View source for LIC-23-0001, Response to Fort Calhoun Station, Unit No. 1 - Review of License Amendment Request to Add License Condition to Include License Termination Plan Requirements - Request for Additional Information

{{Adams
| number = ML23060A197
| issue date = 02/27/2023
| title = Response to Fort Calhoun Station, Unit No. 1 - Review of License Amendment Request to Add License Condition to Include License Termination Plan Requirements - Request for Additional Information
| author name = Barker A
| author affiliation = Omaha Public Power District
| addressee name = 
| addressee affiliation = NRC/NMSS, NRC/NRR, NRC/Document Control Desk
| docket = 05000285, 07200054
| license number = DPR-040
| contact person = 
| case reference number = LIC-23-0001, EPID L-2021-LIT-0000
| document type = Letter type:LIC, Response to Request for Additional Information (RAI)
| page count = 1
| project = EPID:L-2021-LIT-0000
| stage = Request
}}

=Text=
{{#Wiki_filter:10 CFR 50.90 10 CFR 50.82 LIC-23-0001 February 27, 2023 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Fort Calhoun Station (FCS), Unit 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285 Fort Calhoun Station Independent Spent Fuel Storage Installation NRC Docket No. 72-054

==Subject:==
Response to Fort Calhoun Station, Unit No. 1 - FORT CALHOUN STATION, UNIT NO. 1 - REVIEW OF LICENSE AMENDMENT REQUEST TO ADD LICENSE CONDITION TO INCLUDE LICENSE TERMINATION PLAN REQUIREMENTS -
REQUEST FOR ADDITIONAL INFORMATION (EPID L-2021-LIT-0000)

==References:==
: 1.      Letter from OPPD (M. Fisher) to USNRC (Document Control Desk), "License Amendment Request (LAR) 21-01: Revised Fort Calhoun Station License to Add License Condition 3.D to include License Termination Plan Requirements," dated August 3, 2021 (LIC-21-0005) (ML21271A143)
: 2.      Letter from NRC (J. Parrot) to OPPD (T. Via), Fort Calhoun Station, Unit No. 1 -
Review Of License Amendment Request To Add License Condition To Include License Termination Plan Requirements - Request For Additional Information (EPID L-2021-LIT-0000), dated December 20, 2022 (ML22357A066)
By {{letter dated|date=August 3, 2021|text=letter dated August 3, 2021}} (Reference 1) (ML21271A143), Omaha Public Power District (OPPD) submitted a License Amendment Request (LAR) to add a license condition, 3.D, to include License Termination Plan Requirements.
On December 30, 2022 (Reference 2), the NRC provided OPPD with a Request for Additional Information (RAI) regarding the License Termination Plan (LTP) LAR. Attachment 1 of this letter provides the responses to the RAIs.
444 SOUTH 16TH STREET MALL
* OMAHA, NE 681022247

U. S. Nuclear Regulatory Commission LIC-23-0001 Page 2 This letter contains no regulatory commitments.
If you should have any questions regarding this submittal or require additional information, please contact Mrs. Andrea K. Barker, CHP - Regulatory Assurance and Emergency Planning Manager at (531) 226-6051.
Respectfully, Andrea K. Barker, CHP, MHP Regulatory Assurance and Emergency Planning Manager AKB/akb Attachments: 1.) Response to Request for Additional Information c:      S. A. Morris, NRC Regional Administrator, Region IV J. D. Moninger, NRC Deputy Regional Administrator, Region IV J. D. Parrott, NRC Senior Project Manager S. Anderson, NRC Health Physicist, Region IV Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska

LIC-23-0001                                                                      Attachment 1 Page 3 ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION UNIT NO. 1 DOCKET NO. 50-285, 72-054 By application dated August 3, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21271A143), Omaha Public Power District requested changes to the Fort Calhoun Station (FCS) license to add License Condition (LC) 3.D to include License Termination Plan requirements.
To support the U.S. Nuclear Regulatory Commission (NRC) staffs technical review of the LTP license amendment request pursuant to the regulations in Title 10 Code of Federal Regulations (10 CFR) Part 50, please provide the additional information described [as follows]:
TECHNICAL EVALUATION TE-1 Characterization and Application of Fill/Backfill Material Comment: Clarification is needed to describe the characteristics of the proposed clean backfill material and that backfill material will only be applied following the remediation and Final Status Surveys (FSSs) of individual survey units.
Basis: Adequate surveys are needed if the fill materials may contain residual radioactivity, or may be covering residual radioactivity, and must be considered for the survey unit to be compliant with 10 CFR 20.1402 and 10 CFR 20.1501. In multiple places, the LTP states that clean backfill will be used to fill excavations, voids and basements (e.g., Sections 3.3 and 3.5.1.3 of the LTP) to grade level. However, Sections 5.4.1.5 and 6.19 discuss attributing a dose from the proposed backfill materials. Therefore, clarification is needed on what is meant by clean fill materials.
Similarly, the LTP describes in multiple places that backfill and grading will be performed after FSSs.
Request: Define the characteristics of clean backfill material and describe how backfill material will be verified/characterized as clean prior to use. Details related to the selection process for any additional backfill material, beyond what is available from the excavation of the rail spur expansion project, should also be provided. In addition, please verify, and provide a description of the process, by which the final status survey and application of backfill (e.g., when fill/backfill materials are not added until the final status survey is complete) are performed at individual survey units as part of license termination and FSSs.
OPPD Response:
Currently, there are two sources of fill material at FCS: 1) the soil removed from the rail

LIC-23-0001                                                                       Attachment 1 Page 4 spur expansion project and 2) soil from excavations. Soil from the rail spur expansion project is earmarked for use as fill in all site basements. This soil underwent an FSS using sample plan 8200-R, which is available for NRC inspection. Current site practice is that soil from excavations will not be stockpiled and must be used only at the location from which it was excavated. However, two soil stockpiles are available for reuse. The two areas originated from the New Warehouse and the Service Building, both Class 3 structures within a Class 1 land area. The structures were constructed with two to four feet of structural above-grade fill within the elevated slabs. The resulting material was surveyed, controlled, and used as fill in accordance with the guidance of LTP Chapter 5, Section 5.4.1.5. Due to the construction of the structures, a large amount of material remained that is suitable for fill and is currently under FSS control. It is the projects intention to only utilize this material as fill for areas such as slab-on-grade structures, buried pipes, and buried conduit and not in basement structures following FSS. There are currently no plans to import fill material from an off-site source. If the site decides to import fill from an off-site source, the material will be tested at the source (as described in LTP Chapter 5, Section 5.4.1.5) to verify the material is suitable for fill (i.e., the material does not contain elevated levels of naturally occurring radioactive materials that would interfere with performance of FSS).
The word clean is a general term that has been historically used in NRC-approved LTPs to describe fill material that is not expected to contain residual radioactivity, such as soil from non-impacted areas, offsite borrow pits, or concrete demolition debris that had undergone surveys to verify no residual radioactivity above background. The word clean will be removed from the discussion of fill material in Sections 3.3, 3.5.1.3, 3.5.2.3, 3.5.2.4, 3.5.2.5, 3.5.2.7, 3.5.3.1, 6.11.2, 6.11.3, 6.13, and 6.14.2, and clarifying text will be added as shown below via track changes.
FCS LTP Section 3.3, paragraph 4 will be revised as follows:
The basements of the Turbine Building, Containment Building, and Auxiliary Building, and Intake Structure will remain with all interior walls removed, with the exception of the Turbine Building where the turbine pedestal will remain up to three feet below grade. In the Intake Structure, the interior walls will remain at the request of the Army Corp of Engineers. All other structures will be removed in their entirety with the exception of the six above-grade buildings and the Switchyard listed above. Once D&D activities are completed, as well as subsequent FSS, each basement will be backfilled with clean fill material from the rail spur expansion (designated as a Class 3 survey unit 8200) to grade level (approximately 1,004 feet AMSL).
FCS LTP Section 3.5.1.3, paragraph 4 will be revised as follows:
For all excavations created by the removal of buried piping, components or slab-on-grade structures, an RA will be performed prior to backfill. For all sub-grade basements that will remain, a FSS will be performed and, contingent upon the completion of confirmatory surveys and regulatory approval, the basements will be backfilled. If a major sub-grade building has been removed in its entirety, an

