View source for ZS-2022-0001, License Amendment Request for the License Termination Plan

{{Adams
| number = ML22084A058
| issue date = 03/23/2022
| title = License Amendment Request for the License Termination Plan
| author name = Vannoordennen G
| author affiliation = ZionSolutions, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NMSS
| docket = 05000295, 05000304
| license number = DPR-039, DPR-048
| contact person = 
| case reference number = ZS-2022-0001
| document type = Letter type:ZS, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 58
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:*March 23, 2022                                                                        ZS~2022-000 i US. Nuclear Re~latory Commission ATTN: Document Control Desk
    .Washington, D.C ~0555-0001.
Zion Nuclear Power Station, Units 1 and 2 Facility Operating License Nos. DPR-39 and DPR-48 NRC Docket Nos. 50-295 and 50-304 *                *

==Subject:==
License Amendment Request for *the License Termination Plan

==Reference:==
Gerard van Noordennen, Zion.Solutions, Letter to U.S: Nuclear Regulatory Commission, ''Response to Reque.st for Additiona:l 'Information Related to Partial Site R~lease and Recent Site .Survey Activities, and Notification of Foundation Pad Discovery," dated March 8, 2022 Pursuant to 10 CFR 50.90, ZionSolutions hereby requests approvaLofthe enclo'sed amendment to Facility Operating Licenses, D:PR-39 and DPR-48, for Zion Nuclear Power Station (ZNPS),
Units 1 and 2, respectively. J:be enclo.sed license amendment proposes to revise the ZNPS
  .Licens~ Termination Plan (L TP): The proposed revision to the L TP includes incorporation of a
   .Discrete Radioactive Particle (DRP) Survey r1an by reference. In addition, the proposed revision.
prov~des an estimate of the number of DRPs that may hypothetically remain on site and the .
methods for calculating dose and risk from those hypothetical DRPs.
* .'ZionSolutions .submitted the DRP Survey Plan 'to the NRC in a response to a request for additional information (Reference). In addition, ZiqnSolutions comrhitted to provide the Technical Support P,ocument that describes the results of the discrete radioa~tive particle s{uyey and the* associated calculations in the Reference.
* The Enclosure provides a detailed description an,d ev~luation of the proposed changes to the LTP. Attachments t 'and 2 of the enclosure contain-thiproposed markup of the LTP and the retyped clean.pages, respectively:
This letter contains no regulatory commitments .
   .ZionSolutio~ requests approv~l of the proposed license amendment request.(LAR) by M_ay 30,
* 2022. Z'i.onSolutions expects to implement the-proposed amendment.within 30 days of approval oftheLAR.
fu accordance with 10 CFR 50.91 (b), a copy of this LARis. being provided to the State of illinois.
If you should ha;e any questions regarding this submittal, please contact me at (860) 462-9707.
* I declare under penalty. of perjury that the foregoing is true and correct. Executed on March 23, 2022. .                                                                                           .
101 Shiloh Boulevard, Zion* IL 60099 (224) 789-4016 ~ Fax (224) 789-4008
* www.zJonsolubonscompany com

ZionSolutions, LLC
*.zs-2022;0001 Page 2 of 2 Res~tfully,
                                      'D,grt:ally signed by Gerard van Noordennen Gerard van                           DN. cn<>Gerard van Noordennen, o=EnergySohrtJons,
* u=Regulatory Affairs,       '    ,               '
No9rdennen                    , ---""""".,.,J)Vllnnoordenrionoene<gysoluttorw:om, emus Date. 2022.03 23 07-03&deg;03 -04'00' Ge~d   van  Noordennen                          .
Senior Vice Presiqent Regulatory Affairs.
En~rgySolutions, LLC Enqlosu,re: Zion Nuclear Power Station, Units 1 and 2 - License Amendmen~ Req~~:
: License Termination.Plan :
cc:   Kimberly Conway, U.S: NRC Senior Project Manager Regional A~ministr~tor, U.S. NRC, Region III
* Zion Nuclear Power S~tion; Units 1 and 2, License Transfer Service List

