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ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 1 of 156 Enclosure 1 Zion License Termination Plan Response to PAB Zion RAIs, HP Zion RAIs, Environmental Zion RAIs and Comments on TSD 14-022 (Use of In Situ Gamma Spectroscopy for Source Term Survey of End State Structures)
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 2 of 156 PAB Zion RAIs Chapter 2
- 1. NRC Comment: The determination of background values for Cs-137 needs clarification.
Basis: Section 2.3.1 summarizes the three background studies performed at the site (February 2012 ESCSG Study, March 2012 Crib House Study, and July 2012 ZionSolutions Soil Study).
During the first study in February of 2012, 15 samples of concrete, asphalt and soil were collected from the Vertical Concrete Cask Construction area. This study determined that only natural radioactivity was detected. The second study, during which concrete core samples were taken from the Crib House, also determined that only natural radioactivity was detected. In the third study of soil samples from Hosah Park, Cs-137 was identified in addition to naturally occurring radionuclides. In the Hosah Park study, 30 soil sample locations were chosen where static measurements were taken, as well as surface and subsurface volumetric samples. It is postulated that the Cs-137 is attributable to global fallout. The LTP states that while there did appear to be evidence of soil disturbance at Hosah Park, the evidence suggested that this occurred in the past and the land has been undisturbed for a number of years. The study concluded that the majority of the soil samples from Hosah Park were from disturbed soils. The Hosah Park results are summarized in Table 2-10, which shows an average of 0.0626 pCi/g and a maximum of 0.241 pCi/g.
Section 2.3.1.4 of the LTP states The soil sample data compiled in the TSD concludes that the majority of the soil samples taken for the background studies were from disturbed soils. The Hosah Park data as well as the data obtained during the ESCSG study corresponded with documented fallout levels from disturbed soil at sites in Massachusetts, New York and Pennsylvania. Consequently, predicted ranges for background concentrations of Cs-137 were established for disturbed soils as well as undisturbed soils based on literature. These ranges are presented in Table 2-11. The upper Cs-137 concentration for each category was used as the investigation levels for non-impacted open land survey units. The upper Cs-137 concentration for disturbed, non-drainage in Table 2-11 was used as the investigation level for Class 2 and 3 open land area survey units.
Section 2.3.1.4 of the LTP suggests that the background study data from the Hosah Park and ESCSG study corresponded to global fallout levels; the LTP does not provide a basis for why this conclusion was reached even though Cs-137 was not found in the ESCSG Study or Crib House Study. The upper values of global fallout ranges were assigned as investigation levels for Cs-137 when initially characterizing the survey units. Table 2-11 provides Investigative Levels for Cs-137 Based on Background Studies. The upper value for undisturbed soil is 2.8 pCi/g or 0.66 pCi/g for a drainage or non-drainage area respectively. The upper value for disturbed soil in drainage area is 1.67 pCi/g. The upper value for a disturbed, non-drainage area, which was used for Class 2 and 3 open land area survey units, is 0.34 pCi/g. These upper values are well above the average of 0.0626 pCi/g or maximum of 0.241 pCi/g found in Hosah Park. The LTP does
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 3 of 156 not provide a basis for using the upper values of the global fallout ranges from Massachusetts, New York and Pennsylvania, as opposed to the site-specific data.
NUREG 1757 Vol. 2, Rev. 1, Section A.3.2, providing guidance on soil reference areas, states that if there is a choice of possible reference areas with similar soil types, consideration should be given to selecting reference areas that are most similar in terms of other physical, chemical, geological, and biological characteristics. Section A.3.4 provides additional guidance on differences in background areas.
Path Forward:
- a. Provide a basis for why the ESCGG study corresponds to documented global fallout ranges, even though Cs-137 was not found in the ESCSG study.
- b. Provide a basis for assuming the upper values of the global fallout ranges from Massachusetts, New York and Pennsylvania as investigation levels, as opposed to the site-specific background reference data.
Zion Station Restoration Project (ZSRP) Response (1a and 1b) - ZSRP previously submitted a response to RAIs 1a and 1b on January 25, 2016 (ZS-2016-0014).
- c. Provide "Determination of Radionuclide Activity Concentrations in Soils in Non-Impacted Soils Adjacent to the Zion Nuclear Station" (Reference 2-21).
- d. Provide ZionSolutions TSD 13-004, "Examination of Cs-137 Global Fallout In Soils At Zion Station" (Reference 2-22).
ZSRP Response (1c and 1d) - ZSRP previously submitted these reference documents on January 25, 2016 (ZS-2016-0014).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 4 of 156
- 2. NRC Comment: The use of background values in the Final Status Survey compliance determination should be clarified.
Basis: Chapter 2 of the LTP discusses the background studies performed. The LTP (see Section 5.6.4.2 and Section 5.10) states that the Sign Test will be applied for open land survey units and buried piping when demonstrating compliance with the unrestricted release criteria without subtracting background. However, the licensee reserves the option to use the WRS test where background is a significant fraction of the DCGLw. The LTP does not provide enough detail regarding how the decision will be made to use the Sign Test or the WRS test or which background reference area will correspond to each survey unit.
Path Forward:
- a. Provide a detailed example on how the decision to use the Sign Test or the WRS test within a survey unit will be determined.
- b. Provide the background reference area that will be applied in the FSS in each survey unit should the WRS test be used. This may be provided as an addendum to Table 2-3 and Table 2-4 which lists each survey unit, or (assuming reference areas are consistent across types of survey units) in a separate table which categorizes the types of survey units and their corresponding background reference area (e.g., Impacted Structures and Systems will use the Crib House Background Reference Area).
ZSRP Response (2a and 2b) - ZSRP incorporated the Wilcoxon Rank Sum (WRS) Test into the License Termination Plan (LTP) as an option in order to optimize flexibility. However, based upon the radiological characterization and assessments performed to date, background is known to be a small fraction of the unrestricted use criteria for soil, basements and piping.
ZSRP will therefore commit to the exclusive use of the Sign Test for Final Radiation Survey (FRS) data assessment. All references to WRS test will be deleted from the LTP.
The Sign Test will be applied without subtracting background. Therefore NRC RAI 2b, which requests more information regarding background reference areas, is no longer considered applicable.
- c. Provide Energy Services Commercial Services Group Report CS-RS-PN-028, Background Reference Area Report - Zion Nuclear Power Station, - February 2012 (Reference 2-20).
ZSRP Response (2c) - ZSRP previously submitted this reference document on May 12, 2015 (ZS-2015-0084).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 5 of 156
- 3. Comment: The definition of investigation levels in the FSS should be clarified.
Basis: Section 2.3.1.4 of the LTP defines investigation levels for Cs-137 as the upper range of the global fallout values for open land survey units. However, Section 5.7.1.1 states that The investigation levels may be based on the DCGLw, a fraction of the DCGLw, or the DCGLEMC, depending upon the detection capability (instrument and surveyor) to identify radioactivity, and investigation levels were further discussed in Section 5.6.4.6 and Table 5-13. Additionally, Section 5.10 discussing the FSS data assessment states, Survey results will be converted to appropriate units of measure (e.g., dpm/100 cm2, pCi/g) and compared to investigation levels to determine if the action levels for investigation have been exceeded. Measurements exceeding investigation action levels will be investigated. If confirmed within a Class 1 survey unit, the location of elevated concentration may be evaluated using the EMC, or the location may be remediated and re-surveyed. If measurements exceeding investigation action levels are confirmed within a Class 2 or 3 survey unit, in most cases, the entire survey unit will be reclassified and a re-survey performed consistent with the change in classification. It is not clear if the licensee is using the term investigation level consistently when referring to the surveys that were done for initial characterization versus the FSS, and the additional reference to an investigation action level should be clarified.
The investigation levels for FSS are defined in Table 5-13, but this table does not seem to include the investigation level discussed on pg 5-43 In addition, if during the performance of FSS, the analysis of a surface soil sample, or the results of a surface gamma scan indicates the potential presence of residual radioactivity at a concentration of 75 percent of the subsurface DCGLW, then additional biased subsurface soil sample(s) will be taken within the area of concern as part of the investigation.
Path Forward:
- a. Clarify if Table 5-13 includes all the investigation level(s) that will be used in the FSS and add any additional investigation levels not included to Table 5-13.
ZSRP Response (3a) - The statement in section 2.3.1.4 pertains solely to the characterization of non-impacted open land survey units. The last two sentences of section 2.3.1.4 will be revised as follows to clarify that the action levels, previously referred to as investigation levels, applied specifically to characterization; The upper Cs-137 concentration for each category was used as the action level for the characterization of non-impacted open land survey units. The upper Cs-137 concentration for disturbed, non-drainage soil in Table 2-11 was used as the action level for the characterization of Class 2 and 3 open land area survey units.
Also, as a clarification, the last paragraph of LTP section 5.1 will be deleted and replaced with the following text. This was done to specify that the action levels cited for the assessment of
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 6 of 156 Hard-to-Detect (HTD) radionuclides is specific to continuing characterization surveys and do not apply to FRS.
Sufficient characterization samples have been taken from the Containment structures, Auxiliary Building, Turbine Building and Crib House/Forebay concrete to derive the radionuclide mixture and assess the dose impact of HTD radionuclides. The only remaining end-state concrete structure that has not been fully characterized to date is the concrete under the steel liner of the SFP/Transfer Canal. When the underlying concrete of the SFP/Transfer Canal is exposed after the removal of the steel liner, the concrete will be characterized in the same manner as the other end-state concrete structures (surfaces will be scanned and concrete core samples will be taken at the locations of the highest scan results). Continuing characterization will also be performed in several potentially contaminated embedded pipe systems that will be abandoned in place, specifically the floor drains in the 542 foot elevation basement floor of the Auxiliary Building and the Core Spray penetrations between the Containment basements and the Auxiliary Building basement. When the interior surfaces of these pipes systems become accessible, samples will be taken of any loose surface debris in the pipe. In both of these cases, the concrete core and/or debris samples will be analyzed for the presence of HTD radionuclides. If the sample analysis indicates the presence of a significantly different radionuclide mixture from the mixture derived for the Auxiliary Building floor (Table 5-2), then the unique mixture will be documented and applied to the SFP/Transfer canal and/or embedded pipe systems as applicable during survey design and STS.
If a sample and/or measurement is taken on any other end-state structure or embedded pipe system to support decommissioning activities, Radiological Assessments (RA) or Remedial Action Support Surveys (RASS), and the result indicates a SOF in excess of 0.5 based on gamma spectroscopy results, then a sample will be collected at the location of the highest accessible individual measurement and analyzed for HTD radionuclides. If any continuing characterization surveys taken in soil or buried pipe indicate the presence of gamma-emitting radionuclides at concentrations in excess of a SOF of 0.5, then the samples will be analyzed for the presence of HTD radionuclides. In these unlikely situations, if the analysis indicates the presence of HTD radionuclides (other than Ni-63 and Sr-90, which are known to be present) at detectable concentrations, then additional investigation/sampling will be performed.
Based upon the analysis of radionuclide fractions and dose contribution in TSD 14-019, the dose contribution from HTD fractions is expected to be very low in all media (concrete, soil, embedded pipe, buried pipe, penetrations) at a SOF of 0.5 with even the most extreme HTD ratios. In the unlikely situation where these investigation levels are exceeded and one or more HTD radionuclides other than Ni-63 or Sr-90 are positively identified, then the dose impact of the positive HTD radionuclide(s) will be assessed. Additional samples may be collected and analyzed for HTD radionuclides to support the assessment of the dose impact.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 7 of 156 The investigation levels that will be employed during the FSS of impacted open land survey units is presented in section 5.6.4.6 and Table 5-13. Table 5-13 will be revised as follows to ensure consistency with MARSSIM guidance.
Table 5-13 Investigation Levels (Revised)
Classification Scan Investigation Levels Direct Investigation Levels
>DCGLW or >MDCscan if Class 1 MDCscan is greater than > DCGLW DCGLW
>DCGLW or >MDCscan if Class 2 MDCscan is greater than >DCGLW DCGLW
>DCGLW or >MDCscan if Class 3 MDCscan is greater than >0.5 DCGLW DCGLW
- b. Clarify the usage of investigation levels in Chapter 2 of the LTP versus investigation levels as discussed in Section 5.6.4.6.
ZSRP Response (3b) - The investigation levels cited for FSS in section 5.6.4.6 were not used as investigation levels for characterization. The use of the term investigation level in LTP Chapter 2 was intended to convey the implied expectation that an action will be taken when the investigation level was exceeded. The concept however is the same as the action level or investigation level was used to assess the correct classification of the survey unit. In reference to characterization, the term action level and investigation level were used interchangeably in Chapter 2 as they pertained to the same concept. This will be clarified in Chapter 2 by revising all reference to investigation levels for characterization to the more appropriate term action level.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 8 of 156
- 4. Comment: The use of Cs-137 global fallout values as an investigation level during site characterization and initial survey unit classification needs clarification.
Basis: Section 2.3.1.4 of the LTP states that the upper values of global fallout ranges were assumed as background for Cs-137 when characterizing the open land survey units. However, it is not clear how this assumption impacted the initial classification of survey units (if at all).
It is also not clear if the global fallout levels were also assumed as background in characterizing survey units consisting of concrete, asphalt, or piping. NUREG 1757 Vol. 2, Section A.3.3, states that when there are different materials with substantially different backgrounds in a survey unit, the licensee may use a reference area that is a non-impacted room with roughly the same mix of materials as the survey unit. If a survey unit contains several different materials, but one material is predominant or if there is not too great a variation in background among materials, a background from a reference area containing only a single material may still be appropriate. However, the licensee should demonstrate that the selected reference area will not result in underestimating the residual radioactivity on other materials. The LTP suggests that the Hosah Park soil background reference area will be applied for the open land areas. It is not clear if the global fallout values will be assumed for background in other materials such as concrete, asphalt, or piping.
Table 2-31 summarizes the characterization for Class 3 Open Land Survey Units. The footnote on Table 2-31 (pg 2-153) for Survey Unit 10220C (south of the Radiological Restricted Area) indicates that while 41 out of 55 samples exceeded the MDC for Cs-137, with the maximum being 1.14 pCi/g Cs-137, an investigation concluded that the elevated levels of Cs-137 are due to global fallout. However, the details of this investigation for Survey Unit 10220C do not seem to be discussed in the body of the LTP.
Path Forward:
- a. Define the term investigation level as applied during initial classification of survey units.
- b. Clarify how the investigation level may impact whether an open land survey unit area is classified as non-impacted versus impacted.
- c. Evaluate whether initial classifications would change if actual site data were used instead of global fallout data.
- d. Describe the analysis that was performed to conclude that the non-impacted areas are indistinguishable from background.
- e. Provide the investigation details that concluded that the elevated levels of Cs-137 in Survey Unit 10220C are due to global fallout.
ZSRP Response (4a, b, c, d and e) - ZSRP previously submitted responses to RAIs 4a, 4b, 4c, 4d and 4e on January 25, 2016 (ZS-2016-0014).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 9 of 156 Chapter 3
- 5. Comment: More information is needed regarding the backfilling of the basements.
Basis: In Chapter 3, it is stated that the concrete debris suitable for reuse and or/clean fill will be used to backfill the basements and that the top three feet of fill will be soil only. Chapter 3 also states that the concrete debris will be processed to ensure that the pieces are less than 10 inches in diameter. In Chapter 6, it is stated that The specific composition of the backfill has not yet been determined but is expected to be some combination of sand and debris resulting from building demolition that is designated for beneficial reuse as clean hard fill.
Based on the information in the LTP, it is not clear what the composition of the backfill material will be. Additionally, it is not clear what material is included in the clean fill and what type of soil will be used in the backfill. Also, information was not provided on how the backfill material will be placed to ensure that voids are filled adequately.
This information is needed because the projected release and transport of the residual contamination in the basements depends on the type of material used and manner of the backfilling. Additionally, the long term stability of the backfilled basements depends on the way the basements are backfilled.
Path Forward:
- a. Describe the composition of the backfill for the basement, including the relative ratio of materials that will be used. Also describe the method that will be used to ensure that the material is clean and that all voids are filled. If more than one option is being considered for the backfill, please provide a description of all possible options.
ZSRP Response (5a) - ZSRP developed Technical Support Document (TSD)14-005, Backfill Material Specifications, which is provided in Enclosure 2. TSD 14-005 provides specifications for basement fill materials and provides guidance on the selection and use of the material to ensure regulatory and commercial requirements are met and that the end-state condition assumptions remain valid.
In summary, the expected composition of backfill material for basement fill, including the relative ratios of materials and material sizes are Concrete debris from the demolition of buildings above the 588 foot elevation - The only concrete structures that will be considered as acceptable candidates for reuse as clean fill are those where the probability of being contaminated is minimal. The concrete structures that ZSRP believes are acceptable candidates for reuse as fill are the outer shell of the Containment Buildings, the concrete portions of the Turbine Building (including the Steam Tunnels) above the 588 foot, the Crib House and portions of the Forebay above the 588 foot elevation, the Service Building and minor ancillary structures that will be completely demolished such as the Interim Radioactive Waste Storage Facility (IRSF) (which was never
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 10 of 156 used to store radioactive waste), the Mechanical Maintenance Training Center (MMTC) and Warehouse, the Fire Maze complex, the NGET building, the ENC building, the south Warehouse and the North Security Access Gatehouse. Demonstration that plant-derived radioactivity is not present will be accomplished through surveys designed in accordance with NUREG-1575, Supplement 1, Multi-Agency Radiation Survey and Assessment of Materials and Equipment Manual (MARSAME). Candidate concrete surfaces shall be deemed to contain, or be contaminated, with plant-derived radioactivity if radiological surveys and/or sample analyses positively identify plant-derived radioactivity. For solid materials, the required MDCs for scan measurements and smears shall be no greater than the corresponding limits in NRC I.E. Circular No. 81-07. Once concrete has been determined to be radiologically acceptable, the concrete will be demolished and reduced in size to what is referred to in demolition terminology as 10 inch minus.. All metal and rebar will be removed from the concrete. As an estimate, it is projected that approximately 10% of the concrete material will be at or near the 10 inch diameter size, approximately 70% of the material will be sized to less than 3 inches in diameter, and approximately 20% will be fines.
Stockpiled soils from the excavation of subsurface structures and/or systems - On-site clean overburden soils excavated to access and remove subsurface structures and/or systems will be stockpiled and used as basement fill. These soils are addressed in LTP section 5.7.1.6.
This section will be revised for clarification as follows; In several areas, clean overburden soils may be removed and stockpiled on site for use as backfill materials. Prior to reuse, excavated soil will be surveyed to determine its suitability.
ZSRP will demonstrate that the soil is free of detectable plant-derived radioactivity through the use of a graded survey approach. Sufficient radiological surveys will be performed to demonstrate that the soils originating from impacted areas and intended for use as backfill meets the criteria for unconditional release off-site as clean material. The scope of the survey will be designed and documented using DQOs and will be comparable to the rigor of a Final Status Survey. Soils satisfying the criteria for unconditional release may be stockpiled for use as onsite backfill material. Soils with detectable plant-derived radioactivity at concentrations greater than background will not be used as backfill for building basements. These soils may be used to backfill excavation voids outside of the building basement footprints. Stockpiled soils will be controlled using the methods described in section 5.6.3. Scanning requirements and soil sample frequency shall also be determined in accordance with the classification of the area where the soil had originated. Controls will be instituted to prevent mixing of soils from more restrictive survey area classifications (e.g.,
Class 2 material could be used in either Class 1 or 2 areas and Class 1 material could only be used in Class 1 areas).
Clean soil from off-site sources - ZSRP will augment any deficiencies in the availability of on-site material by importing soil from off-site sources. Currently, ZSRP has a contract in place to acquire soil for use as basement fill from the Zion Municipal Landfill that is located
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 11 of 156 approximately 10 miles from ZNPS. This soil is virgin and undisturbed and was sampled for radiological and non-radiological constituents prior to it being deemed acceptable for use as clean backfill material. This soil is described as lean clay with sand (dark grayish brown; 1%
fine gravel, 2.2% coarse sand, 4.7% medium sand, 8.8% fine sand, 54.3% silt, and 28.5%
clay) and silty clay with sand (very dark grayish brown; 9.7% fine gravel, 1.5% coarse sand, 3.4% medium sand, 11.3% fine sand, 55.1% silt, and 19% clay).
Clean concrete debris and soil will be placed in the basement void up to the 588 foot elevation.
The wide range of sizes included in the concrete debris after crushing to 10 inch minus ensures good packing and that voids will be filled. The addition of soil in any ratio will also promote good packing. After placement into the basement, the material will be compacted by a minimum of three passes by a heavy roller or equivalent machinery. The remaining three feet of fill up to the 591 foot elevation (grade) shall consist of only soil (no concrete debris). The overburden soil shall be placed in 18 inch layers and compacted by a minimum of three passes by a heavy roller or equivalent machinery.
In addition, the referenced sentence from Chapter 6 that is cited in the RAI which states The specific composition of the backfill has not yet been determined but is expected to be some combination of sand and debris resulting from building demolition that is designated for beneficial reuse as clean hard fill will be revised to state that the composition of backfill that will be used in building basements is specified in TSD 14-005, Backfill Material Specifications.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 12 of 156
- 6. Comment: More information is needed on the filling of abandoned piping with grout or fill.
Basis: Chapter 3 states that all remaining buried and piping embedded in concrete will be surveyed for compliance with the unrestricted release criteria prior to being isolated, abandoned in place and filled with grout or fill as appropriate. However, it is not clear how the licensee will decide whether to fill the pipes with grout or fill, or to not fill the pipes. A description of the grout or fill that may be used is also not provided.
Path Forward:
- a. Provide a description of the process that will be used to determine whether piping is filled with grout or fill, or not filled.
ZSRP Response (6a) - ZSRP intends to grout or fill all remaining buried pipe that is greater than four feet in diameter and at least three feet below ground with the exception of any system pipe that is to remain in service.
ZSRP does not intend to fill penetrations or embedded pipe with grout. Once compliance with the inventory limit has been demonstrated, ZSRP intends to only plug the openings into embedded pipe systems in order to prevent the possibility of the introduction of any additional inventory into the pipe.
- b. Provide description of any grout or fill that may be used, including the relative ratio of materials that may be included in the fill or grout.
ZSRP Response (6b) - For the larger diameter buried piping, ZSRP intends to use a low density cellular concrete (25 + 5 PCF) as grout. The material is a flowable foam fill that is comprised of pure concrete with no sand or aggregate. As the flowable grout is pumped into a pipe, it is injected with a foam that creates air bubbles. The entire mixture then expands to fill the pipe, including any voids. For all other applications, ZSRP intends to use a more standard grout mix consisting of ASTM C 150 (cement), ASTM C 618 (fly ash), ASTM C 260 (an air entrainment),
ASTM C 494 (Type A water reducer) and ASTM C 33 (fine aggregate).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 13 of 156 Chapter 6
- 7. Comment: Additional information is needed to evaluate the Radionuclides of Concern (ROC).
Basis: Section 6.5.2 of the LTP discusses the Radionuclides of Concern (ROC). The LTP provides an initial suite of radionuclides in Table 6-2. However, it does not provide enough details regarding the starting list of radionuclides and why certain radionuclides were eliminated from the starting list to produce the initial suite. Section 2.3.2 states, Based on the elimination of some of the theoretical neutron activation products, noble gases and radionuclides with a half-life less than 2 years, an initial suite of potential ROC for the decommissioning of the ZNPS was prepared. NUREG 1757 Vol. 2 Appendix O, Question 1 states that, The licensee should also consider historical fuel performance, operational history, and time since shutdown when defining the ROCs.
Path Forward:
- a. Provide additional details regarding the process for defining the initial suite of ROCs, including the consideration of historical fuel performance, operational history, and time since shutdown.
ZSRP Response (7a) - The process for the selection of the initial suite radionuclides is described in detail in TSD 11-001. TSD 14-019 determined the mixture fractions of the initial suite radionuclides, calculated the dose attributable to each radionuclide in the initial suite, identified the insignificant dose contributors and selected the final Radionuclides of Concern (ROCs). ZSRP previously submitted TSD 14-019 and TSD 11-001 on November 12, 2015 (ZS-2015-0163).
In summary, data from literature (NUREG/CR-3474, NUREG/CR-4289 and WINCO-1191) and Zion operational waste assessment data were used to select the initial suite of radionuclides. The literature provided activation and fission product inventories at typical nuclear plants. The operational data provided analytical results of 19 samples of plant media such as resins, smears and sludge that represented a variety of conditions and plant locations. The literature and operational data were used to determine the initial suite radionuclide list using the criteria described in TSD 11-001. During characterization, all of the radionuclides in the initial suite were included in the offsite laboratory analysis.
TSD 14-019 builds on TSD 11-001 by incorporating the results of concrete characterization data to confirm the initial suite of radionuclides, determine the radionuclide mixture fractions for the initial suite, and use the dose assessment results to determine the insignificant dose contributors.
The concrete core data is the most relevant information for determining the site-specific radionuclide mixture fractions for the initial suite radionuclides to remain in the end state. The
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 14 of 156 data was applied after being decay corrected to July 1, 2018 which is the projected license termination date for the Zion plant.
- b. Provide ZionSolutions TSD 11-001, "Potential Radionuclides of Concern during the Decommissioning of Zion Station" (Reference 6-9).
ZSRP Response (7b) - LTP Reference 6-9, TSD 11-001, Potential Radionuclides of Concern during the Decommissioning of Zion Station was previously submitted to the NRC on November 12, 2015 (ZS-2015-0163).
- c. Provide ZionSolutions TSD 14-019, "Radionuclides of Concern for Soil and Basement Fill Model Source Terms" (Reference 4-15).
ZSRP Response (7c) - LTP Reference 4-15, TSD 14-019, Radionuclides of Concern for Soil and Basement Fill Model Source Terms was previously submitted to the NRC on November 12, 2015 (ZS-2015-0163).
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- 8. Comment: Additional information is needed to evaluate the dose contribution from insignificant radionuclides and subsequent adjustments to the basement dose factors and soil DCGLs.
Basis: NRC staff considers radionuclides and exposure pathways that contribute no greater than 10% of the dose criteria (2.5 mrem/yr) to be insignificant contributors. NUREG 1757, Vol. 2, Appendix O, Question 2, states that It is incumbent on the licensee to have adequate characterization data to support and document the determination that some radionuclides may be deselected from further detailed consideration in planning the Final Status Survey (FSS).
Radionuclides that are undetected may also be considered insignificant, as long as the MDCs are sufficient to conclude that the dose contribution is less than 10% of the dose criterion (i.e., with the assumption that the radionuclides are present at the MDCs).
The LTP describes the dose contribution from insignificant ROCs from buildings (Table 6-2 and Table 6-3), and soil (Table 6-19). The dose contribution for ROCs in structures is calculated by multiplying the percent radionuclide activity (for either Containment or Auxiliary Basements in Table 6-2) by the BFM dose factors determined in TSD-14-010. Section 6.5.2.3 states, In TSD 14-019, the dose factors were used to calculate the relative dose (i.e., percentage) from each radionuclide given their respective mixture percentages. The mixture percentages for the initial suite of ROCs for Containment and Auxiliary Basement concrete were developed in TSD 14-019 using the results of the core sample analyses. Section 6.5.2.2 states, The mixture fractions for the non-gamma emitters, or Hard-to-Detect (HTD) radionuclides, were determined by analyzing selected cores from the Containment and Auxiliary Basements that contained the highest radionuclide concentrations based on gamma spectroscopy. The use of cores with higher concentrations was required to ensure that the percentage assigned to HTD radionuclides were not overly influenced by the MDC values which was the only concentration data available for the majority of the HTD radionuclides in the initial suite.
A similar approach was applied for determining the dose contribution from each ROC for soil.
The LTP describes 10 soil samples that were analyzed for HTDs (Section 2.3.5.1 and Section 2.3.5.2). However, section 6.8.2 states, there were very few positive soil sample results identified during characterization and the levels were insufficient to provide a meaningful evaluation of HTD radionuclides. Therefore, the radionuclide mixture for the Auxiliary Basement cores was applied to soil for planning purposes. The percent mixture for the Auxiliary Basement was multiplied by the dose to source ratios for soil in TSD-14-10 to calculate dose contribution from each ROC for soil.
There are several issues with the licensees approach to calculating dose from insignificant radionuclides. First, it is not clear that the values assumed for the concentrations (mixture percent) are appropriate. These should represent the likely concentration of the insignificant radionuclides based on characterization data or detection limits. This information may be included in the reference TSD 14-010 or TSD 14-019. Second, the combined dose from the
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 16 of 156 insignificant radionuclides must not exceed 2.5 mrem/yr. The licensee multiplies the percent activity by a dose to source ratio to calculate a dose and then finds the ratio of each ROC to the total dose. The ratios are presented, but the actual expected doses from each ROC are not provided. It is not clear if the characterization data supports the determination that the dose from all the insignificant radionuclides for all pathways does not exceed 2.5 mrem/yr. Furthermore, it is not clear how the expected annual dose in mrem will be calculated from characterization data.
Clarifying Comments:
The percent annual dose in Tables 6-2 and 6-3 should add up to 100%. The sum is 98.33%.
The dose contribution from insignificant radionuclides for soil DCGLs reported in TSD 14-010 is inconsistent with that reported in the LTP. The LTP states that the insignificant radionuclides contribute 0.171% of the dose from soil. This percentage is based on the Auxiliary concrete mixture fractions listed in Table 6-2 of the LTP, and are calculated using the methods described in TSD 14-019 and Section 6.5.2. TSD-14-010 states that the DCGLs were adjusted to account for the 0.41% dose contribution from insignificant contributors as calculated in TSD-14-019.
ZSRP Response 8 Clarifying Comment - The correct value for the dose percentage attributable to insignificant contributors in soil is 0.171%. The value of 0.41% stated in section 3.2 of TSD 14-010 is a typographical error. The correct value of 0.171% was used to calculate the soil DCGL values in Table 15 of TSD 14-010 and Table 6-27 of LTP Chapter 6. The reported DCGL values are correct. ZSRP checked the total percent annual dose in Tables 6-2 and 6-3 and found that the sum is 100%.
Path Forward:
- a. Provide the technical basis for the Percent Activity for the Containment and Auxiliary Basement listed in Table 6-2 (reference TSD 14-019). For example, show how the core data was aggregated to define the percent radioactivity for each ROC.
ZSRP Response (8a) - LTP Table 6-2 is reproduced from Table 8 of TSD 14-019. Table 8 of TSD 14-019 presents the estimated source term mixture for the Auxiliary Building and Containment basements based on the analysis of concrete cores taken during characterization, decay corrected to July 1, 2018. The approach and concrete core analysis that was used to derive the source terms in the Auxiliary Building and Containments are described in TSD 14-013 and TSD 13-006 respectively. The mixture percentages presented in Table 8 (as well as Table 6-2 of LTP Chapter 6) were derived from the decay corrected estimated activities for each initial suite radionuclide presented in Table 7 of TSD 14-019. The estimated activities are derived from the results of the analysis of concrete core samples obtained during characterization, which is presented in in TSD 14-013 for the Auxiliary Building and TSD 13-006 for the Containments TSD 14-013 estimates the volume of concrete below the 588 foot elevation that will remain in the end-state condition of the Auxiliary Building basement as well as the volume and surface area of potentially radioactively contaminated concrete. The TSD also provides bounding
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 17 of 156 estimates for the concrete source terms based upon concrete cores collected from the 542 foot elevation concrete floor and walls. Concrete cores were collected at locations that exhibited the highest contact dose rates and were therefore biased high.
TSD 13-006 estimates the volume of concrete below the 588 foot elevation in both Containment basements that will remain in the end-state condition as well as the volume and surface area of potentially radioactively contaminated concrete. TSD 13-006 also provides bounding estimates of the residual contamination on the containment liner assuming that all of the interior concrete is removed in accordance with current decommissioning plans. The source term estimates were based upon concrete cores collected from both 568 foot elevation concrete floors inside the liner and from both 541 foot elevation under-vessel floors and walls. As with the Auxiliary Building, concrete cores were collected at locations that exhibited the highest contact dose rates and were therefore biased high.
TSD 13-006 Reactor Building Units 1 & 2 End State Concrete and Liner Initial Characterization Source Terms and Distributions and TSD 14-014, Revision 1, End State Surface Areas, Volumes, and Source Terms of Ancillary Buildings are provided in Enclosure 2.
TSD 14-019, Radionuclides of Concern for Soil and Basement Fill Model Source Terms, was previously provided to NRC on November 12, 2015 (ZS-2015-0163).
Please note that there are trivial differences between the activity concentrations and mixture percentages reported in the corresponding Tables in TSD 14-019, TSD 14-014, TSD 13-006 and Table 6-2 of LTP Chapter 6. These differences are attributable to rounding in the Excel spreadsheets and significant figures used in the decay calculation. The dose consequences of the difference are negligible.
- b. Provide justification for using the cores from the Containment and Auxiliary Basements that contained the highest radionuclide concentrations based on gamma spectroscopy for determining the mixture fractions of non-gamma emitting radionuclides if it is not already contained in the reference requested.
ZSRP Response (8b) - As stated in LTP Chapter 6 section 6.5.2.2, The use of cores with higher concentrations was required to ensure that the percentage assigned to HTD radionuclides were not overly influenced by the MDC values which was the only concentration data available for the majority of the HTD radionuclides in the initial suite. The cores with the highest activity, as indicated by onsite gamma spectroscopy, were sent to an offsite laboratory for analysis of the HTD radionuclides in the initial suite.
The evaluation of HTD analysis results for concrete core samples, as described in the ZSRP response to PAB 8a, showed that essentially all of the HTD radionuclides (with the exception of Ni-63 and Sr-90 which are known to be present) are below or near the MDC with Cs-137 concentrations that range from 1,000 pCi/g to 13,000 pCi/g.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 18 of 156 Higher activity cores are required to develop a reasonably accurate estimate of the mixture percentages for HTD radionuclides because the potential for positive detection of HTD radionuclides (i.e. at concentrations greater than MDC) increases with increasing core activity (predominantly Cs-137). The most accurate radionuclide mixture fractions will be determined if HTD radionuclides are positively detected. If HTD radionuclides are not positively detected, the mixture percentages assigned to the HTD radionuclides are based on MDC values and will be conservative for all cores because by definition the HTD actual activity present is less than the MDC. However, the accuracy of the calculation using MDC values will improve with increasing Cs-137 activity.
Using low activity cores to determine the mixture is technically incorrect in that the mixture percentage assigned to HTD radionuclides will artificially increase as an inverse function of Cs-137 activity. There is no credible mechanism that would cause the HTD mixture percentage to increase with decreasing Cs-137 activity. To determine the most accurate mixture, particularly when the HTD radionuclides are below MDC over a wide range of Cs-137 activity, the highest activity samples must be used. As stated above, since the analysis of the concrete cores showed that the majority of HTD radionuclides are below MDC over the entire range of Cs-137 activity, the calculation of mixture percentage using the highest activity samples remains conservative.
In the Basis discussion, NRC also states the following; Second, the combined dose from the insignificant radionuclides must not exceed 2.5 mrem/yr. The licensee multiplies the percent activity by a dose to source ratio to calculate a dose and then finds the ratio of each ROC to the total dose. The ratios are presented, but the actual expected doses from each ROC are not provided. ZSRP response: The use of the dose ratios does correctly identify the dose from insignificant dose contributors. If one were to prefer seeing the results in terms of dose relative to 25 mrem/yr as opposed to dose ratios, the total dose can be scaled to 25 mrem/yr and the same scaling factor applied to all individual radionuclide doses in the table. A dose ratio of 10%
would then equal 2.5 mrem/yr in the table. However, the insignificant radionuclide dose adjustment factor would not change from that calculated using the dose ratios.
- c. Provide additional consideration that will be utilized, if necessary, to justify consistent relative abundance levels site-wide throughout the decommissioning process, and to justify further reductions in overall contamination during decommissioning. Furthermore, provide basis that decommissioning activities will not result in increased relative concentrations of insignificant radionuclides.
ZSRP Response (8c) - As stated in the response to NRC RAI PAB 13, during continuing characterization, if a sample gamma spectroscopy result exceeds a Sum-of-Fraction (SOF) of 0.5, then the sample will be analyzed for HTD radionuclides. The dose impact of any positively identified HTD radionuclide in a sample with a SOF greater than 0.5 will be evaluated. The dose evaluation will include a review of the radionuclide mixtures and insignificant contributor results provided in the LTP to determine if any adjustments are justified. This review will be
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 19 of 156 conducted during continuing characterization and will be resolved and finalized before FRS begins. No additional HTD analysis will be conducted during FRS.
Regarding the second part of the RAI, there is no credible mechanism for increasing the relative concentrations of insignificant radionuclides during decommissioning. The remediation would either isolate or remove the contaminated material which has been evaluated in the insignificant radionuclide calculation or result in inadvertent mixing with adjacent clean material. Neither of these processes would concentrate one radionuclide preferentially versus another.