LIC-23-0001                                                                        Attachment 1 Page 5 FSS will be performed on the resultant excavation (using the most-restrictive classification of the building prior to demolition) prior to backfill of the excavation.
All void spaces will be backfilled using clean fill material from the rail spur expansion to grade (1,004 AMSL) and in accordance with the Fort Calhoun Excavation and Backfill Requirements.
FCS LTP Section 3.5.2.3, paragraph 6 will be revised as follows:
Once the remaining stainless-steel structure located below 3 feet below grade (1,001 feet AMSL) has satisfactorily undergone FSS and compliance with the unrestricted release criteria has been demonstrated, contingent upon the completion of confirmatory surveys and regulatory approval, the Containment Building void will be backfilled using clean fill material from the rail spur expansion to 1,001 feet AMSL. An impermeable membrane will then be installed on top of the clean fill. The material will be thick enough to protect the soils underneath it from contamination. Three feet of sacrificial fill will then be placed on top of the membrane to approximately the 1,004 feet AMSL elevation.
FCS LTP Section 3.5.2.4, paragraph 5 will be revised as follows:
After the Auxiliary Building basement has satisfactorily undergone FSS (including embedded piping) and compliance with the unrestricted release criteria has been demonstrated, contingent upon the completion of confirmatory surveys and regulatory approval, the Auxiliary Building void will be backfilled using clean fill material from the rail spur expansion to the 1,004 feet AMSL elevation (grade).
FCS LTP Section 3.5.2.5, paragraph 5 will be revised as follows:
After the above-grade portion of demolition has been completed, the temporary seal and gravel will be removed from the electrical vault in Room 505 and an FSS will be performed. Once the electrical vault has satisfactorily undergone FSS and compliance with the unrestricted release criteria has been demonstrated, contingent upon the completion of confirmatory surveys and regulatory approval, the electrical vault void will be backfilled using clean fill material from the rail spur expansion to grade (1,004 feet AMSL).
FCS LTP Section 3.5.2.7, paragraph 3 will be revised as follows:
After excavation is complete, an RA, consisting of gamma scans and soil sampling, will be performed on the excavated footprint prior to backfilling. The void will be backfilled using clean fill material to grade (1,004 AMSL) in accordance with the Fort Calhoun Excavation and Backfill Requirements.
FCS LTP Section 3.5.3.1, paragraph 5 will be revised as follows:
Once the remaining concrete structure located below 3 feet below grade (1,001 feet AMSL) has satisfactorily undergone FSS and compliance with the unrestricted release criteria has been demonstrated, contingent upon the completion of confirmatory surveys and regulatory approval, the Turbine Building void will be backfilled using clean fill material from the rail spur expansion to grade (1,004 feet AMSL) in accordance with the Fort Calhoun Excavation and Backfill Requirements.
FCS LTP Section 6.11.2, paragraphs 2 and 15 will be revised as follows:

LIC-23-0001                                                                        Attachment 1 Page 6 After remediation and FSS are completed, contingent upon the completion of confirmatory surveys and regulatory approval, the Turbine Building, Containment Building, and Auxiliary Building basements will be backfilled to the original grade of 1004 feet AMSL. The Circulating Water Tunnels and Intake Structure will be backfilled with flowable fill (grout) or fill material from the rail spur expansion. All basement walls/floors will therefore have a minimum clean cover thickness of 0.92 m.
The vadose zone thickness of 1.1 m in the site-wide conceptual model (Section 6.7.) is also considered for the BFM insitu scenario. Assuming a 1.1 m vadose zone thickness leads to the conclusion that when the basements are in the as-left geometry at the time of license termination (insitu geometry) all of the walls/floors, other than the top 0.18 m of the walls, are within the saturated zone.
As conservative and simplifying assumption, the 0.18 m portion of wall is also assumed to be in the saturated zone. This leads to the vadose zone being a 0.92 m clean cover. There is no unsaturated zone present in the insitu scenario conceptual model. The excavation and drilling spoils scenarios are not affected by vadose zone thickness.
FCS LTP Section 6.11.3, paragraph 16 will be revised as follows:
In all cases the groundwater concentrations using the actual source term geometry and the more sophisticated MT3DMS transport model are less than the groundwater concentrations calculated by RESRAD using a simplified conceptual model. The maximum ratio for both Cs-137 and Sr-90 is 0.66 in the downstream corner under flow conditions when the well depth is set to 9 m. Placing the well in the center of the basement reduces the ratio to 0.11 and 0.12, respectively. The reduced concentrations with MT3DMS are expected because the well is drawing from a 360 radius as opposed to 100% from the fill immediately adjacent to the well as is assumed in RESRAD. The mixing distance from angles other than zero degrees is effectively longer than 1 m resulting in additional mixing and reduction in groundwater concentration. In addition, the well also draws from areas of clean fill that have yet to be impacted by surface contamination, whichthat further dilutes the concentrations in the well.
FCS LTP Section 6.13, paragraphs 6 and 12 will be revised as follows:
The embedded pipe DSRs are calculated using the deterministic parameters developed for the BFM insitu scenario in Section 6.11.5 with the changes listed in Table 6-20. The uncertainty analysis conducted for the BFM insitu scenario is assumed to apply to embedded pipe given that both involve a fully submerged source term in a basement under a clean cover. The total depth of the fill is assumed to be the same as for the BFM insitu scenario, i.e., 4 m. The cover depth for the embedded pipe scenario is 3.92 m which is the depth of the clean cover over the backfilled basements (0.92 m) plus the 3 m of clean fill assumed to be above the 1 m floor mixing zone.
The conceptual model for the insitu scenario assumes that the residual radioactivity on the internal surfaces of the pipe is instantly released and mixed into a 0.0254 m layer of soil over a contiguous area equal to the total internal surface area of the buried pipe (2,181 m2). No credit is taken for the presence of the pipe to reduce environmental transport. A thin 0.0254 m mixing layer is

LIC-23-0001                                                                     Attachment 1 Page 7 justified because there are no mechanical mixing mechanisms when the pipes are left buried and undisturbed. The pipes are assumed to be located in the soil immediately below the 1.1 m thick vadose zone (see Section 6.14.2) and 100%
submerged in groundwater. The pipe thicknesses are ignored. The source term is therefore a 0.0254 m layer of soil in the saturated zone with a 1.1 m clean cover.
FCS LTP Section 6.14.2, paragraph 4 will be revised as follows:
The conceptual model for the insitu scenario assumes that the residual radioactivity on the internal surfaces of the pipe is instantly released and mixed into a 0.0254 m layer of soil over a contiguous area equal to the total internal surface area of the buried pipe (2,181 m2). No credit is taken for the presence of the pipe to reduce environmental transport. A thin 0.0254 m mixing layer is justified because there are no mechanical mixing mechanisms when the pipes are left buried and undisturbed. The pipes are assumed to be located in the soil immediately below the 1.1 m thick vadose zone (see Section 6.14.2) and 100%
submerged in groundwater. The pipe thicknesses are ignored. The source term is therefore a 0.0254 m layer of soil in the saturated zone with a 1.1 m clean cover.
TE-2 Operational Contamination Control Program and Remedial Action Support Surveys Comment: Additional information is needed regarding how Remedial Action Support Surveys (RASSs) support the operational contamination control program.
Basis: To ensure compliance with 10 CFR 20, Subpart F, Surveys and Monitoring, additional information is needed on how RASSs support the operational contamination control program. Sections 4.2 and 4.3 of the LTP describe the remediation actions and their expected impact on the sites Radiation Protection Program (which includes the contamination control program).
However, the information provided is not sufficient for the NRC to conclude that RASSs will be used to support the contamination control program and are adequate for mitigating any unexpected releases of material to the environment. For example, the LTP does not specify any criteria that would trigger the contamination control program if contamination were unintentionally created or spread by decommissioning or waste processing activities. Of particular interest to the NRC is how the contamination control program addresses the potential creation and spread of discrete radioactive particles (DRPs) or discrete radioactive objects.
Additionally, the proposed methodologies for RASSs are not described in sufficient detail to ensure that DRPs would be identified, remediated, and documented if their presence is suspected.
Request: Add additional information to the LTP with regards to how the contamination control program is supported by RASSs. Describe the survey frequencies (e.g., after each work shift in areas where operations are conducted) and