Zion Nuclear Power Station, Units 1 and 2 Service List cc:
Ken Robuck                                      Steven A. Reynolds President and CEO                               Manager, Nuclear Facility Inspection EnergySolutions                                 Division of Nuclear Safety 299 South Main Street, Suite 1700               Illinois Emergency Management Agency Salt Lake City, UT 84111                        245 W. Roosevelt Road, Units 55 & 56 West Chicago, Illinois 60185 John Sauger President and Chief Nuclear Officer             Kelly F. Grahn ReactorD&D                                      Senior Health Physicist EnergySolutions                                 Illinois Emergency Management Agency 121 W. Trade Street, Suite 2700                 Division of Nuclear Safety Charlotte, NC 28202                             245 W Roosevelt Road, Building 8, Suite 55 West Chicago, 1L 60185 Todd Eiler Project Director                                Dan Eder D&D Engineering/Projects                        :MPH, IPEM, LEHP EnergySolutions                                 Interim Manager 121 W. Trade Street, Suite 2700                 Lake County Emergency Management Agency Charlotte, NC 28202                             1303 N. Milwaukee Avenue Libertyville, 1L 60048-1308 Gerard van Noordennen Senior VP Regulatory Affairs EnergySolutions 121 W. Trade Street, Suite 2700 Charlotte, NC 28202 Russ Workman General Counsel EnergySolutions 299 South Main Street, Suite 1700 Salt Lake City, UT 84111

ZionSolutions, LLC ZS-2022-0001 Enclosure Zion Nuclear Power Station, Units 1 and 2 License Amendment Request:
License Termination Plan

ZS-2022-0001 Enclosure Page 2 of 13 Table of Contents I. 

==SUMMARY==
DESCRIPTION
: 2. DETAILED DESCRIPTION
: 3. TECHNICALEVALUATION
: 4. REGULATORYEVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions
: 5. ENVIRONMENTAL CONSIDERATION
: 6. REFERENCES ATTACHMENTS Attachment I: License Termination Plan Markup Pages : License Termination Plan Clean Pages

ZS-2022-0001 Enclosure Page 3 of 13
: 1. 