- d. Provide the detailed calculations to determine the percent annual doses in Table 6-3 and Table 6-19 if they are not already provided in the reference requested.
ZSRP Response (8d) - The detailed calculations are described in TSD 14-019, (provided to the NRC on November 12, 2015 [ZS-2015-0163]).
- e. Calculate the potential dose (mrem) from the initial suite of 26 ROCs in soil assuming that the radionuclides which were not measureable (above the instrument MDC) are present at their respective MDCs, and assuming other radionuclides are present at levels supported by the characterization data for soil in open land survey units.
ZSRP Response (8e) - The requested dose calculation is provided in the table below. The averages of the HTD results in soil provided in LTP Chapter 2, Table 2-34 were used for the analysis. The surface soil DCGLs for the initial suite radionuclides were from a RESRAD report in Attachment 14 to TSD 14-010, RESRAD Dose Modeling for Basement Fill Model and Soil DCGL and Calculation of Basement Fill Model Dose Factors. The initial suite RESRAD report was previously submitted to the NRC with TSD 14-010 on March 30, 2015 (ZS-2015-0051).
The dose assessment results are provided in the table below.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 20 of 156 Dose Assessment for Initial Suite Radionuclides using Characterization Results for Soil L1-12102- L1-12102- L1-12102- L1-12102- L1-12103- L1-12103- L1-12109- L1-12112- L1-12112- L1-12103-Soil DCGL CJGSSS- CJGSSS- CJGSSS- CQGSSS- CJGSSS- CJGSSS- CJGS-050- CJGSSS- CJGSSS- CJGSSB- Average Dose 0.15 m 0510 0710 0810 0810 1210 1310 SS 0410 0610 0821 pCi/g per 25 (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) mrem/yr mrem/yr H-3 2.51E+00 2.73E+00 2.54E+00 2.48E+00 8.39E-01 2.51E+00 3.96E+00 3.56E+00 3.61E+00 2.69E+00 2.74E+00 4.581E+03 1.50E-02 C-14 8.52E-01 8.10E-01 9.49E-01 8.76E-01 3.36E+00 8.51E-01 1.01E+00 1.07E+00 9.26E-01 9.56E-01 1.17E+00 8.959E+01 3.25E-01 Fe-55 3.53E+00 3.19E+00 3.24E+00 3.47E+00 7.40E+00 3.06E+00 N/A 7.16E+00 6.34E+00 3.24E+00 4.51E+00 3.374E+04 3.35E-03 Ni-59 7.28E+00 7.23E+00 6.65E+00 6.30E+00 4.29E-02 6.30E+00 N/A 6.12E+00 7.71E+00 5.69E+00 5.92E+00 1.094E+04 1.35E-02 Co-60 4.82E-02 4.52E-02 3.00E-02 3.22E-02 6.21E-01 4.29E-02 2.43E-01 8.80E-02 1.22E-01 3.98E-02 1.31E-01 4.734E+00 6.93E-01 Ni-63 6.22E-01 7.06E-01 6.24E-01 6.41E-01 2.67E-01 6.26E-01 1.77E+00 8.36E-01 7.95E-01 6.10E-01 7.50E-01 3.996E+03 4.69E-03 Sr-90 2.90E-01 3.15E-01 2.35E-01 2.99E-01 2.92E-02 3.33E-01 4.70E-01 3.76E-01 3.59E-01 2.95E-01 3.00E-01 1.439E+01 5.21E-01 Nb-94 3.35E-02 2.64E-02 2.21E-02 2.40E-02 4.43E-01 2.85E-02 5.35E-02 7.05E-02 7.16E-02 2.83E-02 8.01E-02 7.507E+00 2.67E-01 Tc-99 4.03E-01 4.38E-01 4.20E-01 3.95E-01 2.97E-02 4.18E-01 2.61E-01 2.61E-01 2.62E-01 6.39E-01 3.53E-01 1.277E+02 6.90E-02 Ag-108m 2.95E-02 3.05E-02 2.03E-02 2.21E-02 8.03E-02 2.91E-02 4.89E-02 7.14E-02 7.52E-02 2.62E-02 4.34E-02 7.400E+00 1.46E-01 Sb-125 9.26E-02 8.40E-02 6.11E-02 6.35E-02 2.78E-02 8.44E-02 1.68E-01 2.19E-01 2.90E-01 7.13E-02 1.16E-01 3.360E+01 8.64E-02 Cs-134 2.92E-02 2.91E-02 2.24E-02 2.41E-02 2.63E-02 2.85E-02 6.04E-02 7.72E-02 9.92E-02 2.72E-02 4.24E-02 7.524E+00 1.41E-01 Cs-137 6.98E-01 3.85E-01 7.98E-02 8.86E-02 3.15E-01 4.57E-01 1.57E-01 2.30E+00 3.39E+00 3.31E-02 7.90E-01 1.576E+01 1.25E+00
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 21 of 156 L1-12102- L1-12102- L1-12102- L1-12102- L1-12103- L1-12103- L1-12109- L1-12112- L1-12112- L1-12103-Soil DCGL CJGSSS- CJGSSS- CJGSSS- CQGSSS- CJGSSS- CJGSSS- CJGS-050- CJGSSS- CJGSSS- CJGSSB- Average Dose 0.15 m 0510 0710 0810 0810 1210 1310 SS 0410 0610 0821 pCi/g per 25 (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) (pCi/g) mrem/yr mrem/yr Eu-152 1.84E-01 2.13E-01 1.61E-01 1.83E-01 8.26E-02 1.76E-01 1.60E-01 6.27E-01 4.74E-01 2.27E-01 2.49E-01 1.074E+01 5.79E-01 Eu-154 1.16E-01 9.27E-02 7.38E-02 7.90E-02 7.11E-02 1.10E-01 8.00E-02 2.23E-01 2.66E-01 9.19E-02 1.20E-01 9.969E+00 3.02E-01 Eu-155 7.42E-02 8.00E-02 6.40E-02 6.32E-02 2.01E-02 7.83E-02 1.43E-01 1.87E-01 2.12E-01 7.41E-02 9.96E-02 3.909E+02 6.37E-03 Np-237 2.43E-02 2.42E-02 2.37E-02 3.22E-02 3.83E-02 2.35E-02 N/A 7.80E-02 7.14E-02 2.58E-02 3.79E-02 8.006E-01 1.18E+00 Pu-238 5.78E-02 3.45E-02 3.33E-02 3.34E-02 3.83E-02 4.30E-02 9.30E-02 7.45E-02 3.95E-02 4.52E-02 4.93E-02 1.617E+02 7.61E-03 Pu-239/240 6.53E-02 4.71E-02 3.33E-02 3.02E-02 6.56E+00 3.88E-02 6.25E-02 6.87E-02 4.49E-02 4.52E-02 7.00E-01 1.456E+02 1.20E-01 Pu-241 8.31E+00 6.13E+00 5.20E+00 5.05E+00 3.57E-02 6.55E+00 N/A 4.89E-01 5.10E-01 7.84E+00 4.46E+00 6.519E+03 1.71E-02 Am-241 6.19E-02 2.91E-02 4.90E-02 3.78E-02 7.03E-02 3.45E-02 4.61E-02 8.29E-02 5.50E-02 4.58E-02 5.12E-02 1.336E+02 9.59E-03 Am-243 3.33E-02 3.90E-02 3.80E-02 4.10E-02 3.44E-02 4.30E-02 N/A 4.51E-02 5.67E-02 4.10E-02 4.13E-02 4.979E+01 2.07E-02 Cm-243/244 5.74E-02 5.40E-02 2.80E-02 4.09E-02 8.39E-01 5.40E-02 N/A 1.22E-01 6.94E-02 3.80E-02 1.45E-01 7.606E+01 4.76E-02 Sum 5.82E+00
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 22 of 156
- f. Calculate the potential dose (mrem) from the initial suite of 26 ROCs in building structures assuming that the radionuclides which were not measureable (above the instrument MDC) are present at their respective MDCs, and assuming other radionuclides are present at levels supported by the characterization data for building structures and systems.
ZSRP Response (8f) - The requested dose calculation is provided in the table below. The Auxiliary Building basement characterization data and BFM Dose Factors were used for the calculation. The characterization data from the Containment basements is not applicable as all of the concrete in the basement will be removed during decommissioning, leaving only the steel liner. For all basements other than Containment, the Auxiliary Building basement concrete contains the highest activity levels by a wide margin and was the basis for the radionuclide mixture calculations for all remaining basements. Note that the SFP/Transfer Canal concrete is an exception in that concrete has not yet been characterized. The characterization of the end-state concrete will be performed following removal of the steel liner from the SFP/Transfer Canal, which will expose the underlying concrete. The characterization data will then be used to assess the radionuclide mixture for the SFP/Transfer Canal end-state concrete.
The calculation was performed using the total inventory estimate for the Auxiliary Building basement provide in TSD 14-019, Table 7 (which is decay corrected to July 2018 as was done in the mixture calculation). BFM Dose Factors were calculated using the DUST results provided in TSD 14-031, BNL Report: Basement Fill Model Evaluation of Maximum Radionuclide Concentrations for Initial Suite of Radionuclides, the Groundwater Exposure Factors provided in TSD 14-010, RESRAD Dose Modeling for Basement Fill Model and Soil DCGL and Calculation of Basement Fill Model Dose Factors Attachment 12, and the Drilling Spoils Dose Factors provided in TSD 14-021, BFM Drilling Spoils and Alternate Exposure Scenarios. The dose assessment results are provided below. TSD 14-031 is provided in Enclosure 2. TSD 14-010 was previously submitted to the NRC on March 30, 2015 (ZS-2015-0051) and TSD 14-021 was previously submitted on November 12, 2015 (ZS-2015-0163).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 23 of 156 Dose Assessment for Initial Suite Radionuclides using Characterization Results for Auxiliary Building basement Concrete Cores GW DF DS DF mrem/yr Radionuclide Total Ci mrem/yr mrem/yr per mCi per mCi H-3 1.46E-03 6.21E-03 0.00E+00 9.06E-03 C-14 3.69E-04 6.49E-02 3.08E-07 0.023938 Fe-55 8.85E-04 8.06E-07 2.97E-10 7.14E-07 Ni-59 4.17E-03 1.35E-04 1.51E-08 0.000561 Co-60 7.60E-03 1.00E-04 1.07E-02 0.082232 Ni-63 1.97E-01 2.86E-04 3.21E-08 0.056312 Sr-90 4.27E-04 3.29E-01 5.84E-05 0.140616 Nb-94 1.07E-04 2.03E-03 1.20E-02 0.001506 Tc-99 1.34E-04 1.48E-01 0.00E+00 0.019822 Ag-108m 1.44E-04 5.47E-03 1.23E-02 0.002565 Sb-125 1.46E-04 1.04E-02 2.75E-03 0.001924 Cs-134 8.63E-05 9.27E-03 6.29E-03 0.001342 Cs-137 6.25E-01 2.64E-02 3.22E-03 18.4961 Eu-152 1.46E-04 5.95E-05 5.02E-03 0.000742 Eu-154 7.89E-05 6.77E-05 5.57E-03 0.000444 Eu-155 6.48E-05 8.01E-06 2.83E-04 1.88E-05 Np-237 3.66E-06 4.92E+01 2.91E-03 0.179937 Pu-238 1.08E-05 2.11E-01 3.97E-05 0.002276 Pu-239 4.47E-06 2.61E-01 4.41E-05 0.001167 Pu-240 4.47E-06 2.61E-01 4.38E-05 0.001166 Pu-241 2.36E-04 4.57E-03 2.97E-06 0.00108 Am-241 1.06E-05 2.58E-01 1.51E-04 0.002733 Am-243 7.95E-06 2.63E-01 1.58E-03 0.002101 Cm-243 2.80E-06 2.59E-02 9.93E-04 7.54E-05 Cm-244 2.58E-06 1.74E-02 2.55E-05 4.5E-05 Sum 1.90E+01
- g. Provide the detailed calculations used to adjust the DCGL values from those provided in the RESRAD summary reports to those listed in Table 6-27.
ZSRP Response (8g) - As stated in LTP Chapter 6, section 6.10, the insignificant contributor dose fraction for soil is 0.171%. Therefore, to calculate the adjusted DCGLs listed in LTP
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 24 of 156 Chapter 6, Table 6-27, the RESRAD results from the reports listed in TSD 14-010 Attachment 10 were multiplied by a factor of (1 - 0.00171) to calculate the adjusted DCGLs. The RESRAD reports were previously submitted to NRC on March 30, 2015 (ZS-2015-0051).
To ensure a conservative DCGL value, the DCGLs provided in Table 6-27 were truncated to the first decimal place as opposed to standard rounding. Because the adjustment factor is very low, i.e., 0.99829, the first decimal place value did not change for most of the radionuclides. To demonstrate this, the adjustment calculations are performed below using three decimal places as reported in RESRAD. As seen in the second table below, the truncated adjusted values are the same as reported in LTP Chapter 6, Table 6-27. The trivial differences after adjustment can be seen in the second and third decimal place.
Soil DCGLs for ROC With/Without Insignificant Contributor Adjustment Surface Soil DCGL without Subsurface Soil DCGL Radionuclide Adjustment for Insignificant without Adjustment for Contributors Insignificant Contributors (pCi/g) (pCi/g)
Co-60 4.734 3.825 Cs-134 7.524 4.930 Cs-137 15.76 8.606 Ni-63 3995 848.6 Sr-90 14.36 1.860 Soil DCGLs for ROC with Insignificant Contributor Adjustment Radionuclide Surface Soil DCGL Subsurface Soil DCGL (pCi/g) (pCi/g)
Co-60 4.726 3.818 Cs-134 7.511 4.922 Cs-137 15.733 8.591 Ni-63 3988.169 847.149 Sr-90 14.335 1.857
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 25 of 156
- h. Provide a technical basis for the IC Dose Adjustment Factors in Equation 6-4 and the detailed calculations used to adjust the Basement Dose Factors in Table 6-18.
ZSRP Response (8h) - BFM Dose Factors are in units of mrem/yr per mCi. Therefore, the IC adjustment factor must increase the BFM Dose Factor. This is the inverse of the soil DCGLs (with units of pCi/g per 25 mrem/yr) which must be decreased to account for the IC as indicated in the ZSRP response to RAI 8g above. Therefore, the adjustment factor is calculated as 1/(1-IC Fraction).
TSD 14-010, (provided to the NRC on March 30, 2015 [ZS-2015-0051]) Attachment 6 provides a copy of the inputs and results of the calculation for the Adjusted BFM Dose Factors. The information in Attachment 6 is reproduced below.
Note that there was a revision to the demolition plan for the Crib House/Forebay after submittal of the LTP that entailed leaving interior walls as opposed to removing them. This results in a decrease in the basement mixing volume as compared to that assumed in the DUST-MS modeling provided in TSD 14-009, Revision 0 and a corresponding increase in the fill and groundwater concentrations calculated in TSD 14-009, Revision 0. The Basement Dose Factors are directly proportional to the fill and groundwater concentrations, which are inversely proportional to the ratio of revised/original mixing volumes. The ratio of the revised/original mixing volumes for the Crib House/Forebay was calculated in TSD 14-014, Revision 1, End State Surface Areas, Volumes, and Source Terms of Ancillary Buildings and was determined to be 0.86. The Crib House/Forebay Basement DFs were therefore adjusted higher by the inverse of 0.86 or a factor of 1.16. TSD 14-010, Revision 1, Attachment 6 includes the change to the Crib House/Forebay Basement DFs and is provided in Enclosure 2. TSD 14-014, Revision 1 is provided in Enclosure 2. TSD 14-009 was submitted to the NRC on March 30, 2015 (ZS-2015-0051). The ZSRP LTP will be revised to incorporate the changes to the Crib House/Forebay BFM Dose Factors.
The calculations of the Basement Dose Factors shown below were made using LTP Chapter 6, Equation 6-4. An example of the Auxiliary Building basement BFM Dose Factor calculation for Cs-137 is described here. First the GW Dose Factor of 2.64E-02 mrem/yr per mCi is added to the Drilling Spoils Dose Factor of 3.22E-03 mrem/yr per mCi for a total of 2.96E-02 mrem/yr per pCi/g. This value is then multiplied by the IC Adjustment Factor of 1.01222 for a final BFM Dose Factor of 3.00E-02 mrem/yr per pCi/g.
The technical basis for the IC adjustment factor is conceptually described above. A numerical demonstration of the IC adjustment is provided here. The allowable inventory of Cs-137 in the Auxiliary Building basement can be calculated as 25 mrem/yr ÷ 3.00E-02 mrem/yr/mCi = 834.6 mCi. The dose from 834.6 mCi is calculated by multiplying this value by the unadjusted dose factor of 3.22E-03 mrem/yr per mCi which results in 24.7 mrem/yr. The difference between 25 and 24.7 is 0.3018 which is 1.207% of 25 mrem/yr and equal to the Auxiliary Building basement IC percentage shown in TSD 14-010, Revision 1. Attachment 6 (reproduced below).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 26 of 156 BFM Groundwater Dose Factors Spent Fuel Pool/Transfer Crib Auxiliary Containment Canals Turbine House/Forebay WWTF (mrem/y per (mrem/y per mCi) (mrem/y per mCi) mCi) (mrem/y per mCi) (mrem/y per mCi) (mrem/y per mCi)
Co-60 1.00E-04 1.14E-02 0.00E+00 2.87E-03 2.45E-03 5.21E-01 Cs-134 9.27E-03 1.98E-01 0.00E+00 4.94E-02 4.22E-02 9.03E+00 Cs-137 2.64E-02 1.57E-01 0.00E+00 3.92E-02 3.35E-02 7.17E+00 Eu-152 5.96E-05 3.87E-03 0.00E+00 9.69E-04 8.29E-04 1.75E-01 Eu-154 6.77E-05 5.62E-03 0.00E+00 1.41E-03 1.20E-03 2.56E-01 H-3 6.21E-03 2.72E-02 0.00E+00 6.80E-03 5.80E-03 1.23E+00 Ni-63 2.86E-04 1.61E-03 0.00E+00 4.01E-04 3.44E-04 7.31E-02 Sr-90 3.29E-01 4.51E+00 0.00E+00 1.13E+00 9.66E-01 2.06E+02 BFM Drilling Spoils Dose Factors (
Reference:
TSD 14-021)
Spent Fuel Crib (1)
Auxiliary Containment(1) Pool/Transfer Turbine (1) WWTF (1)
House/Forebay Canals (mrem/y per (mrem/y per mCi) (mrem/y per mCi) mCi) (mrem/y per mCi) (mrem/y per mCi) (mrem/y per mCi)
Co-60 1.07E-02 2.97E-02 1.58E-01 9.58E-03 1.78E-02 2.26E-01 Cs-134 6.29E-03 1.72E-02 9.41E-02 5.54E-03 1.02E-02 1.31E-01 Cs-137 3.22E-03 7.27E-03 4.83E-02 2.35E-03 4.34E-03 5.57E-02 Eu-152 5.02E-03 1.38E-02 7.46E-02 4.45E-03 8.24E-03 1.05E-01 Eu-154 5.57E-03 1.46E-02 8.25E-02 4.73E-03 8.77E-03 1.12E-01 H-3 0.00E+00 0.00E+00 1.45E-09 0.00E+00 0.00E+00 0.00E+00 Ni-63 3.21E-08 5.57E-08 3.75E-07 1.84E-08 4.11E-08 4.13E-07 Sr-90 5.84E-05 1.30E-04 7.09E-04 4.31E-05 9.30E-05 9.69E-04
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 27 of 156 Basement Dose Factors Basement Dose Factor = (BFM GW Dose Factors+Drilling Spoils Dose Factors)*Insignificant Contributor Dose Adjustment Spent Fuel Crib Auxiliary Containment Pool/Transfer Turbine House/Forebay WWTF Canals (Note 1) (Note 2)
(mrem/y per (mrem/y per mCi) (mrem/y per mCi) mCi) (mrem/y per mCi) (mrem/y per mCi) (mrem/y per mCi)
Co-60 1.10E-02 4.13E-02 1.20E+00 1.26E-02 2.38E-02 7.57E-01 Cs-134 1.57E-02 2.16E-01 7.15E-01 5.56E-02 6.18E-02 9.27E+00 Cs-137 3.00E-02 1.65E-01 3.67E-01 4.21E-02 4.46E-02 7.31E+00 Eu-152 5.14E-03 1.77E-02 5.66E-01 5.48E-03 1.07E-02 2.83E-01 Eu-154 5.70E-03 2.04E-02 6.26E-01 6.21E-03 1.17E-02 3.73E-01 H-3 6.28E-03 2.73E-02 1.10E-08 6.88E-03 6.83E-03 1.25E+00 Ni-63 2.89E-04 1.61E-03 2.85E-06 4.06E-04 4.05E-04 7.40E-02 Sr-90 3.33E-01 4.54E+00 5.38E-03 1.15E+00 1.14E+00 2.09E+02 Insignificant Contributor (IC) Dose Adjustment: IC Dose % IC Dose Adjustment Factor Containment Insignificant Contributor Dose Percentage 0.514% 1.00517E+00 Auxiliary Insignificant Dose Contributor Percentage 1.207% 1.01222E+00 All other Basments use the Auxilairy Insignificant Contributor Percentage 1.207% 1.01222E+00 Note 1: Spent Fuel Pool/Transfer Canal Dose Factor also adjusted higher by a factor of 7.5 to account for dose from large scale excavation scenario (see LTP Chapter 6, section 6.7)
Note 2: The Crib House/Forebay Dose Factor was adjusted higher to account for a revision that lowered the void volume.
The demolition plan for the Crib House was revised to leave the interior walls as opposed to removing all interior walls per the original demolition plan. This resulted in a decrease in the Crib House saturated zone fill mass in the Basement Fill Model with a corresponding increase in the pCi/g and pCi/L values calculated in TSD 14-009. The BFM and DS dose factors are both directly proportional to the pCi/g and pCi/L values.The increase in the pCi/g and pCi/L concentrations are inversely proportional to the ratio of the Revised Volume to the Original Volume for the Crib House/Forebay combined. The revised Crib House/Forebay volume and the ratio of the Revised/Original volume is provided in TSD 14-013, Revision 1.
Ratio of Revised/Original Crib House/Forebay Volume = 8.60E-01
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 28 of 156
- 9. Comment: Additional information is needed on the process for potentially revising the assumed mixture of radionuclides.
Basis: The Auxiliary Basement radionuclide mixture is assumed for several other buildings as well as soil for planning purposes due to lack of characterization data. Section 6.5.2.2 states, As additional radioactive material from different sources (i.e. processed SFP water) is introduced, this could potentially result in a mixture that is different from the Auxiliary Building concrete mixture. The mixture in the SFP/Transfer Canals could also be somewhat different than the Auxiliary Building due to the source of potential contamination, i.e., fuel pool water leaking into the concrete under the liner. However, the mixture in both the Circulating Water Discharge Tunnel and the SFP/Transfer Canal is expected to be primarily Cs-137 as in the other Basements.
Therefore, the Auxiliary Basement mixture is considered reasonable for application to these two structures for planning purposes. The mixtures in these two Basements will be reviewed as continued characterization data is collected from these areas (see LTP Chapter 5, section 5.1).
Path Forward:
- a. Provide information on how the mixtures in the Containment and Auxiliary basements will be reviewed (i.e., what would trigger a recalculation of the dose contribution from insignificant radionuclides and therefore a revision to the adjusted DCGLs).
ZSRP Response (9a) - During the characterization of Zion that has been performed to date, concrete cores collected from the floor of the Auxiliary Building basement have provided the highest activity inventory in any end-state structure. As stated in section 6.5.1.1:
The current Containment basement inventories are not meaningful as a prediction of End State inventories because the vast majority of the contamination is in the concrete which will be completely removed during decommissioning.
Concrete cores taken in other structures (Turbine Building, Steam Tunnels, etc) have indicated minimal activity inventory. In addition, samples of surface and subsurface soils taken in and around Class 1 survey units have not contained significant activity levels. Consequently, due to the absence of a significant source term in soil or in other end-state structures, the radionuclide mixture derived for the Auxiliary Basement concrete was considered to be a reasonably applicable mixture to apply to BFM structures, soils and buried piping for FRS planning and implementation. The one exception is the SFP/Transfer Canal. ZSRP commits in sections 2.3.3.3 and 2.5 to characterize the underlying concrete pad and remaining pool walls once the liner is removed and the underlying concrete is exposed. Section 5.1 states that the characterization will include the acquisition of concrete core samples and that those core samples will be analyzed for the presence of HTD radionuclides. In addition, if a survey measurement taken for continuing characterization, Radiological Assessments or RASS indicates radiological concentrations in excess of a SOF of 0.5 based on gamma spectroscopy results, then a sample will be collected at the location of the highest individual measurement and that sample will be
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 29 of 156 assessed for HTD radionuclides. In this unlikely situation, if the analysis indicates HTD radionuclides (other than Ni-63 and Sr-90) at concentrations exceeding MDC, then additional investigation/sampling will be performed and the dose impact of the identified HTD radionuclides will be assessed. ZSRP considers the possibility of this occurring as very remote but maintains this action as a contingency in the even that something unexpected is encountered.
Sections 6.8.2 and 6.12.1 restate the DCGL investigation levels, but use a value of 10% of the DCGL. This is not correct and will be revised to a SOF of 0.5 consistent with section 5.1.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 30 of 156
- 10. Comment: Additional information is necessary to evaluate the site-specific Kd parameters applied in the RESRAD and DUST-MS models.
Basis: Section 6.9.1 states, Kd values for site soil (sand) were selected based on the review provided by Brookhaven National Laboratory in TSD 14-004 (see Table 6-20). Additionally, Section 6.6.1.1 states that site-specific Kd analyses performed by Brookhaven National Laboratory as documented in two reports, ZionSolutions TSD 14-017, Sorption (Kd)
Measurements on Cinder Block and Grout in Support of Dose Assessments for Zion Nuclear Station Decommissioning, and ZionSolutions TSD 14-020, Sorption (Kd) measurements in Support of Dose Assessments for Zion Nuclear Station Decommissioning. However, TSD 14-004, TSD 14-017, and TSD 14-020 have not been provided.
Also, as described in Section 6.4.4 of the LTP, the Kd values assumed in the calculation of the groundwater exposure factor calculations were the same as those assigned in the DUST-MS model. The text describes the selection process for the Kd values as being conservative.
However, low Kd values generally maximize the release and transport of radionuclides through groundwater, while high Kd values generally maximize the concentration of the radionuclides in soil that is irrigated with contaminated groundwater, which in turn maximizes the concentration in plants and ingestion doses. The use of Kd values that were conservatively determined based on the release from the source term may be non-conservative when used in the analysis of the surface soil concentrations.
Path Forward:
- a. Provide justification that the Kd values selected are appropriate for use in the calculation of the surface soil concentrations.
ZSRP Response (10a) -A sensitivity analysis was conducted to evaluate the effect of using higher Kd values for surface soil in the BFM. The 75th percentiles from the Kd distributions listed in NUREG/CR-6697 were used in the sensitivity analysis. RESRAD was run with the deterministic parameters used to calculate the BFM Dose Factors with one change, i.e. the Kd values were changed from the minimum values used for the BFM to the 75th percentile values.
Groundwater exposure factors were calculated and compared to the groundwater exposure factors listed in Table 6-15 of LTP Chapter 6. The calculation requires inputs from the RESRAD Summary Report and Concentration Report which are both provided in Enclosure 2.
Table 1 lists the RESRAD outputs and the recalculated groundwater exposure factors with the increased Kd values. The exposure factors from LTP Chapter 6, Table 6-15 are also listed and the percent difference between the two groundwater exposure factors is calculated.
As seen in Table 1, there is essentially no difference between the groundwater exposure factors with the increased Kds and those calculated with lower Kds in the LTP. The highest percent difference was for H-3, which was 3.61% lower with the higher Kd value. The percent difference for the remaining radionuclides ranged from -0.09% to 0.22%. These very low
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 31 of 156 differences demonstrate the use of low Kd values in the BFM RESRAD analysis was reasonable and appropriate for calculating dose from surface soil affected by irrigation with well water.
Table 1: Sensitivity Analysis of Surface Soil Kd on Basement Fill Model Dose RESRAD 75th RESRAD 75th Summary Percentile LTP Concentration Percent Nuclide Percentile Report Kd Report mrem/yr Difference Kd mrem/yr mrem/yr per pCi/L per pCi/g pCi/L per pCi/g per pCi/L Co-60 1294 1.936E-02 7.727E-01 2.51E-02 2.50E-02 0.22%
Cs-134 2144 4.079E-02 4.664E-01 8.75E-02 8.75E-02 -0.05%
Cs-137 2144 3.238E-02 4.664E-01 6.94E-02 6.94E-02 0.04%
Eu-152 7268 4.979E-04 1.376E-01 3.62E-03 3.62E-03 -0.04%
Eu-154 7268 7.231E-04 1.376E-01 5.26E-03 5.26E-03 -0.09%
H-3 0.08 1.501E-01 3.515E+03 4.27E-05 4.43E-05 -3.61%
Ni-63 1132 8.645E-04 8.832E-01 9.79E-04 9.78E-04 0.08%
Sr-90 131 8.360E-01 7.662E+00 1.09E-01 1.09E-01 0.10%
- b. Please provide the following references and any key references these documents cite:
ZionSolutions Technical Support Document 14-004, Brookhaven National Laboratory (BNL), Recommended Values for the Distribution Coefficient (Kd) to be used in Dose Assessments for Decommissioning the Zion Nuclear Power Plant ZionSolutions Technical Support Document 14-017, Brookhaven National Laboratory (BNL), Sorption (Kd) Measurements on Cinder Block and Grout in Support of Dose Assessments for Zion Nuclear Station Decommissioning ZionSolutions Technical Support Document 14-020, Brookhaven National Laboratory (BNL), Sorption (Kd) measurements in Support of Dose Assessments for Zion Nuclear Station Decommissioning ZSRP Response (10b) - TSD 14-004, Recommended Values for the Distribution Coefficient (Kd) to be used in Dose Assessments for Decommissioning the Zion Nuclear Power Plant, TSD 14-017, Sorption (Kd) Measurements on Cinder Block and Grout in Support of Dose
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 32 of 156 Assessments for Zion Nuclear Station Decommissioning and TSD 14-020, Sorption (Kd) measurements in Support of Dose Assessments for Zion Nuclear Station Decommissioning were submitted to the NRC on November 12, 2015 (ZS-2015-0163).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 33 of 156
- 11. Comment: Additional information is necessary to evaluate the Soil Area Factors (AFs) listed in Table 5-7, 5-8 and Table 6-28, 6-29.
Basis: The Soil Area Factors are provided in Tables 5-7, 5-8, 6-28, and 6-29. These are calculated using RESRAD for each ROC and for source area sizes ranging from 1 m2 up to the full source area of 64,500 m2. However, not enough information is provided to reproduce the AFs.
Path Forward:
- a. Please provide reference TSD 14-019 and associated RESRAD Summary Reports in TSD 14-011 for calculating AFs.
ZSRP Response (11a) - TSD 14-019, Radionuclides of Concern for Soil and Basement Fill Model Source Terms and TSD 14-011, Soil Area Factors were submitted to the NRC on November 12, 2015 (ZS-2015-0163).
TSD 14-010 included the RESRAD Summary reports used to calculate the Area Factors in LTP Chapter 6 Tables 6-28 and 6-29 and was submitted to the NRC on March 30 2015 (ZS-2015-0051). The RESRAD Summary Reports used to calculate the Area Factors listed in LTP Chapter 5 Tables 5-7 and 5-8 from TSD 14-011 are provided in Enclosure 2.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 34 of 156
- 12. Comment: Additional information is necessary to evaluate the Buried Pipe DCGLs.
Basis: Section 6.12 discusses the development of buried pipe DCGLs, but the conceptual model assumed to derive the DCGLs does not seem consistent with the expected configuration of the residual contamination following decommissioning activities. In the calculation of these DCGL values, it is assumed that the pipe is filled with soil that has an activity of radionuclides equal to the Surface Soil DCGL values. These volumetric DCGL values were then converted to an areal value on the interior surface of the pipe, which is effectively equivalent to assuming a dilution of the material on the surface of the pipe over the whole volume of the pipe. The pipes that remain on site at the time of license termination may not be filled with soil (or grouted), so it is not clear that this conceptual model applies. In addition, the dose to an individual may be different from a thin skin of contamination on the surface of the pipe than from a volumetric source containing the same amount of activity.
The conceptual model attempts to account for the potential dose from the pipes being excavated and reused through the buried pipe excavation scenario. The amount of material assumed to be exhumed is limited to a 20 m length of pipe (based on a basement which is 10 m x 20 m).
However, the possibility of an intruder encountering more than one pipe in a single basement is not discussed. Also, an Area Factor is applied in the buried pipe excavation scenario. The NRC staff does not view this application of the Area Factor to be appropriate.
The buried pipe DCGL derivation attempts to account for the contribution of the pipes to the dose for the resident through the In situ Scenario. The In situ Scenario dose is calculated by assuming the inventory of the pipes is in a continuous layer 1m from the surface, which is not consistent with the expected configuration of the pipes. Hypothetically placing the inventory in the pipes closer to the surface may be conservative for the external pathway, but it may not be conservative for the groundwater dose. Section 6.12.2.2 of the LTP states that a number of the buried pipes have a very low potential for being contaminated based on their associated systems.
Therefore, the licensee limits the inventory to that within 9.6 m3 of piping in comparison to over 100 m3 of pipe volume that could remain and is listed in Table 2-27 of the LTP. However, if the total volume of piping in Table 2-27 is expected to remain onsite, then it should be considered in the DCGL calculation.
NUREG 1757 Vol. 2 defines Derived Concentration Guideline Levels (DCGLs) as Radionuclide-specific concentration limits used by the licensee during decommissioning to achieve the regulatory dose standard that permits the release of the property and termination of the license. The DCGL is the concentration for each radionuclide which would result in doses less than or equal to the release criteria of 25 mrem/yr. It is not clear that the buried pipe DCGLs in Table 6-32 would result in doses less than or equal to 25 mrem/yr should all the pipes in Table 2-27 be left onsite and released according to the DCGLs in Table 6-32.
The licensee seems to confuse the DCGL approach (developing or using DCGLs and performing an FSS to demonstrate that the DCGLs have been met) with the dose assessment approach
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 35 of 156 (characterizing the siteafter remediation, if necessaryand performing a dose assessment) for piping. The licensee should refer to NUREG 1757 Vol. 2, Section 2.5 for guidance related to demonstrating compliance using dose assessment methods versus DCGLs and Final Status Surveys. The LTP provides DCGLs to be applied for pipes, but Equation 6-5 suggests that a maximum dose from piping will be calculated, which would indicate that the licensee intends to use the dose assessment approach.
Path Forward:
- a. If the licensee desires to follow the DCGL approach using the DCGL values in Table 6-32, provide additional dose modeling calculations showing that if all the piping to remain onsite were released at the DCGL values, the dose to the critical group would be equal to or less than 25 mrem/yr for each radionuclide.
ZSRP Response (12a)
The calculation of buried pipe DCGLs was revised to include all buried pipe, including electrical conduit and boxes, service water lines, ice melting pipe, etc., that are essentially non-impacted and were excluded from the original calculation. In addition, a more conservative approach was applied that included two in-situ scenarios, one that assumes the buried pipe is in the unsaturated zone and a second that assumed the pipe is in the saturated zone. The Buried Pipe DCGL was based on the more conservative of the two in-situ scenarios combined with an excavation scenario.
The revised approach and calculations are described in TSD 14-015, Revision 1, which is provided in Enclosure 2. The revised DCGLs are listed below.
Revised Zion Buried Pipe DCGLs (Reference TSD 14-015 Revision 1)
Buried Pipe DCGL Radionuclide (dpm/100 cm2)
Co-60 3.02E+04 Cs-134 5.33E+04 Cs-137 1.23E+05 Ni-63 1.15E+08 Sr-90 7.84E+04 The revised DCGLs are slightly lower but similar to the original DCGLs for all radionuclides except Sr-90. The revised Sr-90 Buried Pipe DCGL was about five times lower due to the very conservative inclusion of an in-situ scenario that assumes all buried pipe is located in the saturated zone.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 36 of 156
- b. If the licensee desires to follow the dose assessment approach, provide reasonable cleaning goals for pipes that can be used in conjunction with the dose assessment approach and a detailed example of how the dose from piping will be calculated during surveys.
ZSRP Response (12b) - A dose assessment approach was not proposed in the LTP and will not be used.
- c. Provide TSD 14-015, "Buried Pipe Dose Modeling & DCGLs" (Reference 6-4).