LIC-23-0001                                                                   Attachment 1 Page 8 methods that will be used for operational surveys for waste generation/processing activities. Information on the methods should include appropriate detectors and methods suitable for identifying the potential presence of contamination including DRPs or radioactive objects that may be generated during decommissioning. A summary of the remedial actions that will be taken in the event DRPs or objects are present outside of a controlled operational area should also be included.
OPPD Response:
An RAI clarification call between the NRC and OPPD was held on January 12, 2023.
Based on the discussion, OPPD is revising the language in LTP Section 5.4.1.11, as follows:
FCS LTP Section 5.4.1.11 will be replaced with the following:
OPPD has prepared a site procedure to describe appropriate methods and detectors suitable for identifying DRPs and/or discrete radioactive objects that may be identified or generated during decommissioning. The procedure describes the measures for conducting contamination control surveys and the handling of DRPs if they are discovered. The procedure requirements ensure timely and prudent actions are taken in response DRPs.
The information below is provided to answer the RAI request. OPPD does not intend to revise the LTP with this information.
Action levels established in the procedures are similar to standard radiation protection practices in the industry for contamination. There are Red Zones for areas greater than 500,000 dpm and buffer zones, or Yellow Zones. Surveys are performed prior to and during working activities in designated areas with the highest potential to create contamination that could be spread including the Reactor Cavity, Fuel Transfer Canal, and Spent Fuel Pool area. Surveys are performed either directly or indirectly with a GM instrument or ion chamber. Controls, such as structural and localized barriers, and the use of tacky pads to minimize contamination migration are used in Red and Yellow Zones, per industry standards.
Waste originating from the Containment Building is transferred through the Containment Waste Structure (CWS), a large, temporary structure directly adjacent to, but outside, Containment. Waste originating from the demolition of the Radwaste and Auxiliary Building inside the Deconstruction Area (DA) is contained behind a berm, loaded into waste transfer trucks, and deposited in the Waste Processing Structure (WPS). The WPS is a large temporary structure located near the rail spur and outside the DA where the waste is loaded into railcars. Trucks are surveyed for removable contamination when leaving the Decontamination Area, where they are loaded, and when they depart

LIC-23-0001                                                                     Attachment 1 Page 9 the WPS after unloading there. The WPS and CWS are surveyed daily while in use.
Weekly samples are obtained where larger rock was placed near the Sally Port to remove materials from truck tires prior to exiting the DA and outside the heavy haul path. Boundary surveys of the DA/RCA and surveys within the RCA are performed daily with NaI detectors. The haul route taken by the trucks is surveyed by plastic scintillators after each truck, and by NaI detectors periodically throughout the day when the haul road is in use. Samples are taken from outside the WPS and inside and outside the DA berm weekly. Areas identified as being contaminated during the surveys and samples are remediated and taken to the WPS for loading into a railcar, with post-removal surveys being performed to confirm removal of the elevated area.
TE- 3 Basement Survey Unit Size Limits Comment: Section 5.2.2 of the LTP indicates that basement survey units have no size limit regardless of classification.
Basis: To do adequate surveys of structures with residual radioactivity in compliance with 10 CFR 20.1501, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) guidance (NUREG-1575, Revision 1) indicates that Class 1 structural survey units have a size limit of 100 m2 and Class 2 structural survey units have a size limit of 1,000 m2. Also, lower structure classification survey units include the floor and lower 2 meters of the structural walls which are most likely to have been impacted by facility operations.
Request: Provide justification for indicating that there are no survey unit size limits in basement survey units or revise Section 5.2.2 to be consistent with the structural survey unit size guidance in MARSSIM and NUREG-1757, Vol 2.
OPPD Response:
A revision to Section 5.2.2 is shown below to clarify that survey unit areas for backfilled basement surfaces are developed in accordance with the recommendation of MARSSIM Section 4.6.2, which states:
Survey units should be limited in size based on classification, exposure pathway modeling assumptions, and site-specific conditions.
The survey unit areas for structures in MARSSIM Table 4.1 do not apply to the backfilled basement surfaces. The suggested survey unit areas provided in MARSSIM Table 4.1 for Structures and Land Areas apply to the Building Occupancy Scenario and Resident Farmer Scenario, respectively. NUREG-1757, Vol 2. Rev 1, Section A.3 states that the 100 m2 structural survey unit area represents a room of normal size and is consistent with commonly used dose modeling codes. In the case of the 100 m2 area, the commonly used code is RESRAD-Build which models building occupancy. The exposure pathways for the Building Occupancy Scenario do not apply to the Basement Fill Model (BFM).

LIC-23-0001                                                                    Attachment 1 Page 10 The approach described above is consistent with accepted industry practice that the MARSSIM structure survey unit sizes are based on the building occupancy scenario, which is different from the BFM.
Additional discussion on the differences between the building occupancy and BFM exposure pathways is provided here. The Building Occupancy Scenario exposure pathways are described in NUREG 5512, Vol 1, Residual Radioactive Contamination from Decommissioning, Pacific Northwest Laboratories, 1992, and include the following:
external exposure to penetrating radiation from surface sources inhalation of resuspended surface contamination inadvertent ingestion of surface contamination These pathways are not plausible for a backfilled basement at the time of license termination. The Building Occupancy exposure pathways are, therefore, not included as exposure pathways in Table 6.2 of the LTP, Compliance Scenario Environmental Pathways and Exposure Pathways. The BFM includes three exposure pathways for basement contamination, which are described below and listed as sources 13, 14, and 15 in the LTP Table 6.2.
The surface and volumetric source terms in backfilled basement walls, floors, and ceilings must be transported from the basement to the ground surface before exposure to the average member of the critical group can occur. The source term can be transported from the basement surfaces to the ground surface land areas by three pathways:
: 1. from the basement surface to groundwater to open land area via well water (in situ scenario, Source No. 13 in Table 6.2)
: 2. from basement surface to drilling spoils to open land area (drilling spoils scenario, Source No. 14 in Table 6.2)
: 3. by excavating concrete and depositing the concrete on open land areas (excavation scenario, Source No. 15 in Table 6.2)
The BFM requires contamination to be transported from the basement to an open land area before exposure to the critical group can occur. Therefore, the survey unit areas for backfilled basement surfaces are based on the open land survey unit areas in MARSSIM Table 4.1.
Class 1 or Class 2 basement survey unit sizes can exceed the open land area sizes listed in MARSSIM Table 4.1, but the direct measurement frequency (i.e., samples per area) must be equal to or greater than that derived assuming survey unit sizes of 2,000 m2 and 10,000 m2, respectively. For example, if 18 direct measurements are required in a Class 1 survey unit, the minimum allowable sample frequency would be 2,000 m2/18 or 1 sample per 111 m2. If the basement survey unit were 3,000 m2, the number of

