 1= mean, 2 = standard deviation Bounded lognormal-n: 1= mean, 2 = standard deviation, 3 = minimum, 4 = maximum Truncated lognormal-n: 1= mean, 2 = standard deviation, 3 = lower quantile, 4 = upper quantile Bounded normal: 1 = mean, 2 = standard deviation, 3 = minimum, 4 = maximum Beta: 1 = minimum, 2 = maximum, 3 = P-value, 4 = Q-value Triangular: 1 = minimum, 2 = mode, 3 = maximum Uniform: 1 = minimum, 2 = maximum e Sm-148 an ND-144 not listed in RESRAD FGR 11 DCF file e Haley and Aldrich, Inc., "Hydrogeological Investigation Report La Crosse Boiling Water Reactor, Dairyland Power Cooperative, Genoa, WI January 2015 f ZionSolutions Technical Support Document 14-003, Conestoga Rovers & Associates (CRA) Report, Zion Hydrogeologic Investigation Report g Argonne National Laboratory, Users Manual for RESRAD Version 6, ANL/EAD 4, July 2001 6-136