ZSRP Response (12c) - TSD 14-015, Revision 1 Buried Pipe Dose Modeling & DCGLs is provided in Enclosure 2.
- d. Provide TSD 14-021 Basement Fill Model (BFM) Drilling Spoils and Alternate Exposure Scenarios" and associated RESRAD summary files.
ZSRP Response (12d) - TSD 14-021, Basement Fill Model (BFM) Drilling Spoils and Alternate Exposure Scenarios was submitted to the NRC on November 12, 2015 (ZS-2015-0163). Associated RESRAD summary files were submitted with TSD 14-010 and provided to the NRC on March 30, 2015 (ZS-2015-0051).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 37 of 156
- 13. Comment: The use of background in calculating dose for each survey unit to determine whether to analyze for HTD radionuclides should be clarified.
Basis: The LTP (see Section 6.8.2, Section 6.12.1, and Section 5.1) states that if levels of residual radioactivity are encountered in a survey unit for open land, buried piping, or structures that exceeds 10% of the 25 mrem/yr Dose Criterion (2.5 mrem/yr), then samples will be analyzed for HTD radionuclides. The LTP does not provide details on how the dose will be calculated and how background radioactivity is treated in the dose calculation (e.g., whether background is subtracted and what that background value is assumed to be for the various survey units).
Path Forward:
- a. Provide an example of how the dose will be calculated to determine whether to test for HTD radionuclides in each type of survey unit or media (e.g., structures, open land, piping). If background is subtracted, provide a technical basis for the background values that will be applied.
ZSRP Response (13a) - The investigation levels cited in section 5.1, section 6.8.2 and section 6.12.1 pertain specifically to continuing characterization only. ZSRP does not propose to analyze for HTD radionuclides during the performance of FRS (STS or FSS). ZSRP will infer the presence of Ni-63 and Sr-90 using the surrogate approach specified in section 5.2.4. H-3 will be added as an additional HTD radionuclide that will be inferred using the surrogate approach for the STS of the Containment basements. The scaling factors that will be used are presented in Table 5-2 (and Table 6-3 for the Containment basements).
Also, as a clarification, the last paragraph of section 5.1 has been deleted and replaced with the following text. This was done to specify that the action levels cited for the assessment of HTD radionuclides is specific to continuing characterization surveys and do not apply to FRS.
Sufficient characterization samples have been taken from the Containment structures, Auxiliary Building, Turbine Building and Crib House/Forebay concrete to derive the radionuclide mixture and assess the dose impact of HTD radionuclides. The only remaining end-state concrete structure that has not been fully characterized to date is the concrete under the steel liner of the SFP/Transfer Canal. When the underlying concrete of the SFP/Transfer Canal is exposed by the removal of the steel liner, the concrete will be characterized in the same manner as the other end-state concrete structures (surfaces will be scanned and concrete core samples will be taken at the locations of the highest scan results). Continuing characterization will also be performed in several potentially contaminated embedded pipe systems that will be abandoned in place, specifically the floor drains in the 542 foot elevation basement floor of the Auxiliary Building and the Core Spray penetrations between the Containment basements and the Auxiliary Building basement. When the interior surfaces of these pipes systems become accessible, samples will be taken of any loose surface debris in the pipe. In both of these cases, the concrete core and/or
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 38 of 156 debris samples will be analyzed for the presence of HTD radionuclides. If the sample analysis indicates the presence of a significantly different radionuclide mixture from the mixture derived for the Auxiliary Building floor (Table 5-2), then the unique mixture will be documented and applied to the SFP/Transfer canal and/or embedded pipe systems as applicable during survey design and STS.
If a sample and/or measurement is taken on any other end-state structure or embedded pipe system to support decommissioning activities, Radiological Assessments (RA) or Remedial Action Support Surveys (RASS), and the result indicates a SOF in excess of 0.5 based on gamma spectroscopy results, then a sample will be collected at the location of the highest accessible individual measurement and analyzed for HTD radionuclides. If any continuing characterization surveys taken in soil or buried pipe indicate the presence of gamma-emitting radionuclides at concentrations in excess of a SOF of 0.5, then the samples will be analyzed for the presence of HTD radionuclides. In these unlikely situations, if the analysis indicates the presence of HTD radionuclides (other than Ni-63 and Sr-90, which are known to be present) at detectable concentrations, then additional investigation/sampling will be performed.
Based upon the analysis of radionuclide fractions and dose contribution in TSD 14-019, the dose contribution from HTD fractions is expected to be very low in all media (concrete, soil, embedded pipe, buried pipe, penetrations) at a SOF of 0.5 with even the most extreme HTD ratios. In the unlikely situation where these investigation levels are exceeded and one or more HTD radionuclides other than Ni-63 or Sr-90 are positively identified, then the dose impact of the positive HTD radionuclide(s) will be assessed. Additional samples may be collected and analyzed for HTD radionuclides to support the assessment of the dose impact.
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- 14. Comment: The methodology that will be used to calculate the compliance dose is not clear.
Basis: Equation 6-5 describes how the compliance dose will calculated from the various source terms at the ZNPS. However, it is not clear how the dose from the source terms will be calculated from the final site FSS data. In particular, it is not clear how the max soil and max buried pipe doses will be calculated from the FSS data and the DCGL values for soil and piping.
It is also not clear how the dose from elevated areas will be accounted for in the soil dose.
Path Forward:
- a. Provide additional details regarding the methodology that will be used to calculate the compliance dose, including the equations that will be used to calculate the maximum soil dose and the maximum piping dose from the FSS data.
ZSRP Response (14a) - Demonstrating compliance with the dose based release criterion will be accomplished for each FRS (STS or FSS) survey unit.
For FSS (applicable to soil and buried pipe), compliance with the dose based release criterion is demonstrated for each FSS survey unit if the survey unit passes the Sign Test as stated in section 5.10.3.2, if the Elevated Measurement Comparison (if applicable) as stated in section 5.10.4 (Equation 5-13) is less than unity and, if the SOF for the mean activity for each ROC is less than unity as stated in section 5.10.3.1. The process for demonstrating compliance for a STS survey unit is described in section 5.5.4. For STS (applicable to basements, penetrations and embedded pipe), the Sign Test will be also be used to demonstrate compliance with the dose based release criterion. If the Sign Test passes and the Mean Inventory Fraction is less than unity, then the survey unit will pass STS. If the Sign Test fails, or if the Mean Inventory Fraction exceeds one, then the survey unit will fail STS. If a basement has multiple STS survey units (i.e. floors, walls, embedded pipe/penetrations), then each survey unit must pass the Sign Test and the sum of the Mean Inventory Fractions for each must also be less than unity.
To ensure conservatism and consistency with the bounding, screening approach to dose modeling used by ZSRP, the maximum dose from all media (basements, soil, buried pipe, and groundwater) will be summed, notwithstanding the fact that it is very unlikely that the Average Member of the Critical Group (AMCG) would be exposed to residual radioactivity in all media simultaneously. The dose from the maximum individual survey unit for each media will be used in the compliance calculation regardless of the physical location of the survey unit or time of maximum exposure. The compliance dose calculation is performed using Equation 6-5 (which is equivalent to Equation 5-3 in LTP Chapter 5) which is reproduced below.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 40 of 156 Compliance Dose = Max BFM + Max Soil + Max Buried + Max GW where:
Compliance Dose = Dose to Resident Farmer Critical Group (mrem/yr)
Max BFM = Maximum dose from Basements (mrem/yr), including dose from embedded piping and penetrations Max Soil = Maximum dose from open land survey units (mrem/yr)
Max Buried = Maximum dose from buried piping (mrem/yr)
Max GW = Maximum dose from radionuclides identified in existing groundwater (none expected)
After it has been demonstrated that each FRS survey unit passes the Sign Test and has a mean SOF less than one, then the dose attributable to the survey unit is calculated in accordance with ZionSolutions procedure ZS-LT-300-001-004, Final Radiation Survey Data Assessment provided in Enclosure 2. The maximum dose for all survey units for a given media is identified and used as an input to Equation 6-5. For example, the FSS results for all soil survey units will be compared and the maximum dose value of all soil survey units selected and used for the Max Soil term in Equation 6-5. The same process will be followed for the basements, buried pipe and existing groundwater. The maximum doses identified for all media will be summed as indicated in Equation 6-5. This dose must be less than the 25 mrem/yr dose criterion.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 41 of 156
- 15. Comment: Information on the process for determining the total basement inventory from FSS data is needed.
Basis: In the LTP, the licensee proposes to calculate the dose from the residual radioactivity in the basements from the basement dose factors provided in Table 6-18. These basement dose factors are used to calculate the dose per total inventory in the each basement. The determination of the total inventory is therefore an important factor in the determination of the projected dose. However, details on how the FSS data will be used to generate a total inventory of residual contamination is not provided in the LTP. The specific equations that will be used to determine the total inventory were not provided. It is not clear how the licensee intends to use individual data points to develop the total inventory. For example, it is not clear how each data point will be factored into the total inventory. Additionally, it is not clear how elevated areas and volumetric contamination will be accounted for. It is also not clear how the inventory will be developed for areas that are below the detection limit.
Path Forward:
- a. Describe how the inventory for each basement will be developed from FSS data, and provide a sample calculation, including the use of any scaling factors.
- b. Include a description of how each data point will be factored into the total inventory and the manner in which elevated areas and volumetric contamination will be accounted for.
- c. Also, describe how the inventory for areas that are below the detection limits will be assessed.
ZSRP Response (15a, b and c) - The process that will be used to assess STS data and demonstrate compliance with the unrestricted release criteria has been formalized in ZionSolutions procedure ZS-LT-300-001-004, Final Radiation Survey Data Assessment.
ZionSolutions TSD 14-022, Use of In-situ Gamma Spectroscopy for Source Term Survey of End State Structures describes the ISOCS geometry that will be used for STS, the capability of the selected geometry to detect ROC activity at depth, and the sensitivity to non-uniform areal distribution. As described in detail in ZionSolutions TSD 14-022, (submitted to the NRC on May 27, 2015 (ZS-2015-0078), the activity in any elevated areas contained in an ISOCS measurement Field-of-View (FOV) will be conservatively accounted for by using the efficiency factors determined based on the uniform contamination assumptions described in the TSD. No additional consideration of elevated areas is necessary since total activity is the only data required to determine dose in the BFM. Also, the efficiency calculations for ISOCS STS measurements conservatively account for contamination at depth through a detailed evaluation of the activity depth profile in concrete core samples (also described in TSD 14-022).
ZSRP will demonstrate compliance with the dose-based release criteria using BFM Dose Factors, and corresponding BILs, by assessing total inventory in each structure through the survey/measurement of a reasonable and risk-informed areal coverage based on a graded
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 42 of 156 approach. However, instead of basing the areal coverage on the expected fraction of the DCGL, the coverage is based on the expected fraction of the BIL, which represents the activity inventory corresponding to the 25 mrem/yr dose criteria in the BFM. Note that the BIL is calculated by inverting the BFM Dose Factors (with units of mrem/yr per mCi) and multiplying by 25 mrem/yr.
As clarification, the following paragraphs will be added to section 5.5.4:
After a sufficient number of ISOCS measurements are taken in a STS unit in accordance with the areal coverage requirements specified in Table 5-11, the data will be summarized, including any judgmental or investigation measurements. The measured activity for each gamma-emitting ROC (and any other gamma radionuclide identified at levels greater than the ISOCs MDC) will be recorded (in units of pCi/m2). Background will not be subtracted from any measurement.
Using the radionuclide mixture fractions applicable to the survey unit, an inferred activity will be derived for each applicable HTD ROC. ZSRP will infer the presence of Ni-63 and Sr-90 using the surrogate approach specified in section 5.2.4. For the STS of the Containment basements, H-3 will be added as an additional HTD radionuclide that will be inferred. The scaling factors that will be used are presented in Table 5-2 (and Table 6-3 for the Containment basements). A sum of fractions (SOF) calculation will be performed for each measurement by dividing the reported concentration by the applicable BIL for each ROC, after converting the BIL to the same units as the ISOCS measurement (pCi/m2). The individual ROC fractions will then be summed to provide a total SOF value for the measurement.
As described in section 5.10.3.2, the Sign Test will be used to evaluate the remaining residual radioactivity in each survey unit against the dose criterion. The SOF for each measurement will be used as the weighted sum for the Sign Test. If the Sign Test demonstrates that the mean activity of the survey unit is less than the BIL at a 5% Type 1 error rate, then the mean of all the total SOFs for each measurement in a given survey unit (designated as the Mean Inventory Fraction) is calculated.
If the Sign Test fails, or if the Mean Inventory Fraction in a basement exceeds one, then the survey unit will fail STS. If a survey unit fails STS, then the STS survey unit may be reclassified, additional remediation will be performed and the STS performed again.
In situations where there are multiple survey units in a STS basement (e.g. the Auxiliary Building will have three STS units, one for the walls, one for the floor and one for any embedded piping and/or penetrations that may remain in the end-state), the fraction for each ROC will be calculated by dividing the reported ISOCS activity level (pCi/m2) by an allocated fraction of the applicable BIL that is selected for each survey unit apriori to the performance of STS. Note that the sum of the allocated BIL fractions for each survey unit applicable to a given basement must equal one. The upper bound of the gray region for the Sign Test will be the allocated fraction of the BIL for each STS survey unit in the basement, as opposed to the full BIL in the case of a basement that has only one STS survey unit.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 43 of 156 For example, if a basement has two STS units, and 20% of the BIL is allocated to the first survey unit and 80% of the BIL allocated to the second survey unit, then the SOF for the first survey unit will be calculated using 20% of the BIL value and the SOF in the second survey unit calculated using 80% of the BIL value. The SOF for each measurement will be calculated using the allocated fraction, as opposed to the full BIL, and the resulting fraction used as the Weighted Sum (Ws) when performing the Sign Test.
During the data assessment, if the actual mean inventory in a STS unit exceeds the allocated fraction, or the survey unit fails the Sign Test, then the allocated fraction may be increased as long as the allocated fractions in the other STS units in the basement are decreased accordingly such that the revised fractions still sum to one. However, the Sign Test must be passed for all survey units in a given basemen at the final allocated fraction selected. After all survey units are demonstrated to pass the Sign Test, the Mean Inventory Fractions for each of the survey units are summed.
If a combination of allocated fractions cannot be selected that results in all survey units in a given basement passing the Sign Test, or if the sum of the Mean Inventory Fractions of all STS units in a basement exceeds one, then the survey unit will fail STS. If a survey unit fails STS, then additional remediation will be performed and the STS performed again.
A sample calculation illustrating compliance in a single STS unit is as follows:
The STS survey unit for the Turbine Building will be used for the first example calculation.
Table 5-11 specifies that 14 ISOCS measurements will be taken in this Class 3 STS survey unit, resulting in an areal coverage of approximately 392 m2 or 3% of the total surface area in the basement. If the gamma-emitting ROC is not detected by the measurement, then the MDC for that ROC will be used for the calculation of Mean Inventory Fraction. This example assumes that 14 measurements were obtained at random locations resulting in the following survey population in units of pCi/m2. Please note that these are hypothetical concentrations and do not reflect the actual concentrations of residual radioactivity in the Turbine Building concrete:
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 44 of 156 Co-60 Cs-134 Cs-137 (pCi/m2) (pCi/m2) (pCi/m2)
Measurement #01 1.14 E-01 2.33 E-02 1.27 E-01 Measurement #01 9.62 E-02 1.88 E-02 2.99 E-01 Measurement #03 1.00 E-01 9.87 E-03 1.08 E+00 Measurement #04 3.68 E+00 1.28 E-01 1.27 E-01 Measurement #05 2.11 E-01 1.52 E-01 1.76 E-01 Measurement #06 1.35 E+01 8.22 E+00 3.97 E+01 Measurement #07 5.11 E+01 5.33 E+01 7.85 E+02 Measurement #08 2.22 E+00 8.44 E-01 7.16 E+01 Measurement #09 9.45 E-01 1.07 E-02 1.18 E+00 Measurement #10 3.33 E+01 9.78 E+01 6.54 E+02 Measurement #11 7.14 E-01 8.38 E-02 2.27 E-01 Measurement #12 8.23 E-01 5.78 E-01 1.29 E+00 Measurement #13 5.16 E-01 3.66 E-01 1.55 E+00 Measurement #14 5.14 E+01 4.33 E+00 2.98 E+02 These measurements are then assessed in a Preliminary Data Assessment as follows:
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use STS Unit B306100AF Description Turbine Building Basement Classification STS Class 3 Measured Inferred Activity Total Location FOV BIL/Area Ratio to Activity Activity per ROC Fraction Activity Sample ID ROC SOF Cs-137 of BIL X-Axis Y-Axis (m2) (pCi/m2) (pCi/m2) (pCi/m2) (pCi/m2) (pCi/m2)
Co-60 1.14E-01 1.14E-01 1.33E+08 8.57E-10 Cs-134 2.33E-02 2.33E-02 3.02E+07 7.72E-10 Measurement #01 Example Example 28 Cs-137 1.27E-01 1.27E-01 3.98E+07 3.19E-09 4.89E-09 3.04E-01 Ni-63 0.315 4.00E-02 4.00E-02 4.13E+09 9.68E-12 Sr-90 0.001 8.43E-05 8.43E-05 1.46E+06 5.77E-11 Co-60 9.62E-02 9.62E-02 1.33E+08 7.23E-10 Cs-134 1.88E-02 1.88E-02 3.02E+07 6.23E-10 Measurement #02 Example Example 28 Cs-137 2.99E-01 2.99E-01 3.98E+07 7.51E-09 9.02E-09 5.08E-01 Ni-63 0.315 9.41E-02 9.41E-02 4.13E+09 2.28E-11 Sr-90 0.001 1.98E-04 1.98E-04 1.46E+06 1.36E-10 Co-60 1.00E-01 1.00E-01 1.33E+08 7.52E-10 Cs-134 9.87E-03 9.87E-03 3.02E+07 3.27E-10 Measurement #03 Example Example 28 Cs-137 1.08E+00 1.08E+00 3.98E+07 2.71E-08 2.88E-08 1.53E+00 Ni-63 0.315 3.40E-01 3.40E-01 4.13E+09 8.23E-11 Sr-90 0.001 7.17E-04 7.17E-04 1.46E+06 4.91E-10 Co-60 3.68E+00 3.68E+00 1.33E+08 2.77E-08 Cs-134 1.28E-01 1.28E-01 3.02E+07 4.24E-09 Measurement #04 Example Example 28 Cs-137 1.27E-01 1.27E-01 3.98E+07 3.19E-09 3.52E-08 3.98E+00 Ni-63 0.315 4.00E-02 4.00E-02 4.13E+09 9.68E-12 Sr-90 0.001 8.43E-05 8.43E-05 1.46E+06 5.77E-11 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 45 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Measured Inferred Activity Total Location FOV BIL/Area Ratio to Activity Activity per ROC Fraction Activity Sample ID ROC SOF Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 2.11E-01 2.11E-01 1.33E+08 1.59E-09 Cs-134 1.52E-01 1.52E-01 3.02E+07 5.03E-09 Measurement #05 Example Example 28 Cs-137 1.76E-01 1.76E-01 3.98E+07 4.42E-09 1.11E-08 5.95E-01 Ni-63 0.315 5.54E-02 5.54E-02 4.13E+09 1.34E-11 Sr-90 0.001 1.17E-04 1.17E-04 1.46E+06 8.00E-11 Co-60 1.35E+01 1.35E+01 1.33E+08 1.02E-07 Cs-134 8.22E+00 8.22E+00 3.02E+07 2.72E-07 Measurement #06 Example Example 28 Cs-137 3.97E+01 3.97E+01 3.98E+07 9.97E-07 1.39E-06 7.39E+01 Ni-63 0.315 1.25E+01 1.25E+01 4.13E+09 3.03E-09 Sr-90 0.001 2.64E-02 2.64E-02 1.46E+06 1.81E-08 Co-60 5.11E+01 5.11E+01 1.33E+08 3.84E-07 Cs-134 5.33E+01 5.33E+01 3.02E+07 1.76E-06 Measurement #07 Example Example 28 Cs-137 7.85E+02 7.85E+02 3.98E+07 1.97E-05 2.23E-05 1.14E+03 Ni-63 0.315 2.47E+02 2.47E+02 4.13E+09 5.98E-08 Sr-90 0.001 5.21E-01 5.21E-01 1.46E+06 3.57E-07 Co-60 2.22E+00 2.22E+00 1.33E+08 1.67E-08 Cs-134 8.44E-01 8.44E-01 3.02E+07 2.79E-08 Measurement #08 Example Example 28 Cs-137 7.16E+01 7.16E+01 3.98E+07 1.80E-06 1.88E-06 9.72E+01 Ni-63 0.315 2.25E+01 2.25E+01 4.13E+09 5.46E-09 Sr-90 0.001 4.75E-02 4.75E-02 1.46E+06 3.26E-08 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 46 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Measured Inferred Activity Total Location FOV BIL/Area Ratio to Activity Activity per ROC Fraction Activity Sample ID ROC SOF Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 9.45E-01 9.45E-01 1.33E+08 7.11E-09 Cs-134 1.07E-02 1.07E-02 3.02E+07 3.54E-10 Measurement #09 Example Example 28 Cs-137 1.18E+00 1.18E+00 3.98E+07 2.96E-08 3.77E-08 2.51E+00 Ni-63 0.315 3.71E-01 3.71E-01 4.13E+09 8.99E-11 Sr-90 0.001 7.83E-04 7.83E-04 1.46E+06 5.37E-10 Co-60 3.33E+01 3.33E+01 1.33E+08 2.50E-07 Cs-134 9.78E+01 9.78E+01 3.02E+07 3.24E-06 Measurement #10 Example Example 28 Cs-137 6.54E+02 6.54E+02 3.98E+07 1.64E-05 2.03E-05 9.91E+02 Ni-63 0.315 2.06E+02 2.06E+02 4.13E+09 4.98E-08 Sr-90 0.001 4.34E-01 4.34E-01 1.46E+06 2.97E-07 Co-60 7.14E-01 7.14E-01 1.33E+08 5.37E-09 Cs-134 8.38E-02 8.38E-02 3.02E+07 2.77E-09 Measurement #11 Example Example 28 Cs-137 2.27E-01 2.27E-01 3.98E+07 5.70E-09 1.40E-08 1.10E+00 Ni-63 0.315 7.14E-02 7.14E-02 4.13E+09 1.73E-11 Sr-90 0.001 1.51E-04 1.51E-04 1.46E+06 1.03E-10 Co-60 8.23E-01 8.23E-01 1.33E+08 6.19E-09 Cs-134 5.78E-01 5.78E-01 3.02E+07 1.91E-08 Measurement #12 Example Example 28 Cs-137 1.29E+00 1.29E+00 3.98E+07 3.24E-08 5.84E-08 3.10E+00 Ni-63 0.315 4.06E-01 4.06E-01 4.13E+09 9.83E-11 Sr-90 0.001 8.56E-04 8.56E-04 1.46E+06 5.87E-10 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 47 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Measured Inferred Activity Total Location FOV BIL/Area Ratio to Activity Activity per ROC Fraction Activity Sample ID ROC SOF Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 5.16E-01 5.16E-01 1.33E+08 3.88E-09 Cs-134 3.66E-01 3.66E-01 3.02E+07 1.21E-08 Measurement #13 Example Example 28 Cs-137 1.55E+00 1.55E+00 3.98E+07 3.89E-08 5.58E-08 2.92E+00 Ni-63 0.315 4.88E-01 4.88E-01 4.13E+09 1.18E-10 Sr-90 0.001 1.03E-03 1.03E-03 1.46E+06 7.05E-10 Co-60 5.14E+01 5.14E+01 1.33E+08 3.86E-07 Cs-134 4.33E+00 4.33E+00 3.02E+07 1.43E-07 Measurement #14 Example Example 28 Cs-137 2.98E+02 2.98E+02 3.98E+07 7.49E-06 8.18E-06 4.48E+02 Ni-63 0.315 9.38E+01 9.38E+01 4.13E+09 2.27E-08 Sr-90 0.001 1.98E-01 1.98E-01 1.46E+06 1.35E-07 Number of Measurements Taken in Survey Unit 14 Corresponding Areal Coverage based on FOV 392 m2 Maximum SOF Observed 2.23E-05 Mean of Total Activity Results 1.97E+02 pCi/m2 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 48 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Calculation of Mean Activity and Dose for Turbine Building Basement STS Total Total Mean Basement Mean Mean Dose Mean Dose Mean Activity Dose Activity per ROC for STS Inventory per ROC Factors ROC per ROC Unit Fraction for STS (mrem/yr Unit (pCi/m2) (mCi) (mrem/yr) (mrem/yr) per mCi)
Co-60 1.13E+01 1.66E-04 1.26E-02 2.10E-06 Cs-134 1.18E+01 1.74E-04 5.56E-02 9.67E-06 Cs-137 1.32E+02 1.94E-03 4.21E-02 8.19E-05 9.54E-05 3.81E-06 Ni-63 4.17E+01 6.12E-04 4.06E-04 2.48E-07 Sr-90 8.79E-02 1.29E-06 1.15E+00 1.48E-06 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 49 of 156
ZS-LT-300-001-004 Attachment 14 Revision 1 Sign Statistical Test Information Use Survey Area: No. 06100
Description:
Turbine Building Basement Survey Unit: No. 06100
Description:
Turbine Building Basement Classification: 3 Type I ()Error: 0.5 Number of Samples (n): 14 Weighted
- 1-Ws Sign Sum (Ws) 1 4.89E-09 1.00 +1 2 9.02E-09 1.00 +1 3 2.88E-08 1.00 +1 4 3.52E-08 1.00 +1 5 1.11E-08 1.00 +1 6 1.39E-06 1.00 +1 7 2.23E-05 1.00 +1 8 1.88E-06 1.00 +1 9 3.77E-08 1.00 +1 10 2.03E-05 1.00 +1 11 1.40E-08 1.00 +1 12 6.19E-09 1.00 +1 13 5.58E-08 1.00 +1 14 8.18E-06 1.00 +1 Critical Value (Table I.3 of MARSSIM) = 10 Number of Positive Differences (S+) = 14 The survey unit (meets) (does not meet) the acceptance criteria Prepared by (RE):
(Print Name) (Signature) (Date)
Peer Reviewed by (RE):
(Print Name) (Signature) (Date)
As this survey unit passes the Sign Test, and the Mean Inventory Fraction is less than one and, there are no other STS units in the Turbine Building basement, this survey unit would pass STS.
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ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 51 of 156 A sample calculation illustrating compliance in a basement with multiple STS units is as follows:
For this example, the Auxiliary Building walls and the Auxiliary Building floor will be used as two STS units in an end-state basement. Again, please note that these are hypothetical concentrations and do not reflect the actual residual concentration of radioactivity in the Auxiliary Building wall or floor concrete. Table 5-11 specifies that 100% areal coverage is required for the Auxiliary Building floor that, assuming a FOV of 28 m2, would equate to 93 ISOCS measurements assuming no FOV overlap. For the purposes of this example, it is assumed that 15 ISOCS measurements will provide 100% areal coverage. For the Auxiliary Building walls, Table 5-11 specifies a Class 2 areal coverage requirement of 10% adjusted to 14 ISOCS measurements, resulting in an areal coverage of approximately 392 m2 or 10% of the total surface area of the walls. The expected fraction of the inventory in the walls is assumed to be 20% of the total inventory in the Auxiliary Building basement and the expected fraction of the inventory in the floor is assumed to be 80% of the total inventory in the Auxiliary Building basement.
This example assumes that the following 14 measurements were obtained at random locations on the Auxiliary Building walls, resulting in the following survey population in units of pCi/m2; Co-60 Cs-134 Cs-137 (pCi/m2) (pCi/m2) (pCi/m2)
Measurement #W01 2.83 E+01 5.47 E+00 2.87 E+02 Measurement #W01 1.75 E+00 3.33 E+00 8.88 E+03 Measurement #W03 6.84 E+02 7.77 E+01 1.10 E+05 Measurement #W04 2.87 E+00 2.48 E+00 2.10 E+01 Measurement #W05 3.79 E+01 8.65 E+01 2.87 E+04 Measurement #W06 2.11 E+01 2.11 E+01 2.65 E+03 Measurement #W07 1.02 E+01 1.00 E+01 2.97 E+02 Measurement #W08 9.71 E+00 1.29 E+01 9.53 E+03 Measurement #W09 1.00 E+01 1.87 E+00 3.33 E+02 Measurement #W10 2.87 E+01 1.97 E+01 1.02 E+02 Measurement #W11 6.46 E+01 6.52 E+00 1.00 E+03 Measurement #W12 5.77 E+01 2.12 E+01 9.00 E+04 Measurement #W13 5.55 E+01 1.11 E+01 7.46 E+04 Measurement #W14 2.11 E+01 4.44 E+00 4.46 E+02
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 52 of 156 This example also assumes that the following 15 measurements were obtained at random locations on the Auxiliary Building floor, resulting in the following survey population in units of pCi/m2; Co-60 Cs-134 Cs-137 (pCi/m2) (pCi/m2) (pCi/m2)
Measurement #F01 3.25 E+02 9.87 E+01 1.27 E+04 Measurement #F01 3.88 E+02 6.54 E+01 2.99 E+03 Measurement #F03 2.45 E+03 3.21 E+02 1.08 E+03 Measurement #F04 6.67 E+02 1.36 E+01 1.27 E+03 Measurement #F05 8.88 E+02 2.45 E+01 1.76 E+05 Measurement #F06 1.40 E+01 3.78 E+01 3.97 E+04 Measurement #F07 1.89 E+03 4.96 E+04 7.85 E+06 Measurement #F08 1.87 E+05 7.86 E+03 7.16 E+06 Measurement #F09 6.58 E+01 1.02 E+01 1.18 E+01 Measurement #F10 7.12 E+04 1.36 E+02 6.54 E+05 Measurement #F11 2.20 E+02 1.78 E+02 2.27 E+06 Measurement #F12 7.12 E+03 6.66 E+01 1.29 E+05 Measurement #F13 3.21 E+02 1.36 E+01 1.55 E+07 Measurement #F14 4.56 E+04 5.00 E+03 2.98 E+07 Measurement #F15 7.89 E+03 5.87 E+02 2.98 E+06 These measurements are then assessed in a following Preliminary Data Assessments. The BIL for the Auxiliary Building for each ROC is as follows; Auxiliary Nuclide (mCi)
Co-60 2.28E+03 Cs-134 1.59E+03 Cs-137 8.35E+02 Ni-63 8.64E+04 Sr-90 7.50E+01 The BIL values are converted to units of pCi/m2 by converting mCi to pCi and dividing by the surface area in the STS survey unit. The subsequent BIL/Area value for each ROC for the Auxiliary Building walls is derived from 20% of the total inventory (or BIL) value and the BIL/Area value for each ROC for the Auxiliary Building floor is derived from 80% of the total inventory (or BIL) value. The resulting SOF for each measurement calculated using the expected fraction of the inventory in each STS unit was then used to execute the Sign Test. This provides assurance that the mean fraction for each STS unit is less than the expected inventory fraction at a 95% confidence level.