LIC-23-0001                                                                     Attachment 1 Page 11 direct measurements would be 3,000/111 or 27 measurements. The same rationale would apply to a Class 2 survey unit. Regardless of the survey unit size, the scan percentage is as recommended in MARSSIM.
FCS LTP Section 5.2.2, paragraph 5 and Table 5-1 will be revised as follows:
Survey units are limited in size based on classification, exposure pathway modeling assumptions, and site-specific conditions. The surface area limits, used in establishing the initial set of survey units, are provided in Table 5-1 for above-ground structures, and land areas, and basement surfaces. The area limits for above-ground structures refer to the floor area, and not the total surface area (floors plus walls and ceilings). The area for the basement surface applies to the total surface area (floors, plus walls and ceilings). This is consistent with the guidance in Table A.1 of Appendix A to NUREG-1757 and MARSSIM. The surface area limits given in Table 5-1 will also be used should the need arise to establish any new survey units beyond the initial set given in Chapter 2.
Table 5-1 FCS Survey Unit Surface Area Limits Survey Unit Type/Classification                      Suggested Surface Area Limit Class 1:
Above-ground Structures (floor area)  100 m2 Land Areas                                  2,000 m2 Basements (all surfaces combined)           2,000 m2 no limit Class 2:
Above-ground Structures (floor area) 100 to 1,000 m2 Land Areas                                 2,000 m2 to 10,000 m2 Basements (all surfaces combined)          2,000 m2 to 10,000 m2 no limit Class 3:
Above-ground Structures (floor area) no limit Land Areas                                 no limit Basements (all surfaces combined)          no limit Class 1 or Class 2 basement survey unit sizes can exceed the open land area sizes listed in Table 5-1, but the direct measurement frequency (i.e., samples per area) must be equal to or greater than that derived assuming survey unit sizes of 2,000 m2 and 10,000 m2, respectively. For example, if 18 direct measurements are required in a Class 1 survey unit, the sample frequency could not be less

LIC-23-0001                                                                     Attachment 1 Page 12 than 2,000 m2/18 = 1 sample per 111 m2. If the basement survey unit were 3,000 m2 the number of direct measurements would be 3,000/111 = 27. If unit size is less than 1,000 m2 the sample frequency increases to 1000/18 = 1 sample per 55 m2. The same rationale would apply to a Class 2 survey unit. Regardless of the survey unit size, the scan percentage using ISOCS is as recommended in MARSSIM.
TE-4 Isolation and Control Measures Comment: Additional information is needed on the isolation and control measures that have or will be established for survey units once decommissioning activities have been completed and a decision to perform FSSs has been made.
Basis: Additional clarifications are needed about the isolation and control program and the FSS data to ensure it will be adequate to demonstrate compliance with 10 CFR 20, Subparts E and F. Section 5.2.3 of the LTP describes the measures for isolation and control that the licensee will take prior to initiating FSS and will maintain during and after FSS. This section also describes the area surveillances that will be performed after FSS to provide assurance that these areas remain unchanged.
The NRC uses FSS data to reach reasonable assurance that a site has been adequately characterized and meets the unrestricted release criteria. It is important for this data to reflect the final site conditions. For example, if a survey unit were to become re-contaminated after the FSS was performed, the FSS data would no longer be applicable. It is important that the controls that have already been put in place, as well as those that will be in place, ensure that no activities will occur within a survey unit that could potentially cause the radiological conditions to change once a decision to transition to isolation has been made.
Additional clarification is needed on:
a) how isolation and control measures will be managed for open land areas that do not have a positive access control. It is not clear from the LTP whether the isolation and control program for these areas will be primarily through the surveillances or if the other isolation and control measures described in the LTP also apply to these areas. For example, it is not clear whether the measures such as instructing individuals to contact the LT/FSS group before performing work in the area and the tracking of entries into the area will apply to the open land areas.
b) when the licensee intends to stop using isolation and control measures relative to when the license for the FCS site (minus the ISFSI) is terminated. Section 5.2.3.2 indicates that the isolation and control measures will remain in force until there is minimal risk of recontamination from decommissioning or the survey unit has been released from the license. However, the next paragraph

LIC-23-0001                                                                     Attachment 1 Page 13 states that To provide additional assurance that survey units that have successfully undergone FSS remain unchanged until final site release. It is not clear from the LTP whether any movement of materials across survey units (e.g., regrading of the site) would be allowed to occur prior to the release of the site or if the isolation and control measures would prevent this.
c) the method that will be used to determine the rigor needed for the surveillance surveys. It is unclear from the LTP how the surveillances will be documented and whether this information will be included in the FSS reports.
Request: Please provide additional information on the controls that are or will be in place to ensure that no activities that could potentially contaminate the survey unit will occur once a decision to transition to isolation has been made. If investigation surveys (e.g., Radiological Assessments (RAs)) may be used for justification that no reperformance of an FSS is necessary, clarify how documentation of such surveys will supplement previously performed FSSs.
OPPD Response:
a) Physical isolation and control (I&C) measures such as ropes, gates, jersey barriers, or doors will be used to ensure that no activities that could potentially contaminate a survey unit once a decision has been made to transition that survey unit to isolation. These physical I&C measures are implemented for all open land or structure survey units when they are ready for FSS, with the exception of survey units where it is not practical to construct a barrier (e.g., large Class 3 open land survey units with no close proximity to major D&D activities or shared land areas such as farmland or state park land).
In cases of survey units where a physical barrier is impractical and there is a low risk for cross-contamination, alternate I&C measures will be implemented. These include the use of signs that require contact with the LT/FSS group if entry is requested. Personnel wishing to enter an LT/FSS I&C area should contact the LT/FSS group, sign in / sign out on an I&C log, and undergo the process of frisking of personnel and equipment in and out of the area. Exemptions for certain processes can be given to people who pose little to no risk for contaminating a controlled area (e.g., state park workers, farmers). A surveillance is not relied upon as an I&C measure, rather it is an activity performed to verify that I&C measures are being maintained.
b) FCS LTP Section 5.2.3.2, paragraph 2 will be revised as follows to clarify when I&C measures can be lifted for a survey unit:
To provide additional assurance that survey units that have successfully undergone FSS remain unchanged until final site release there is minimal risk of recontamination from decommissioning or the survey unit has been released

LIC-23-0001                                                                           Attachment 1 Page 14 from the license, documented routine surveillances of the completed survey units, or other control measures, will be performed on a semi-annual basis. The surveillances will be performed in areas following FSS completion to monitor for indications of recontamination and verification of postings and access control measures. These routine surveillances will consist of:
Movement of materials across survey units, e.g., regrading of the site, that are under I&C would be allowed to occur prior to the release of the site, given that the proper protocols are followed regarding surveillance. Final site grading will not be performed until after license termination.
c) The method for performing surveillances is detailed in Section 4.4 of FCSD-RA-LT-303, Final Status Survey Isolation and Control Measures,1 which is available for inspection.
Any additional rigor to a surveillance (e.g., additional sampling or scanning surveys) is decided by the LT/FSS Manager on a case-by-case basis.
Surveillances outside of the routine are deemed special surveillances and usually include additional rigor based on the situation (e.g., excavations after FSS, site grading after FSS, or other work performed in survey units under I&C).
Investigation surveys (e.g., radiological assessments or surveillances) may be used for justification that no reperformance of an FSS is necessary. The documentation of such surveys does not directly supplement previously performed FSSs. Surveillances are documented with an attachment to FCSD-RA-LT-302, as described above. This information is not included in survey unit release records, as the surveillances occur after FSS is completed. It would be impractical to revise submitted release records that are under technical review twice a year to include surveillance information. However, the surveillance survey documentation is available to NRC for inspection and can be provided to the NRC upon request. If there is something discovered during a surveillance that challenges the data reported in the survey unit release record, FSS in that survey unit is reperformed, and a revision to the existing release record is prepared.
TE-5 Surrogate and Insignificant Contributor Quality Control (QC) Checks Comment: Clarification is needed regarding which surveys the licensee will perform to assess the insignificant contributors and surrogate ratios.
Basis: This information is needed to assure compliance with 10 CFR 20.1402 (i.e.,
that the total dose to potential future site occupants is less than the dose criteria) and 1 Omaha Public Power District, FCSD-RA-LT-303, Revision 0, Final Status Survey Isolation and Control Measures