==SUMMARY==
DESCRIPTION In February 1998, in accordance with 10 CFR 50.82(a)(l)(i) and (ii) (Reference 1),
Exelon Generating Company LLC (Exelon) notified the Nuclear Reg&#xb5;latory Commission (NRC) of the permanent cessation of operations at the Zion Nuclear Power Station (ZNPS). The NRC transferred Facility Operating License Numbers DPR-39 and DPR-48 on September 1, 2010, from Exelon to ZionSolutions, LLC (ZionSolutions) (Reference 2) so ZionSolutions could begin active decommissioning of the site.
10 CFR Section 50.82(a)(9) requires that a licensee submit an application for the termination of the site's Part 50 license. The application for termination of the license must be accompanied or preceded by a license termination plan (LTP) to be submitted for NRC approval. The L TP describes the process used to meet the requirements for terminating the 10 CFR Part 50 license and releasing the site for unrestricted use. The L TP is a supplement to the plant's UFSAR or an equivalent document (Defueled Safety Analysis Report (DSAR)).
The NRC issued Amendment Nos. 191 and 178 to the ZNPS licenses on September 28, 2018 (Reference 3). The amendments revised the ZNPS, Units 1 and 2, licenses to add License Condition 2.C.(17). The license condition incorporated the LTP into the ZNPS license and specified limits on changes ZionSolutions could make to the L TP without prior NRC review and approval.
Currently, Chapter 5 of the ZNPS L TP provides the Final Status Survey (FSS) Plan for the Zion Station Restoration Project (ZSRP). Chapter 6 of the ZNPS LTP provides the methods used by the ZSRP to demonstrate compliance with the dose criterion in 10 CFR 20.1402. The proposed amendment would add Subsection 5.4.7, Survey Considerations for Suspected Discrete Radioactive Particle Areas, to define discrete radioactive particles (DRPs) and incorporate the DRP Survey Plan, for the detection and remediation ofDRPs, into the L TP by reference. In addition, the proposed amendment would revise L TP Subsection 6.4.6, Alternate Scenarios, to add a fourth scenario which would consider the less likely but plausible (LLBP) scenario where a hypothetical DRP is ingested, inhaled, or deposited on the skin. L TP Subsection 6.4.6 would refer to a new attachment within Chapter 6, Attachment 5, titled Less Likely but Plausible Scenario for Exposure to Hypothetical Discrete Radioactive Particles. The proposed attachment would provide an estimate of the number of DRPs that may hypothetically remain on site following survey completion and the methods for calculating dose and risk from the hypothetical DRPs.
: 2. DETAILED DESCRIPTION As presented in Subsection 6.1 of the LTP, the site release criteria for the ZSRP are the two radiological criteria for unrestricted use specified in 10 CFR 20.1402:
* Dose Criterion: the residual radioactivity that is distinguishable from background radiation results in a Total Effective Dose Equivalent (TEDE) to an Average Member of the Critical Group (AMCG) that does not exceed 25 millirem/year (mrem/yr), including that from groundwater sources of drinking water; and
* ZS-2022-0001 Enclosure Page 4 of 13
* As Low As Reasonable Achievable (ALARA) Criterion: the residual radioactivity has been reduced to levels that are ALARA.
The Final Status Survey (FSS) Plan for the ZSRP is presented in Chapter 5 of the LTP.
The purpose of the FSS plan is to describe the methods used in planning, designing, conducting, and evaluating the FSS. The FSS Plan describes the final survey process used to demonstrate that the ZNPS facility and site comply with the radiological criteria for unrestricted use specified in 10 CFR 20.1402 (i.e., the Dose Criterion and the ALARA Criterion). ZionSolutions has concluded that because the survey methodology to detect DRPs is different from the survey techniques described in Chapter 5 of the LTP, a revision to the L TP is necessary.
ZionSolutions proposes to amend Chapter 5 of the LTP by incorporating a new section, Subsection 5.4.7, titled Survey Considerations for Suspected Discrete Radioactive Particle Areas. This section would define DRPs as specks ofradioactive material identified usually as either activated corrosion product such as cobalt-60, or an irradiated fuel fragment exhibiting greater than 10,000 corrected counts per minute using a beta-sensitive GM pancake detector (100,000 dpm). In addition, the Survey Plan for Discrete Radioactive Particle Identification and Remediation, 7.S-LT-07 (the "DRP Survey Plan"),
is proposed to be incorporated by reference into the L TP. The survey plan describes the special survey techniques used for the detection of DRPs in open land survey units where DRPs are suspected.
Chapter 6 provides the methods used by the ZSRP to demonstrate compliance with the Dose Criterion. LTP Subsection 6.4 provides a general overview of the dose Modeling approach. Dose Modeling is performed to demonstrate that remaining residual radioactivity does not result in a dose exceeding 25 mrem/yr as required by the Dose Criterion. The four potential sources ofresidual radioactivity discussed in Subsection 6.4 are backfilled basements, buried pipe, soil, and groundwater. Alternate scenarios, which were qualitatively considered, are provided in Subsection 6.4.6, Alternate Scenarios. The alternate scenarios studied involved changes to the "as left" geometry of the backfilled basements.
ZionSolutions proposes to amend Chapter 6 of the LTP by revising Subsection 6.4.6 to add a fourth alternate scenario which considers the LLBP scenario where a hypothetical DRP is ingested, inhaled, or deposited on the skin. Additionally, Subsection 6.4.6 would refer to a new attachment, Attachment 5, Less Likely but Plausible Scenario for "Exposure to Hypothetical Discrete Radioactive Particles. L TP Chapter 6, Attachment 5, is proposed to provide an estimate of the number ofDRPs that may hypothetically remain on site and the methods for calculating dose and risk from those hypothetical DRPs. The proposed amendment states the results of the DRP Survey Plan will be provided in Technical Support Document 22-001, Discrete Radioactive Particle Survey Report (the "DRP Survey Report").