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use STS Unit B205119W Description Auxiliary Building Basement Walls Classification STS Class 2 Measured Inferred Activity Total Location FOV Ratio to BIL/Area Fraction Sample ID ROC Activity Activity per ROC SOF Activity Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 2.83E+01 2.83E+01 1.17E+08 2.43E-07 Cs-134 5.47E+00 5.47E+00 8.13E+07 6.73E-08 Measurement #W01 Example Example 28 Cs-137 2.87E+02 2.87E+02 4.27E+07 6.72E-06 7.10E-06 4.11E+02 Ni-63 0.315 9.03E+01 9.03E+01 4.42E+09 2.04E-08 Sr-90 0.001 1.91E-01 1.91E-01 3.83E+06 4.97E-08 Co-60 1.75E+00 1.75E+00 1.17E+08 1.50E-08 Cs-134 3.33E+00 3.33E+00 8.13E+07 4.10E-08 Measurement #W02 Example Example 28 Cs-137 8.88E+03 8.88E+03 4.27E+07 2.08E-04 2.10E-04 1.17E+04 Ni-63 0.315 2.79E+03 2.79E+03 4.42E+09 6.33E-07 Sr-90 0.001 5.89E+00 5.89E+00 3.83E+06 1.54E-06 Co-60 6.84E+02 6.84E+02 1.17E+08 5.87E-06 Cs-134 7.77E+01 7.77E+01 8.13E+07 9.56E-07 Measurement #W03 Example Example 28 Cs-137 1.10E+05 1.10E+05 4.27E+07 2.58E-03 2.61E-03 1.45E+05 Ni-63 0.315 3.46E+04 3.46E+04 4.42E+09 7.84E-06 Sr-90 0.001 7.30E+01 7.30E+01 3.83E+06 1.90E-05 Co-60 2.87E+00 2.87E+00 1.17E+08 2.46E-08 Cs-134 2.48E+00 2.48E+00 8.13E+07 3.05E-08 Measurement #W04 Example Example 28 Cs-137 2.10E+01 2.10E+01 4.27E+07 4.92E-07 5.52E-07 3.30E+01 Ni-63 0.315 6.61E+00 6.61E+00 4.42E+09 1.50E-09 Sr-90 0.001 1.39E-02 1.39E-02 3.83E+06 3.64E-09 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 53 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Measured Inferred Activity Total Location FOV Ratio to BIL/Area Fraction Activity Activity per ROC Activity Sample ID ROC SOF Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 3.79E+01 3.79E+01 1.17E+08 3.25E-07 Cs-134 8.65E+01 8.65E+01 8.13E+07 1.06E-06 Measurement #W05 Example Example 28 Cs-137 2.87E+04 2.87E+04 4.27E+07 6.72E-04 6.81E-04 3.79E+04 Ni-63 0.315 9.03E+03 9.03E+03 4.42E+09 2.04E-06 Sr-90 0.001 1.91E+01 1.91E+01 3.83E+06 4.97E-06 Co-60 2.11E+01 2.11E+01 1.17E+08 1.81E-07 Cs-134 2.11E+01 2.11E+01 8.13E+07 2.60E-07 Measurement #W06 Example Example 28 Cs-137 2.65E+02 2.65E+02 4.27E+07 6.21E-06 6.71E-06 3.91E+02 Ni-63 0.315 8.34E+01 8.34E+01 4.42E+09 1.89E-08 Sr-90 0.001 1.76E-01 1.76E-01 3.83E+06 4.59E-08 Co-60 1.02E+01 1.02E+01 1.17E+08 8.75E-08 Cs-134 1.00E+01 1.00E+01 8.13E+07 1.23E-07 Measurement #W07 Example Example 28 Cs-137 2.97E+02 2.97E+02 4.27E+07 6.96E-06 7.24E-06 4.11E+02 Ni-63 0.315 9.35E+01 9.35E+01 4.42E+09 2.12E-08 Sr-90 0.001 1.97E-01 1.97E-01 3.83E+06 5.14E-08 Co-60 9.71E+00 9.71E+00 1.17E+08 8.33E-08 Cs-134 1.29E+01 1.29E+01 8.13E+07 1.59E-07 Measurement #W08 Example Example 28 Cs-137 9.53E+03 9.53E+03 4.27E+07 2.23E-04 2.26E-04 1.26E+04 Ni-63 0.315 3.00E+03 3.00E+03 4.42E+09 6.79E-07 Sr-90 0.001 6.33E+00 6.33E+00 3.83E+06 1.65E-06 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 54 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Measured Inferred Activity Total Location FOV Ratio to BIL/Area Fraction Activity Activity per ROC Activity Sample ID ROC SOF Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 1.00E+01 1.00E+01 1.17E+08 8.58E-08 Cs-134 1.87E+00 1.87E+00 8.13E+07 2.30E-08 Measurement #W09 Example Example 28 Cs-137 3.33E+02 3.33E+02 4.27E+07 7.80E-06 7.99E-06 4.50E+02 Ni-63 0.315 1.05E+02 1.05E+02 4.42E+09 2.37E-08 Sr-90 0.001 2.21E-01 2.21E-01 3.83E+06 5.77E-08 Co-60 2.87E+01 2.87E+01 1.17E+08 2.46E-07 Cs-134 1.97E+01 1.97E+01 8.13E+07 2.42E-07 Measurement #W10 Example Example 28 Cs-137 1.02E+02 1.02E+02 4.27E+07 2.39E-06 2.90E-06 1.83E+02 Ni-63 0.315 3.21E+01 3.21E+01 4.42E+09 7.27E-09 Sr-90 0.001 6.77E-02 6.77E-02 3.83E+06 1.77E-08 Co-60 6.46E+01 6.46E+01 1.17E+08 5.54E-07 Cs-134 6.52E+00 6.52E+00 8.13E+07 8.02E-08 Measurement #W11 Example Example 28 Cs-137 1.00E+03 1.00E+03 4.27E+07 2.34E-05 2.43E-05 1.39E+03 Ni-63 0.315 3.15E+02 3.15E+02 4.42E+09 7.12E-08 Sr-90 0.001 6.64E-01 6.64E-01 3.83E+06 1.73E-07 Co-60 5.77E+01 5.77E+01 1.17E+08 4.95E-07 Cs-134 2.12E+01 2.12E+01 8.13E+07 2.61E-07 Measurement #W12 Example Example 28 Cs-137 9.00E+04 9.00E+04 4.27E+07 2.11E-03 2.13E-03 1.18E+05 Ni-63 0.315 2.83E+04 2.83E+04 4.42E+09 6.41E-06 Sr-90 0.001 5.97E+01 5.97E+01 3.83E+06 1.56E-05 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 55 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Measured Inferred Activity Total Location FOV Ratio to BIL/Area Fraction Activity Activity per ROC Activity Sample ID ROC SOF Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 5.55E+01 5.55E+01 1.17E+08 4.76E-07 Cs-134 1.11E+01 1.11E+01 8.13E+07 1.37E-07 Measurement #W13 Example Example 28 Cs-137 7.46E+04 7.46E+04 4.27E+07 1.75E-03 1.77E-03 9.82E+04 Ni-63 0.315 2.35E+04 2.35E+04 4.42E+09 5.32E-06 Sr-90 0.001 4.95E+01 4.95E+01 3.83E+06 1.29E-05 Co-60 2.11E+01 2.11E+01 1.17E+08 1.81E-07 Cs-134 4.44E+00 4.44E+00 8.13E+07 5.46E-08 Measurement #W14 Example Example 28 Cs-137 4.46E+02 4.46E+02 4.27E+07 1.04E-05 1.08E-05 6.12E+02 Ni-63 0.315 1.40E+02 1.40E+02 4.42E+09 3.18E-08 Sr-90 0.001 2.96E-01 2.96E-01 3.83E+06 7.72E-08 Number of Measurements Taken in Survey Unit 14 Corresponding Areal Coverage based on FOV 392 m2 Maximum SOF Observed 2.61E-03 Mean of Total Activity Results 3.06E+04 pCi/m2 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 56 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Calculation of Mean Activity and Dose for Auxiliary Building Basement Walls Total Total Mean Basement Mean Mean Dose Mean Dose Mean Activity Dose Activity per ROC for STS Inventory per ROC Factors ROC per ROC Unit Fraction for STS (mrem/yr Unit (pCi/m2) (mCi) (mrem/yr) (mrem/yr) per mCi)
Co-60 7.38E+01 2.89E-04 1.10E-02 3.18E-06 Cs-134 2.03E+01 7.94E-05 1.57E-02 1.25E-06 Cs-137 2.32E+04 9.07E-02 3.00E-02 2.72E-03 2.75E-03 1.10E-04 Ni-63 7.29E+03 2.85E-02 2.89E-04 8.25E-06 Sr-90 1.54E+01 6.02E-05 3.33E-01 2.00E-05 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 57 of 156
ZS-LT-300-001-004 Attachment 14 Revision 1 Sign Statistical Test Information Use Survey Area: No. 05119
Description:
Auxiliary Building Basement Survey Unit: No. 05119
Description:
Auxiliary Building Basement Walls Classification: 2 Type I ()Error: 0.5 Number of Samples (n): 14 Weighted
- 1-Ws Sign Sum (Ws) 1 7.10E-06 1.00 +1 2 2.10E-04 1.00 +1 3 2.61E-03 1.00 +1 4 5.52E-07 1.00 +1 5 6.81E-04 1.00 +1 6 6.71E-06 1.00 +1 7 7.24E-06 1.00 +1 8 2.26E-04 1.00 +1 9 7.99E-06 1.00 +1 10 2.90E-06 1.00 +1 11 2.43E-05 1.00 +1 12 4.95E-07 1.00 +1 13 1.77E-03 1.00 +1 14 1.08E-05 1.00 +1 Critical Value (Table I.3 of MARSSIM) = 10 Number of Positive Differences (S+) = 14 The survey unit (meets) (does not meet) the acceptance criteria Prepared by (RE):
(Print Name) (Signature) (Date)
Peer Reviewed by (RE):
(Print Name) (Signature) (Date)
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 58 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use STS Unit B205100F Description Auxiliary Building Basement Floor Classification STS Class 1 Measured Inferred Activity Total Location FOV Ratio to BIL/Area Fraction Sample ID ROC Activity Activity per ROC SOF Activity Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 2.83E+01 2.83E+01 1.17E+08 2.43E-07 Cs-134 5.47E+00 5.47E+00 8.13E+07 6.73E-08 Measurement #F01 Example Example 28 Cs-137 2.87E+02 2.87E+02 4.27E+07 6.72E-06 7.10E-06 4.11E+02 Ni-63 0.315 9.03E+01 9.03E+01 4.42E+09 2.04E-08 Sr-90 0.001 1.91E-01 1.91E-01 3.83E+06 4.97E-08 Co-60 1.75E+00 1.75E+00 1.17E+08 1.50E-08 Cs-134 3.33E+00 3.33E+00 8.13E+07 4.10E-08 Measurement #F02 Example Example 28 Cs-137 8.88E+03 8.88E+03 4.27E+07 2.08E-04 2.10E-04 1.17E+04 Ni-63 0.315 2.79E+03 2.79E+03 4.42E+09 6.33E-07 Sr-90 0.001 5.89E+00 5.89E+00 3.83E+06 1.54E-06 Co-60 6.84E+02 6.84E+02 1.17E+08 5.87E-06 Cs-134 7.77E+01 7.77E+01 8.13E+07 9.56E-07 Measurement #F03 Example Example 28 Cs-137 1.10E+05 1.10E+05 4.27E+07 2.58E-03 2.61E-03 1.45E+05 Ni-63 0.315 3.46E+04 3.46E+04 4.42E+09 7.84E-06 Sr-90 0.001 7.30E+01 7.30E+01 3.83E+06 1.90E-05 Co-60 2.87E+00 2.87E+00 1.17E+08 2.46E-08 Cs-134 2.48E+00 2.48E+00 8.13E+07 3.05E-08 Measurement #F04 Example Example 28 Cs-137 2.10E+01 2.10E+01 4.27E+07 4.92E-07 5.52E-07 3.30E+01 Ni-63 0.315 6.61E+00 6.61E+00 4.42E+09 1.50E-09 Sr-90 0.001 1.39E-02 1.39E-02 3.83E+06 3.64E-09 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 59 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Measured Inferred Activity Total Location FOV Ratio to BIL/Area Fraction Sample ID ROC Activity Activity per ROC SOF Activity Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 8.88E+02 8.88E+02 7.04E+08 1.26E-06 Cs-134 2.45E+01 2.45E+01 4.91E+08 4.99E-08 Measurement #F05 Example Example 28 Cs-137 1.76E+05 1.76E+05 2.58E+08 6.83E-04 6.91E-04 2.32E+05 Ni-63 0.315 5.54E+04 5.54E+04 2.67E+10 2.08E-06 Sr-90 0.001 1.17E+02 1.17E+02 2.32E+07 5.05E-06 Co-60 1.40E+01 1.40E+01 7.04E+08 1.99E-08 Cs-134 3.78E+01 3.78E+01 4.91E+08 7.70E-08 Measurement #F06 Example Example 28 Cs-137 3.97E+04 3.97E+04 2.58E+08 1.54E-04 1.56E-04 5.23E+04 Ni-63 0.315 1.25E+04 1.25E+04 2.67E+10 4.68E-07 Sr-90 0.001 2.64E+01 2.64E+01 2.32E+07 1.14E-06 Co-60 1.89E+03 1.89E+03 7.04E+08 2.68E-06 Cs-134 4.96E+04 4.96E+04 4.91E+08 1.01E-04 Measurement #F07 Example Example 28 Cs-137 7.85E+06 7.85E+06 2.58E+08 3.04E-02 3.09E-02 1.04E+07 Ni-63 0.315 2.47E+06 2.47E+06 2.67E+10 9.26E-05 Sr-90 0.001 5.21E+03 5.21E+03 2.32E+07 2.25E-04 Co-60 1.87E+05 1.87E+05 7.04E+08 2.66E-04 Cs-134 7.86E+03 7.86E+03 4.91E+08 1.60E-05 Measurement #F08 Example Example 28 Cs-137 7.16E+06 7.16E+06 2.58E+08 2.78E-02 2.83E-02 9.61E+06 Ni-63 0.315 2.25E+06 2.25E+06 2.67E+10 8.45E-05 Sr-90 0.001 4.75E+03 4.75E+03 2.32E+07 2.05E-04 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 60 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Measured Inferred Activity Total Location FOV Ratio to BIL/Area Fraction Activity Activity per ROC Activity Sample ID ROC SOF Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 6.58E+01 6.58E+01 7.04E+08 9.35E-08 Cs-134 1.02E+01 1.02E+01 4.91E+08 2.08E-08 Measurement #F09 Example Example 28 Cs-137 1.18E+01 1.18E+01 2.58E+08 4.58E-08 1.60E-07 9.15E+01 Ni-63 0.315 3.71E+00 3.71E+00 2.67E+10 1.39E-10 Sr-90 0.001 7.83E-03 7.83E-03 2.32E+07 3.38E-10 Co-60 7.12E+04 7.12E+04 7.04E+08 1.01E-04 Cs-134 1.36E+02 1.36E+02 4.91E+08 2.77E-07 Measurement #F10 Example Example 28 Cs-137 6.54E+05 6.54E+05 2.58E+08 2.54E-03 2.66E-03 9.32E+05 Ni-63 0.315 2.06E+05 2.06E+05 2.67E+10 7.72E-06 Sr-90 0.001 4.34E+02 4.34E+02 2.32E+07 1.87E-05 Co-60 2.20E+02 2.20E+02 7.04E+08 3.13E-07 Cs-134 1.78E+02 1.78E+02 4.91E+08 3.63E-07 Measurement #F11 Example Example 28 Cs-137 2.27E+06 2.27E+06 2.58E+08 8.80E-03 8.90E-03 2.99E+06 Ni-63 0.315 7.14E+05 7.14E+05 2.67E+10 2.68E-05 Sr-90 0.001 1.51E+03 1.51E+03 2.32E+07 6.51E-05 Co-60 7.12E+03 7.12E+03 7.04E+08 1.01E-05 Cs-134 6.66E+01 6.66E+01 4.91E+08 1.36E-07 Measurement #F12 Example Example 28 Cs-137 1.29E+05 1.29E+05 2.58E+08 5.00E-04 5.16E-04 1.77E+05 Ni-63 0.315 4.06E+04 4.06E+04 2.67E+10 1.52E-06 Sr-90 0.001 8.56E+01 8.56E+01 2.32E+07 3.70E-06 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 61 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Measured Inferred Activity Total Location FOV Ratio to BIL/Area Fraction Activity Activity per ROC Activity Sample ID ROC SOF Cs-137 of BIL 2 2 2 2 2 X-Axis Y-Axis (m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m ) (pCi/m2)
Co-60 3.21E+02 3.21E+02 7.04E+08 4.56E-07 Cs-134 1.36E+01 1.36E+01 4.91E+08 2.77E-08 Measurement #F13 Example Example 28 Cs-137 1.55E+07 1.55E+07 2.58E+08 6.01E-02 6.07E-02 2.04E+07 Ni-63 0.315 4.88E+06 4.88E+06 2.67E+10 1.83E-04 Sr-90 0.001 1.03E+04 1.03E+04 2.32E+07 4.44E-04 Co-60 4.56E+04 4.56E+04 7.04E+08 6.48E-05 Cs-134 5.00E+03 5.00E+03 4.91E+08 1.02E-05 Measurement #F14 Example Example 28 Cs-137 2.98E+07 2.98E+07 2.58E+08 1.16E-01 1.17E-01 3.92E+07 Ni-63 0.315 9.38E+06 9.38E+06 2.67E+10 3.52E-04 Sr-90 0.001 1.98E+04 1.98E+04 2.32E+07 8.54E-04 Co-60 7.89E+03 7.89E+03 7.04E+08 1.12E-05 Cs-134 5.87E+02 5.87E+02 4.91E+08 1.20E-06 Measurement #F15 Example Example 28 Cs-137 2.98E+06 2.98E+06 2.58E+08 1.16E-02 1.17E-02 3.93E+06 Ni-63 0.315 9.38E+05 9.38E+05 2.67E+10 3.52E-05 Sr-90 0.001 1.98E+03 1.98E+03 2.32E+07 8.54E-05 Number of Measurements Taken in Survey Unit 15 Corresponding Areal Coverage based on FOV 420 m2 Maximum SOF Observed 1.17E-01 Mean of Total Activity Results 5.86E+06 pCi/m2 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 62 of 156
ZS-LT-300-001-004 Attachment 12 Revision 1 STS Preliminary Survey Data Summary Information Use Calculation of Mean Activity and Dose for Auxiliary Building Basement Floor Total Total Mean Basement Mean Mean Dose Mean Dose Mean Activity Dose Activity per ROC for STS Inventory per ROC Factors ROC per ROC Unit Fraction for STS (mrem/yr Unit (pCi/m2) (mCi) (mrem/yr) (mrem/yr) per mCi)
Co-60 2.17E+04 5.63E-02 1.10E-02 6.19E-04 Cs-134 4.27E+03 1.11E-02 1.57E-02 1.74E-04 Cs-137 4.44E+06 1.15E+01 3.00E-02 3.45E-01 3.49E-01 1.40E-02 Ni-63 1.40E+06 3.62E+00 2.89E-04 1.05E-03 Sr-90 2.95E+03 7.63E-03 3.33E-01 2.54E-03 ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 63 of 156
ZS-LT-300-001-004 Attachment 14 Revision 1 Sign Statistical Test Information Use Survey Area: No. 05100
Description:
Auxiliary Building Basement Survey Unit: No. 05100
Description:
Auxiliary Building Basement Floors Classification: 1 Type I ()Error: 0.5 Number of Samples (n): 15 Weighted
- 1-Ws Sign Sum (Ws) 1 5.04E-05 1.00 +1 2 1.24E-05 1.00 +1 3 8.37E-06 1.00 +1 4 5.95E-06 1.00 +1 5 6.91E-04 1.00 +1 6 1.56E-04 1.00 +1 7 3.09E-02 0.97 +1 8 2.83E-02 0.97 +1 9 1.60E-07 1.00 +1 10 2.66E-03 1.00 +1 11 8.90E-03 0.99 +1 12 5.16E-04 1.00 +1 13 6.07E-02 0.94 +1 14 1.17E-01 0.88 +1 14 1.17E-02 0.99 +1 Critical Value (Table I.3 of MARSSIM) = 11 Number of Positive Differences (S+) = 15 The survey unit (meets) (does not meet) the acceptance criteria Prepared by (RE):
(Print Name) (Signature) (Date)
Peer Reviewed by (RE):
(Print Name) (Signature) (Date)
As the Sign Test for both STS units passed at the BIL equal to the expected fraction of the inventory, and the sum of the Mean Inventory Fraction is less than one (1.10E-04 + 1.40E-02 =
1.41E-02), these survey units would pass STS ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 64 of 156
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 65 of 156
- 16. Comment: Justification is needed for the conceptual model assumed in the basement fill model.
Basis: Many aspects of the conceptual model assumed in the determination of the dose from residual radioactivity in the basements are conservative (e.g., the assumption that the water in the basements can support a well). However, other aspects of the conceptual model may be non-conservative.
The release of the radionuclides is assumed to be either instantaneous for loose surface contamination or a diffusion controlled release for volumetric contamination. The DUST-MS model partitions the released radionuclides between the aqueous and solid phases based on the sorption coefficient (Kd value) for the radionuclide. The Kd approach to modeling the sorption of radionuclides assumes that the medium is a porous medium and that the surface area is available for sorption. It is not clear if this model is appropriate for basements that are backfilled with concrete debris with a diameter of up to 10 inches. The transport of radionuclides in a basement with this configuration may be more like transport through fractures than flow through porous media. In transport through fractures, there may be less sorption, and consequently higher aqueous concentrations, than in transport through a porous medium because less surface area is available for sorption. Tables 11 to 15 in TSD 14-009 show that the model calculated a significant fraction of the radionuclides to be sorbed rather than being in solution for all radionuclides but H-3. The modeling approach assumed in DUST-MS may therefore underestimate the aqueous concentrations in the basements.
Because the conceptual model used in the DUST-MS calculations contains both conservatisms and potential non-conservatisms, the overall conservatism of the basement fill model is not clear.
Path Forward:
- a. Provide a quantitative evaluation that demonstrates that the DUST-MS modeling approach is either appropriate for the expected end state of the backfilled basements or that the DUST-MS modeling approach bounds the potential dose due to the residual radioactivity in the basements.
ZSRP Response (16a) - ZSRP believes that the primary basis for this NRC comment is a question regarding the application of a Kd approach in the Basement Fill Model (BFM). In the Basis discussion, NRC states the following:
The Kd approach to modeling the sorption of radionuclides assumes that the medium is a porous medium and that the surface area is available for sorption. It is not clear if this model is appropriate for basements that are backfilled with concrete debris with a diameter of up to 10 inches. The transport of radionuclides in a basement with this configuration may be more like transport through fractures than flow through porous media. In transport through fractures, there may be less sorption, and consequently higher aqueous concentrations, than in transport through a porous medium because less surface area is available for sorption.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 66 of 156 There are several reasons, as listed below, why the Kd approach for the partitioning of radionuclides between the solid and liquid phases in each individual basement is valid, appropriate, and conservative for the dose assessment of basements backfilled with 10 inch minus crushed concrete.
- 1) The sizing of concrete debris to 10 inch minus is performed with a grinding/crushing machine that produces debris with a wide range of sizes with the maximum dimension of a singular piece of concrete debris having a diameter of 10 inches. The resulting wide range of sizes, down to very small pieces and fines, is actually the ideal geometry to provide low porosity and good compaction. An example of concrete on the Zion site that has been crushed and sized to the 10 inch minus specification is shown in the picture below:
Concrete on the Zion site that has been crushed to a 10 inch minus specification showing the distribution of sizes after crushing.
- 2) As stated above, concrete debris crushed to a 10 inch minus specification includes a wide range of pieces that are sized smaller than 10 inch in diameter that allows for good compaction during backfill. Fracture systems are characterized by a network of small connected fractures in relatively low porosity systems (e.g. granite) that allow the majority of
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 67 of 156 flow to occur through the fractures. The backfill will be a randomly packed, porous material that will not present long interconnected flow paths as in a fracture system. The Kd approach is fully appropriate to concrete crushed to a 10 inch minus specification. The wide range of sizes provides tight packing with extensive surface area available for sorption.
- 3) The RAI refers to flow but the BFM is a static equilibrium model with no flow assumed to occur between basements or to outside soil. Radionuclides released from concrete are assumed to equilibrate in the water/mass matrix of the fill within each basement separately.
- 4) Assuming a static model with no flow between basements is conservative. If flow and mixing between basements were included, for example, between the Auxiliary Building basement and the Turbine Building basement, the BFM Dose Factor (mrem/yr per mCi) for the combined basements would decrease by about a factor of 2 compared to the Dose Factors provided in LTP Chapter 6 due to essentially doubling the mixing volume. Theoretically this would not make a difference in the activity allowed to remain in each basement since the decreased Dose Factor would be offset by the increased concrete surface area of the combined basements. However, in practice this would allow the acceptable total residual radioactivity inventory in the Auxiliary Building basement to increase by about a factor of 2 compared to the inventory currently allowed. This would occur because characterization and operational history indicate with high certainty that the residual radioactivity inventory in the Turbine Building basement is a very small fraction of that contained in the Auxiliary Building basement. In essence, the source term would be entirely from the Auxiliary Building basement but the mixing volume would increase by a factor of 2 after including the Turbine Building basement. Therefore, the allowable inventory in the Auxiliary Building basement would increase in proportion to the increased volume added by the Turbine Building basement. This example demonstrates the conservatism of assuming no flow between basements.
- 5) Assuming no flow to soil outside of the basements is conservative. As stated in the response to RAI 21, if water is assumed to flow from the basements to the adjacent surrounding soil and a hypothetical well is installed in the adjacent soil as opposed to inside the basement, the Cs-137 groundwater dose (the predominant radionuclide) would be lower by approximately a factor of 100 as compared to the Auxiliary Building basement dose.
As an illustration, there are several conservative assumptions in the BFM which are listed below:
- 1) The assumption that water well will be installed on the site.
It is unlikely that within the next 100 years a land use would occur that included an onsite water well which is prohibited by local municipal code.
- 2) Drilling water well within a basement.
If the local laws were ignored, it is unlikely that anyone would drill through the backfill (concrete construction debris fill and concrete floor) to install a well.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 68 of 156 If a well were installed, the water in the basements would be non-potable due to the high pH (>10) that will occur from leaching of the concrete construction debris.
- 3) Conservative RESRAD parameter selection process that applies 75th or 25th percentile values for sensitive parameters.
- 4) Minimum Kd values were used which maximizes the calculated concentrations in the basement fill water and the corresponding groundwater dose.
- 5) The assumption in the calculation of BFM Dose Factors that there is no water mixing between basements is conservative. As an example, given the actual expected source term based on characterization results, assuming mixing between the Auxiliary Building basement and Turbine Building basement would reduce the actual dose from Cs-137 (the predominant radionuclide) by approximately a factor of 2 compared to the modeled dose.
- 6) The assumption in the calculation of the BFM Dose Factors that the well is installed in the basement as opposed to down gradient soil is conservative. As demonstrated in the ZSRP response to RAI PAB 21, assuming that the well is drilled into down gradient soil, which is more likely than into the basement fill, would reduce dose by approximately a factor of 100 as compared to the Auxiliary Building basement dose.
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- 17. Comment: The basis for the parameter values assumed for the mixing volumes and surface areas for the basements is needed.
Basis: Tables 1 and 2 in TSD 14-009 describe the mixing volume and surface area assumed for each building in DUST-MS. An email from Farr was cited as the reference for these volumes, but this reference was not provided as part of the LTP submittal. The NRC staff needs information on the basis for these values to evaluate the DUST-MS modeling results.
Path Forward:
- a. Provide the basis for the mixing volumes and surface areas for the basements assumed in the DUST-MS model.
- b. Provide the following document: Farr, H.C., Re: New Volumes e-mail 9/24/14 to T.
Sullivan ZSRP Response (17a and 17b) - A copy of the email from Farr, H.C., Re: New Volumes September 24, 2014 to T. Sullivan was provided to the NRC on November 12, 2015, (ZS-2015-0163).
The bases for the void volumes provided in the email are in three TSDs:
TSD 14-013 Zion Auxiliary Building End State Estimated Concrete Volumes, Surface Areas, and Source Terms TSD 14-014 End State Surface Areas, Volumes, and Source Terms of Ancillary Buildings TSD 13-005 Unit 1 & 2 Reactor Building Estimated End State Concrete and Liner Volumes and Surface Areas TSD 14-013, TSD 14-014, and TSD 13-005 are provided in Enclosure 2.
The table below cross references the values reported in the referenced email from Farr to the tables in the various TSDs where the data is reported. The values in the September 24, 2014 email were from final drafts of the three TSDs listed above.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 70 of 156 Cross Reference between Values Provided in Email Referenced in TSD 14-009 and the Tables in TSDs13-005, 14-014, and 14-013 that Provide the Final Values for the Zion Basements Email Water Total Table Void Reference TSD and Table Number &
Surface Volume per Discrepancy Between Void Volume in Structure Area ft2 Item ft3 TSD 14-009 and the Final TSD (if any) Reason Unit 1 Containment TSD 13-005 Table 25 and Table 34, and Liner Only 2.97E+04 2.31E+05 TSD 14-014 Rev 1 Table 60 N/A Unit 2 Containment TSD 13-005 Table and TSD 14-014 Rev Liner Only 2.97E+04 2.31E+05 1 Table 60 N/A TSD 14-013 Tables 2 & 3 and TSD 14-Auxiliary Building 7.00E+04 1.00E+06 014 Table 60 N/A Turbine Bld, Main Minor error found Steam Tunnel, Diesel TSD 14-014 Table 60 Total Surface during final review Oil 1.59E+05 9.23E+05 Area 1.60E+05 ft2 of TSD.
Unit 1 Discharge Tunnel 2.62E+04 1.06E+05 TSD 14-014 Table 60 N/A Unit 2 Discharge Tunnel 2.62E+04 1.06E+05 TSD 14-014 Table 60 N/A TSD 14-014 Rev. 0 Table 57. Rev 1 Calculation revised Table 60 Changed to Surface Area to after change to 1.49E+05 ft2, Water Table Void Spec to demolition plan in Crib House and 923,738 ft3, Concrete Volume June of 2015 Forebay 7.47E+04 1.08E+06 to 5.05E+05 ft3 Waste Water Treatment Facility 1.21E+04 5.08E+03 TSD 14-014 Table 60 N/A TSD 14-014 Table 60 Surface Area Minor error found Changed to 7.78E+03 ft2 for surface during final review Spent Fuel Building 8.40E+03 7.35E+03 area of TSD.
As seen in the table above, there were slight differences in the surface areas reported in the email and the surface areas in the Final Revision 0 of the TSDs for the Turbine, Main Steam Tunnel, Diesel Oil and Spent Fuel Building structures as result of minor errors found during the final reviews. However, the volumes remained the same and therefore no adjustment to the BFM Dose Factors is necessary. Note that the methods used to calculate the BFM Dose Factors were independent of the surface area because a unity source term approach was used in the DUST-MS modeling.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 71 of 156 Additional Information: The approach for the demolition of the Crib House and Forebay was revised after submittal of LTP Revision 0 which resulted in an increase in the number of interior walls to remain. The change resulted in increased surface area and decreased void volume. As stated above, the BFM Dose Factors are independent of surface area but are inversely proportional to void volume. Therefore, the BFM Dose Factors for the Crib House and Forebay will be recalculated in the revised LTP to be submitted following completion of the RAI process.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 72 of 156
- 18. Comment: Additional information is needed on the groundwater and hydrological parameters.
Basis: Section 6.6.1 of the LTP states Based on current demolition plans, there will be no connection between the basements and surrounding groundwater and the End State configuration provides no route for groundwater ingress into the basements, leaving only rainwater infiltration as the source of water in the fill. Section 6.6.1.2 states the projected equilibrium water elevation in the Basements was evaluated in TSD 14-032. The water level is driven by the location, elevation and size of existing penetrations between the Basements and between the Basements and outside ground. In TSD 14-09, it is stated that the mixing volume is calculated assuming that the water level in the basements is equal to the natural water table elevation outside of the basements (i.e., 579 feet), which is the minimum long term level that could exist in the basements. Based on these descriptions, it is not clear if there is connectivity between the basements and surrounding groundwater.
Also, TSD 14-032 and TSD 14-006, which contains key hydrological parameters, were not provided to the NRC.
Path Forward:
- a. Describe the connectivity between the basements and surrounding groundwater in the end state configuration.
ZSRP Response (18a) - An excel spreadsheet describing all penetrations between basements, including the drawing number references is provided in Enclosure 2. The spreadsheet provides the inputs used for the evaluations in TSD 14-032, Conestoga Rovers & Associates Report Simulation of the Post-Demolition Saturation of Foundation Fill Using a Foundation Water Flow Model
- b. Provide ZionSolutions Technical Support Document 14-032, Conestoga Rovers &
Associates Report, Simulation of the Post-Demolition Saturation of Foundation Fill Using a Foundation Water Flow Model ZSRP Response (18b) - TSD 14-032, Conestoga Rovers & Associates Report, Simulation of the Post-Demolition Saturation of Foundation Fill Using a Foundation Water Flow Model was submitted to the NRC on November 12, 2015 (ZS-2015-0163).
- c. Provide 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.
ZSRP Response (18c) - TSD 14-006, Evaluation of Hydrological Parameters in Support of Dose Modeling for the Zion Restoration Project was submitted to the NRC on May 12, 2015 (ZS-2015-0084) as Reference 26 of TSD 14-003.
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- 19. Comment: More information is needed on the development of the BFM drilling spoils dose factors.
Basis: Chapter 6 of the LTP describes the drilling spoils dose factor calculations, but the details of this analysis are included in TSD 14-021, which was not provided to the NRC.
From the description in Chapter 6, it is not clear how the source term concentration in the basements with diffusion controlled release was calculated. Section 6.6.7 states The BFM Drilling Spoils scenario addresses one of the BFM exposure pathways listed in section 6.5.4 by calculating the dose from residual radioactivity in fill material (resulting from release from surfaces to clean fill after backfill) which is brought to the surface during the installation of a well in the basement. The activity remaining in the concrete surfaces, if any, is also included in the drilling spoils source term. The text states that the source term for the BFM Drilling Spoils scenario is the average concentration in fill, and remaining in concrete, at the time of maximum groundwater concentration. Later the text states to ensure conservatism and consistency in the BFM source term, the maximum fill concentrations (which occur at the time of maximum groundwater concentrations) are applied for each radionuclide regardless of when the maximum occurs. From these descriptions, it is not clear how the residual contamination remaining in concrete is incorporated into the source term.
Path Forward:
- a. Describe how the residual radioactivity in the concrete is included in the source term for the drilling spoils.
ZSRP Response (19a) - The BFM Drilling Spoils scenario is described in detail in TSD 14-021, section 3.1. A summary of the source term assumptions is provided below.
The source term for the BFM Drilling Spoils scenario is calculated in one of two ways depending on whether the release of residual radioactivity from basement concrete in the BFM Groundwater scenario is assumed to be diffusion controlled or instant release. As discussed in LTP Chapter 6 and TSD 14-010 (which was previously submitted to NRC), diffusion controlled release was assumed for the Auxiliary Building basement and SFP/Transfer Canal. Instant release was assumed for the remaining basements.
For instant release, the Drilling Spoils source term is contained entirely in the saturated fill material. After instant release, all residual radioactivity is assumed to be in the fill/water with no residual radioactivity remaining in the basement concrete. The water supply well is drilled through the fill material down to the basement floor and the residual radioactivity in the fill is brought to the ground surface with the drilling spoils.
For diffusion controlled release, a fraction of the Drilling Spoils source term is contained in the fill material after release from the concrete and a fraction of the source term remains in the concrete. The relative fractions in fill and concrete vary as a function of time as diffusion controlled release occurs. The Drilling Spoils fill and concrete source term in the two basements
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 74 of 156 with diffusion controlled release (Auxiliary Building and SPF/Transfer Canal) is calculated at the time of maximum radionuclide concentrations in the fill material (and interstitial pore space water) in order to be consistent with the time of maximum source term that is applied to the BFM Groundwater scenario. The total residual radioactivity inventory in the fill and concrete, at the time of maximum fill and water concentrations, is assumed to be brought to the surface with the drilling spoils.
The time when maximum fill and water concentrations occur is a function of the diffusion coefficient and half-life of each radionuclide. This is demonstrated in the table below which is reproduced from TSD 14-009, Brookhaven National Laboratory: Evaluation of Maximum Radionuclide groundwater Concentrations for Basement Fill Model (submitted to NRC on March 30, 2015 [ZS-2015-0051]).
Auxiliary Building Peak Groundwater Concentrations (pCi/L) per unit source of 1 pCi/m2 and diffusion controlled release from 0.5 inch of contaminated concrete.
Peak Peak Diffusion Time to Peak Radioactivity Radioactivity Peak Sorbed Coefficient Kd Peak Concentration in Solution Sorbed Concentration Nuclide (cm2/s) (ml/g) (years) pCi/L pCi pCi pCi/g H-3 5.00E-07 0 0.1 9.10E-04 6503 0.0 0.00E+00 Co-60 4.10E-11 223 4 2.60E-08 0.2 249 5.80E-09 Ni-63 1.10E-09 62 37 1.90E-06 13.6 5051 1.18E-07 Sr-90 5.20E-10 2.3 21 1.96E-05 140.1 1933 4.51E-08 Cs-134 3.00E-09 45 1.5 6.89E-07 4.9 1329 3.10E-08 Cs-137 3.00E-09 45 14 2.47E-06 17.7 4766 1.11E-07 Eu-152 5.00E-11 95 10 1.07E-07 0.8 440 1.03E-08 Eu-154 5.00E-11 95 6 8.38E-08 0.6 341 7.96E-09 Note that the time to peak concentration ranges from 0.1 to 37 years after license termination.
However, as stated in LTP Chapter 6, section 6.5.4, to ensure conservatism and consistency in the BFM source term, the maximum fill concentrations (which occur at the time of maximum groundwater concentrations) are applied for each radionuclide regardless of when the maximum occurs. This statement is intended to recognize that at any given time the concentrations in the fill, considering all radionuclides combined, will be less than the total assumed in the BFM source terms (Groundwater and Drilling Spoils) since the maximum values actually occur at
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 75 of 156 different times for each radionuclide. But the primary reason for applying the source term at the time of maximum fill concentrations is consistency with the BFM Groundwater scenario.
Although inconsistent with the BFM Groundwater scenario, the Drilling Spoils source term for the diffusion controlled release basements (Auxiliary Building and SFP//Transfer Canal) would be maximized at time equal to zero. However, this would reduce the combined BFM Dose Factors (see LTP Chapter 6, Table 6-18) which is the sum of both the Groundwater and Drilling Spoils Dose. Note that decreasing the BFM Dose Factors increases the allowable total inventory levels. For example, the primary source term for all basements is Cs-137 in the Auxiliary Building basement. As seen in the table above, the maximum fill and water concentrations for Cs-137 in the Auxiliary Building basement occur after 14 years. Assuming that drilling occurs at time equal to zero, the Drilling Spoils source term and dose factor would increase by 27% due to reduced decay. However, the Groundwater Dose Factor would decrease by at least 27%
attributed to decay because of the time dependent nature of diffusion controlled release.
Because the Groundwater Dose Factor is approximately an order of magnitude greater than the Drilling Spoils Dose Factor (see LTP Chapter 6 Tables 6-16 and 6-17), the combined BFM Dose Factor for Cs-137 in the Auxiliary Basement would be reduced by at least 21% if the source term were calculated at time equal to zero as opposed to the time of maximum fill and groundwater concentrations. The 21% value assumes that the Drilling Spoils Dose Factor is increased by 27%
and the Groundwater Dose Factor in decreased by 27% with the two dose factors then summed as in LTP Chapter 6 to calculate a hypothetical BFM Dose Factor at time equal to zero.
- b. Provide ZionSolutions Technical Support Document 14-021 Basement Fill Model (BFM)
Drilling Spoils and Alternate Exposure Scenarios.
ZSRP Response (19b) - TSD 14-021, Basement Fill Model (BFM) Drilling Spoils and Alternate Exposure Scenarios was submitted to the NRC on November 12, 2015 (ZS-2015-0163).
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- 20. Comment: Additional details on the basement fill model elevated area consideration are needed.
Basis: In Chapter 6 of the LTP, it is stated that the maximum concentrations that could remain in the Basements are limited by the implementation of the open air demolition limits described in TSD 10-002. These limits include:
less than 2 mR/hr beta-gamma total surface contamination on contact with structural concrete, and less than 1,000 dpm/100cm2 beta-gamma loose surface contamination.
The loose surface contamination limit is based on the loose surface contamination levels prior to demolition. It is possible that the loose surface contamination levels could increase after demolition due to dust generated during the demolition process.
Also, Chapter 6 of the LTP does not provide sufficient details on how the maximum concentrations assumed in the worst-case drilling spoils scenario were generated. These details may be in TSD-14-021, which the NRC staff requested in a separate RAI.
Path Forward:
- a. Provide a justification that the loose surface contamination levels will remain within these limits after demolition. Include information on any actions that will be taken to minimize the dust present and any measurements that will be performed to evaluate the final, post-demolition levels of loose surface contamination.
ZSRP Response (20a) - There is no requirement that the loose surface contamination levels be less than the open air demolition limits after demolition is completed. The limits were designed to ensure that offsite public dose is maintained below applicable criteria during structure demolition without construction of an enclosure, i.e., in open air. However, the potential for significant levels of removable contamination to be present after demolition is low. For example, for the Auxiliary Building basement walls, a very conservative estimate of Cs-137 inventory was provided in LTP Chapter 5, section 5.5.2.1.1 as being 0.15 Ci. The total volume of concrete in the Auxiliary basement walls was determined to be 3.62E+05 ft3 in TSD 14-003, Zion Auxiliary Building End State Estimated Concrete Volumes, Surface Areas, and Source Terms,(submitted on May 12, 2015 [ZS-2015-0084]), Table 3. Based on these values, assuming uniform mixing with the wall volume during demolition and that a 1mm dust layer would remain after debris is removed and the area is cleaned/vacuumed, loose contamination levels would be nominally calculated as 218 dpm/100 cm2.