LIC-23-0001                                                                       Attachment 1 Page 15 10 CFR 20.1501 (i.e., that the site has been adequately characterized). At minimum, the additional site characterization surveys and QC sampling during FSSs should include determinations that the surrogate ratios and insignificant contributor data in each survey unit are consistent with the data sets utilized to establish values in the LTP (see Table 5-16 and section 6.15 of the LTP).
Clarification is also needed for the actions that will be taken if the data in any survey unit is not consistent with the established values. Section 5.2.6.2 of the LTP describes the development of surrogate radionuclide ratios (Table 5-16) while other sections (e.g.,
Sections 5.2.5 and 6.15) describe the mixture ratios, dose contribution, and selection of the insignificant contributors from the initial suite of site radionuclides. In Section 5.2.5 of the LTP, the licensee commits to analyzing 10 percent of the continuing characterization samples for all media samples collected in a survey unit for the initial suite of radionuclides and provides a method to assure that the appropriate surrogate and insignificant contributors are considered. Section 5.2.6.2 also commits to directly measuring C-14 in all soil samples. However, the LTP provides insufficient details on whether FSS QC samples or other survey types (e.g., RA, RASS), or other media, will also address the surrogate/inferred radionuclides or the insignificant contributors and whether the established values are appropriate for the survey unit.
Because the insignificant contributors and inferred radionuclides are not directly measured, it is necessary to verify that assumptions made based on initial characterization data hold true throughout the site during and after decommissioning and that the dose considerations for these components of potential dose are appropriately addressed.
In addition, the technical evaluation performed in RSCS TSD 21-043, Rev 1, should be updated after all relevant site characterization data and final status survey QC data are available and included with the data set in the TSD and the impact to all FSSs determined.
Request: Update the LTP sections dealing with future site characterization (e.g.,
Sections 1.5.2, 2.2.2,, 2.2.6, 5.2.5), QC during FSSs (e.g., Sections 5.2.6.2, 5.6.3, 5.7.1.), as well as other sections pertaining to other survey types, as appropriate, to explicitly include data evaluations to validate that the insignificant contributors and surrogate ratios are correct and what actions the licensee will take if the data indicate inconsistency with the previously established values.
Clarify for which surveys and for which media verification of the surrogate ratios and insignificant contributors will occur. Please confirm that the evaluation performed in RSCS TSD 21-043 will be updated after all applicable future characterization and FSS QC data is included in the data set and that the impact of the newly updated radionuclide mixture fractions, insignificant contributors, and surrogate ratios will be assessed for all FSSs.

LIC-23-0001                                                                            Attachment 1 Page 16 OPPD Response:
The data evaluations to validate that the insignificant contributors and surrogate ratios are correct and what actions the licensee will take if the data indicate inconsistency with the previously established values are described in LTP Sections 5.2.5 and 5.2.6.8 for each relevant survey type. In summary, 10% of the media samples collected during continuing characterization and RA/RASSs will be analyzed for the full initial suite of radionuclides to verify IC dose and HTD to surrogate radionuclide ratios are unchanged from those reported in TSD 21-043, Radionuclides of Concern in Support of the Fort Calhoun License Termination Plan.2 For FSS, all media samples collected in basements (10% of the ISOCS locations) will be analyzed for HTD radionuclides (Sr-90 and C-14) to verify the HTD to surrogate radionuclide ratios are unchanged from those reported in TSD 21-043.
The results of the IC dose and surrogate ratio evaluations will be included in the release record of the subject survey unit rather than as revisions to TSD 21-043.
The sections noted in the request are governed by LTP sections 5.2.5 and 5.2.6.8; therefore, no additions to the other referenced LTP sections are needed.
TE-6 Data Quality Objectives for Groundwater Monitoring Comment: More information is needed on the Data Quality Objectives (DQOs) for groundwater monitoring during decommissioning.
Basis: The LTP does not provide information on the DQOs for the groundwater monitoring and whether they are sufficient to provide reasonable assurance that the final conditions of the site are consistent with the dose criteria in 10 CFR 20.1402 and that the site has been adequately characterized per 10 CFR 20.1501. Section 2.4 of the LTP describes surface and groundwater characterization that has previously been performed and Section 5.4.1.10 describes the survey methodology that will be used for groundwater. Section 5.4.1.10 states that the assessments of residual radioactivity in groundwater at the site will be via groundwater monitoring wells installed at FCS and that the ongoing monitoring includes the Radiological Environmental Monitoring Program (REMP), Radiological Groundwater Protection Program, and National Pollutant Discharge Elimination System monitoring. In some cases, the information needed to demonstrate compliance with the decommissioning regulations is different than the information needed during operations, and the groundwater program may need to be 2 Radiation Safety and Control Service, TSD 21-043, Revision 1, Radionuclides of Concern in Support of the Fort Calhoun License Termination Plan

LIC-23-0001                                                                    Attachment 1 Page 17 adjusted accordingly. Groundwater monitoring during and following decommissioning activities should include residual radioactivity identified in the groundwater during decommissioning, if any, and investigations of subsurface residual radioactivity identified prior to or during decommissioning that could be associated with groundwater contamination. These investigations generally consider the conceptual site model, and transport time and spatial length scales for relevant radionuclides. The NRC provides guidance on Surface Water and Groundwater Characterization as part of decommissioning in NUREG-1757, Vol 2, Rev 2.
Sections 5.2.6.6 and 6.18 of the LTP describe the methodology that the licensee will use to determine the dose from groundwater contamination, if any, that exists at the time of license termination. However, neither of these sections describes how the radionuclides of concern (ROCs) for groundwater will be determined and whether they are expected to differ from the ROCs identified for other media. Additionally, it is not clear if the concentrations of all the ROCs will be evaluated as part of the groundwater monitoring program. As part of developing DQOs for groundwater monitoring, the licensee should describe how they will determine the ROCs for the groundwater pathway. The DQOs should be sufficient to determine the concentration of the ROCs in the groundwater and their associated dose.
Additionally, the DQOs for the groundwater monitoring are expected to be sufficient to allow comparison to the U.S. Environmental Protection Agency (EPA) Maximum Contaminant Levels, and monitoring should be sufficient to statistically demonstrate that residual radioactivity concentrations in water are stable or decreasing (e.g., use of the Mann-Kendall test). For reference, see the Memorandum of Understanding (MOU) between EPA and NRC for Consultation and Finality on Decommissioning and Decontamination of Contaminated Sites (Agencywide Documents Access and Management System [ADAMS] Accession No. ML073090532) and the paper written on implementing the MOU (ADAMS Accession No. ML051380168).
Request: Update the LTP to include DQOs for groundwater monitoring similar to the DQOs established for FSS sampling. The DQOs should address the initial suite of radionuclides established in the LTP (Table 5-2) and should provide sufficient information to provide reasonable assurance that the groundwater conditions remaining at the site at the time of license termination are consistent with the NRC regulations.
OPPD Response:
LTP Section 5.4.1.10 will be revised to include DQOs for groundwater monitoring that address the initial suite of ROCs. The LTP revision provides sufficient information to provide reasonable assurance that the groundwater conditions remaining at the site at the time of license termination are consistent with the NRC regulations. The DQOs for