ZS-2022-0001 Enclosure Page 5 of 13 In summary, the proposed changes add Subsection 5.4.7 to define DRPs and to incorporate the DRP Survey Plan into the L TP by reference. In addition, the proposed amendment would revise L TP Subsection 6.4.6 to add a fourth alternate scenario which would consider the LLBP scenario where a hypothetical DRP is ingested, inhaled, or deposited on the skin. L TP Chapter 6 would contain a new attachment that would provide an estimate of the number of DRPs that may hypothetically remain on site following survey completion and the methods for calculating dose and risk from those hypothetical DRPs.
The proposed changes to the L TP are shown on the marked-up pages included in Attachment 1. The clean retyped pages are provided in Attachment 2.
: 3. TECHNICAL EVALUATION DRPs were periodically identified on the Zion site during decommissioning operations as well as during Final Status Survey. ZionSolutions concluded that the survey techniques and dose calculations for DRPs were sufficiently different from those described in the LTP. Therefore, a change to the LTP is proposed to incorporate a DRP Survey Plan, and to add the LLBP scenario for exposure to hypothetical DRPs.
DRP Survey Plan As there is a lack of specific NRC guidance on planning and conducting surveys for DRPs, ZionSolutions devised a DRP Survey Plan. The objective of the DRP Survey Plan is to identify and remediate all DRPs that are detected. This approach provides reasonable assurance that there will be no known DRPs remaining on the Zion site that could pose an unacceptably high risk to a member of the public.
The survey approach relies on proven technology and incorporates available industry experience in conducting field surveys. The Data Quality Objective Process and the equipment used to implement the plan are described in detail in the DRP Survey Plan. A discussion on survey techniques and the areas surveyed is presented below.
Survey Techniques and Detailed Investigations The survey approach uses a towed array of six detectors mounted on a utility terrain vehicle and attached to a counter data logger (the "towed array"). The unit is equipped with a single GPS receiver and antenna combined with a high-accuracy inertial measurement unit (IMU). The data logger, GPS receiver, and the IMU are integrated using an on-board tablet or laptop computer running the scanning software. Any time the investigation level is exceeded during a towed array survey, a detailed investigation is performed. The investigation is designed to identify and remove DRPs that potentially were the source of the elevated reading by the towed array.
The towed array has significant advantages over hand scanning. It provides the ability to cover a large area far more efficiently than hand scanning. This method is consistent with the proposed guidance contained in the draft revision to NUREG-157 5 (MARSSIM)
(Reference 4), where NRC staff was involved. The draft revision is informative as it recognizes advances in technology. These advances have encouraged the use of automated scanning as a viable option for large survey areas.