After remediation and demolition are complete, loose material will be removed by mechanical methods and/or vacuuming to provide a clean work area to perform STS. Although loose contamination levels are expected to be low after demolition there is no specific requirement that loose contamination be limited to a certain level in order to apply the BFM Dose Factors. As described in LTP Chapter 6, the BFM conceptual model assumes that 100% of residual
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 77 of 156 radioactivity is instantly released for all basements except the Auxiliary Building and SFP/Transfer Canal. For these two basements a diffusion controlled release is assumed due to contamination being identified at depth during characterization of the Auxiliary Building basement and the expectation that contamination will be found at depth in the concrete under the SFP/Transfer Canal steel liner after the liner is removed.
For basements with instant release the dose is independent of the ratio of loose to fixed residual radioactivity levels. For the two basements with diffusion controlled release (Auxiliary Building and SFP/Transfer Canal) there is a potential to underestimate the dose from loose contamination which would be assumed to instantly release and mix with the saturated fill as opposed to releasing by diffusion from concrete. However, the potential dose impact from any loose contamination remaining is insignificant as described in the example below.
For Cs-137, the predominant radionuclide, the maximum water concentrations under the diffusion controlled release scenario is 74% of the maximum water concentration that could occur under an instant release scenario. Assuming that loose contamination would instantly mix, the resulting water concentrations attributable to the loose contamination would be 1/0.74 = 1.35 times higher than projected based on diffusion of the same inventory from a 0.5 inch thickness of concrete.
However, the total inventory attributable to potential loose contamination is low. For example, assuming that the level of loose radioactivity is 1,000 dpm/100 cm2 over 100% of wall and floor surfaces in the Auxiliary Building basement, the total inventory of the loose radioactivity would be 0.3 mCi. From LTP Chapter 5, Table 5-9 the Basement Inventory Limit (BIL) for Cs-137 in the Auxiliary Building basement, i.e., the inventory corresponding to 25 mrem/yr, is 835 mCi.
As shown above, the dose from the loose contamination could be at the most underestimated by a factor of 0.35. The underestimated dose would then be calculated as 0.3 mCi ÷ 835 mCi*25 mrem/yr *0.35 which equals 0.003 mrem/yr. From this calculation it is demonstrated that even at a completely unrealistic loose contamination level of 100,000 dpm/100 cm2, the potential dose would be underestimated by 0.3 mrem/yr which is insignificant.
Regardless of the insignificant dose consequences of loose contamination, standard good radiological control practices will be applied to reduce and minimize levels of loose contamination after demolition and remediation. This would include manual cleaning and vacuuming to optimize the environment for ease of access and controls during performance of the STS surveys.
- b. Provide details on the derivation of the maximum concentration assumed in the worst-case drilling spoils scenario.
ZSRP Response (20b) - A detailed explanation of the derivation of the maximum concentration for the worst case drilling spoils assessment is provided in TSD 14-021, Section 4 which was submitted on November 12, 2015 (ZS-2015-0163).
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- 21. Comment: The basis for assumptions in the sensitivity analysis performed for an outside well receptor is needed.
Basis: In TSD 14-009, a sensitivity analysis was performed for a well receptor located outside of the basements. In this analysis, the peak concentrations in a receptor well 2 m outside the turbine building were calculated. This analysis considered source terms from the auxiliary and turbine buildings with contamination levels of 1 pCi/m2 and 0.001 pCi/m2 respectively. TSD 14-009 states that the well location was selected to be the closest place to put a well outside of the auxiliary building, which is the building that will have the highest levels of residual contamination. It is not clear if sources from other basements, such as the containment building basements, could affect the groundwater at this location. Also, it is not clear how the assumed concentrations will compare to the end state concentrations in those basements.
Path Forward:
- a. Provide a justification for the source terms and concentrations included in the outside well receptor analysis. If the containment buildings, or any other basements, could contribute to the groundwater concentration at the well, provide a sensitivity analysis that estimates the maximum projected groundwater concentration including these source terms.
ZSRP Response (21a) - The source terms used in the well receptor sensitivity analysis in TSD 14-009, Revision 0 were based on general but conservative assumptions as to the inventory of residual radioactivity remaining in the Auxiliary Building and Turbine Building basements at the time of license termination. The objective of the sensitivity analysis was to provide a nominal estimate of the ratio of the water concentrations in a well assumed to be located in soil immediately down gradient of the Turbine Building basement to the water concentrations in the Auxiliary Building basement calculated using the BFM bounding assumption that the well is located in the Auxiliary Building basement (which is projected to contain the highest levels of residual radioactivity at license termination).
The sensitivity analysis was performed for all ROC. The activity ratio between the Auxiliary Building Basement and the Turbine Building basement was determined based on Cs-137 characterization data and assumed to apply to all radionuclides. This is reasonable given that Cs-137 is the dominant dose contributor and that Cs-137 was the only positively identified radionuclide in the Turbine Building basement.
The source term for the well receptor sensitivity analysis was revised to accommodate the inclusion of the Containment Building basements as requested in the RAI. The full details of the revised DUST-MS well receptor sensitivity analysis are provided in Appendix B to TSD 14-009, Revision 1 which is provided in Enclosure 2. A detailed summary of the source term assumptions and calculations is provided below.
The initial analysis presented in TSD 14-009, Revision 0 assumed that the activity in the Turbine Building basement was lower than the Auxiliary Building basement by a nominal factor of 1.0E-
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- 03. The 1.0E-03 value is the estimated ratio of the actual Cs-137 inventory in the Turbine basement (based on characterization data) to the Turbine Building Basement Inventory Limit (BIL) for Cs-137, (i.e., the activity that would result in 25 mrem/yr). The addition of the Containment Building basements required a slight revision to the method for calculating the source term for the well receptor sensitivity analysis. This included a more detailed calculation of the Turbine Building inventory and BIL ratio as described below From LTP Chapter 2, Table 2-25, the average concentration of Cs-137 from the analysis of biased cores from the Turbine Building basement floor was 15.4 pCi/g. This activity was found only in several sporadic localized areas but for the purposes of this sensitivity analysis, the 15.4 pCi/g concentration was conservatively assumed to be uniformly distributed over the entire basement floor. From TSD 14-021 Revision 0, Table 2, the Turbine Building basement floor surface area is 48,576 ft2 or 4,512 m2. The total estimated Cs-137 inventory in mCi was determined to be 2.12 mCi. This value was determined by assuming a Cs-137 concentration of 15.4 pCi/g uniformly distributed over the entire floor area to a depth of 0.5 inch, with a concrete density of 2.4 g/cm3.
LTP Chapter 6, Table 6-16, lists the BFM Groundwater Dose Factors. Since the well sensitivity analysis addresses the groundwater scenario only (not Drilling Spoils), the Groundwater Dose Factors are used as opposed to the full Basement Dose Factors in LTP Chapter 6, Table 6-18.
From LTP Chapter 6, Table 6-16, the BFM Groundwater Dose Factor for Cs-137 in the Turbine Building basement is 3.92E-02 mrem/yr per mCi. Therefore, the Cs-137 BIL for the groundwater scenario is calculated as (1/3.92E-02 mrem/yr per mCi) x 25 mrem/yr = 637 mCi.
Given that the Cs-137 inventory in the Turbine Building basement concrete was conservatively estimated at 2.12 mCi, the ratio of the estimated inventory to the BIL is 2.12/637 = 3.32E-03.
This value was approximated as 1.0E-03 in the original well receptor sensitivity analysis.
The Turbine BILs for all ROCs were multiplied by the 3.32E-03 fraction to determine the Turbine Building inventory in the well sensitivity analysis. The inventory for all ROC in the Containment Building and Auxiliary Building basements were conservatively assumed to be at the maximum allowable level, i.e., the BIL. The resulting total inventories for each ROC in each basement applied to the well sensitivity analysis are provided in the table below. Note that both the BIL and final adjusted inventories are listed for the Turbine Building basement.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 80 of 156 Basement Inventories Applied in Well receptor Sensitivity Analysis Auxiliary Containment Turbine Adjusted BIL BIL BIL Turbine Radionuclide mCi mCi mCi mCi H-3 4.03E+03 9.20E+02 3.68E+03 1.22E+01 Co-60 2.50E+05 2.19E+03 8.72E+03 2.90E+01 Ni-63 8.75E+04 1.56E+04 6.23E+04 2.07E+02 Sr-90 7.59E+01 5.54E+00 2.21E+01 7.34E-02 Cs-134 2.70E+03 1.27E+02 5.06E+02 1.68E+00 Cs-137 9.48E+02 1.59E+02 6.37E+02 2.12E+00 Eu-152 4.20E+05 6.46E+03 2.58E+04 8.57E+01 Eu-154 3.69E+05 4.45E+03 1.78E+04 5.90E+01 The Base Case DUST-MS modeling used to support the calculation of BFM Dose Factors assumed a uniform contamination level of 1 pCi/m2 in each basement. For comparison with the Base Case, the units of inventory for the well sensitivity analysis were therefore directly scaled from mCi to pCi. The objective of the sensitivity study was to show the ratio of the concentration at the receptor well to the concentration in the Auxiliary Building basement.
Scaling does not impact this objective.
The BIL values are based on the entire void space volume of the basements that will remain at Zion. Therefore, as described in detail in TSD 14-009 Revision 1 Attachment B, the inventory of the Auxiliary Building basement and Turbine Building basement required reduction to accommodate the geometry adjustments required for the DUST-MS well receptor sensitivity modeling. This reduction is the ratio of the modeled volume (see TSD 14-009, Appendix B, Table 2) to the actual volume (see TSD 14-009, Appendix B, Table 1). The Auxiliary Building basement inventory is therefore reduced by a factor of 0.51 and the Turbine Building basement by a factor of 0.55. The resulting inventory used in the well receptor simulation is provided below.
Inventory used in the DUST-MS Well Receptor Sensitivity Model Nuclide Auxiliary Containment Turbine pCi pCi pCi H-3 2.07E+03 9.20E+02 6.75E+00 Co-60 1.28E+05 2.19E+03 1.60E+01 Ni-63 4.49E+04 1.56E+04 1.14E+02 Sr-90 3.90E+01 5.54E+00 4.05E-02 Cs-134 1.38E+03 1.27E+02 9.28E-01 Cs-137 4.87E+02 1.59E+02 1.17E+00 Eu-152 2.15E+05 6.46E+03 4.73E+01 Eu-154 1.89E+05 4.45E+03 3.26E+01
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 81 of 156 The final table below is the source term used in the well sensitivity analysis. The units of pCi/m2 are used for consistency with the 1 pCi/m2 unit source term used in the Base Case. As stated above, the units applied have no bearing on the final result which is the ratio of well water concentration in a well placed outside of the basements to a well placed in the Auxiliary Building basement. The table below is generated by dividing the pCi values in the table above by the surface area in each basement.
Inventory per unit surface area in each basement.
Nuclide Auxiliary Containment Turbine pCi/m2 pCi/m2 pCi/m2 H-3 6.19E-01 3.34E-01 8.32E-04 Co-60 3.84E+01 7.92E-01 1.97E-03 Ni-63 1.35E+01 5.64E+00 1.41E-02 Sr-90 1.17E-02 2.01E-03 5.00E-06 Cs-134 4.15E-01 4.59E-02 1.15E-04 Cs-137 1.46E-01 5.78E-02 1.44E-04 Eu-152 6.45E+01 2.34E+00 5.84E-03 Eu-154 5.67E+01 1.61E+00 4.02E-03 The results of the well receptor sensitivity analysis including the Containment Building basement are provided in TSD 14-009 Revision 1, Attachment B, Table 9, and are reproduced in the table below. In summary, the concentration of Cs-137 in a well located in soil outside of the Turbine Building basement is 1% of the concentration calculated for the Auxiliary Building basement in the bounding BFM groundwater scenario which assumes a well is located within the basements.
This corresponds to a reduction in groundwater dose by about a factor of 100 since the majority of the dose is from Cs-137. The addition of the Containment Building basements with assumed maximum contamination levels does not impact the concentrations in the receptor well. There are slight differences in the results for the low Kd radionuclides such as Sr-90 and H-3 but the concentrations in the receptor well are lower for all radionuclides and the actual inventory of Sr-90 and H-3 are very low relative to Cs-137 and thus their dose in much lower.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 82 of 156 Ratio Well Turbine Receptor to Time to Containment Aux Bldg Bldg Well Auxiliary peak Basement (pCi/L) (pCi/L) (pCi/L) (pCi/L) (years)
H-3 5.6E-04 5.6E-04 4.2E-04 3.9E-04 0.70 1.1 Co-60 9.93E-07 9.95E-07 3.5E-09 1.7E-14 1.7E-08 16.8 Ni-63 2.54E-05 2.57E-05 3.3E-06 1.9E-06 0.07 286 Sr-90 2.3E-07 2.4E-07 1.7E-07 1.6E-07 0.65 13.6 Cs-134 2.83E-07 2.85E-07 1.0E-09 1.7E-13 0.00 5 Cs-137 3.6E-07 3.6E-07 4.9E-09 1.8E-09 0.01 167 Eu-152 6.9E-06 6.9E-06 2.5E-08 2.5E-10 0.00 21.2 Eu-154 4.7E-06 4.7E-06 1.7E-08 3.8E-11 0.00 15.3
- b. Additionally, provide a description of the process that will be used to re-evaluate this analysis if final inventory numbers for the basements are higher than those assumed in this analysis.
ZSRP Response (21b) - The source terms for the Containment and Auxiliary Building basements in the receptor well sensitivity analysis are assumed to be at the maximum allowable levels and by definition cannot be greater. For the Turbine Building basement, there is no condition envisioned where the source term could be significantly higher given the conservative application of actual characterization data. In addition, this evaluation is intended as a sensitivity analysis to quantify conservatism in the BFM bounding model and is not a part of the dose assessment to demonstrate compliance with the dose criterion. Therefore, no future re-evaluation of the source term or this sensitivity analysis is considered necessary.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 83 of 156 HP Zion RAIs Chapter 2
- 1. Comment: Section 2.2.1 (Data Quality Objectives)
Section 2.2.1 of the LTP indicates that direct measurements and scans of concrete and surface soils were also made using the same instruments and Minimal Detectable Concentrations (MDC) as will be employed for FSS. However, survey methodologies described throughout Chapter 2 of the LTP indicate that alarm set points were utilized during characterization (generally described as being set at the observed background plus the Minimum Detectable Count Rate for the instrument). The calculation of Scan MDCs for FSS in Section 5.8.4.3 of the LTP (Beta-Gamma Scan Measurement Minimum Detectable Concentration) utilizes an equation based on signal detection theory and how a human observer theoretically processes the audible input and then makes decisions. Additional details are needed on the calculation of Scan MDCs used during characterization and the applicability of those MDCs when instrument alarm set points are used.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.2 (Scoping and Characterization Surveys), licensees should provide:
A description of the field instruments and methods that were used for measuring concentrations and the sensitivities of those instruments and methods Section O.3.3.1 (Alarm Set-Points and the MARSSIM Scanning MDC Calculation) of NUREG-1757, Vol. 2, Rev. 1, notes that:
There have been a number of instances where FSS procedures have implemented the use of various detectors coupled to data logging instruments. These instruments in several cases were set to alarm at a pre-determined count rate action level that is calculated to correspond to the DCGLW, rather than relying on the surveyor listening to the audible response. Although this may be an acceptable practice, with the provision of an adequate technical basis, the MARSSIM scan MDC equations are no longer appropriate.
Path Forward:
- a. Provide additional details on the calculation of Scan MDCs for instruments used during characterization, and evaluate the appropriateness of those MDCs when alarm set points are used.
ZSRP Response (HP-2.1a) - The statement in section 2.2.1 of the LTP that direct measurements and scans of concrete and surface soils were also made using the same instruments and Minimal Detectable Concentrations (MDC) as will be employed for FSS was intended to convey that the same types of instrumentation that ZSRP intended to use to demonstrate compliance during the FRS phase of the project, and the inherent sensitivity that those instruments provided, were used to acquire characterization data during the
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 84 of 156 characterization survey phase of the project. However, the Data Quality Objectives for characterization are significantly different. For FRS, the sensitivity of the instruments used is primarily driven by the release criteria that must be achieved to demonstrate compliance. The principal study question that must be answered by the characterization is whether or not a survey unit is classified correctly.
Radiation detection and measurement instrumentation for characterization were selected to provide both reliable operation and adequate sensitivity to detect the Co-60 and Cs-137 at levels sufficiently below the action levels established for characterization. The instruments and detectors selected for scanning were capable of detecting the activity to a MDC of 50% of the applicable action level. This was an administrative guideline only and not necessarily a limit.
As an action level for characterization for the scanning of open land areas, ZSRP employed the NRC soil screening levels provided in NUREG/CR-5512 for Co-60 and Cs-137. These values are 3.8 and 11.0 pCi/g. respectively. Assuming nuclide fractions of 0.1 and 0.9 for Co-60 and Cs-137 and using the surrogate formula from section 11.2 of NUREG-1505 A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys, the surrogate value for Cs-137 is 8.3 pCi/g and 50% of the surrogate value is 4.15 pCi/g. Using the spreadsheets from ZionSolutions TSD# 11-004, Ludlum Model 44-10 Detector Sensitivity, maintaining the detector end-cap of the 2x 2 NaI detector three inches from the soil surface at a scan speed will be 0.25 m/s (~10 inches per second) produced a scan MDC sufficient to detect 50% of the characterization action levels for soil up to a background count rate of about 8,000 cpm. ZionSolutions TSD 11-004, Ludlum Model 44-10 Detector Sensitivity is provided in .
During the performance of characterization, particularly in Class 3 and non-impacted areas, ZSRP wanted to flag all potential scans that indicated activity in excess of background for additional investigation. Consequently, alarm set points for scanning were set at the mean observed background plus the Minimum Detectable Count Rate (MDCR) for the detector.
MDCR was calculated using an index of sensitivity of 1.38 and an interval of one second. In accordance with the characterization survey package instructions, gamma scans included the surveyor listening for an audible change (increase) in the instruments count rate during the scan operation. If the survey area was subject to high noise, then the use of earphones was considered. Considerations for conducting audible surveys and responding to increased count rates included, but was not limited to, slowing the scan speed over the area where the increasing count rate was detected, defining the region of increased count rate and flagging the area for further investigation.
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- 2. Comment: Section 2.2.1 (Data Quality Objectives), Section 2.3.4 (Non-Impacted Open Land Areas), and Section 2.3.5 (Impacted Open Land Areas)
Section 2.2.1 of the LTP indicates that direct measurements and scans of concrete and surface soils were also made using the same instruments and Minimal Detectable Concentrations (MDC) as will be employed for FSS, and that volumetric samples that exhibited the highest activity were sent to an off-site laboratory for analysis of Hard-to-Detect (HTD) radionuclide(s). This discussion indicates that HTD radionuclides are only measured during characterization when an elevated direct measurement was encountered, but it is not clear that a surrogate correlation has been established between HTD and easy-to-detect radionuclides, or that these radionuclides are co-located at the site. As such, it is not evident that a sufficient number of HTD samples were taken during characterization to establish a surrogate or to adequately characterize the radiological status of the site. For example, Section 2.3.4.1 of the LTP discusses surface soil sampling in non-impacted areas and indicates that of the total number of surface soil samples taken and analyzed, Cs-137 was identified at concentrations greater than the MDC of the instrument in 106 surface soil samples, and that no other potential plant-derived radionuclides were positively identified. However, it is not evident from the discussion in Section 2.3.4.1 and the corresponding results summary in Table 2-29 (Non-Impacted Open Land Survey Units -
Characterization Survey Summary) that an analysis of these soil samples was performed for radionuclides other than Cs-137 and Co-60. A similar question exists for the discussion of soil sampling related to Class 2 and 3 impacted areas, as presented in Section 2.3.5 of the LTP and Tables 2-31 and 2-32 (though footnotes associated with Class 3 areas in Table 2-31 indicate that some samples were analyzed for H-3, Fe-55, Ni-63 and Sr-90).
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.2 (Scoping and Characterization Surveys), licensees should provide:
A description and justification of the survey measurements for impacted media (for example, building surfaces, building volumetric, surface soils, subsurface soils, surface water, ground water, sediments, etc., as appropriate),
The justification for considering areas to be non-impacted, A discussion of why the licensee considers the characterization survey to be adequate to demonstrate that it is unlikely that significant quantities of residual radioactivity have gone undetected.
Path Forward:
- a. Describe the basis for only analyzing for hard-to-detect radionuclides when an elevated direct measurement was encountered during characterization.
- b. Describe the analytical measurements that were performed to ensure potential plant derived radionuclides other than Cs-137 and Co-60 would be positively identified in both non-impacted and impacted areas.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 86 of 156 ZSRP Response (HP 2.2a and 2.2b) - ZSRP previously submitted a response to RAIs 2.2a and 2.2b on January 25, 2016 (ZS-2016-0014).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 87 of 156
- 3. Comment: Section 2.3.1.2. (Crib House Concrete Study)
Section 2.3.1.2 of the LTP indicates that in March of 2012, ZSRP commenced an assessment of the concrete 594 foot and 559 foot elevation floors and lower walls of the Crib House, and that the DQOs established for this survey were to establish a background threshold range for volumetric concrete at ZNPS, evaluate the basement foundations and floors of the Crib House for the presence of volumetric radiological contamination and to provide a sufficient quantity and quality of uncontaminated concrete media representative of the Basement Fill concrete to an off-site vendor for the derivation of distribution coefficients for the radionuclides of concern.
Additional clarification is needed on the usage of the Crib House as a background area as Section 2.1.5 of the LTP indicates the Crib House is considered an impacted area, which may not be appropriate for a background area.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.4 (Final Status Survey Design), licensees should provide:
A description of the background reference areas and materials, if they will be used, and a justification for their selection.
Path Forward:
- a. Describe the purpose of the background threshold range. Clarify if these data are being utilized as background and if so, justify the usage of background data from a radiologically impacted area.
ZSRP Response (HP 2.3a) - Section 2.3.1, 1st paragraph states; In March and April of 2012, ZionSolutions conducted a comprehensive background study of non-contaminated concrete by acquiring and analyzing concrete core samples taken from the 559 foot elevation and 594 foot elevation of the Crib House. This study was conducted to support the eventual evaluation of concrete demolition debris as clean hard fill.
To clarify this statement, ZSRP does not intend to use the concrete data from the Crib House as a media background for STS. ZSRP intends to only use the Sign Test as the statistical test that will be employed to demonstrate compliance during FRS (FSS and STS). All reference to the use of the WRS Test will be deleted from the LTP.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 88 of 156
- 4. Comment: Section 2.3.3 (Impacted Structures and Systems) and Section 2.3.3.1 (Unit 1 and Unit 2 Containments)
Section 2.3.3 of the LTP notes that the exposed steel lined walls and floor from the Unit 1 and Unit 2 Containment Buildings (after all interior concrete is removed) and the below-grade structural concrete outside of the liner will remain after demolition and will be subjected to Source Term Surveys. The basic decommissioning end-state for each containment building is further discussed in Section 2.3.3.1, where it is noted that the exposed metal liner will be remediated to levels commensurate with inventory limits in the Basement Fill Model (BFM) that represent the annual dose criterion for unrestricted release specified in 10 CFR 20.1402.
Additional details are needed on characterization of below-grade concrete walls outside of the liner and on steps that may be taken if elevated areas are encountered during remediation.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.2 (Scoping and Characterization Surveys), licensees should provide:
A description and justification of the survey measurements for impacted media (for example, building surfaces, building volumetric, surface soils, subsurface soils, surface water, ground water, sediments, etc., as appropriate)
The survey results including tables or charts of the concentrations of residual radioactivity measured Path Forward:
- a. Provide details of any current characterization of the below-grade concrete walls outside of the liner and any future characterization plans. Describe the steps that will be taken to investigate any elevated areas found during remediation.
ZSRP Response (HP 2.4a) - The basic decommissioning end-state for each Containment building will consist of the walls and floors below 588 foot elevation. Interior concrete walls and floors except for the 568 foot elevation floor and the under-vessel incore walls and floor will be removed. In addition, the 3 feet of concrete on the 568 foot elevation floor will be removed to expose the metal liner.
During characterization, concrete core samples were taken from the Unit 1 Bio-Shield and from the 541 foot elevation floor of each unit to assess the activation radionuclide profile at depth.
The Bio-Shield core sample was used to quantify the potential depth of neutron activation of the structural concrete surrounding the reactor, assuming that the maximum neutron flux occurred at the reactor centerline (elevation 572 foot 9-inches). A 6 foot core was obtained from the outer surface of the Unit 1 Bio-Shield inward toward the core center line. Analysis of the core indicates that Eu-152 and Eu-154, which are indicators of the neutron activation of concrete, were not detected in concentrations greater than their respective MDC until a depth of 47.5 inches (from the exterior of the Bio-Shield inward toward the core center line). The concrete core samples taken from the 541 foot elevation floor of each unit were taken to a depth
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 89 of 156 of 18 inches. Analysis of the 541 foot elevation cores indicates minimal concentrations of Eu-152 and Eu-154 detected at depth. These characterization results support the conclusion that neutrons were successfully attenuated by the shielding around the reactors and consequently, there is minimal potential for the activation of concrete in areas outside the Bio-Shield or below the 3 feet of concrete above the liner. Consequently, it is highly unlikely that the concrete under the liner is activated.
Also during characterization, eleven locations surrounding both Containment Buildings were selected for subsurface soil sampling using Geo-probe to depths up to 3-meters below grade.
While detectable plant-derived radioactivity was positively detected in these samples, the concentrations were at very low levels and not indicative of system leakage or a breach of containment.
For continuing characterization, ZSRP intends to sample the soils surrounding the Containment Buildings that are exposed by demolition activities. Of particular interest are the soils located between each Containment and the Turbine Building, as the HSA indicates the potential for soil contamination in these areas from documented historical spill incidents. To date, ZSRP has characterized this soil to the extent possible. However, subsurface concrete structures and utilities are hindering continued investigation in certain areas. As these soils become accessible, additional soil samples will be taken to assess residual radioactivity in the subsurface soil.
ZSRP believes, based upon the results of concrete core samples and the subsurface soil sampling performed to date, that the probability of contamination or activation of Containment Building concrete beneath the liner or exterior to the Containment is very low. ZSRP also contend that the STS survey that will be performed on the interior surface of each Containment basement after the removal of the concrete floor above the liner will be sufficient to demonstrate compliance with the total inventory limit specified in the BFM.
Section 5.3.4.4 will be revised for clarification as follows; ZSRP has performed sufficient characterization of end-state concrete structures to assess the current residual radioactivity concentration, radionuclide mixture and to ensure the correct classification of each STS survey unit. The only remaining structure that has not been fully characterized to date is the underlying concrete of the SFP and Transfer Canal. Additional characterization may also be performed if necessary in the interiors of select buried or embedded piping and penetrations that will remain. When the underlying concrete in the SFP and Transfer Canal becomes accessible, then the exposed concrete will be characterized. If it is necessary to ensure that a buried or embedded pipe is classified correctly, then the interior surfaces of the buried or embedded pipe will be characterized as they become accessible.
ZSRP also contends that sufficient characterization has been performed of the surface and subsurface soils surrounding ZNPS. Several areas do remain that will need assessment, all pertaining to subsurface soils in Class 1 areas with accessibility contingent upon the demolition
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 90 of 156 of standing structures. Of particular interest are the soils located between each Containment and the Turbine Building, as the HSA indicates the potential for soil contamination in these areas from documented historical spill incidents. As this subsurface soil becomes accessible, it will also be characterized. All surface soil at ZNPS has been adequately characterized and additional characterization of surface soil is not anticipated during continuing characterization.
Examples of other areas where characterization may be performed if necessary but have been deferred due to accessibility include soils under structures, soils under concrete or asphalt coverings and the interior and exterior of both Containment domes. Characterization surveys may be performed as these structures or subsurface soil become accessible and additional characterization data will be collected as necessary, evaluated and stored with-other radiological survey data in a survey history file for the survey unit. In previously inaccessible soils and structures where historical information, process knowledge or operational survey data indicate that the area is classified correctly and no significant concentrations of residual radioactivity is identified or anticipated, then survey design for FRS will be use a coefficient of variation of 30% as a reasonable value for sigma () in accordance with the guidance in MARSSIM, section 5.5.2.2.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 91 of 156
- 5. Comment: Section 2.3.4.1 (Surface Soils)
Section 2.3.4.1 of the LTP indicates that approximately 21% of the surveyed areas were considered inaccessible, which was defined as an area where personnel or vehicle transit was inhibited by the presence of standing water, marsh or wet-lands, thick underbrush, trees or natural grasses where clearing would be prohibitive. Clarification is needed on the adequacy of these surveys to demonstrate that it is unlikely that significant quantities of residual radioactivity have gone undetected.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.2 (Scoping and Characterization Surveys), licensees should provide:
A discussion of why the licensee considers the characterization survey to be adequate to demonstrate that it is unlikely that significant quantities of residual radioactivity have gone undetected, A discussion of how they were surveyed or why they did not need to be surveyed for areas and surfaces that were considered to be inaccessible or not readily accessible.
Path Forward:
- a. Provide justification that the samples from accessible areas adequately represent or bound the areas that were inaccessible and why the characterization survey is adequate to demonstrate that it is unlikely that significant quantities of residual radioactivity have gone undetected from these areas.
ZSRP Response (HP 2.5a) - The Zion Station Historical Site Assessment (HSA) was performed and documented in August of 1999 by Exelon (submitted on November 3 2015 [ZS-2015-0160]).
The HSA is a detailed investigation that collected existing information (from the start of operational activities) regarding the use, handling, and storage of radioactive material at the Zion Nuclear Station and its surroundings.
Based upon the information compiled in the HSA, several large outlying open land survey units received a MARSSIM classification as non-impacted. In accordance with the definition in MARSSIM, non-impacted areas have no reasonable potential for residual contamination because historical information indicates there was no known impact from site operations. These include the outlying open land areas of the site as well as contiguous areas that have no impact from site operations based upon the location(s) of licensed operations, site use, topography, site discharge locations, and other site physical characteristics. These areas are not required to be surveyed for demonstrating compliance beyond surveys performed to validate the basis for the classification.
There is no guidance provided to define the scope and extent of any surveys performed to achieve this objective beyond the requirement to demonstrate that non-impacted areas at the site have not been adversely impacted by decommissioning operations.
The areas referred to in section 2.3.4.1 are open land survey units classified as non-impacted.
The non-impacted open land areas include all of the surrounding Exelon owned land outside of
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 92 of 156 the footprint of the 87 acre, fence-enclosed Industrial Area as well as tracts of land that are owned by the Town of Zion and adjacent businesses.
ZSRP performed radiological surveys to validate the basis for the non-impacted classification of these open land survey areas. The surveys focused primarily on surface soils. The survey consisted of static measurements using a Canberra In Situ Object Counting System (ISOCS),
investigative gamma scans using a Ludlum Model 2350-1 data logger paired with a Ludlum Model 44-10 Sodium Iodide (NaI) detector and the collection of soil samples for analysis. The soil sampling locations were selected based on a random design to ensure an unbiased survey.
Analysis times were set to achieve the required MDCs based on the Cs-137 background due to global fallout as set forth in TSD 13-004, Examination of Cs-137 Global Fallout in Soils at Zion Station. The results of the ISOCS and soil sample measurements were compared to the appropriate background category.
Survey design specified that the minimum number of random measurements in each survey unit were adjusted to approximate one measurement location for every 2,000 m2 of land area. If a randomly selected location was found to be either inaccessible or unsuitable (e.g. a portion of the surface area in the instruments Field-of-View (FOV) was covered in standing water), then the location was adjusted to the closest adjacent suitable location. Due to the topography of the open land area in question, which is composed mostly of marsh land and native grasses with several areas of forest and dense brush, random locations were relocated frequently. The statement in section 2.3.4.1 of the LTP that states that approximately 21% of the surveyed areas were considered inaccessible is the consequence of this process.
Of the 691,913 square meters of surface area classified as non-impacted, 9,378 square meters were scanned by a Model 2350 paired with a Model 44-10 NaI detector and 6,608 square meters were assessed using the ISOCS. This equates to an areal coverage of approximately 2% of the total non-impacted surface area. In addition to the random locations selected, the survey design also allowed for biased scans and surface soil samples to be obtained at any location where visual cues indicated the probability of a potential burial site. It should be noted that no areas fitting this description were identified, either during the pre-survey walkdown of the areas or during the performance of the survey itself.
During the characterization of non-impacted survey units, if the result of an investigation positively identified detectable plant derived radionuclides or Cs-137 at concentrations greater that the upper background level as specified in TSD 13-004 (submitted on January 25 2016 [ZS-2016-0014]), then the survey unit or a portion of the survey unit would have been reclassified as impacted. 236 ISOCS measurements and 166 surface soil samples were taken during this survey. No plant-derived radionuclide other than Cs-137 was identified in any sample or measurement taken. For this sample population (ISOCS measurements and soil sample analysis), the value for Cs-137 averaged 0.21 pCi/g and ranged from non-detectable to a maximum reading of 0.57 pCi/g. No measurement or sample result taken during the
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 93 of 156 characterization of the non-impacted survey units prompted the acquisition of investigation samples as no investigation level was exceeded.
ZSRP believes that the scope and extent of the surveys performed in the non-impacted classified open land areas were more than sufficient to demonstrate that the non-impacted classification was appropriate for the open land areas in question and that it is highly unlikely, based upon the findings and conclusions of the HSA, combined with the results of the characterization, that plant-derived radioactivity resides in these areas.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 94 of 156 Chapter 4
- 1. Comment: Section 4.4.2 (ALARA Analysis for Remediation of Basement Structures)
Section 4.4.2 of the LTP states that in summary, the vast majority of residual radioactivity remaining in the structures after the open air demolition criteria is met and after all concrete is removed from the Containment Building basements will be located in the 542 foot elevation floor of the Auxiliary Building, and that the ALARA assessment for the remediation of basement structures will focus on the 542 foot elevation floor of the Auxiliary Building as this is the location were the greatest benefit of concrete remediation could be achieved. However, previous statements in Section 4.4.2 indicate that the lower 13 foot (~4 m) concrete bottom of the Spent Fuel Pool (SFP) and the Transfer Canal will remain following building demolition, and it is acknowledged that contamination is expected in these areas although characterization has not yet taken place. Clarification is needed on why the 542 concrete is considered to be bounding, and on the plans for additional ALARA analyses.
Basis: As discussed in the acceptance criteria/information to be submitted described in NUREG-1757, Vol. 2, Rev. 1, Section 6 (ALARA Analyses), the information supplied by the licensee should be sufficient to allow NRC staff to fully understand the basis for the licensees conclusion that projected dose limit/residual radioactivity concentrations are ALARA.
Path Forward:
- a. Justify the nature of the 542 foot elevation as a bounding condition for concrete contamination in light of uncharacterized impacted areas elsewhere.
- b. Describe the plans for any additional ALARA analyses after further characterization is completed.
ZSRP Response (HP 4.1a and 4.1b) - ZSRP contends that sufficient characterization has been performed of the Containments, Auxiliary Building, Turbine Building and Crib House/Forebay concrete end-state structures to estimate the source term, derive the radionuclide mixture and assess the dose impact of HTD radionuclides. Of these structures, surveys indicate that the vast majority of residual radioactivity remaining in the structures after all concrete is removed from the Containment Building basements will be located in the 542 foot elevation floor of the Auxiliary Building. The only other remaining end-state concrete structures that have not been fully characterized to date are the underlying concrete of the SFP and Transfer Canal. ZSRP commits to performing characterization of the bottom 12 feet of the SFP and adjoining Transfer Canal once the liner is removed and the underlying concrete is exposed. Following an assessment of the results of the characterization, a cost benefit analysis will be performed to determine if the remaining concrete will be remediated and abandoned in place or completely removed. This is stated in section 2.3.3.3 of the LTP. If the characterization indicates significant concentrations of residual radioactivity in the underlying concrete of the SFP and
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 95 of 156 Transfer Canal and, the decision is made to remediate the remaining concrete and leave it in place, then ZSRP will perform and document a separate ALARA analysis for the remediation of the SFP and Transfer Canal concrete or, evidence that the ALARA analysis of the 542 foot floor of the Auxiliary Building basement is still bounding.
As clarification, the following sentence will be added to the end of the 5th paragraph of section 4.4.2; If the future characterization of the underlying concrete of the SFP and Transfer Canal indicates significant concentrations of residual radioactivity and, if the remaining concrete will be remediated and abandoned in place, then ZSRP will perform and document a separate ALARA analysis for the remediation of the SFP and Transfer Canal concrete or, evidence that the ALARA analysis of the 542 foot floor of the Auxiliary Building is still bounding.