LIC-23-0001                                                                        Attachment 1 Page 18 groundwater monitoring will be in accordance with FCSD-CH-104, Ground Sampling and Analysis Process.3 The DQOs for groundwater monitoring that will be added to LTP Section 5.4.1.10 are:
The method MDC shall be in accordance with CH-ODCM-0001, Off-Site Dose Calculation Manual.
If a parameter currently being analyzed is not listed in the ODCM, the program required MDC will be determined from appropriate reference documents (i.e.,
EPA). The rationale for its establishment will be listed in Appendix B with its value.
The contract laboratory shall achieve the required method uncertainty for each radionuclide as listed in Attachment 2 and as established per Reference 6.7.
Uncertainty (Umr) for 95% confidence level should be set at Action Level-MDC/3.
The contract laboratory will be maintained as an approved supplier on the OPPD or EnergySolutions Approved Supplier List (ASL). The procurement of materials, equipment, and services for characterization and FSS will be performed in a controlled manner, which will ensure compliance with applicable regulatory requirements, procedures, quality assurance standards and regulations. Service requests will be reviewed for technical adequacy and, in order to assure confidence with services provided. Instrument calibration and laboratory analysis services for characterization and FSS, will be procured as QL-2 level services from vendors on the ASL. Additionally, regular vendor performance reviews, audits and/or surveillances of these contractors may be performed to provide an adequate level of assurance that the quality activities are being effectively performed.
Each batch of samples shall, as an average, achieve the required method uncertainty for each analyte (i.e., a batch of samples analyzed for tritium would have an average method uncertainty  566 pCi/L).
All sentinel and surface water samples will be analyzed for gamma and tritium at a minimum.
All results greater than a 95% confidence detection threshold (2.33  counts) are to be evaluated to determine if they are part of the overall background distribution or positive activity.
Acceptance criteria on laboratory control samples per radionuclide are identified in the vendors applicable procedures, typically +/- 30%.
Each full quarterly batch of samples will contain a duplicate, matrix spike, and laboratory control sample. Every year at least one spike will be performed on a contaminated well, and one non-contaminated, if possible.
3 Omaha Public Power District, FCSD-CH-104, Revision 3, Ground Sampling and Analysis Process

LIC-23-0001                                                                    Attachment 1 Page 19 Matrix spikes will be performed for tritium and a hard-to-detect radionuclide by gamma spectroscopy.
Minimal yields for tracers or carriers are as follows:
55           63           90 Matrix              FE           Ni           Sr Groundwater          90%          60%         60%
TE-7 Surveys of Excavations Comment: Surveys (including FSSs and RASSs) of excavations should include the sidewalls as well as the footprint of the excavation (see NUREG-1757, Vol 2, Section G.3.2.1).
Basis: The FSS should incorporate all exposed surfaces in an excavation to be compliant with 10 CFR 20.1501. In Section 5.4.1.4 of the LTP (as well as other sections), the licensee states that an FSS) designed as an open land survey will be performed of excavations. Clarify that this means both the excavation footprint as well as any sidewalls of the excavation will be subject to the FSS and RASS, as appropriate.
Request: Clarify in Section 5.4.1.4 (and others as appropriate) that exposed sidewalls of excavations will be incorporated into any RASS or FSS.
OPPD Response:
To clarify that the exposed sidewalls should be included in any FSS or RA/RASS, FCS LTP Section 5.4.1.4 will be revised as follows:
Any soil excavation created to expose or remove a potentially contaminated basement structure will be subjected to FSS prior to backfill. The FSS will be designed as an open land survey, ensuring the inclusion of the footprint and any exposed sidewalls, using the classification of the removed structure in accordance with Section 5.3 of the LTP using the Operational DCGLs for surface soils or subsurface soils (depending on the thickness of contamination) as the release criterion.
During decommissioning of FCS, any surface or subsurface soil contamination that is identified by continuing characterization or operational radiological surveys that is in excess of the Base Case DCGLs for each of the potential ROC as presented in Table 5 7 will be remediated. The remediation process will include performing a RASS of the open excavations, including sidewalls, in accordance with procedures. The RASS will include scan surveys and the collection of soil samples during excavation to gauge the effectiveness of remediation, and to identify locations requiring additional excavation. The scan surveys and the collection of and subsequent laboratory analysis of soil samples will be performed in a manner that is intended to meet the DQOs of FSS. The data obtained during the RASS is expected to provide a high degree of confidence

LIC-23-0001                                                                        Attachment 1 Page 20 that the excavation, or relevant portion of the excavation, meets the criterion for the unrestricted release of open land survey units. Soil samples will be collected to depths at which there is high confidence that deeper samples will not result in higher concentrations. Alternatively, an NaI detector or intrinsic germanium detector of sufficient sensitivity to detect residual radioactivity at the Operational DCGL can be used to scan the exposed soils in an open excavation to identify the presence or absence of soil contamination, and the extent of such contamination. If the detector identifies the presence of contamination at a significant fraction of the Operational DCGL, additional confirmatory investigation and analyses of soil samples of the suspect areas will be performed.
TE-8 Discrete Radioactive Particle Definition and Surveys Comment: Justify the definition of DRPs of concern and actions to be taken in event of DRP release.
Basis: A dose basis and activity level for DRPs of concern that are anticipated to remain on site at the conclusion of decommissioning are needed for the NRC staff to assess the potential dose impact of any DRPs to future site occupants and compliance with decommissioning regulations in 10 CFR 20.1402. In addition, justification that the survey methodology is sufficient to detect DRPs that have an activity level of concern and the additional information on survey techniques is needed to ensure surveys would be compliant with 10 CFR 20.1501.
Section 5.4.1.11 of the LTP provides a description of survey considerations for areas that are suspected to have DRP contamination. In this section, DRPs are defined as specks of radioactive material identified usually as either activated corrosion product such as cobalt-60, or an irradiated fuel fragment exhibiting greater than 10,000 corrected counts per minute (100,000 dpm). Due to issues such as what instruments are being utilized, how instruments are set up, detector/source geometry, shielding materials that may be present, etc. basing a definition on instrument response is inadequate.
Section 5.4.1.11 of the LTP provides information on the survey methods that the license will use in suspected DRP areas. However, this survey information does not include a justification that the methodology is sufficient to detect DRPs that have an activity level of concern. Also, the description of the survey methodology does not specify if alarms are going to be used to identify DRPs during scanning and, if so, what the instrument set points will be.
Also, the LTP does not appear to address radioactive objects which are larger in volume than the typical DRP but of a lesser volume than would typically be considered for materials and equipment. Information is needed on whether the same approach will be used for these discrete radioactive objects as will be used for DRPs. The LTP does not clarify if the DRPs or radioactive objects will be considered and evaluated