ZS-2022-0001 Enclosure Page 6 of 13 Areas inaccessible to the towed array are surveyed using hand-scanning methods. The protocols for gamma scanning with hand-held detectors are delineated in the DRP Survey Plan. Any elevated activity detected by hand scanning would result in a detailed investigation.
Survey Areas Thirty-six survey units were selected for surveying using the DRP Survey Plan. These areas were selected because they were considered at higher risk for the presence of DRPs based on decommissioning project experience (e.g., a location where clean concrete demolition debris was temporarily stored or transported through), or where previous surveys identified particles or areas of elevated activity. The DRP Survey Plan specifies that areas within the scope of the survey will be expanded, as necessary, as a result of investigations performed, to bound areas of elevated radioactivity or as a result of remediation.
The rationale for not including certain areas in the DRP Survey Plan is that there is a low potential for the presence of DRPs in those areas. This assumption is based on process knowledge and the results of previous radiological surveys, including PSS and ORISE surveys.
LLBP Scenario for Exposure to Hypothetical DRPs The proposed revision to the LTP provides an estimate of the number of DRPs that may hypothetically remain on site and the methods for calculating dose and risk from those hypothetical DRPs. The DRP activity assumed, and the results of the dose and risk calculations will be provided in a DRP Survey Report.
Justification for Designation as a LLBP Scenario In the proposed markup to Chapter 6, the DRP exposure scenario is evaluated as an LLBP scenario due to the low probability ofDRP exposure occurring. Treating the low-probability DRP exposure as an LLBP scenario is consistent with the approach approved in the Zion LTP for assessing the low-probability scenario of the well driller contacting the Auxiliary Building drains, which was also designated as an LLBP scenario. To justify the designation as an LLBP scenario, the probabilities of DRP ingestion and inhalation are compared to the probability of drilling into the Auxiliary Building drains, which was accepted by the NRC as not likely. The probability of a drill contacting the Auxiliary Building drains is l .5xl 0-3
* The lifetime probability of a future site resident ingesting or inhaling a DRP, assuming one DRP remains on site, is 1.6 xl0- 8 and 1.7 x10- 10 ,
respectively. These values, which assume a lifetime exposure period of 30 years, are much lower than that of the probability of drilling into the Auxiliary Building drains.
The final lifetime probability of ingesting or inhaling a DRP accounts for the projection of all DRPs hypothetically remaining on site. The estimate of the number ofDRPs that could hypothetically remain was calculated by multiplying the number of DRPs identified during execution of the DRP Survey Plan (i.e., one DRP, assumed to be contained in 1.0 cm of soil) by 30.5 (i.e., assumption that DRPs are limited to first 30.5 cm of soil). Therefore, the estimate of the number of unidentified DRPs that hypothetically remain is 31. The lifetime probabilities of a future resident on the Zion site

ZS-2022-0001 Enclosure Page 7 of 13 ingesting or inhaling a hypothetical DRP, assuming 31 DRPs hypothetically remain, is projected to be 4.8x10-7 and 5.2x10-9, respectively.
The probability of a DRP being deposited on the skin during the lifetime of a future site resident is 1.2xl o-7 . The dermal deposition parameters, used in the equation to derive the probability, are from the Environmental Protection Agency's Risk Assessment Guidance for Superfund (Reference 5). The single DRP lifetime probability is multiplied by 31, the number of hypothetical DRPs projected to remain on the site, to derive a final lifetime probability of 3.7xl0-6.
In accordance with Section 5.5.2 ofNUREG-1757 (Reference 6), an LLBP scenario is used to " ... better risk-inform the decision" and to ensure that" ... unacceptably high risks would not result," but are not considered compliance scenarios. ZionSolutions' approach is to demonstrate that the risk from the hypothetically remaining DRPs is not unacceptably high. Consistent with the designation as an LLBP exposure scenario, the dose from the hypothetical DRPs will not be added to the compliance dose.
Acceptance Criteria As no NRC guidance currently exists concerning the dose and/or risk assessment of DRPs, ZionSolutions proposes to perform five separate assessments to demonstrate that unacceptably high risks would not result from hypothetical exposure to a DRP. The dose and risk assessments include calculating the following:
* Expectation dose (mrem/yr) from ingestion and inhalation
* TEDE (mrem/yr) from DRP Activity Distributed in Soil
* Lifetime fatal cancer risk from ingestion and inhalation
* EDE (mrem) from skin exposure
* Lifetime fatal cancer risk from SDE due to skin exposure The final dose and risk assessments from DRP ingestion, inhalation, and skin exposure will be calculated, using the equations in the proposed amendment, and provided in the DRP Survey Report. The proposed amendment to LTP Chapter 6 includes the addition of Table A5-4, Acceptance Criteria for Demonstrating that "unacceptably high risks would not result" from the LLBP DRP Exposure Scenario. Table A5-4 is shown below. The acceptance criteria proposed in the table are not intended to be the absolute limits at which the LLBP DRP scenario is assessed. Meeting the acceptance criteria demonstrates that unacceptably high risks do not result; however, higher values may be acceptable on a case-by-case basis. Higher values which are determined to be acceptable will be justified in the DRP Survey Report.