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- 2. Comment: Section 4.4.2.1 (ALARA Analysis Equation)
It is noted in Section 4.4.2.1 of the LTP that for the ALARA analysis for the remediation of basement structures, the equation from section 4.4.1.5 for the ratio of the concentration to the DCGLw when the total cost (CostT) is set equal to the dose averted is modified, and that the denominator must be summed and the individual dose contribution normalized to account for the multiple detectable radionuclides that are present in the radionuclide distribution for the Auxiliary Building. Several points within the calculation are unclear and are discussed as follows:
It appears from the discussion after Equation 4-10 that the final calculation is total cost (CostT) divided by the total benefit of averted dose. However, it is not evident that Equation 4-10 was properly rearranged to perform this calculation.
In Equation 4-10, summation to account for multiple radionuclides of concern is only performed in the denominator, but a decay constant () which would differ for each radionuclide of concern is in the numerator (and would need to be factored into each summation entity).
The label of dose factor in Table 4-3 and the associated reference to Table 4-1 appear to be incorrect. The values actually used in Table 4-3 and Table 4-1 appear to be inventory limits rather than dose factors.
It is not clear how the activity fractions in Table 4-3 relate to the fraction of the 25 mrem/year dose limit, and it is similarly not clear how normalization of the activity fraction times the inventory limit (as shown in Table 4-3) relates to the dose limit. The definition of fi shown under Equation 4-10 indicates that fi is the product of the Basement Inventory Levels for the Auxiliary Building for each individual Radionuclide of Concern (ROC) (from Chapter 5, Table 5-9) normalized to one, which does not seem to include the activity fraction. The steps in this calculation should be clarified. It would seem appropriate to simply set fi to the respective fraction of the dose limit represented by each radionuclide inventory.
As noted in the August 16, 2007 Federal Register Notice (72 FR 46102), Consolidated Decommissioning Guidance; Notice of Revision to, Withdrawal of Portions of, and Process for Updating NUREG-1757, the discussion on discount rates was withdrawn. As such, no discount rate should be used for the analysis, or discount rates could be applied as described in NUREG/BR-0058 (Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission), which includes a consideration for intergenerational consequences and recommends providing supplemental information and sensitivity/uncertainty analyses whenever the values of key attributes can range widely.
The area (A) of 10000 m2 used to calculate population density is based on the size of the resident farmer reference area. It would be more appropriate to consider the actual footprint of the basement areas being evaluated.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 97 of 156 Basis: As discussed in the acceptance criteria/information to be submitted described in NUREG-1757, Vol. 2, Rev. 1, Section 6 (ALARA Analyses), the information supplied by the licensee should be sufficient to allow NRC staff to fully understand the basis for the licensees conclusion that projected dose limit/residual radioactivity concentrations are ALARA.
Path Forward:
Re-evaluate the ALARA analysis presented in Section 4.4.2.1 to ensure that appropriate parameters are used and that there is a clear correlation to the unrestricted use dose limit of 25 mrem/year.
- a. It appears from the discussion after Equation 4-10 that the final calculation is total cost (CostT) divided by the total benefit of averted dose. However, it is not evident that Equation 4-10 was properly rearranged to perform this calculation.
ZSRP Response (HP 4.2a) - The following is Equation 4-10 from LTP Chapter 4:
( )( + )
=
($2,000) ( )( )(0.025)()()(1 (+) )
The following is equation (N-8) from NUREG-1757, Volume 2, Revision 1, Appendix N.
Conc CostT r DCGLW $2,000 PD 0.025 F A 1 e ( r ) N where:
PD = population density for the critical group scenario in people/m2; A = area being evaluated in square meters (m2);
0.025 = annual dose to an average member of the critical group from residual radioactivity at the Derived Concentration Guideline Level (DCGLW) concentration in rem/y; F = effectiveness, or fraction of the residual radioactivity removed by the remediation action; Conc = average concentration of residual radioactivity in the area being evaluated in units of activity per unit area for buildings or activity per unit volume for soils; DCGLW = derived concentration guideline equivalent to the average concentration of residual radioactivity that would give a dose of 25 mrem/y to the average member of the critical group, in the same units as Conc; r = monetary discount rate in units per year;
= radiological decay constant for the radionuclide in units per year; and N = number of years over which the collective dose will be calculated.
For multiple radionuclides the denominator is summed over all radionuclides as shown below:
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 98 of 156 Conc CostT DCGLW n 1 e ( r i ) N i $2 ,000 PD 0.025 f i F A r i
where:
i = radionuclide i,;
n = total of all radionuclides, and fi = dose fraction of radionuclide i Consequently, this calculation can then be expressed in the same manner as Equation 4-10 by multiplying the numerator and denominator by the variable ( + ). In Equation 4-10 in Chapter 4, the i subscript was not added to the lambda symbol but the calculation was performed applying the decay factor to each radionuclide and then summing. Equation 4-10 will be revised to correct this omission.
- b. In Equation 4-10, summation to account for multiple radionuclides of concern is only performed in the denominator, but a decay constant () which would differ for each radionuclide of concern is in the numerator (and would need to be factored into each summation entity).
ZSRP Response (4.2b) - In the equation 4-10 in Chapter 4, the i subscript was not added to the lambda symbol. Equation 4-10 will be revised to correct this omission. However, in the calculation presented in Table 4-3, the decay constant for each ROC summed in the equation was correctly used in the numerator.
- c. The label of dose factor in Table 4-3 and the associated reference to Table 4-1 appear to be incorrect. The values actually used in Table 4-3 and Table 4-1 appear to be inventory limits rather than dose factors.
ZSRP Response (4.2c) - The reviewer is correct. Table 4-3 will be revised to change Dose Factor to Inventory Limit.
- d. It is not clear how the activity fractions in Table 4-3 relate to the fraction of the 25 mrem/year dose limit, and it is similarly not clear how normalization of the activity fraction times the inventory limit (as shown in Table 4-3) relates to the dose limit. The definition of fi shown under Equation 4-10 indicates that fi is the product of the Basement Inventory Levels for the Auxiliary Building for each individual Radionuclide of Concern (ROC) (from Chapter 5, Table 5-9) normalized to one, which does not seem to include the activity fraction. The steps in this calculation should be clarified. It would seem appropriate to simply set fi to the respective fraction of the dose limit represented by each radionuclide inventory.
ZSRP Response (4.2d) - The inventory limits as specified in Table 4-1 and Column J of Table 4-3 represent the hypothetical activity concentration of each radionuclide that would equate to a dose of 25 mrem/yr in the BFM. This is the same concept as the DCGL, in that when activity for
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 99 of 156 a ROC is present at the DCGL concentration, this would equate to a dose of 25 mrem/yr. As the dose consequence for each ROC is different, the activity fraction, or radionuclide mixture presented in Column I creates a fraction of the BIL activity based upon the ROC abundance, assuming that the total dose from all ROC total 25 mrem/yr. The normalization process in Column L converts the activity fraction to a dose fraction, which equates to the variable (fi) in the equation.
- e. As noted in the August 16, 2007 Federal Register Notice (72 FR 46102), Consolidated Decommissioning Guidance; Notice of Revision to, Withdrawal of Portions of, and Process for Updating NUREG-1757, the discussion on discount rates was withdrawn. As such, no discount rate should be used for the analysis, or discount rates could be applied as described in NUREG/BR-0058 (Regulatory Analysis Guidelines of the U.S. Nuclear Regulatory Commission), which includes a consideration for intergenerational consequences and recommends providing supplemental information and sensitivity/uncertainty analyses whenever the values of key attributes can range widely.
ZSRP Response (4.2e) - Upon review of 72 FR 46102, the ALARA calculation in Table 4-3 will be revised using a value of 0 for the variable r. The revised Table 4-3 is provided below.
- f. The area (A) of 10,000 m2 used to calculate population density is based on the size of the resident farmer reference area. It would be more appropriate to consider the actual footprint of the basement areas being evaluated.
ZSRP Response (4.2f) - The area of 10,000 m2 is appropriate for this calculation as it represents the resident farmer scenario. The dose pathway for the basements is water. Consequently, if water was hypothetically pumped from a well placed in the center of the Aux Building basement, water would not be confined to the footprint of the basement.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 100 of 156 LTP Table 4-3 Revised ALARA Analysis for Volumetrically Contaminated Subsurface Structures - Auxiliary Building 542 ft.
A = 10,000 m2, r = 0.00 yr-1, N = 1,000 yr, PD = 0.0004 person/m2 Fraction of Activity removed by remedial action (F) = 1 Column A Column B Column C Column D Column E Column F Column G Column H Column I Column J Column K Column L Column M
[1-e-(r+)N] fi Column Half-Life Activity Inventory (Columns I Cost Nuclide (r+) (r+)N e-(r+)N 1-e-(r+)N /(r+) K divided (yrs)a (yr-1)a Fractionb Limita x J) Benefit by sum Co-60 5.27E+00 1.31E-01 1.31E-01 1.31E+02 7.77E-58 1.00E+00 7.60E+00 0.92% 2.28E+03 2.10E+01 9.79E-04 $1.49 Ni-63 9.60E+01 7.22E-03 7.22E-03 7.22E+00 7.33E-04 9.99E-01 1.38E+02 23.71% 8.76E+04 2.08E+04 9.70E-01 $26,845.08 Sr-90 2.91E+01 2.38E-02 2.38E-02 2.38E+01 4.54E-11 1.00E+00 4.20E+01 0.05% 7.50E+01 3.75E-02 1.75E-06 $0.01 Cs-134 2.06E+00 3.36E-01 3.36E-01 3.36E+02 7.94E-147 1.00E+00 2.97E+00 0.01% 1.59E+03 1.59E-01 7.42E-06 $0.00 Cs-137 3.02E+01 2.30E-02 2.30E-02 2.30E+01 1.06E-10 1.00E+00 4.35E+01 75.32% 8.35E+02 6.29E+02 2.94E-02 $255.65 (CostB Check Sum 100% Sum 1.24E+03 1.00E+00 $27,102.24
)
(A result < 1 would justify remediation whereas a result > 1 would demonstrate that residual radioactivity is Conc/
19.09 ALARA) DCGL Cost (in dollars) of remedial action (CostT) = $517,407.37 (a) From Table 4-1 (b) From Table 4-2
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 101 of 156 Chapter 5
- 1. Comment: Section 5 (Final Radiation Survey Plan)
Section 5 discusses the beneficial reuse of concrete that meets the non-radiological definition of clean concrete demolition debris and where radiological surveys demonstrate that the concrete is free of plant derived radionuclides above background, and also notes that radiological surveys will be performed in accordance with the guidance of NUREG-1575, Supplement 1, Multi-Agency Radiation Survey and Assessment of Materials and Equipment Manual (MARSAME)
(Reference 5-8). Additional clarification is needed on the radiological status of concrete structures which may be reused as fill material and the design of surveys of those materials.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.4 (Final Status Survey Design), licensees should provide:
A brief overview describing the FSS design A description of the background reference areas and materials, if they will be used, and a justification for their selection Path Forward:
- a. Provide additional details on the design of surveys that will be performed on reuse concrete, the radiological status of those structures that are candidate for reuse, the determination of background, and the release criteria being used.
ZSRP Response (5.1a) - LTP section 2.2 states; The decommissioning approach for ZSRP also calls for the beneficial reuse of concrete from building demolition as clean fill. The only concrete structures that will be considered are those where the probability of being contaminated is minimal. Characterization in this case will consist of an in-situ assessment of the concrete under consideration to ascertain if the structure concrete is an acceptable candidate. Demonstration that plant-derived radioactivity is not present will be accomplished with a separate survey that will be designed in accordance with NUREG-1575, Supplement 1, Multi-Agency Radiation Survey and Assessment of Materials and Equipment Manual (MARSAME).
LTP section 2.3.1.2 states; In March of 2012, ZSRP commenced an assessment of the concrete 594 foot and 559 foot elevation floors and lower walls of the Crib House. The DQOs established for this survey were to establish a background threshold range for volumetric concrete at ZNPS, evaluate the basement foundations and floors of the Crib House for the presence of volumetric radiological contamination, and to provide a sufficient quantity and quality of uncontaminated concrete media representative of the Basement Fill concrete to an off-site vendor for the derivation of distribution coefficients for the radionuclides of concern.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 102 of 156 The results of the background study for concrete are presented in Table 2-9. The concrete structures that ZSRP believes are acceptable candidates for reuse as fill are the outer shell of the Containment Buildings, the concrete portions of the Turbine Building (including the Steam Tunnels) above the 588 foot, the Crib House, portion of the Forebay above the 588 foot elevation, the Service Building and minor ancillary structures that will be completely demolished such as the Interim Radioactive Waste Storage Facility (IRSF) (which was never utilized to store radioactive waste), the Mechanical Maintenance Training Center (MMTC) and Warehouse, the Fire Maze complex, the NGET building, the ENC building, the south Warehouse and the North Security Access Gatehouse. The disposition of these buildings and the potential reuse of the concrete from their demolition are addressed in LTP sections 3.2.3, 3.3.2, 3.3.4, 3.3.7.2 and 3.3.7.5.
Surveys will be performed of candidate structural concrete prior to demolition. Unconditional release surveys will be designed and performed in accordance with ZionSolutions procedure ZS-LT-400-001-001, Unconditional Release Materials, Equipment and Secondary Structures, provided in Enclosure 2. This procedure provides guidance for performing Unconditional Release Surveys (URS) on secondary side structures, systems, and miscellaneous Materials and Equipment (M&E) prior to demolition activities, and documents that the material is suitable for unconditional release from, or for reuse at ZNPS. In accordance with the procedural requirements, material shall be deemed to contain, or be contaminated, with plant-related radioactivity if radiological surveys and/or sample analyses positively identify plant-related radioactivity. For solid materials, the required MDCs for scan measurements and smears shall be no greater than the corresponding limits in NRC I.E. Circular No. 81-07. For the analysis of volumetric activity in solids/sludge using gamma spectroscopy analyses, the analysis MDCs shall be no greater than the MDCs derived from ODCM Chapter 12, Table 12.5-3 for sediments.
If, during the performance of the unconditional release survey, residual radioactivity is positively detected, then the candidate concrete will be disqualified as an acceptable material for reuse as fill and will be controlled and properly disposed of as waste. Concrete that is demonstrated as acceptable for reuse as fill will be controlled through the demolition process and stockpiled as fill. The same posting and access control methods as specified in LTP section 5.6.3 and ZionSolutions procedure ZS-LT-300-001-003, Isolation and Control for FRS (provided in ) will be used to maintain control of the stockpiled material up to the time it is used as fill. Also the filled voids containing the reused material will be subject to FSS as an open land survey unit commensurate with the classification of the area in which it resides.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 103 of 156
- 2. Comment: Section 5.1 (Radionuclides of Concern and Mixture Fractions), Section 5.7.1.5.2 (Sampling of Subsurface Soils during FSS) and Section 5.7.1.5.3 (Sampling of Subsurface Soils below Structure Basement Foundations)
Section 5.1 of the LTP states that as currently inaccessible soils, structures and piping systems are made accessible and are surveyed, either by characterization or by Remedial Action Support Survey (RASS), the survey results will be reviewed to ensure that the suite of ROC and radionuclide mixture derived for the Auxiliary Building is applicable.
Section 5.7.1.5.2 of the LTP notes that the HSA as well as the results of the extensive characterization of subsurface soils in the impacted area surrounding the Zion facility have shown that there is minimal residual radioactivity in subsurface soil, and that consequently, Zion proposes to perform minimal subsurface sampling during FSS. However, Chapter 2 of the LTP has also indicated that the assessment of potential subsurface soil contamination is not currently complete, and that the survey of many inaccessible or not readily accessible subsurface soils or surfaces has been deferred. Examples of areas where surveys are deferred were provided as follows: soils under structures, soils under concrete or asphalt coverings, structural wall and floor surfaces in the basements of structures that will remain and be subjected to FRS, the remaining surfaces of the SFP and Transfer Canal after liner removal, the interiors of embedded and/or buried pipe that may remain and the interior and exterior of both Containment domes. Additionally, Section 5.7.1.5.3 of the LTP indicates that, prior to license termination, it will be necessary to ascertain the radiological conditions of sub-slab soils beneath several areas that will remain at the time of license termination (the foundation walls and basement floors below the 588 foot elevation of the Unit 1 Containment, Unit 2 Containment, Auxiliary Building, Turbine Building, Crib House/Forebay, WWTF and remnants of the SFP) to demonstrate suitability for unrestricted release.
Clarification is needed on the manner in which FSS/remediation plans and strategies will be revised in the event that contaminated soils, structures and piping systems are found during additional characterization or FSS sampling activities Basis: As discussed under Characterization Surveys in NUREG-1757, Vol. 2, Rev. 1, Section 4.2 (Scoping and Characterization Surveys), the objectives of characterization surveys include:
Determining the nature and extent of residual radioactivity Evaluating remediation alternatives (e.g., unrestricted use, restricted use, onsite disposal, offsite disposal)
Developing input to the FSS design Path Forward:
- a. Describe how remediation plans/strategies and the FSS design will be revised in the event that contaminated soils, structures and piping systems are found during additional characterization, RASS, or FSS sampling activities.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 104 of 156 ZSRP Response (5.2a) - ZSRP contends that sufficient characterization samples have been taken of the Containment structures, Auxiliary Building, Turbine Building and Crib House/Forebay concrete to estimate the source term, derive the radionuclide mixture and assess the dose impact of HTD radionuclides. The only remaining structure that has not been fully characterized to date is the underlying concrete of the SFP and Transfer Canal. When the underlying concrete in the SFP and Transfer Canal becomes accessible, the concrete surfaces will be characterized.
ZSRP also contends that sufficient characterization has been performed of the surface and subsurface soils surrounding ZNPS. Several areas do remain that will need assessment, all pertaining to subsurface soils in Class 1 areas with accessibility contingent upon the demolition of standing structures. As with the concrete, as this soil becomes accessible, it will also be characterized. All surface soil at ZNPS has been adequately characterized and additional characterization of surface soil is not anticipated during continuing characterization.
ZSRP will revise section 5.1 of LTP Chapter 5 to read as follows:
Sufficient characterization samples have been taken of the Containment structures, Auxiliary Building, Turbine Building and Crib House/Forebay concrete to derive the radionuclide mixture and assess the dose impact of HTD radionuclides. The only remaining end-state concrete structure that has not been fully characterized to date is the concrete under the steel liner of the SFP/Transfer Canal. When the underlying concrete of the SFP/Transfer Canal is exposed by the removal of the steel liner, the concrete will be characterized in the same manner as the other end-state concrete structures (surfaces will be scanned and concrete core samples will be taken at the locations of the highest scan results). Continuing characterization will also be performed in several potentially contaminated embedded pipe systems that will be abandoned in place, specifically the floor drains in the 542 foot elevation basement floor of the Auxiliary Building and the Core Spray penetrations between the Containment basements and the Auxiliary Building basement. When the interior surfaces of these pipes systems become accessible, samples will be taken of any loose surface debris in the pipe. In both of these cases, the concrete core and/or debris samples will be analyzed for the presence of HTD radionuclides. If the sample analysis indicates the presence of a significantly different radionuclide mixture from the mixture derived for the Auxiliary Building floor (Table 5-2), then the unique mixture will be documented and applied to the SFP/Transfer canal and/or embedded pipe systems as applicable during survey design and STS.
If a sample and/or measurement is taken on any other end-state structure or embedded pipe system to support decommissioning activities, Radiological Assessments (RA) or Remedial Action Support Surveys (RASS), and the result indicates a SOF in excess of 0.5 based on gamma spectroscopy results, then a sample will be collected at the location of the highest accessible individual measurement and analyzed for HTD radionuclides. If any continuing characterization surveys taken in soil or buried pipe indicate the presence of gamma-emitting
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 105 of 156 radionuclides at concentrations in excess of a SOF of 0.5, then the samples will be analyzed for the presence of HTD radionuclides. In these unlikely situations, if the analysis indicates the presence of HTD radionuclides (other than Ni-63 and Sr-90, which are known to be present) at detectable concentrations, then additional investigation/sampling will be performed.
Based upon the analysis of radionuclide fractions and dose contribution in TSD 14-019, the dose contribution from HTD fractions is expected to be very low in all media (concrete, soil, embedded pipe, buried pipe, penetrations) at a SOF of 0.5 with even the most extreme HTD ratios. In the unlikely situation where these investigation levels are exceeded and one or more HTD radionuclides other than Ni-63 or Sr-90 are positively identified, then the dose impact of the positive HTD radionuclide(s) will be assessed. Additional samples may be collected and analyzed for HTD radionuclides to support the assessment of the dose impact.
Section 2.3.3.3 of LTP Chapter 2 states the following pertaining to the SFP and Transfer Canal:
The only portion of the building that resides below the 588 foot elevation is the bottom 12 feet of the SFP and adjoining Transfer Canal. As part of the building demolition, the steel liner will be removed from the SFP and Transfer Canal. Once the liner is removed and the underlying concrete is exposed, additional characterization surveys will be performed to assess the radiological condition of the underlying concrete pad and remaining pool walls. Following an assessment of the results of the survey, a cost benefit analysis will be performed to determine if the concrete will be remediated and left in place or removed. Any remaining concrete surfaces will be remediated to levels commensurate with inventory limits developed with the BFM that represent the dose criterion for unrestricted release specified in 10 CFR 20.1402.
Consequently, as with other end-state structures, if the characterization indicates that any significant concentrations of residual radioactivity reside in the underlying concrete of the SFP and Transfer Canal and, the decision is made to leave the remaining concrete in place, then the surfaces will be remediated to the open air demolition criteria presented in ZionSolutions TSD 10-002, Technical Basis for Radiological Limits for Structure/Building Open Air Demolition (provided on 11/12/2015 [ZS-2015-0163]) and then subjected to STS using the Basement Dose Factors for the Fuel Building to demonstrate compliance.
Section 2.3.3.7 of LTP Chapter 2 states the following pertaining to embedded and buried pipe:
Several sections of embedded and buried piping located below the 588 foot elevation have been designated to remain following demolition as part of the end-state condition for the structures in which they reside. At the time of LTP submittal, the interior surfaces of most of these sections of pipe were not accessible. As decommissioning progresses and access is achieved, radiological surveys will be performed to assess if any remediation is necessary, to confirm the radiological distribution inside of the pipe and to assess the dose from residual radioactivity remaining in the pipes.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 106 of 156 For pipe embedded in concrete, the pipe interiors will be remediated to levels commensurate with inventory limits developed with the BFM that represent the dose criterion for unrestricted release specified in 10 CFR 20.1402. In most cases, these sections of pipe will primarily consist of penetrations through the remaining concrete walls of the structure. A list of penetrations that is anticipated to remain as part of the end-state condition of Unit 1 and Unit 2 Containment is presented in Table 2-26. As decommissioning progresses and access is achieved to the interior of these pipe sections, STS will be performed in accordance with section 5.5.5 of this LTP.
For pipe buried in soil, the pipe interiors will be remediated to levels less than the site-specific DCGLs presented in section 5.2.3 of this LTP. FSS surveys will be performed as described in section 5.5.1.8. A list of the sections of buried piping that will be abandoned in place and surveyed during FSS is presented in Table 2-27.
Section 5.7.1.5 of LTP Chapter 5 states the following pertaining to subsurface soils:
During decommissioning of Zion, any subsurface soil contamination that is identified by continuing characterization or operational radiological surveys that is in excess of the site specific (DCGLw) for each of the potential ROC as presented in Table 5-2 will be remediated.
The remediation process will include performing RASS of the open excavations in accordance with section 5.4.2 of this FRS Plan. 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 rigors of FSS. The data obtained during the RASS is expected to provide a high degree of confidence that the excavation, or portion of the excavation, meets the criterion for the unrestricted release of open land survey units.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 107 of 156
- 3. Comment: Section 5.1 (Radionuclides of Concern and Mixture Fractions), Section 5.2.4 (Surrogate Radionuclides)
As previously discussed in an RAI for Chapter 2, Hard-to-Detect (HTD) radionuclides were only measured during characterization when an elevated direct measurement was established, and it is not clear that a surrogate correlation had been established between HTD and easy-to-detect radionuclides. Similar methodology is presented in Chapter 5 for use during FSS. Without an established surrogate in place NRC staff cannot assess the viability of this relationship and cannot evaluate whether it is an acceptable method to quantify HTD radionuclides during FSS.
This relationship is discussed several times in Chapter 5 as follows:
Section 5.1 of the LTP states the concrete samples taken from the underlying concrete of the SFP and Transfer Canal will be analyzed for the presence of Hard-to-Detect (HTD) radionuclides. For other structural surfaces, embedded piping and penetrations included in the BFM, if surveys indicates that the potential dose from residual radioactivity in a structure could exceed 10% of the dose limit (2.5 mrem/yr), then the samples will be analyzed for the presence of HTD radionuclides. If surveys indicate the presence of gamma-emitting radionuclides at concentrations greater than 50% of a DCGL in soils or buried pipe, then the samples will be analyzed for the presence of HTD radionuclides.
Section 5.2.4 of the LTP states as previously discussed in section 5.1, the radionuclide mixture for Auxiliary Building concrete developed in TSD 14-019 and listed in Table 5-2 are the scaling factors that will be used to determine the surrogate relationship, but further notes that once appropriate scaling factors are determined, the DCGL of the measured radionuclide is modified to account for the represented radionuclide(s) according to the following equation from section 4.3.2 of MARSSIM.
Sections 5.7.1.5.2 and 5.7.1.5.3, which discuss subsurface soil sampling, also indicate that if gamma spectrometry analysis indicates the presence of plant-derived gamma emitting radionuclides at concentrations greater than the DCGLw, then the sample will also be analyzed for the potential presence of HTD ROC.
Additional clarification is needed on the establishment of surrogate relationships at the site to determine the acceptability of proposed sampling strategies.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.2 (Scoping and Characterization Surveys), licensees should provide:
For sites, areas, or buildings with multiple radionuclides, a discussion justifying the ratios of radionuclides that will be assumed in the FSS or an indication that no fixed ratio exists and each radionuclide will be measured separately (note that this information may be developed and refined during decommissioning and licensees may elect to include a plan to develop and justify final radionuclide ratios in the DP).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 108 of 156 Path Forward:
- a. Describe the correlation between direct measurement surveys and the presence of HTD radionuclides, and provide adjusted DCGLs that will be used during FSS to account for HTD radionuclides. However, if the licensee intends to develop and refine surrogate ratios during decommissioning a plan to develop and justify final radionuclide ratios should be provided in the LTP, which should include an evaluation of the acceptability of the currently proposed sampling strategies (it may prove useful to consider discussions in MARSSIM Section 4.3.2 on DCGLs and the use of surrogate measurements).
ZSRP Response (5.3a) - ZSRP contends that sufficient characterization samples have been taken of the Containment structures, Auxiliary Building, Turbine Building and Crib House/Forebay concrete to adequately derive the radionuclide mixture and assess the dose impact of HTD radionuclides. The only remaining end-state concrete structure that has not been fully characterized to date is the concrete under the steel liner of the SFP/Transfer Canal. When the underlying concrete of the SFP/Transfer Canal is exposed by the removal of the steel liner, the concrete will be characterized in the same manner as the other end-state concrete structures (surfaces will be scanned and concrete core samples will be taken at the locations of the highest scan results).
Continuing characterization will also be performed in several potentially contaminated embedded pipe systems that will be abandoned in place, specifically the floor drains in the 542 foot elevation basement floor of the Auxiliary Building and the Core Spray penetrations between the Containment basements and the Auxiliary Building basement. When the interior surfaces of these pipes systems become accessible, samples will be taken of any loose surface debris in the pipe. In both of these cases, the concrete core and/or debris samples will be analyzed for the presence of HTD radionuclides. If the sample analysis indicates the presence of a different radionuclide mixture from the mixture derived for the Auxiliary Building floor (Table 5-2), then the unique mixture will be documented and applied to the SFP/Transfer canal and/or embedded pipe systems as applicable during survey design and STS.
Elevated measurements were not detected in any characterization sample taken outside of a Class 1 area at Zion. ZSRP contends that there is no reasonable or plausible scenario at Zion where a HTD ROC would be present in any dose significant concentration without the presence of a plant-derived gamma emitting ROC. In the cases where gamma emitting radionuclides were detected in any sample at a significant concentration greater than MDC, then the sample was analyzed for HTDs ROC. This occurred in only 9 surface soil samples and 1 subsurface soil sample, all taken in Class 1 soils. The result of these analyses is presented in Table 2-34 in Chapter 2.
The mixture percentages for the initial suite of radionuclides for Containment and Auxiliary Basement concrete were developed in TSD 14-019 (provided on November 12 2015 [ZS-2015-0163]) using the results of concrete core sample analyses. The mixture fractions for the HTD
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 109 of 156 ROC were determined by analyzing concrete core samples that contained the highest radionuclide concentrations. This ensured that the mixture percentages were not overly influenced by the MDC values of radionuclides that were not positively detected. Concrete core samples and measurements taken from other end-state structures (e.g. Turbine Building), as well as surface and subsurface soils taken from impacted areas indicated very low concentration of residual radioactivity. As current characterization data indicates the lack of any significant residual radioactivity concentrations in other structures or soil, the accurate determination of mixture percentages or, surrogates for HTD ROC is not feasible. Consequently, given the lack of available data and the very low levels of residual radioactivity expected to remain, the radionuclide mixture for the Auxiliary Building was considered to be a reasonable mixture to apply to structures other than the Auxiliary Building, soils, buried piping, embedded piping and penetrations.
During the performance of surveys to support decommissioning activities, or during Radiological Assessments and/or RASS, if a sample and/or measurement is taken with a result that indicates a SOF in excess of 0.5 based on gamma spectroscopy results, then a sample will be collected and analyzed for HTD radionuclides.
For FSS and STS, ZSRP does not propose to analyze for HTD radionuclides. ZSRP will infer the presence of Ni-63 and Sr-90 using the surrogate approach specified in section 5.2.4. H-3 will be added as an additional HTD radionuclide that will be inferred using the surrogate approach for the STS of the Containment basements. The scaling factors that will be used are presented in Table 5-2 (and Table 6-3 for the Containment basements).
As a clarification, the last paragraph of section 5.1 will be deleted and replaced with the following text. This was done to specify that the action levels cited for the assessment of HTD radionuclides is specific to continuing characterization surveys and do not apply to FRS.
Sufficient characterization samples have been taken of the Containment structures, Auxiliary Building, Turbine Building and Crib House/Forebay concrete to derive the radionuclide mixture and assess the dose impact of HTD radionuclides. The only remaining end-state concrete structure that has not been fully characterized to date is the concrete under the steel liner of the SFP/Transfer Canal. When the underlying concrete of the SFP/Transfer Canal is exposed by the removal of the steel liner, the concrete will be characterized in the same manner as the other end-state concrete structures (surfaces will be scanned and concrete core samples will be taken at the locations of the highest scan results). Continuing characterization will also be performed in several potentially contaminated embedded pipe systems that will be abandoned in place, specifically the floor drains in the 542 foot elevation basement floor of the Auxiliary Building and the Core Spray penetrations between the Containment basements and the Auxiliary Building basement. When the interior surfaces of these pipes systems become accessible, samples will be taken of any loose surface debris in the pipe. In both of these cases, the concrete core and/or debris samples will be analyzed for the presence of HTD radionuclides. If the sample analysis
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 110 of 156 indicates the presence of a different radionuclide mixture from the mixture derived for the Auxiliary Building floor (Table 5-2), then the unique mixture will be documented and applied to the SFP/Transfer canal and/or embedded pipe systems as applicable during survey design and STS.
If a sample and/or measurement is taken on any other end-state structure or embedded pipe system to support decommissioning activities, Radiological Assessments (RA) or Remedial Action Support Surveys (RASS), and the result indicates a SOF in excess of 0.5 based on gamma spectroscopy results, then a sample will be collected at the location of the highest accessible individual measurement and analyzed for HTD radionuclides. If any continuing characterization surveys taken in soil or buried pipe indicate the presence of gamma-emitting radionuclides at concentrations in excess of a SOF of 0.5, then the samples will be analyzed for the presence of HTD radionuclides. In these unlikely situations, if the analysis indicates the presence of HTD radionuclides (other than Ni-63 and Sr-90, which are known to be present) at detectable concentrations, then additional investigation/sampling will be performed.
Based upon the analysis of radionuclide fractions and dose contribution in TSD 14-019, the dose contribution from HTD fractions is expected to be very low in all media (concrete, soil, embedded pipe, buried pipe, penetrations) at a SOF of 0.5 with even the most extreme HTD ratios. In the unlikely situation where these investigation levels are exceeded and one or more HTD radionuclides other than Ni-63 or Sr-90 are positively identified, then the dose impact of the positive HTD radionuclide(s) will be assessed. Additional samples may be collected and analyzed for HTD radionuclides to support the assessment of the dose impact.
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- 4. Comment: Section 5.4.3 (Field Screening Methods for RASS of Below-Grade Structural Surfaces)
Section 5.4.3 of the LTP indicates that:
All structural surfaces will be remediated to the open air demolition limits prior to structural demolition. Confirmatory radiological surveys will be performed using approved procedures following remediation and prior to demolition to ensure that contamination levels are acceptable.
The radiological surveys will include extensive surveys on the structural surfaces (walls and floors) located below the 588 foot elevation that will remain. These surveys will be performed using conventional gamma instruments in typical scanning and measurement modes. Scanning coverage for pre-remediation surveys on structures prior to open air demolition could include up to 100% of the accessible surface area depending on the contamination potential. Consequently, the pre-remediation surveys performed to prepare building surfaces for open air demolition will provide confidence that structural surfaces that have significant elevated activity will be removed. Once remediation is complete, structural surfaces located above the 588 foot elevation and non-load-bearing interior concrete walls below the 588 foot elevation that contain detectable residual radioactivity will be demolished, reduced in size, packaged and shipped off-site to a licensed disposal facility.
This discussion is provided under the heading of Field Screening Methods for RASS of Below-Grade Structural Surfaces. Remedial Action Support Surveys (RASS) are intended to demonstrate that remediation is complete (i.e., the dose criterion has been met and areas are ready for FSS). Clarification should be provided on how these surveys meet that goal. It is also not clear how the open air demolition limits relate to the unrestricted release criterion of 25 mrem/yr.
Basis: As discussed under Areas of Review in NUREG-1757, Vol. 2, Rev. 1, Section 4.3 (Remedial Action Support Surveys), the purpose of the review of the description of the remedial action support surveys is to verify that the licensee has designed these surveys appropriately and to assist the licensee in determining when remedial actions have been successful and that the FSS may commence.
Path Forward:
- a. Clarify how the surveys described in Section 5.4.3 of the LTP will assist the licensee in determining when remedial actions have been successful to meet the unrestricted release criterion and that the FSS may commence.
ZSRP Response (5.4a) - The open air demolition limits have no direct correspondence to the 25 mrem/yr unrestricted use dose criterion. However, TSD 14-019, Table 7 provides a conservative estimate of the current inventory in the Auxiliary Building basement prior to remediation to open air demolition limits. As shown in the ZSRP response to RAI PAB 8f, the estimated dose to the AMCG, assuming the current inventory in the Auxiliary basement, before any remediation is
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 112 of 156 performed, is 19 mrem/yr. Therefore, ZSRP has high confidence that after surveys and remediation are performed to meet the open air demolition limits, remaining residual radioactivity will result in a dose below 25 mrem/yr. The post remediation RASS surveys that will demonstrate the open air demolition criteria are met will also serve to confirm that the survey unit is ready for STS.
- b. Provide ZionSolutions Technical Support Document 10-002, Technical Basis for Radiological Limits for Structure/Building Open Air Demolition.
ZSRP Response (5.4b) - TSD 10-002, Technical Basis for Radiological Limits for Structure/Building Open Air Demolition was submitted to the NRC on November 12, 2015 (ZS-2015-0163).
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- 5. Comment: Section 5.5 (Source Term Survey) and Section 5.5.2.1 (STS Areal Coverage)
Section 5.5 of the LTP states that the BFM is a mixing model with a source term based on total inventory that is independent of concentration levels and areal distribution of residual radioactivity, and that the typical scan coverage guidance as presented in MARSSIM, which relies on concentration based DCGLW and DCGLEMC values, is not directly applicable. Section 5.5 further notes that the standard approach in MARSSIM for calculating AFs in conjunction with DCGLW values to determine the acceptability of elevated areas of activity does not apply to the BFM inventory source term which is independent of concentration levels and areal distribution.
Additional discussion in Section 5.5.2.1 notes that a secondary consideration [of STS areal coverage] is the potential for the presence of small areas of elevated radioactivity in a STS survey unit that could exceed the BIL, and that the areal coverage of the STS should be commensurate with the probability that a small area of elevated radioactivity could exist within a STS survey unit in a concentration exceeding the BIL and the likelihood that such an area would be detected by the STS ISOCS measurements.
The stated threshold for areas of elevated activity is the BIL, which was defined as a maximum activity level for an entire survey unit. This implies that a single elevated area of activity can exceed the total maximum value assumed for the entire survey unit. It is additionally unclear how the BIL is correlated to 25 mrem per year for comparison to the unrestricted release criterion.