LIC-23-0001                                                                    Attachment 1 Page 21 even if composed of materials not directly related to plant operations and not specifically excluded from 10 CFR 20.1402 (e.g., TENORM).
Finally, Section 5.4.1.11 states that a condition report will be generated for each DRP encountered. However, the LTP does not appear to commit to performing contamination control surveys to identify and disposition DRPs in a timely manner to avoid potential secondary transport and if DRP remediation efforts will be complete prior to any additional soil disturbing activities in the affected area.
Request:
a) Please propose a dose basis and activity level for DRPs of concern if such are anticipated to be part of the residual radiation present at the conclusion of decommissioning.
b) Describe the method for surveying for DRPs that demonstrate that it will be sufficient to identify DRPs of concern.
c) If using alarms to identify DRPs during scanning, justify the proposed alarm set point for the surveys and what instruments the set points will apply to.
d) Describe the procedures, if any, that will apply to surveys for radioactive objects which are larger in volume than the typical DRP but of a lesser volume than would typically be considered for materials and equipment (e.g., activated concrete pebble/rock).
e) Describe how DRPs or radioactive objects will be considered and evaluated including if they are composed of materials not directly related to plant operations and not specifically excluded from 10 CFR 20.1402 (e.g., TENORM).
f) If a release of DRPs does occur, describe the contamination control surveys to be used to identify and disposition DRPs, and how they will be done to avoid potential secondary transport. Also describe how DRP remediation efforts will be done in relation to any soil disturbing activities in the affected area. Describe how the activities will be documented.
OPPD Response:
See response to TE-2.
CLARIFICATION RAIs CL-1 Criteria for Determining when Changes to the LTP Require NRC Approval Comment: The LTP change criteria listed in the proposed license amendment do not include all the criteria listed in NUREG-1700, Revision. 2.
Basis: To evaluate that the LTP demonstrates meeting the requirements of 10 CFR Part 20, Subpart E, Radiological Criteria for License Termination, NRC uses NUREG-1700, Standard Review Plan for Evaluating Nuclear Power Reactor License

LIC-23-0001                                                                   Attachment 1 Page 22 Termination Plans. NUREG-1700 recognizes that the licensee may make changes to the LTP following its approval by the NRC. However, NUREG-1700 identifies that there are areas of the LTP that cannot be changed without prior NRC approval because it would represent a change to the approved license termination methodology. To control this change process, NUREG-1700 contains a list of LTP areas that cannot be changed without prior NRC approval and states that the LTP should include a provision that addresses how changes are made to the LTP after approval.
The license amendment request (LAR) proposes to add license condition (LC) 3.D to the FCS license. The LC would approve the LTP and identify areas of the LTP that cannot be changed without NRC approval. The LAR states that the proposed LC is in accordance with NUREG- 1700, Revision 1, Appendix 2, which is a list of NRC identified areas of the LTP that would require prior NRC approval to be changed. The proposed LC essentially captures the list in NUREG-1700, Revision 1, Appendix 2.
However, NUREG-1700 has been updated (Revision 2) and contains a list, in Appendix B, that is more comprehensive than that in Revision 1. The list from Revision 2 contains two additional areas of evaluation that, if changed in the LTP, would need approval by NRC. These areas are: 1) a change to the approach used to demonstrate compliance with dose criteria, and 2) a change to the dose assessment parameter values or pathway dose conversion factors that would lead to a lower calculated dose than what was approved in the LTP.
Request: For proposed LC 3.D, please clarify that the proposed list of LTP areas that cannot be changed without NRC approval does not include the two additional areas described in NUREG- 1700, Revision 2, Appendix B or, revise the proposed LC to reflect those two additional areas. If those two additional evaluation areas are not to be included in proposed condition 3.D, please describe how any changes to those two areas would be controlled in the LTP.
OPPD Response:
OPPD proposes to revise the LTP to include the two additional evaluation areas described in NUREG-1700, Revision 2, Appendix B.
The complete list to be included in FCS LTP Section 1.6, paragraph 2 is as follows:
OPPD is also submitting a proposed amendment to the FCS license that adds a license condition that establishes the criteria for determining when changes to the LTP require prior NRC approval. Changes to the LTP require prior NRC approval when the change:
requires Commission approval pursuant to 10 CFR 50.59, results in significant environmental impacts not previously reviewed,

LIC-23-0001                                                                    Attachment 1 Page 23 detracts or negates the reasonable assurance that adequate funds will be available for decommissioning, decreases a survey unit area classification (i.e., impacted to not impacted, Class 1 to Class 2, Class 2 to Class 3, or Class 1 to Class 3 without providing NRC a minimum 14-day notification prior to implementing the change in classification),
increases the DCGLs and related minimum detectable concentrations (for both scan and fixed measurement methods),
increases in the radioactivity level, relative to the applicable DCGL, at which an investigation occurs, changes the statistical test applied to one other than the Sign test, or increases in the Type I decision error, changes the approach used to demonstrate compliance with the dose criteria (e.g., change from demonstrating compliance using DCGLs to demonstrating compliance using a dose assessment that is based on final concentration data), or changes parameter values or pathway dose conversion used to calculate the dose such that the resultant dose is lower than in the approved LTP and if a dose assessment is being used to demonstrate compliance with the dose criteria.
CL-2 Clarification of Figures Showing Random Sample and Scan Locations Comment: The information shown on the figures need to be clarified.
Basis: The Survey Unit Random Sample and Scan Locations figures referenced in Sections 2.3.2.1.1-2.3.2.1.3, are inconsistent in the information conveyed. Survey Unit 8100 Figure 2-26 has a legend that includes locations for walkover grids, Figure 2-28 (SU 8200) has no legend, and Figure 2-30 (SU 8300) and others show a legend with only sample locations. The dot within a circle indicates a soil sample location per the legend in the lower right corner of Figure 2-30 but does not state if the circle represents the scan area like the legend for Figure 2-26. The title of Figure 2-26 indicates that it includes "Scan Locations." Figure 2-30 is believed to convey the same information but has a legend in the lower left showing the circled dot as a sample location rather than as a random sample location. Consistent survey data presentation in figures is required under 10 CFR 20.1501(a)(2) and is needed by NRC staff to better understand the information in the figures.

LIC-23-0001                                                                    Attachment 1 Page 24 Request: The licensee should provide updated figures with consistent legends and data consistent with the legend and titles for those figures that indicates the meaning of each mark on the figure and that corresponds to the text describing the figure.
OPPD Response:
All figures will be reviewed and updated with consistent legends and data consistent with the legend and titles for those figures that indicates the meaning of each mark on the figure and that corresponds to the text describing the figure as necessary through a revision to the LTP.
CL-3 Scan Alarms Comment: Provide a description for the Scan Alarms indicated in Table 2-83.
Basis: Table 2-83, Turbine Building Scan Results, indicates that at four locations there were Scan Alarms. Clarification is needed to demonstrate compliance with 10 CFR 20.1501(a)(2) on the adequacy of surveys for residual radioactivity.
Request: Please provide a description on what the Scan Alarms are.
OPPD Response:
Alarm set points for scanning were based on the mean of five 1-minute static background measurements plus the MDCRsurveyor. Static measurements were obtained at the highest value identified during the scanning surveys, including areas where the alarm set points have been exceeded. Action Levels for static measurements were one-half of the screening values presented in Tables H.1 and H.2 of NUREG-1757, Volume 2, Revision 1, Consolidated Decommissioning Guidance, Characterization, Survey and Determination of Radiological Criteria for Co-60, the most restrictive of the radionuclides of concern for FCS for confirmation of the classification of the Turbine Building structure (Class 3).
CL-4 Inconsistent Low-Level Waste Volumes Comment: The low-level radioactive waste volumes given in two places in the LTP are different from each other and both are different than the volume given in the Site-Specific Decommissioning Cost Estimate (DCE).
Basis: The requirement in 10 CFR 50.82(a)(9)(ii)(F), for an updated site-specific estimate of remaining decommissioning costs in the LTP, requires an estimate of low-level radioactive waste volumes from decommissioning so that waste disposal costs can be estimated as part of the site-specific DCE.
Page 3-21 of the LTP states the total volume of radioactive waste for disposal has been estimated at 3,222,861 cubic feet. The sum of the volumes of waste in Class A, Class B/C and Greater than Class C in Table 3-4 is 3,334,679 cubic feet. Table 6-5 of

LIC-23-0001                                                                   Attachment 1 Page 25 the DCE, that was provided in the June 15, 2022, LTP RAI response, contains a breakdown of Class A, B/C and Greater than Class C waste. The total volume of those wastes is 2,499,928 cubic feet. Table 6-5 of the DCE, that was provided in the June 15, 2022, LTP RAI response, contains a breakdown of Class A, B/C, and Greater than Class C waste. The total volume of those wastes is 2,499,928 cubic feet.
Request: Clarify the difference between these estimates of the total quantity of radioactive and/or provide a consistent projected volume for each class of radioactive waste.
OPPD Response:
The waste volumes provided in Chapter 3, Section 3.6.2, paragraph 2 and LTP Table 3-4 will be updated to provide consistent projected waste volumes for each class of radioactive waste and to reflect current waste projections.
FCS LTP Section 3.6.2 will be revised as follows:
Table 3 4 provides projections of waste classifications and quantities that will be generated by the decommissioning of FCS. As OPPD has elected to institute an approach commonly referred to as rip & ship verses performing significant on-site decontamination activities, the total volume of radioactive waste for disposal has been estimated at 3,222,8613,334,898 cubic feet. Actual waste volumes and classifications may vary. The vast majority of waste will be loaded into 8-120A or 3-60B casks, gondola rail cars, or articulating bulk cars and shipped to Clive, Utah. Other radioactive waste will go to the licensed WCS facility in Andrews County, Texas also by gondola railcar or truck. GTCC waste will remain on-site at the ISFSI.

LIC-23-0001                                                                Attachment 1 Page 26 FCS LTP Table 3-4 will be replaced with the following:
The waste estimate is based upon current existing site and building drawings and plant system data provided by OPPD combined with EnergySolutions decommissioning experience to establish plant systems and building inventories. These inventories, EnergySolutions Unit Cost Factors, and other plant data were utilized to generate required waste volumes, weights, and classifications. The current waste estimate is an update to Table 6-5 of the DCE, that was provided in the June 15, 2022, LTP RAI response. Waste volume estimates can be expected to change as the project progresses.

LIC-23-0001                                                                    Attachment 1 Page 27 CL-5 Methods Anticipated to be Used to Determine Ambient Background Comment: Clarify the methods anticipated to be used to determine ambient background as discussed in section 5.2.4 of the LTP.
Basis: Section 5.2.4, Reference Areas and Materials, does not specifically state what methods would be used to determine ambient background when performing surveys. This information is necessary to demonstrate compliance with 10 CFR 20.1501(a)(2) on the adequacy of surveys for residual radioactivity.
Request: The licensee should describe the process for obtaining ambient background levels during surveys. A specific consideration for staff would be a discussion on how ambient background might change in a survey unit and how a revised ambient background would be determined.
OPPD Response:
FCS LTP Section 5.2.4 will be revised to include the following information for obtaining ambient background levels:
Backgrounds are typically collected at various locations within the area to be surveyed. Background is typically established as the mean of 5 one-minute static measurements from the various locations. Background measurements with gamma detectors are collected at a minimum height of 6 inches above the ground. Backgrounds with beta instrumentation are taken at waist height with the detector facing away from the materials to be surveyed.
Material specific backgrounds may be required for complex survey units or areas (e.g., floor tiles or fire extinguishers containing NORM). In these instances, the background measurements may be collected by placing the detector on its side on the material surface and recording the mean of five one-minute static measurements.
Ambient background levels can change within survey units. Examples of this occurring are when a survey is spread over several days and radon becomes more prevalent at certain points of the day or when there is a concentrated area of NORM in the survey unit. When there is a significant rise or fall in background, a new ambient background is established using the process described above.
These changes are made at the professional judgment of the LT/FSS Supervisors.
CL-6 Verification of As Low As is Reasonably Achievable (ALARA) Parameters

LIC-23-0001                                                                   Attachment 1 Page 28 Comment: The ALARA evaluation are using parameters inconsistent with the values present in the references/citations.
Basis: In Section 4.4.1.2 of the LTP, the licensee references a transportation fatal accident rate that appears inconsistent with the value in the cited document. The licensee should verify it is using the correct values in the reference document, or more recent document if available, to perform its evaluations.
Request: Revise the ALARA evaluations, as appropriate, to be consistent with the cited values/parameters after verifying the values being cited.
OPPD Response:
The units for the fatal transport rate referenced within Table 4-1 of the LTP, 3.8 x 10-8 per hour, is incorrect. NUREG-1496, Volume 2, Appendix B, Table A.1 shows the correct value and units to be 3.8 x 10-8 per kilometer. The correction will be applied where applicable through revision of the LTP; however, the ALARA calculation will not change, since the numerical value for fatal transport rate is unchanged.
CL-7 In Situ Object Counting System (ISOCS) Use Comment: Clarification is needed on how the licensee will address ISOCS field of view (FOV) uncertainties as discussed in sections 5.3 through 5.5 of the LTP. Verify that the RASS performed prior to ISOCS measurements will ensure a relatively homogenous radiation fluence is present in the ISOCS FOV (i.e., no DRPs or hot spots will be present).
Basis: Section 5.3, Final Status Survey Design, Section 5.4, Final Status Survey Implementation, and Section 5.5, Final Status Survey Data Assessment of the LTP address the use of ISOCS measurements for RASS and FSS. NUREG-1700, Section 5, Final Radiation Survey Plan, calls for a demonstration that the in-situ sample measurements with field instruments, and the associated survey methods, have adequate sensitivity as part of the FSS design. Accordingly, the LTP should address survey areas where inconsistent geometry is present (e.g., corners, junctions of walls, etc.) relative to ISOCS measurements and what determines when an ISOCS measurement location is unsuitable. This is needed to ensure an adequate survey is conducted consistent with the requirements of 10 CFR 20.1501.
Request: Update the appropriate discussions in the LTP regarding use of ISOCS and how prior to the use of ISOCS it is ensured that no DRPs or elevated areas are present in the ISOCS FOV. Describe how if an elevated area is present, the ISOCS may still be used to investigate the area relative to its FOV of the elevated area. Also discuss the possibility of utilizing combinations of in situ and conventional measurements as part of the overall FSS design and how ISOCS measurements will be conservatively performed when uncertain geometries are being measured.
OPPD Response:

LIC-23-0001                                                                   Attachment 1 Page 29 A new subsection to LTP Section 5.4 will be added to discuss post-demolition surveys. The new subsection will contain the following language:
Following demolition of Class 1 structures with basements, after all debris is removed and the floors are cleaned, a RASS will be performed to ensure that any individual ISOCS measurement will not exceed the BcDCGLwf from Table 5-5 during FSS. The survey will be performed using hand-held beta-gamma instrumentation as presented in Table 5-21 in typical scanning and measurement modes.
Areas greater than the BcDCGLwf will be remediated and resurveyed to ensure the elevated area is successfully remediated. Areas exceeding the OpDCGLwf will be bounded and investigated. Investigation may include the use of ISOCS with a reduced FOV. The reduced FOVs and/or overlapping FOVs are able to identify any elevated areas.
The information below is provided to answer the RAI request. OPPD does not intend to revise the LTP with this information.
Combinations of in situ and conventional measurements as part of the overall FSS design will not be utilized. As described above, hand-held instruments, and possibly ISOCS, will be used to perform RASSs post-demolition.
Complex geometries such as corners will be conservatively modeled. Corners will be modeled as a planer circle or similar, with the distance from the detector to the wall conservatively set to the furthest portion of the walls in the FOV. This results in an overestimation of activity since the perpendicular wall would result in a higher efficiency in the modeling (which was not utilized) and a higher count rate when measurements are obtained, resulting in a conservative concentration for the geometry.}}