Justification should be provided for the usage of the BIL as the threshold value for small areas of elevated radioactivity in the survey design. Clarification should be provided on the treatment of multiple elevated areas in one survey unit.
Basis: As discussed in the Final Status Survey Design described in NUREG-1757, Vol. 2, Rev. 1, Section 4.4, two objectives of the FSS are to:
Demonstrate that the potential dose from residual radioactivity is below the release criterion for each survey unit, and Demonstrate that the potential dose from small areas of elevated activity is below the release criterion for each survey unit Path Forward:
- a. Describe the correlation of the BIL to a 25 mrem per year release criterion.
ZSRP Response (5.5a) - As stated in LTP Chapter 5, section 5.5.2.1, For the purpose of STS design, the Basement Dose Factors were used to calculate a hypothetical maximum inventory level the ROC in each basement as listed in Table 5-9. These calculated values, which are designated as Basement Inventory Levels (BIL), do not represent the inventory levels expected to
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 114 of 156 remain at license termination. An example of how the BIL is calculated is presented below using Cs-137 in the Auxiliary Building Basement.
From LTP Chapter 5, Table 5-3, the BFM Dose Factor for Cs-137 in Auxiliary Building basement is 3.00E-02 mrem/yr per mCi. To calculate the maximum hypothetical inventory level (BIL) for Cs-137, i.e., the inventory that would result in 25 mrem/yr to the AMCG, the BFM Dose Factor is inverted and multiplied by 25 mrem/yr. Table 5-9 lists the BIL for Cs-137 in the Auxiliary Building basement as 835 mCi. The calculation of this BIL is shown below (note that the calculation as shown results in 833 mCi which is due to rounding error; the 835 mCi value listed in Table 5-9 is correct):
1
() = 25 = 835 3.00 02
- b. Provide justification for the usage of the BIL as the threshold value for small areas of elevated radioactivity in the survey design.
ZSRP Response (5.5b) - The statement that the BIL is the upper threshold for a small area of activity is intended to simply acknowledge that the dose to the AMCG is independent of the area containing the residual radioactivity and that the theoretical maximum inventory that could remain in an individual elevated area is the BIL. In practice this would not occur because all of the activity would not be in one area.
The purpose of the discussion in LTP Chapter 6 section 5.5.2.1 is simply to provide a threshold, i.e. the BIL, against which to judge the potential for a significant elevated area to be present.
The areal coverage provided by the STS in-situ gamma spectroscopy measurements is commensurate with the potential for a significant elevated area to be present (i.e., one approaching 25 mrem/yr or some significant fraction of 25 mrem/yr), consistent with a graded, risk-informed approach.
- c. Describe how multiple areas in one survey unit will be accounted for if the threshold value for a single elevated area is the maximum activity level for the entire unit.
ZSRP Response (5.5c) - As stated in LTP Chapter 5, section 5.5, The BFM is a mixing model with a source term based on total inventory that is independent of concentration levels and areal distribution of residual radioactivity. The total inventory of residual radioactivity must be less than the BIL and is accounted for by summing the inventory from all basement areas including any elevated areas. There is no direct analogy to the standard MARSSIM process where a DCGLEMC is used in conjunction with a DCGLw where an elevated area represents a dose that must be accounted for separately and in addition to the average dose based on the mean of the measurements in a given survey unit. As described in TSD 14-022, Use of In-Situ Gamma Spectroscopy for Source Term Survey of End State Structures (submitted to the NRC on May 27, 2015 [ZS-2015-0078]), the ISOCS measurement method for STS conservatively accounts for
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 115 of 156 elevated areas that may be randomly distributed within the ISOCs field of view (FOV) in the measurement of total activity for compliance with the BFM inventory limits.
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- 6. Comment: Section 5.5.1 (Instruments Selected for Performing STS)
Section 5.5.1 of the LTP notes that the Canberra ISOCS has been selected as the primary instrument that will be used to perform STS, and that direct beta measurements taken on the concrete surface will not provide the data necessary to determine the residual radioactivity inventory at depth in concrete and therefore, would have to be augmented with core sampling.
Section 5.5.1 further indicates that the Auxiliary Building 542 foot elevation floor (and possibly the remaining concrete of the SFP/Transfer Canal) is the only structural surface that will remain in the end-state where excessive variability in the geometry of residual radioactivity has been detected at depth, and that concrete core sampling and/or scan surveys may also be used to confirm the areal and depth distribution of activity in concrete in support of ISOCS geometry assumptions and sensitivity analysis.
Additional clarification is needed on the design of core sampling surveys and the methods being utilized to determine the areal and depth distribution of activity in concrete.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.4 (Final Status Survey Design), licensees should provide:
A brief overview describing the FSS design A description of scanning instruments, methods, calibration, operational checks, coverage, and sensitivity for each media and radionuclide Path Forward:
- a. Describe how surveys will be designed to utilize supplementary core sampling along with ISOCS measurements, including the determination that a representative number of core samples will be taken.
ZSRP Response (5.6a) - Additional core samples would be a part of supplemental characterization and not FRS. There are no plans at this time to acquire any additional concrete core samples from end-state concrete for characterization with the exception of the SFP/Transfer Canal concrete when exposed following the removal of the steel liner. TSD 14-022, Use of In-Situ Gamma Spectroscopy for Source Term Survey of End State Structures (submitted to the NRC on May 27, 2015 [ZS-2015-0078]) describes in detail the development of calibration geometries for the Auxiliary Building basement based on core sample data. TSD 14-022 also selects calibration geometries for the remaining basements, with the exception of the SFP/Transfer Canal end-state concrete, which has not yet been characterized.
With the exception of the characterization of the SFP/Transfer Canal end-state concrete, the decision to obtain any additional concrete core samples from other end-state concrete structures would be made only if a condition were encountered during development of the STS survey design where the condition of the concrete surfaces following any remediation or demolition activity appeared to be significantly inconsistent with the depth profile and geometries assumed
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 117 of 156 in TSD 14-022. If any additional cores were collected, they would be based on engineering judgment to identify the depth profile and geometry in the area under investigation to ensure that the ISOCS efficiency calibration assumptions described in TSD 14-022 remain conservative or to produce new efficiency calibrations if necessary.
The collection of supplemental characterization concrete core samples of the end-state SFP/Transfer Canal concrete when exposed will follow the same approach used for characterization of the other basements end-state concrete. The objective of core sampling will be to determine the contamination depth profile in the area being investigated. The locations where contamination is expected to be present at the greatest depth are those with the highest surface dose rates which will be identified by scan surveys. Therefore, core samples would be taken at biased locations that exhibited the highest activity. This approach ensures that the areas with the greatest contamination at depth are identified, which is the primary information required to demonstrate that the ISOCS efficiency calibration geometries are conservative. As described in detail in TSD 14-022, ISOCS efficiencies decrease with increased contamination at depth.
- b. Justify the usage of scan surveys as a method to confirm the depth distribution of activity in concrete.
ZSRP Response (5.6b) - Scan surveys would not be used to confirm the depth distribution of activity but only to identify the area and boundaries of elevated activity that can then be subjected to core sampling to determine the depth distribution.
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- 7. Comment: Section 5.5.2 (STS Survey Units)
Section 5.5.2 of the LTP indicates that:
STS will be designed and documented in the same manner as a traditional FSS and performed in accordance with approved procedures and in compliance with FSS quality requirements in the QAPP. The BFM is not sensitive to the concentration or areal distribution of residual radioactivity. Therefore, there is no intrinsic survey unit size limitation for the basement structures analogous to MARSSIM recommended survey unit sizes, which are based on a building occupancy scenario.
It is indicated that there is no intrinsic survey unit size limitation for basement structures. As such, areas indicated as Class 1 and 2 (as noted in Table 5-10) do not correlate to recommended survey areas from MARSSIM.
Basis: MARSSIM Section 4.6 states that the limitation on survey unit size for Class 1 and Class 2 areas ensures that each area is assigned an adequate number of data points, and that the rationale for selecting a larger survey unit area should be developed using the DQO Process (Section 2.3) and fully documented.
Path Forward:
- a. Describe the rationale for selecting a larger survey unit area and how the DQO process was used for that purpose.
ZSRP Response (5.7a) - MARSSIM section 4.6 states that survey unit size should be limited based on classification, exposure pathway modeling assumptions and site-specific conditions.
As stated in section 5.5.2, the BFM is not sensitive to the concentration or areal distribution of residual radioactivity. Therefore, there is no intrinsic survey unit size limitation for the basement structures analogous to MARSSIM recommended survey unit sizes, which are based on a building occupancy scenario. The statement in section 5.5.2 is consistent with MARSSIM guidance regarding justification for alternate survey unit sizes.
For Class 1 basement STS units, no additional DQO analysis regarding the adequate number of data points is necessary because 100% of the surface areas in Class 1 STS units will be subjected to quantitative measurement with ISOCS. No greater number of data points could be acquired regardless of survey unit size. In addition, the statistical parameters and tests proposed for the Class 1 STS units are directly from MARSSIM DQO guidance.
For Class 2 STS units, the MARSSIM DQO process for determining sample size was followed and documented in LTP Chapter 5. Consequently, additional DQO documentation of the statistical parameters and tests were not considered necessary. The only remaining issue to address is whether there will be an adequate number of data points in the Class 2 STS units with areas that exceed the suggested survey unit sizes in MARSSIM.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 119 of 156 This statement regarding adequate number of data points in MARSSIM was based on qualitative deliberations by the MARSSIM committee which centered on concerns as to whether very small sample sizes using typical 100 cm2 gas flow proportional counters provided adequate areal coverage for quantitative measurements. This concern was predominantly addressed by the use of scan surveys to augment the small areal coverage represented by the quantitative measurements with 100 cm2 detectors.
The concern of small areal coverage by quantitative measurements, i.e. adequate number of data points, is significantly reduced by the use of ISOCS which provides quantified measurements with much larger areal coverage than was deemed acceptable in MARSSIM. For example, consider the smallest recommended structural Class 2 survey unit size of 100 m2. A statistical sample size of 14 measurements would correspond to 1400 cm2 areal coverage by quantitative methods using the industry standard 100 cm2 gas flow proportional counters use at the time of MARSSIM development, or 1.4% of the survey unit surface area. At the maximum recommended Class 2 survey unit size of 1000 m2 this would represent 0.14% areal coverage.
The largest Class 2 survey unit size for Zion basements is 3,912 m2 for the Auxiliary basement walls. However, the 14 ISOCS measurements that will be collected from the Auxiliary Building basement wall STS unit to satisfy the statistical DQO requirements corresponds to quantitative measurement areal coverage of 10% of the survey unit. The areal coverage by ISOCS is higher by a factor of approximately 70 times than what would be achieved using MARSSIM recommended sample sizes and industry standard measurement methods in use at the time. The increased areal coverage of quantified measurements proposed in the LTP Chapter 5 coupled with the sample size justification using MARSSIM recommended DQO methods and statistical parameters provides a robust demonstration that the proposed number of data points are adequate and, in fact, exceed what is considered acceptable by the MARSSIM guidance.
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- 8. Comment: Section 5.5.2.1 (STS Areal Coverage)
Section 5.5.2.1 of the LTP notes that:
For the purpose of STS design, the Basement Dose Factors were used to calculate a hypothetical maximum inventory level the ROC in each basement as listed in Table 5-9. These calculated values, which are designated as Basement Inventory Levels (BIL), do not represent the inventory levels expected to remain at license termination. The actual mean inventory determined by performing the STS will be used as the source term in the BFM to determine the dose to the AMCG. As described below, the actual inventory expected to remain in the basements is a small fraction of the BIL.
The BIL is used during STS survey design to determine a reasonable areal coverage based upon the theoretical potential of exceeding the inventory level based upon characterization survey data.
The primary consideration for determining STS areal coverage is the potential for the presence of a residual radioactivity inventory in a STS survey unit that could exceed the BIL. Survey areal coverage can be a low percentage of the total wall and floor surface area in survey units that have a low potential for exceeding the BIL. Conversely, in areas with a high potential for approaching the BIL, a higher percentage of coverage is justified.
Clarification is needed on the manner in which scanning coverage is determined, and on the consideration of judgmental scans as discussed in MARSSIM Section 2.5.5.
Basis: With regard to scan coverage, MARSSIM Section 2.5.5 (Developing an Integrated Survey Design) discusses the judgmental nature of scans for Class 3 and sometimes Class 2 areas (dependent on the potential for portions of the survey unit to exhibit a higher probability for areas of elevated activity). MARSSIM 2.5.5 states:
The level of scanning effort should be proportional to the potential for finding areas of elevated activity: in Class 2 survey units that have residual radioactivity close to the release criterion a larger portion of the survey unit would be scanned, but for survey units that are closer to background scanning a smaller portion of the survey unit may be appropriate. Class 2 survey units have a lower probability for areas of elevated activity than Class 1 survey units, but some portions of the survey unit may have a higher potential than others. Judgmental scanning surveys would focus on the portions of the survey unit with the highest probability for areas of elevated activity. If the entire survey unit has an equal probability for areas of elevated activity, or the judgmental scans don't cover at least 10% of the area, systematic scans along transects of the survey unit or scanning surveys of randomly selected grid blocks are performed.
Class 3 areas have the lowest potential for areas of elevated activity. For this reason, MARSSIM recommends that scanning surveys be performed in areas of highest potential (e.g., comers, ditches, drains) based on professional judgment. This provides a qualitative level of confidence
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 121 of 156 that no areas of elevated activity were missed by the random measurements or that there were no errors made in the classification of the area.
Path Forward:
- a. Describe how the determination of the percentage and location of the areal coverage via ISOCs will consider the need for judgmental measurements in Class 2 and 3 areas.
ZSRP Response (5.8a) - The following clarifying text will be added to the end of LTP section 5.5.2.1.1 and section 5.5.2.1.2; In addition to the prescribed areal coverage, additional judgmental measurements may be collected at locations with higher potential for containing elevated concentrations of residual radioactivity based on professional judgment.
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- 9. Comment: Section 5.5.2.2 (STS Sample Size Determination)
Section 5.5.2.2 of the LTP notes that:
Following the guidance in MARSSIM, the Type I decision error that was used for this calculation was set at 0.05 and the Type II decision error was set at 0.05. The upper boundary of the gray region was set at the BFM Inventory Limit. The Lower Bound of the Gray Region (LBGR) was set at the expected fraction of the BFM Inventory Limit in the STS survey unit. The expected fraction of the BIL in both Class 2 STS survey units was set at 50% and the expected fraction of the BIL in both Class 3 STS survey units was set at 1%. The standard deviation of the concrete core samples taken in the Turbine Building was used for sigma () in the STS survey units for the Turbine Building, Crib House/Forebay, WWTF and Circulating Water Intake and Discharge Tunnels.
It is not clear how the standard deviation of concrete core samples was utilized for the purpose of calculating the relative shift. The upper and lower bounds of the gray region are in terms of total inventory, but there is no description of how the standard deviation of core samples is related to a standard deviation of total inventory or how consistent units are maintained for the purpose of this calculation.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.4 (Final Status Survey Design), licensees should provide:
A summary of the statistical tests that will be used to evaluate the survey results, including the elevated measurement comparison, if Class 1 survey units are present; a justification for any test methods not included in MARSSIM; and the values for the decision errors ( and )
with a justification for values greater than 0.05 Path Forward:
- a. Clarify the usage of the parameters used to develop the relative shift.
- b. Describe how the same units and correlation to total inventory are maintained.
ZSRP Response (5.9a and 5.9b) - The BIL for Cs-137 for the Turbine Building is 5.94E+11 pCi. If this amount of activity were present in the Turbine Building basement at the time of STS, it would equate to 25 mRem. Consequently, it is analogous to a DCGL in this regard. 1% of 5.94E+11 pCi is 5.94E+09 pCi. During characterization, a series of concrete core samples were taken in the Turbine Building. Cs-137 was positively detected at concentrations of 0.127 pCi/g to 45.2 pCi/g. The standard deviation of this sample population was 15.71 pCi/g. The pCi/g value was converted to total activity by multiplying 15.71 pCi/g by the total mass of the end-state Turbine Building concrete (assuming a contamination depth of 0.5 inches) of 4.47 E+08 grams which equals 7.02E+09 pCi. Consequently, the formula for relative shift then becomes the at 1% LBGR divided by sigma or (5.94E+11 pCi - 5.94E+09 pCi)/7.02E+09 pCi producing
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 123 of 156 a relative shift of 83.7. Consequently, the relative shift was adjusted to 3, resulting in a sample population of 14.
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- 10. Comment: Section 5.5.4 (STS Data Assessment and Application of Results to BFM)
Section 5.5.4 of the LTP discusses the usage of the Sign Test for STS results as follows:
As described in section 5.10.3.3, the Sign Test will be used to evaluate the remaining residual radioactivity in each survey unit against the dose criterion using the Basement Dose Factors. If the Sign Test fails, the expected fraction of the BIL assumed in survey design as specified in section 5.5.2.1, can be adjusted to higher values that allow the test to pass. This assumes that the total mean inventory in the entire basement, considering all survey units is below 25 mrem/yr. In addition, if a building has multiple STS survey units (e.g. the Auxiliary Building will have three (3) STS survey units, one for the wall, one for the floor, and one for the embedded piping/penetrations remaining in the Auxiliary Building), the sum of the inventory fractions for each STS survey unit in the building must be less than one. If the sum exceeds one, then an investigation will be performed which may result in the acquisition of additional measurements, the reclassification of the STS survey unit, or additional remediation.
Several points in this discussion require clarification, as discussed below:
It is not clear that the expected fraction of the BIL assumed in survey design was discussed in Section 5.5.2.1.
The implementation of the Sign Test is not fully described, and would benefit from an example calculation. For example the upper bound of the gray region (as discussed in Section 5.2.2) is set to the BIL (which was previously defined as the hypothetical maximum inventory level of the ROC in each basement). The implication is that any sample can include activity up to the hypothetical maximum inventory level of each basement. It does not seem logical to assume every sample can potentially be at the limit if each sample must be added together for a final comparison to that same limit as the survey unit maximum.
The discussion in Section 5.5.4 discusses the fraction of the BIL used in survey design and notes a comparison among inventory fractions, which suggests that the Sign Test may be comparing individual results against a fraction of the BIL.
A DCGL value which would normally be used as the release criterion in the MARSSIM Sign Test, would be correlated to the TEDE to an average member of the critical group that does not exceed 25 mrem per year. As previously noted, there is no description of the dose assessment associated with the BIL, and the correlation to 25 mrem per year.
With regard to the statement that if the Sign Test fails, the expected fraction of the BIL assumed in survey design as specified in section 5.5.2.1, can be adjusted to higher values that allow the test to pass, simply changing design values after a survey unit has failed in order to make the survey unit pass does not seem appropriate. If the licensee intends to use MARSSIM survey methodology, the DQO process should be utilized to determine why a survey unit failed, and consideration should be given to the necessity for additional measurements and re-survey of the survey unit.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 125 of 156 Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.4 (Final Status Survey Design), licensees should provide:
A summary of the statistical tests that will be used to evaluate the survey results, including the elevated measurement comparison, if Class 1 survey units are present; a justification for any test methods not included in MARSSIM; and the values for the decision errors ( and )
with a justification for values greater than 0.05 Path Forward:
- a. Provide additional details on all of the parameters that will be used in survey design and in the Sign Test for STS data assessment.
- b. Clarify how the BIL (or a fraction of the BIL) will be used in the assessment. A complete example would be useful.
ZSRP Response (5.10a and 5.10b) - An example calculation for a hypothetical STS performed in the Turbine Building and Auxiliary Building basement has been provided in ZSRP Response to RAI PAB 15a. As clarification, the second and third paragraph in section 5.5.4 will be revised as follows; After a sufficient number of ISOCS measurements are taken in a STS unit in accordance with the areal coverage requirements specified in Table 5-11, the data will be summarized, including any judgmental or investigation measurements. The measured activity for each gamma-emitting ROC (and any other gamma radionuclide identified at levels greater than the ISOCs MDC) will be recorded (in units of pCi/m2). Background will not be subtracted from any measurement.
Using the radionuclide mixture fractions applicable to the survey unit, an inferred activity will be derived for each applicable HTD ROC. ZSRP will infer the presence of Ni-63 and Sr-90 using the surrogate approach specified in section 5.2.4. For the STS of the Containment basements, H-3 will be added as an additional HTD radionuclide that will be inferred. The scaling factors that will be used are presented in Table 5-2 (and Table 6-3 for the Containment basements). A sum of fractions (SOF) calculation will be performed for each measurement by dividing the reported concentration by the applicable BIL for each ROC, after converting the BIL to the same units as the ISOCS measurement (pCi/m2). The individual ROC fractions will then be summed to provide a total SOF value for the measurement.
As described in section 5.10.3.2, the Sign Test will be used to evaluate the remaining residual radioactivity in each survey unit against the dose criterion. The SOF for each measurement will be used as the weighted sum for the Sign Test. If the Sign Test demonstrates that the mean activity of the survey unit is less than the BIL at a 95% Type 1 error rate, then the mean of all the total SOFs for each measurement in a given survey unit (or Mean Inventory Fraction) is calculated. If the Sign test fails, or if the Mean Inventory Fraction in a basement exceeds one, then the survey unit will fail STS. If a survey unit fails STS, then the STS survey unit may be reclassified, additional remediation will be performed and the STS performed again.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 126 of 156 In situations where there are multiple survey units in a STS basement (e.g. the Auxiliary Building will have three STS units, one for the walls, one for the floor and one for any embedded piping and/or penetrations that may remain in the end-state), the fraction for each ROC will be calculated by dividing the reported ISOCS activity level (pCi/m2) by an allocated fraction of the applicable BIL that is selected for each survey unit apriori to the performance of STS. Note that the sum of the allocated BIL fractions for each survey unit applicable to a given basement must equal one. The upper bound of the gray region for the Sign Test will be the allocated fraction of the BIL for each STS survey unit in the basement, as opposed to the full BIL in the case of a basement that has only one STS survey unit.
For example, if a basement has two STS units, and 20% of the BIL is allocated to the first survey unit and 80% of the BIL allocated to the second survey unit, then the SOF for the first survey unit will be calculated using 20% of the BIL value and the SOF in the second survey unit calculated using 80% of the BIL value. The SOF for each measurement will be calculated using the allocated fraction, as opposed to the full BIL, and the resulting fraction used as the Weighted Sum (Ws) when performing the Sign Test During the data assessment, if the actual mean inventory in a STS unit exceeds the allocated fraction, or the survey unit fails the sign test, then the allocated fraction may be increased as long as the allocated fractions in the other STS units in the basement are decreased accordingly such that the revised fractions still sum to one. However, the Sign test must be passed for all survey units in a given basemen at the final allocated fraction selected. After all survey units are demonstrated to pass the Sign test, the Mean Inventory Fraction for each of the survey units are summed.
If a combination of allocated fractions cannot be selected that results in all survey units in a given basement passing the Sign test, or if the sum of the Mean Inventory Fractions of all STS units in a basement exceeds one, then the survey unit will fail STS. If a survey unit fails STS, then additional remediation will be performed and the STS performed again.
- c. Evaluate whether or not Section 5.5.2.1 was the appropriate reference on the expected fraction of the BIL assumed in survey design.
ZSRP Response (5.10c) - As part of the proposed clarifying revisions to section 5.5.4, this reference has been deleted.
- d. Clarify what is meant by the statement that if the Sign Test fails, the expected fraction of the BIL assumed in survey design as specified in section 5.5.2.1, can be adjusted to higher values that allow the test to pass, and describe how the DQO process will be utilized to determine why a survey unit failed and to determine the appropriate path forward.
ZSRP Response (5.10d) - As part of the proposed clarifying revisions to section 5.5.4, this text will be deleted.
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- 11. Comment: Section 5.5.5 (Methods for STS of Embedded Piping and Penetrations) and Section 5.7.1.8 (Buried Piping)
Section 5.5.5 of the LTP discusses surveys of the interior of embedded pipe or penetration sections that cannot be accessed directly, and notes the following:
A static measurement will be acquired at a specified distance traversed into the pipe. This distance will be determined as a DQO based on the contamination potential in the pipe. The detector output will represent the gamma activity in gross cpm. This gamma measurement value in cpm will then be converted to dpm using an efficiency factor based on the calibration source.
The total activity in dpm will adjusted for the assumed total effective surface area commensurate with the pipe diameter, resulting in measurement results in units of dpm/100 cm2.
Buried piping is discussed in Section 5.7.1.8 of the LTP where it is noted that radiological evaluations for piping or drains that cannot be accessed directly will be performed via measurements made at traps and other appropriate access points where the radioactivity levels are deemed to either bound or be representative of the interior surface radioactivity levels providing that the conditions within the balance of the piping can be reasonably inferred based on those data.
Additional clarification is needed on the survey design for embedded and buried piping measurements.
Basis: Per the acceptance criteria described in NUREG-1700, Rev. 1, Section 5 (Final Status Survey Plan), licensees should provide:
Methods for surveying embedded piping Challenges related to Embedded and Buried Piping Characterization are also discussed in Appendix O of NUREG-1757, Vol. 2, Rev. 1, where it is concluded that it is incumbent on the licensee to develop and document a comprehensive approach to embedded pipe and buried piping characterization that accounts for limitations and uncertainties, taking into account MARSSIM guidance in developing the related DQOs. Appendix O also notes that, to be found acceptable, the methods should each address the following nine issues:
o radionuclides of interest and chosen surrogate, o levels and distribution of contamination, o internal surface condition of the piping, o internal residues and sediments and their radiation attenuation properties, o removable and fixed surface contamination, o instrument sensitivity and related scan and fixed minimum detectable concentrations, o piping geometry and presence of internally inaccessible areas/sections, o instrument calibration, and o data quality objectives (DQOs).
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 128 of 156 Path Forward:
- a. Describe the comprehensive approach to embedded and buried piping characterization that accounts for limitations and uncertainties, taking into account MARSSIM guidance in developing the related DQOs. Consideration should be given to the nine issues discussed in Appendix O NUREG-1757, Vol. 2, Rev. 1, and presented in the RAI basis.
- b. Describe how a 100% surface scan will be performed for Class 1 impacted piping in accordance with MARSSIM.
- c. Describe how the number of discrete measurements will be determined.
ZSRP Response (5.11a, 5.11b and 5.11c) - The ZSRP approach for the radiological survey of pipe system interiors involves the insertion of various sized NaI detectors appropriately sized for use in various piping diameters. A simple push-pull methodology is used, whereby the position of the detector in the piping system can be easily determined in a reproducible manner.
The NaI detectors are configured in a fixed geometry relative to the surveyed surface, thus creating a situation where a defensible efficiency can be calculated. The detectors are then deployed into the actual pipe and timed measurements are acquired at intervals commensurate with the contamination potential of the pipe. A conservative area of detection is assumed for each pipe size. It is also conservatively assumed that any activity is uniformly distributed in the area of detection. For example, if the pipe to be surveyed is potentially contaminated (i.e.
commensurate with a MARSSIM Class 1 classification), then a static measurement is taken at one foot intervals. Based upon the area of detection for the detector used, this will conservatively provide 100% areal coverall of the pipe interior surfaces. For pipe systems that are not potentially contaminated (i.e. commensurate with a MARSSIM Class 2 or Class 3 classification), measurements may be taken at intervals greater than one foot to provide sufficient areal coverage based on the classification of the pipe. A surrogate correction based upon the radionuclide distribution present in the pipe is applied to the gamma emission to account for the presence of other non-gamma emitting radionuclides in the mixture.
Typically, the first step of the survey process is the performance of a video inspection of the piping system to be surveyed. A miniature camera is inserted into the pipe and a recording is made as the camera is advanced through the system. The camera is maneuvered through the pipe by the manipulation of fiber-composite rods which are manually pushed or pulled to provide locomotion to the camera. This inspection provides an unambiguous record of the physical condition of the piping and identifies any obstructions which may affect the ability to advance the detector.
For the performance of the radiological surveys, it is anticipated that a 1 x 1 detectors (NaI or CsI) will be used for pipe sizes ranging from 2-inch to 8-inch in diameter and 3 x 3 NaI detectors for pipes sizes exceeding 10-inches in diameter. Detector, cable and instrument combinations will be calibrated as a unit using NIST traceable sources.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 129 of 156 Instrument efficiency is developed by placing a flexible radiological plane source (based upon the predominate gamma emitting radionuclide nuclide of concern) into various sized pipe models, similar to the geometry that will be encountered in various diameter pipes and taking measurements with the selected detector. Using the known source activity, it is then possible to calculate an efficiency factor for various scenarios in each size of pipe. This efficiency is specific to a particular detector inside of a particular sized pipe while exposed to the appropriate gamma energy. Each detector will have an efficiency factor to convert the cpm readout of the NaI detector to dpm for each applicable diameter of piping. Subsequently, each measurement result taken inside the actual pipe will represent a radiological contamination activity over a total effective surface area commensurate with the pipe diameter. The results will be conservative because there will be some additional instrument response from contamination located in the pipe at distances greater than the maximum distance. The weighted average efficiency calculation is designed to theoretically limit the conservatism to about 20% or less.
Daily prior to use and daily following use, each detector will undergo an Operational Response Check in accordance with procedure. The Daily Operational Response Check will ensure that the detector is working properly. The Daily Operational Response Check compares response to background and response to check sources for the ranges established.
A background value will also be determined for the detector/instrument combination to be used prior to deployment. The background value will be obtained at the location where the pre-use response check of the instrument was performed. The background value is primarily used to ensure that the detector has not become cross-contaminated.
Once the detector is determined to be fully functional, it is deployed to the field for inserting into the targeted piping. A static one-minute measurement will be acquired at the pre-determined interval for the areal coverage to be achieved. The NaI output will represent the gamma activity in gross cpm for each foot of piping traversed. This measurement value in cpm will then be converted to dpm using the efficiency factor.
The total activity in dpm will be adjusted for the assumed total effective surface area commensurate with the pipe diameter, resulting in measurement results in units of dpm/100 cm2.
This activity value will then be corrected for the radionuclide mixture. This measurement result will then represent a commensurate and conservative gross measurement that can be compared to the buried pipe DCGLs, or converted to units of inventory for STS.
Currently, ZSRP is evaluating the accessibility of the embedded floor drain systems in the concrete floor of the Auxiliary Building 542 foot elevation. Emergent concerns pertaining to potential groundwater intrusion into the Auxiliary Building basement if this pipe was removed from the concrete has caused ZSRP to revise the decommissioning approach for these systems. The original plan was to excavate the Auxiliary Building floor drain systems and dispose of the material as radioactive waste. However, there are concerns that the structural integrity of the basement floor may be compromised if the system is removed. ZSRP is in the
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 130 of 156 process of assessing a practicable method for determining the total activity inventory that is defensible and bounding. If the approach proposed for demonstrating compliance in this system is different from the more traditional approach previously described, then ZSRP will document the process that will be used in a TSD.
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- 12. Comment: Section 5.6 (Final Status Survey Design)
Section 5.6 of the LTP discusses FSS design and notes that any minor above-grade structures that will remain on the site (e. g. ISFSI Warehouse, Sewage Lift Station, etc.) will be released for unrestricted use using the survey approach from MARSAME.
Surveys of structures remaining on site may not be within the scope of MARSAME.
Basis: Per MARSAME Section 1.1 (Purpose and Scope of MARSAME) fixed buildings and structures are not considered within the scope of MARSAME. MARSAME Section 1.1 states the following with regard to materials within the scope of MARSAME:
MARSAME provides technical information on approaches for planning, implementing, assessing, and documenting surveys to determine proper disposition of M&E. Release (including clearance) and interdiction are types of disposition options in MARSAME.
Detailed descriptions of these disposition options are provided in Chapter 2. Examples of M&E include metals, concrete, tools, equipment, piping, conduit, furniture, and dispersible bulk materials such as trash, rubble, roofing materials, and sludge. Liquids, gases, and solids stored in containers (e.g., drums of liquid, pressurized gas cylinders, containerized soil) are also included in the scope of this document.
MARSAME further indicates the following with regard to the types of materials that are outside the scope of MARSAME:
Radionuclides or radioactivity on workers or members of the public are outside the scope of the document, as are liquid and gaseous effluent releases and real property (e.g., fixed buildings and structures, surface and subsurface soil remaining in place).
Path Forward:
- a. Describe the implementation of MARSAME for minor above-grade structures and justify that such structures are within the scope of MARSAME, or re-evaluate the survey methodology.
- b. Describe the release criteria to be utilized for structures that will remain.
ZSRP Response (5.12a and 5.12b) - In the end-state condition, all major structures at ZNPS will be demolished to 3 feet below grade. All portions of the structures located above three feet below grade will be demolished, dismantled and disposed of as a waste stream. Structural material from impacted contaminated structures such as the interior of the Containment structures, Auxiliary Building and Fuel Handling Building will be demolished, properly packaged and disposed of as radioactive waste.
The decommissioning approach for ZSRP also calls for the beneficial reuse of concrete from building demolition as clean fill. The only concrete structures that will be considered are those where the probability of residual contamination is minimal. The concrete structures that ZSRP believes are acceptable candidates for reuse as fill are the outer shell of the Containment
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 132 of 156 Buildings, the concrete portions of the Turbine Building (including the Steam Tunnels) above the 588 foot, the Crib House, portion of the Forebay above the 588 foot elevation, the Service Building and minor structures such as the Interim Radioactive Waste Storage Facility (IRSF), the Mechanical Maintenance Training Center (MMTC) and Warehouse, the Fire Maze complex, the NGET building, the ENC building, the south Warehouse and the North Security Access Gatehouse. Surveys will be performed of candidate structural concrete prior to demolition.
These unconditional release surveys will be designed and performed in accordance with ZionSolutions procedure ZS-LT-400-001-001, Unconditional Release of Materials, Equipment and Secondary Structures (provided in Enclosure 2). This procedure describes the protocols to be used to design, perform, control, evaluate and document radiological surveys performed on structures, systems and materials to demonstrate compliance with the unconditional release criteria and, subsequent release from ZNPS as an uncontaminated or clean material or, for reuse at ZNPS. This procedure and the methods described within are based on the graded survey approach guidance are established using guidance found in NUREG-1575, Supplement 1, Multi-Agency Radiation Survey and Assessment of Materials and Equipment Manual (MARSAME).
In accordance with the procedure, material shall be deemed to contain, or be contaminated, with plant-related radioactivity if radiological surveys and/or sample analyses positively identify plant-related radioactivity. For solid materials, the required MDCs for scan measurements and smears shall be no greater than the corresponding limits in NRC I.E. Circular No. 81-07. For the analysis of volumetric activity in solids/sludge using gamma spectroscopy analyses, the analysis MDCs shall be no greater than the MDCs derived from ODCM Chapter 12, Table 12.5-3 for sediments. If, during the performance of the unconditional release survey, residual radioactivity is positively detected, then the material will be disqualified as an acceptable material for unconditional release or as reuse and will be controlled and properly disposed of as waste.
In addition to the structures that will be demolished and reused as clean fill, there are several minor structures at ZNPS that will not be demolished and disposed of as waste but rather, will remain in the end-state condition to support activities at the site post-decommissioning. These minor structures are specified in the Asset Sale Agreement as the Warehouse constructed to support the ISFSI, the Sewage Lift Station, the active switchyard and minor valve and conduit boxes. ZSRP proposes to demonstrate that these minor structures are acceptable for release as uncontaminated or clean material using the graded survey approach described in procedure ZS-LT-400-001-001. As with material that would be released off-site, material shall be deemed to contain, or be contaminated, with plant-related radioactivity if radiological surveys and/or sample analyses positively identify plant-related radioactivity. For solid materials, the required MDCs for scan measurements and smears shall be no greater than the corresponding limits in NRC I.E. Circular No. 81-07. If, during the performance of the unconditional release survey, residual radioactivity is positively detected, then ZSRP will perform further assessments. In the case of a structure that must remain (such as the ISFSI warehouse or switchyard), the structure
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 133 of 156 may be decontaminated, reclassified and/or resurveyed. If the structure is not critical (such as a valve or conduit box), it may be completely removed and properly disposed of as radioactive waste.
ZSRP contends that the approach described above is appropriate for the minor structures described and is more conservative than using an adjusted gross DCGL to demonstrate compliance with a dose-based release criterion, even using the default screening criteria provided in NUREG-1757, Appendix H. In addition, ZSRP contend that there is precedence for the use of this approach for the release of minor structures that remain in the end-state condition. In 2007, the Haddam Neck decommissioning released the Emergency Operations Facility as an end-state structure using the unconditional release criteria verses a dose-based criteria. This approach was approved and accepted by the NRC.
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- 13. Comment: Section 5.6.2.5 (Develop a Decision Rule)
Section 5.6.2.5 of the LTP states that; if the SOF is greater than or equal to unity (1), then the survey unit does not meet the criteria for unrestricted release and, further action, including but not limited to additional remediation, survey unit reclassification, additional data collection, or dose assessment will be taken.
It is not clear how a dose assessment will be utilized or how any of the potential actions, other than remediation, would affect an SOF level that is greater than 1.
Basis: As discussed in the Final Status Survey Design described in NUREG-1757, Vol. 2, Rev. 1, Section 4.4, two objectives of the FSS are to:
Demonstrate that the potential dose from residual radioactivity is below the release criterion for each survey unit, and Demonstrate that the potential dose from small areas of elevated activity is below the release criterion for each survey unit Path Forward:
- a. Clarify what is meant by "dose assessment," and
- b. Describe how the stated potential actions, other-than remediation, would affect an SOF level greater than 1 in order to demonstrate the survey unit release criterion is met.
ZSRP Response (5.13a and 5.13b) - To clarify, the second bullet of section 5.6.2.5 will be revised to read as follows; If the SOF is greater than or equal to unity (1), then the survey unit does not meet the criteria for unrestricted release. Additional remediation followed by FSS redesign and resurvey will be performed.
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- 14. Comment: Section 5.6.4.2.1 (WRS Test Sample Size)
Section 5.6.4.2.1 of the LTP states that Equation 5-1 of NUREG-1507 may alternatively be used to calculate the number of sampling and measurement locations. This is likely a misprint that was intended to refer to Equation 5-1 of NUREG-1575.
Basis: Equation 5-1 in NUREG-1507 deals with total efficiency of a distributed source.
Path Forward:
- a. Correct this statement to refer to the intended reference.
ZSRP Response (5.14a) - The reviewer is correct, the correct reference is equation 5-1 of NUREG-1575 verse NUREG-1507. However, this section will be revised to completely remove reference to the WRS test. Consequently, this reference will be deleted.
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- 15. Comment: Section 5.6.4.6 (Investigation Process) and Table 5-13 (Investigation Levels)
Table 5-13 of the LTP indicates that the direct investigation level for a Class 1 area is a level that is greater than the DCGLEMC. This investigation level does not appear to be consistent with MARSSIM guidance.
Basis: MARSSIM guidance indicates that direct measurements should be flagged for investigation when the DCGLw or a statistical parameter-based value is exceeded (MARSSIM Table 5.8). MARSSIM also notes that For a Class 1 survey unit, measurements above the DCGLw are not necessarily unexpected. However, a measurement above the DCGLw at one of the discrete measurement locations might be considered unusual if it were much higher than all of the other discrete measurements. Thus, any discrete measurement that is both above the DCGLw and above the statistical-based parameter for the measurements should be investigated further. Any measurement, either at a discrete location or from a scan, that is above the DCGLEMC should be flagged for further investigation.
Path Forward:
- a. Describe how an investigation level comparison to the DCGLw and a parameter-based value will be incorporated into the Zion FSS methods, per MARSSIM guidance.
ZSRP Response (5.15a) - LTP Table 5-13 will be revised as follows to ensure consistency with MARSSIM guidance.
Table 5-13 Investigation Levels Direct Investigation Classification Scan Investigation Levels Levels
>DCGLW or >MDCscan if Class 1 MDCscan is greater than > DCGLW DCGLW
>DCGLW or >MDCscan if Class 2 MDCscan is greater than >DCGLW DCGLW
>DCGLW or >MDCscan if Class 3 MDCscan is greater than >0.5 DCGLW DCGLW
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- 16. Comment: Sections 5.6.4.6.1 (Remediation and Reclassification)
Reclassification of survey units is discussed in Section 5.6.4.6.1 of the LTP as follows:
If an individual survey measurement (scan or direct) in a Class 2 survey unit exceeds the DCGLw, the survey unit, or portion of the survey unit, will be investigated, and if necessary, be reclassified to a Class 1 area and the survey re-designed and re-performed accordingly. If an individual survey measurement in a Class 3 survey unit exceeds 50 percent of the DCGLw, the survey unit, or portion of a survey unit, will be investigated, and if determined to exceed the DCGL, reclassified to a Class 1 or a Class 2 survey unit and the survey re-designed and re-performed accordingly. If the elevated survey measurement (>DCGL) is confirmed by investigation, but cannot be thoroughly described as an isolated condition, (i.e., it cannot be demonstrated with great certainty that this condition does not exist elsewhere in the survey unit) the survey unit will be reclassified. If the result cannot be duplicated, the population of the individual and average measurement results with respect to the DCGL will be reviewed, and if the variability does not suggest the initial classification was inappropriate, the survey unit will not be reclassified.
This section describes an allowance for isolated areas of elevated contamination above a DCGL to occur and not result in a reclassification of Class 2 or 3 areas. This is not consistent with MARSSIM guidance.
Basis: MARSSIM discusses reclassification as follows:
(MARSSIM 4.4) - As a survey progresses, reevaluation of this classification may be necessary based on newly acquired survey data. For example, if contamination is identified in a Class 3 area, an investigation and reevaluation of that area should be performed to determine if the Class 3 area classification is appropriate. Typically, the investigation will result in part or all of the area being reclassified as Class 1 or Class 2. If survey results identify residual contamination in a Class 2 area exceeding the DCGL or suggest that there may be a reasonable potential that contamination is present in excess of the DCGL, an investigation should be initiated to determine if all or part of the area should be reclassified to Class 1.
(MARSSIM 5.5.3.1) - Class 2 Areas. Surface scans are performed over 10 to 100% of structure surfaces. Generally, upper wall surfaces and ceilings should receive surface scans over 10 to 50% of these areas. Locations of scanning survey results above the investigation level are identified and investigated. If small areas of elevated activity are confirmed by this investigation, all or part of the survey unit should be reclassified as Class I and the survey strategy for that survey unit redesigned accordingly.
Path Forward:
- a. Describe how reclassification will be accomplished in accordance with guidance from MARSSIM Section 4.4 and Section 5.5.3.1.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 138 of 156 ZSRP Response (5.16a) - The first portion of the second paragraph of section 5.6.4.6.1 will be revised as follows; If an individual survey measurement (scan or sample) in a Class 2 survey unit exceeds the DCGLw, the survey unit, or portion of the survey unit, will be investigated. If small areas of elevated activity are confirmed by this investigation or, suggests that there may be a reasonable potential that contamination is present in excess of the DCGLw, then all or part of the survey unit will be reclassified as Class 1 and the survey strategy for that survey unit redesigned accordingly. If an individual survey measurement in a Class 3 survey unit exceeds 50 percent of the DCGLw, the survey unit, or portion of a survey unit, will be investigated. If the investigation confirms residual radioactivity in excess of 50 percent of the DCGLw, then the survey unit will be reclassified to a Class 1 or a Class 2 survey unit and the survey re-designed and re-performed accordingly.
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- 17. Comment: Section 5.6.4.6.2 (Resurvey)
Section 5.6.4.6.2 of the LTP discusses remediation and resurvey during FSS as follows:
If remediation is required in only a small area of a Class 1 survey unit, any replacement measurements or samples required will be made within the remediated area at randomly selected locations following verification that the remediation activities did not affect the remainder of the unit. Re-survey will be required in any area of a survey unit is affected by subsequent remediation activities. Additional guidance regarding the failure and re-survey of a survey unit and is provided in section 8.5.3 of MARSSIM.
Section 8.5.3 of MARSSIM is referred to, but there is no indication of how MARSSIM guidance will be utilized in the event that a small area is remediated in a Class 1 survey unit.
Basis: Section 8.5.3 of MARSSIM discusses the usage of the DQO process and includes examples for Class 1 survey units where the nonparametric statistical tests are both failed and passed.
Path Forward:
- a. Describe how MARSSIM guidance and the DQO process will be implemented if small areas of elevated radioactivity require remediation in Class 1 survey units.
ZSRP Response (5.17a) - Section 5.7 of ZionSolutions procedure ZS-LT-300-001-004, Final Radiation Survey Data Assessment (provided in Enclosure 2) details the corrective actions for Final Radiation Survey failure. In accordance with the procedure, if the survey unit did not pass FRS, then Attachment 16, Corrective Actions for Final Radiation Survey Failure or an equivalent form will be used. Attachment 9 of procedure ZS-LT-300-001-004 provides potential corrective actions that may be taken as corrective actions for the FSS failure of a survey unit.
These corrective actions may require remediation, reclassification, and/or resurvey. Attachment 9 is reproduced below. Also, as clarification, the last paragraph of section 5.6.4.6.2 will be revised as follows to reflect the procedure; If remediation is required in only a small area of a Class 1 survey unit (defined as an Elevated Radioactivity Fraction (fEMC) that exceeds unity in 5% or less of the survey unit area), then additional measurements will be taken to determine the effectiveness of the remediation and FSS will be re-performed using the same survey design. If remediation is required in a larger area of a Class 1 survey unit (defined as an fEMC that exceeds unity in greater than 5% of the survey unit area), then the FSS will be restarted under a new survey design. Additional guidance regarding the failure and re-survey of a survey unit and is provided in section 8.5.3 of MARSSIM.
ZS-LT-300-001-004 Attachment 9 Revision 1 Remediation, Reclassification and Resurvey Actions Information Use REMEDIATION Remediation Criteria Proposed Remediation
- 1) The EMC SOF for survey unit is less than or equal to unity (1) None Class 1 FSS Survey 2) The Elevated Radioactivity Fraction (fEMC) exceeds unity in 5% or less of the survey unit area. Spot Remediation & Resurvey Unit
- 3) The Elevated Radioactivity Fraction (fEMC) exceeds unity in greater than 5% of the survey unit area. General Remediation and Restart FSS
- 1) The mean inventory fraction (total mean dose for the survey unit divided by the dose criterion of 25 mrem/yr) is greater than or equal to one.
Class 1 STS Unit General Remediation and Restart STS
- 2) The sum of the mean inventory fractions for each STS unit contained within a building basement is greater than or equal to one.
RECLASSIFICATION Reclassification Criteria Proposed Action The extent of the elevated area relative to the total Reclassify only the bounded discrete area area of the survey unit is minimal and the source of of elevated activity to Class 1.
the residual radioactivity is known One or several survey measurements (scan, sample or Class 2 Survey Unit direct measurement) exceed 50% of the dose criterion The extent of the elevated area relative to the total Reclassify 2,000 m2 for soils or 100 m2 or a DCGLw or a portion of the survey unit is area of the survey unit is minimal and the source of for structures around the area of elevated remediated. the residual radioactivity is unknown activity as Class 1.
The extent of the elevated area relative to the total Reclassify the entire survey unit as Class area of the survey unit is significant. 1.
The extent of the elevated area relative to the total Reclassify the area of elevated activity to One or several survey measurements (scan, sample or area of the survey unit is minimal Class 1 and create a Class 2 buffer zone direct measurement) exceed 50% of the dose criterion of appropriate size around the area.
or a DCGLw or a portion of the survey unit is Reclassify the area of elevated activity to remediated. The extent of the elevated area relative to the total Class 1 and create a Class 2 buffer zone area of the survey unit is significant. of appropriate size around the area.
Class 3 Survey Unit The extent of the elevated area relative to the total Reclassify the area of elevated activity to area of the survey unit is minimal Class 2.
One or several survey measurements (scan, sample or direct measurement) exceed 1% of the dose criterion For soils, reclassify 10,000 m2 around the or 50% of a DCGLw The extent of the elevated area relative to the total area of elevated activity to Class 2. For area of the survey unit is significant. structures, reclassify 1,000 m2 around the area of elevated activity to Class 2.
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ZS-LT-300-001-004 Attachment 9 Revision 1 Remediation, Reclassification and Resurvey Actions Information Use RESURVEY Resurvey Criteria Proposed Action Re-scan remediated area; collect The extent of the elevated area relative to the total replacement systematic population area of the survey unit is minimal. samples/measurements within the The survey unit has been remediated. remediated area using random selection.
Class 1 Survey Unit The extent of the elevated area relative to the total Resurvey entire survey unit using a new area of the survey unit is significant. survey design.
Survey unit has been reclassified from a Class 2 Increase scan or areal coverage to 100%.
survey unit. No remediation was performed. Additional statistical samples are not required.
The area of the new Class 1 survey unit relative to the Increase scan or areal coverage in Class 2 area of the initial Class 2 survey unit is minimal and survey unit.
Survey unit has been divided to accommodate a new no statistical samples were affected.
Class 2 Survey Unit Class 1 survey unit.
Statistical sample population was affected by the Increase scan or areal coverage in Class 2 reclassification. survey unit and resurvey entire survey unit using a new survey design.
Increase scan or areal coverage in Class 3 The area of the new Class 2 survey unit relative to the survey unit, collect replacement area of the initial Class 3 survey unit is minimal. systematic population Class 3 Survey Unit Survey unit has been divided to accommodate a new samples/measurements within the Class 2 survey unit. remediated area using random selection.
The area of the new Class 2 survey unit relative to the Resurvey entire survey unit using a new area of the initial Class 3 survey unit is significant. survey design.
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- 18. Comment: Section 5.7.1.4.1 (Gamma Scans of Surface Soils)
Section 5.7.1.4.1 of the LTP indicates that In situ gamma spectroscopy may also be effectively substituted for scanning surveys, and that when this approach is used, the orientation of the detector, FOV, and a priori MDC will be documented and approved as a DQO during FSS design and survey package development.
In situ gamma spectroscopy is also listed in Table 5-14 (Typical FSS Survey Instrumentation) along with text in Section 5.8.1 (Instrument Selection) indicating that Other measurement instruments or techniques may be utilized. The acceptability of additional or alternate instruments or technologies for use in the FSS will be justified in a technical basis evaluation document prior to use. Technical basis evaluations for alternate final status survey instruments or techniques will be provided for NRC review 30 days prior to use.
While in situ gamma spectroscopy is listed in Table 5-14, NRC staff notes that no technical justification for this technology has been provided with regard to scanning of soils.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.4 (Final Status Survey Design), licensees should provide:
A description of scanning instruments, methods, calibration, operational checks, coverage, and sensitivity for each media and radionuclide A description of the instruments, calibration, operational checks, sensitivity, and sampling methods for in situ sample measurements, with a demonstration that the instruments and methods have adequate sensitivity Path Forward:
- a. Provide a technical basis for in situ gamma spectroscopy usage for soil scanning, or commit to providing this justification for NRC review 30 days prior to use.
ZSRP Response (5.18a) - It is very unlikely that ISOCS would be used in place of scanning in open land survey units and was only included to provide flexibility. The last two sentences in the first paragraph of section 5.7.1.4.1 will be deleted and ISOCS is no longer considered an option for scanning of open lands.
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- 19. Comment: Section 5.7.1.5. (Subsurface Soils)
Surveys and sampling of subsurface soils during Remedial Action Support Surveys (RASS) are discussed in Section 5.7.1.5 of the LTP as follows:
The data obtained during the RASS is expected to provide a high degree of confidence that the excavation, or 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, a NaI detector or intrinsic germanium detector of sufficient sensitivity to detect residual radioactivity at the DCGLw may be utilized to identify the presence or absence of subsurface contamination, and the extent of such contamination.
It is not clear how a NaI detector or intrinsic germanium detector will be utilized to identify the presence or absence of subsurface contamination, and the extent of such contamination.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.4 (Final Status Survey Design), licensees should provide:
A description of scanning instruments, methods, calibration, operational checks, coverage, and sensitivity for each media and radionuclide A description of the instruments, calibration, operational checks, sensitivity, and sampling methods for in situ sample measurements, with a demonstration that the instruments and methods have adequate sensitivity Path Forward:
- a. Describe the technical basis for the usage of a NaI detector or intrinsic germanium detector to identify the presence or absence of subsurface contamination at depth and to establish the extent of such contamination.
ZSRP Response (5.19a) - The paragraph cited pertains to the use of a NaI to scan the exposed surfaces of an open excavation that are below the ground surface. This is stated in the second sentence of the paragraph which states; The remediation process will include performing RASS of the open excavations in accordance with section 5.4.2 of this FRS Plan.
ZSRP does not propose to use a NaI detector to determine the presence of subsurface soil contamination. To clarify, the 7th sentence of the 3rd paragraph of section 5.7.1.5 will be revised as follows; Alternatively, a NaI detector or intrinsic germanium detector of sufficient sensitivity to detect residual radioactivity at the DCGLw may be utilized to scan the exposed soils in an open excavation to identify the presence or absence of soil contamination, and the extent of such contamination.
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- 20. Comment: Section 5.7.1.6 (Stored Excavated Soils)
It is noted in Section 5.7.1.6 of the LTP that in several areas, clean overburden soils may be removed and stockpiled on site for use as backfill materials, and that prior to reuse, excavated soil will be surveyed to determine its suitability. It is further noted that scanning requirements and soil sample frequency shall also be determined in accordance with the classification of the area where the soil had originated, and that controls will be instituted to prevent mixing of soils from more restrictive survey area classifications (e.g., Class 2 material could be used in either Class 1 or 2 areas and Class 1 material could only be used in Class 1 areas).
Additional details should be provided on the characterization methodology and instrumentation that will be utilized for soils reused from radiologically impacted areas. The NRC staff notes that other licensees have been allowed to reuse excavated soils from impacted areas on site, but only after a demonstration that surveys and characterization of soils are comparable to the rigor of a Final Status Survey. This process can be complex, and has generally necessitated the excavation and survey of soils via automated sorting systems or in systematic lifts where the depth of soil is limited to a height that can be adequately scanned.
Basis: Per the acceptance criteria/information to be submitted described in NUREG-1757, Vol.
2, Rev. 1, Section 4.4 (Final Status Survey Design), licensees should provide:
A brief overview describing the FSS design A description of scanning instruments, methods, calibration, operational checks, coverage, and sensitivity for each media and radionuclide A description of the instruments, calibration, operational checks, sensitivity, and sampling methods for in situ sample measurements, with a demonstration that the instruments and methods have adequate sensitivity Path Forward:
- a. Provide additional details on the methodologies/instrumentation for the reuse of soils from radiologically impacted areas, and demonstrate that surveys will be performed to the rigor of a Final Status Survey.
ZSRP Response (5.20a) - As stated in section 5.7.1.6, ZSRP is making a commitment to only use soils with no detectable plant-derived radioactivity at concentrations greater than background for use as backfill for building basements. ZSRP will demonstrate that the soil is free of detectable plant-derived radioactivity through the use of a graded survey approach. Sufficient radiological surveys will be performed to demonstrate that the soils originating from impacted areas and intended for use as backfill meets the criteria for unconditional release off-site as clean material. Once that is determined, the soil will be stockpiled and controlled for reuse as backfill.
If the surveys demonstrate the presence of detectable plant-derived radioactivity, then the soil will be properly disposed of as radioactive waste or, if the soil is to remain on-site, it will be surveyed in accordance with MARSSIM. Once the material has been deposited into the
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 145 of 156 excavation or used as ground cover, it will be surveyed again during FRS to the classification of the area in which it is placed to demonstrate compliance with the dose-based release criteria.
As clarification, the following sentences will be added to the first paragraph of section 5.7.1.6; ZSRP will demonstrate that the soil is free of detectable plant-derived radioactivity through the use of a graded survey approach. Sufficient radiological surveys will be performed to demonstrate that the soils originating from impacted areas and intended for use as backfill meets the criteria for unconditional release off-site as clean material. The scope of the survey will be designed and documented using DQOs and will be comparable to the rigor of a Final Status Survey
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- 21. Comment: Section 5.7.1.6 (Stored Excavated Soils)
Section 5.7.1.6 of the LTP states that soils satisfying the criteria for unconditional release may be stockpiled for use as onsite backfill material. Additional details are needed on the control of cross-contamination when soils are stockpiled for re-use.
Basis: Per the acceptance criteria described in NUREG-1700, Rev. 1, Section 5 (Final Status Survey Plan), licensees should provide:
Access control procedures to control recontamination of clean areas Path Forward:
- a. Describe the administrative and engineering controls that will be maintained to prevent contamination of stockpiled soils from ongoing site remediation activities, or conversely, the contamination of non-impacted areas in the event that impacted soils are stockpiled in a non-impacted area of the site.
ZSRP Response (5.21a) - The access controls procedures that will be used to control recontamination of clean areas is described in section 5.6.3. For clarification, the following will be added to the end of the 1st paragraph of LTP section 5.7.1.6:
Stockpiled soils will be controlled using the methods described in section 5.6.3.
The measures described in this section will also apply to stockpiled soils that have been determined to be free of detectable plant-derived radioactivity and designated for use as backfill for excavations. A procedure has been developed for the implementation of the controls described. Posting and access control of areas that have been turned-over for FSS are presented in procedure ZS-LT-300-001-003, Isolation and Control for FRS (provided in ). This procedure provides detailed instructions for the control measures described in section 5.6.3 including isolation and control measures, walk-downs, turnover surveys, posting, access control as well as routine and special surveillances.
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- 22. Comment: Section 5.10.2.1 (Data Validation)
In Section 5.10.2.1 of the LTP, one of the data validation actions is to ensure that the instrumentation MDC for fixed or volumetric measurements was below the DCGLw or if not, it was below the DCGLEMC for Class 1, below the DCGLw for Class 2 and below 0.5 DCGLw for Class 3 survey units. This statement is not consistent with MARSSIM, which states that for direct measurements and sample analyses, minimum detectable concentrations (MDCs) less than 10% of the DCGL are preferable while MDCs up to 50% of the DCGL are acceptable.
Basis: Instrumentation selection is discussed in the MARSSIM Roadmap, where it is noted that for direct measurements and sample analyses, minimum detectable concentrations (MDCs) less than 10% of the DCGL are preferable while MDCs up to 50% of the DCGL are acceptable.
Path Forward:
- a. Update the review criteria and direct measurement MDCs to be consistent with MARSSIM.
ZSRP Response (5.22a) - Section 5.8.1, 2nd paragraph is the section of the LTP that pertains to the instrument sensitivity requirements for instrument selection as addressed in the MARSSIM Roadmap. Section 5.10.2.1 pertains to the a priori scan MDC requirement for detection at the investigation level commensurate with the classification of the area surveyed in accordance with MARSSIM section 5.5.2.6 and, to ensure during data assessment that the scan MDC was sufficient for the applicable investigation level. In section 5.8.1, the following clarification will be added after the 2nd sentence of the 2nd paragraph; The target MDC for field instruments is the maximum acceptable value. The actual MDCs expected to be used during FSS will be much lower.
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- 23. Comment: Section 5.10.5 (Data Conclusions)
Section 5.10.5 of the LTP indicates that if a data assessment determines that additional samples are necessary to provide sufficient power one course of action might be to determine the number of additional samples and collect them at random locations, and that this method may increase the Type I error, therefore agreement with the regulator will be necessary prior to implementation.
NRC staff concurs that agreement with the regulator will be necessary prior to implementation of a plan to collect additional samples in a failed survey unit. The use of two-stage or double sampling should also be considered as part of the DQO process when developing the design of the FSS.
Basis: NUREG-1757, Vol. 2, Rev. 1, states in Section A.7.5 (Use of Two-Stage or Double Sampling) that the use of two-stage or double sampling should be considered as part of the DQO process when developing the design of the FSS. Further guidance is provided in Appendix C of that volume, which states that:
The term double sampling will be used to refer to the case when the survey design is a one-stage design, but allowance is made for a second set of samples to be taken if the retrospective power of the test using the first set of samples does not meet the design objectives. Such allowance, if given, should be specifically mentioned in preparing the Data Quality Objectives (DQOs) and in advance of any sampling and analysis. During the DQO process, double sampling could be considered as an option in setting the Type I error rates.
Path Forward:
- a. In accordance with NUREG-1757 guidance, describe how the usage of two-stage or double sampling will be considered as part of the DQO process when developing the design of the FSS.
ZSRP Response (5.23a) - The last four sentences of the 4th paragraph of section 5.10.5 will be deleted.
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Section 5.11.2 of the LTP discusses information that will be provided in the FSS and STS final reports. However, there is no description of how ALARA practices were employed to achieve final activity levels.
Basis: NUREG-1757, Vol. 2, Rev. 1 Section 4.5.2 discusses acceptance criteria for final status survey reports and notes that a description of how ALARA practices were employed to achieve final activity levels should be provided.
Path Forward:
- a. Commit to providing a description of how ALARA practices were employed to achieve final activity levels, per NUREG-1757.
ZSRP Response (5.24a) - An 8th bullet will be added to section 5.11.2 that states the following:
Description of how ALARA practices were employed to achieve final activity levels.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 150 of 156 Environmental Zion RAIs Chapter 8 Basis: Pursuant to 10 CFR 50.82 (a)(9)(ii)(G), the LTP must include a supplement to the environmental report, pursuant to 10 CFR 51.53, describing any new information or significant environmental change associated with the licensees proposed termination activities.
- 1. Section 8.6.3.4 (Water Quality)
Path Forward:
Provide examples of appropriate Best Management Practices (BMPs).
ZSRP Response (8.1.) - Examples of BMPs regarding soil erosion and sedimentation control include the use of silt fencing and hay bales to prevent materials from migrating from work areas to surface waters. In addition, filter bags are used within catch basins to prevent soils and sediments from entering the storm drain system. Routine management observations are conducted site-wide to proactively monitor all site activities and to ensure BMPs are being utilized.
- 2. Section 8.6.3.5 (Air Quality)
Path Forward:
Provide examples of appropriate BMPs; equipment will be maintained to minimize not prevent increased exhaust emissions.
ZSRP Response (8.2.) - As the site has transitioned from a Lifetime Operating Permit (LOP) to a Registration of Smaller Sources (ROSS) program, there are minimal sources of air emissions produced onsite. Some examples of BMPs regarding air quality at the site include performing periodic inspections of the aboveground gasoline tank to make sure no leaks are present and that there are spill kits nearby. In addition, due to the nature and security of the site, vehicles are not allowed to remain idling in most demolition areas. All equipment utilized onsite will be maintained to minimize exhaust emissions based on manufactures recommendations.
Further BMPs for air quality include using dust suppression techniques (using water mist) to mitigate impacts from demolition activities.
- 3. Section 8.6.3.7(Terrestrial Ecology)
Path Forward:
Provide specific BMPs.
ZSRP Response (8.3.) - Best management practices for preventing invasive species will include the planting of acceptable seed mix after top soil or fill is placed. This will be completed to limit the potential for invasive species to take root. Seed mix will we watered as needed and assessed to ensure plants have sufficient germination.
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- 4. Section 8.6.3.11 (Socioeconomic Impacts)
Path Forward:
Provide break out into categories, (e.g., demographics, workforce, housing, public services, and infrastructure).
ZSRP Response (8.4.) - Please see reference documentation provided in Enclosure 2 for the town of Zion regarding socioeconomic categories. As discussed, due to the nature of the specialized project, and because the facility has not been in operation since 1998, this demolition and dismantlement project has a minimal socioeconomic impact on the town of Zion.
- 5. Section 8.6.3.13 (Cultural, Historic and Archaeological Resources)
Path Forward:
Provide a copy of the letter from the State of Illinois, Department of Conservation stating that no archaeological, architectural or historic resources are evident on the Zion Nuclear Power Station site.
ZSRP Response (8.5.) - The letter from the State of IL was submitted on November 12 2015 (ZS-2015-0163).
- 6. Section 8.6.3.15 (Noise)
Path Forward:
Provide a list of equipment types (with manufacturers spec sheets) that will be used on site during demolition and dismantling of facility. Was a pre-construction noise survey conducted?
ZSRP Response (8.6.) - A pre-construction noise survey has not been documented. However, prior to construction, the plant was not active and did not have active heavy equipment.
Zion is using and has used multiple types of equipment to perform demolition and dismantling activities. Examples of our larger equipment are listed below:
- High reach stacker
- Aerial lifts
- Heavy Lift Rail System
- Manitowoc
- Chelino
- Hoe rams Additionally smaller fork lifts, skid-steers, and jack hammers of various types are routinely used.
Spec Sheets are attached.
ZionSolutions complies with the noise ordinances from the City of Zion which limit noise during evening hours.
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- 7. Section 8.6.3.16 (Irretrievable Resources)
Path Forward:
Are there any resources that were considered irreversible?
ZSRP Response (8.7.) - Irretrievable resources that would occur during the decommissioning process are the materials used to decontaminate the facility (e.g., rags, solvents, gases, and tools),
and fuel used for construction machinery and for transportation of materials to and from the site.
These resource commitments are considered to be minor and are neither detectable nor destabilizing.
- 8. Section 8.6.3.17 (Traffic and Transportation)
Path Forward:
Provide the number of workers that will be required to dismantle and decommission the facility; ZSRP Response (8.8.) - A timeline, which shows the number of workers required to dismantle and decommission the facility, is provided in Enclosure 2. As seen on the chart, we have passed the peak number of workers on site and the number of workers is steadily decreasing as we get closer to major building demolition.
Will the on-site parking be adequate for the workers; ZSRP Response (8.8.) - Based on the forecasting of work activities required to complete the demolition and dismantlement of the facility, adequate on-site parking will always be available for all personnel.
What roads will workers and haul vehicles use from I-94; ZSRP Response (8.8.) - As workers travel to the site from a very large geographical area, it is difficult to assess travel pathways for the entire staff. Waste shipment travel is dictated by the carrier per IAW 49CFR. The site does not have a traffic plan as it is not required.
The only entrance to the site is from Shiloh Boulevard. This road is lightly traveled at most times of day. Shiloh Boulevard connects to Sheridan Road which is a major road running parallel to the Lake Michigan shoreline. The closest access to I-94 from Sheridan Road is via Illinois Route 173.
How often will waste be shipped to Clive, UT by rail; and ZSRP Response (8.8.) - While it is difficult to assess waste shipments as they are variable, it is likely that waste will be shipped to Clive, UT by rail on a weekly basis.
How many rail cars will constitute a shipment?
ZSRP Response (8.8.) -For the ZSRP, each rail car constitutes a shipment. From the fourth quarter of 2015 until the end of the decommissioning project, an estimated 1,059 rail cars (gondola or flatbed) will be shipped.
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- 9. Section 8.6.3.18 (Placement of Clean Construction Demolition Debris (CCDD) and Sand Mix in Major Building Basements: Terrestrial Ecology and Transportation)
Path Forward:
How much CCDD and how much sand mix will be required?
ZSRP Response (8.9.) - Currently, the total backfill required to complete the decommissioning project is estimated to be approximately 210,000 cubic yards. The estimated CCDD to be used for backfill material is approximately 76,000 cubic yards and the estimated soil fill is approximately 134,000 cubic yards. There is approximately 35,000 cubic yards of soil currently stored on site for use as backfill material, so approximately 99,000 cubic yards of soil will be imported to complete the backfill of the site.
How many trucks will be required, and what size trucks will be used?
ZSRP Response (8.9.) - It is uncertain what size trucks will be used to import the soil to be used for backfill material at this time, or if the same size truck will be used throughout the project. If we assume each truck can deliver approximately 13 cubic yards of material, Zion would need to bring in approximately 7,615 trucks of soil to receive 99,000 cubic yards of backfill.
Where will materials be stockpiled?
ZSRP Response (8.9.) - Stockpiled materials that will be used as backfill material that are currently onsite are stored at the south end of the former parking lot, in the southeast portion of the Radiologically Restricted Area. Soils that are imported to the site for use as backfill material will also be stockpiled there.
What type of equipment will be used to prepare the CCDD?
ZSRP Response (8.9.) - To prepare the CCDD materials onsite, heavy machinery will be used (i.e. excavators) to process and size the materials.
To what size will it be prepared; to what depth will it be spread? and, ZSRP Response (8.9.) - The CCDD materials will be prepared to pieces that are 10 inches in diameter or less prior to being utilized as backfill material. The CCDD materials will be used to fill in voids to a maximum depth of 3 feet below ground surface.
Will there be a top soil and vegetation cover?
ZSRP Response (8.9.) - Per the asset sale agreement, lease agreement, and environmental permits, areas that require backfill will contoured to blend in with adjoining property. Up to nine inches of top soil will be placed in some areas that will be seeded to promote vegetative growth as necessary to achieve final stabilization.
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- 10. Table 8-1 Path Forward:
What document do the sections refer to?
ZSRP Response (8.10.) - Table 8-1 provides a summary of Environmental Impacts for the Zion Decommissioning Project, explained within Chapter 8 of the LTP.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 155 of 156 Comments on TSD 14-022 (Use of In Situ Gamma Spectroscopy for Source Term Survey of End State Structures)
- 1. Page 5 of 92 Comment: On page 5 of 92, it is mentioned that 100 percent of the surfaces will be measured using ISOCS. Due to the circular field of view (FOV), to truly achieve 100 percent measurement coverage there will have to be some amount of overlap between the FOVs of multiple measurements. Additional details are needed on how overlap between FOVs has been considered.
Basis: Circle packing theorems can be investigated to evaluate whether overlap is necessary.
For example, a hexagonal packing arrangement of measurement FOVs will account for 91 percent of the area, with no overlap. This combined with survey data can be inputs to a FOV overlap decision. The amount of overlap (if any) should be determined by the likelihood of missing a high activity spot which falls outside the FOV of neighboring measurements. Data to feed this decision could be derived from survey data.
Path Forward:
- a. Provide additional information as to how the 100 percent measurement requirement will be satisfied, and how the requirement for overlap of FOVs will be determined.
ZSRP Response (TSD 14-022-1) - The initial plan was to apply a triangular grid pattern with a spacing that resulted in overlap such that 100% of the surface would be covered by an ISOCS measurement. However, a preliminary survey design indicated that the number of measurements required in the Class 1 Auxiliary Building basement floor to accomplish 100% coverage by overlap was nearly double that required for the closest circle packing with no overlap.
Therefore, the approach to grid spacing and areal coverage in the two Class 1 survey unit areas (Auxiliary Basement floor and SFP/Transfer Canal) was revised.
A triangular grid spacing will be used which results in hexagonal packing. As stated in the NRC comment, hexagonal packing results in 9% of the survey unit area being outside of an ISOCs FOV. This gap will be addressed by establishing an investigation level of 0.75 SOF value for individual measurements. If a given measurement exceeds a SOF of 0.75, then the five gap areas surrounding the given measurement FOV will be specifically targeted for additional ISOCS measurement to ensure that significant activity is not missed.
- 2. Page 14 of 92, Table 1 Comment: It was previously stated in the document that the highest level of Cs-137 activity is located within the first one-half inch of concrete. As the core analysis data shows in Table 1, the ratio of activity contained in the first one-half inch to the activity contained in the second half-inch of the core is greater than 3 in all but three core samples.
ZionSolutions, LLC ZS-2016-0022: Enclosure 1 Page 156 of 156 The three areas which show deep Cs-137 penetration (Samples B105103-CJFCCV-001, B105105-CJFCCV-001, and B105106-CJFCCV-001) exhibit a longer drop-off than the other samples. For these areas the activity at depth may need to be addressed differently than others, as the document mentions in Note 2 on Table 6 on page 26 of 92, and additionally in the conclusion in Section 3.5.
Additional details are needed on the quality assurance (QA) methods to verify the geometry application for areas where activity at depth may need to be addressed differently than others.
Basis: Areas with activity depth profiles that differ from the majority of core samples may have to be treated differently, which is indicated at several points within the document. One of the keys to the measurement campaigns success is ensuring that the correct geometries are applied to the correct areas. If there are going to be specific cases such as the pipe ducts called out, it must be ensured within the program flow that the specific geometries for those areas are applied to the measured spectra.
Path Forward:
- a. Provide details on the QA methods for verifying the geometry application in areas where activity at depth may need to be addressed differently than others.
ZSRP Response TSD 14-022 ZionSolutions Procedure ZS-LT-300-001-001 Final Radiation Survey Package Development (provided in Enclosure 2) Section 4.2.7 states, The Canberra In-Situ Object Counting System (ISOCS) has been selected as the primary instrument that will be used to perform STS. ZionSolutions Technical Support Document (TSD)14-022, Use of In-Situ Gamma Spectroscopy for Source Term Survey of End State Structures (Reference 6.5) provides the initial justification for selecting a reasonably conservative geometry for efficiency calibrations for the ISOCS based on the physical conditions of the remediated surface and the depth and distribution of activity in the concrete surface. Review post remediation conditions and surveys to determine if the geometry of remaining residual radioactivity has significantly changed from that assumed in TSD 14-022. If the geometry appears to be significantly different from that which was assumed in TSD 14-022, then inform the C/LT Manager.
If the Radiological Engineer identifies a suspect condition, the C/LT manager will evaluate the conditions on a case-by-case basis. Actions could include additional core samples and/or a modification to the efficiency calibration geometry to be consistent with geometry encountered in the field.
It is important to note that after an ISOCS spectrum is collected, the efficiency calibration can be adjusted as necessary to ensure the activity measurement is sufficiently conservative. If there are questions as to the geometry, sensitivity analyses can be conducted over a range of potential geometries at any time after the measurement is collected to ensure that a conservative efficiency geometry is applied.