ZS-2022-0001 Enclosure Page 8 of13 Table AS-4 Acceptance Criteria for Demonstrating that "unacceptably high risks would not result" from the LLBP DRP Exposure Scenario Lifetime DRP Exposure             Probability of Assessment Criteria          Acceptance Criteria Pathway              DRP Exposure Occurrin2 Ingestion and                NA 1 Expectation Dose               <25 mrem/yr Inhalation Ingestion. 4.8x10-7     Fatal Cancer Risk from Ingestion and Inhalation: 5.2xI0-9       DRP inhalation and                 <4x10"""'
Inhalation lngestion 2 Ingestion and                NA 3            TEDE from DRP Activity
                                                                                        <25 mrem/yr Inhalation                                    Distributed in Soil Skin                  3.7xl0~               Skin Dose (EDE)                  <25 mrem Skin                  3.7xl0~         Fatal Cancer Risk from SDE              <4x10"""'
: 1)   The expectation dose calculation includes annual probability of exposure occurring
: 2)   The 4xI0"""' fatal cancer risk is the risk corresponding to the 25 mrem/yr unrestricted use criterion as stated in SECY-97-046A
: 3)   The exposure to a small soil area of distributed activity is assumed to occur annually in the same manner as other small soil elevated areas Summary The DRP Survey Plan provides confidence that any remaining DRPs that might pose an unacceptably high risk to an AMCG have been identified and remediated. The DRP Survey Plan supplements the PSS Plan. In accordance with NUREG-1757, the evaluation of the LLBP scenario for exposure to hypothetical DRPs ensures that "unacceptably high risks would not result," but are not considered compliance scenarios. Accordingly, the dose from the hypothetical DRPs will not be added to the ZSRP compliance dose. Rather, the hypothetical DRP dose will be used to better risk inform the decision to terminate the license.
Since the L TP continues to be based on NRC guidance and establishes the methodology ZionSolutions will use to meet license termination criteria, this proposed license amendment is appropriate to allow completion of the ZNPS decommissioning project and license termination.
: 4. REGULATORY EVALUATION The proposed amendment has been evaluated to determine whether applicable regulations and requirements continue to be met. ZionSolutions has determined that the proposed changes do not require any exemptions or relief from regulatory requirements.

ZS-2022-0001 Enclosure Page 9 of 13 4.1   Applicable Regulatory Requirements/Criteria 4.1.1    10 CFR 50.82, Termination of License 10 CPR 50.82(a)(6)
Licensees shall not perform any decommissioning activities, as defined in 10 CPR 50.2, that -
(i) Foreclose release of the site for possible unrestricted use; (ii) Result in significant environmental impacts not previously reviewed; or (iii) Result in there no longer being reasonable assurance that adequate funds will be available for decommissioning.
The proposed changes add a new section in Chapter 5 to define DRPs and to incorporate the DRP Survey Plan into the L TP by reference. In addition, the proposed amendment would revise L TP Subsection 6.4.6 to add a fourth scenario which would consider the LLBP scenario where a hypothetical DRP is ingested, inhaled, or deposited on the skin. L TP Chapter 6 would contain a new attachment that would provide an estimate of the number ofDRPs that may hypothetically remain on site following survey completion and the methods for calculating dose and risk from the hypothetical DRPs.
Therefore, the proposed changes do not impact the criteria listed above in 10 CPR 50.82(a)(6).
10 CPR 50.82(a)(9)
All power reactor licensees must submit an application for termination of license.
The application for termination of license must be accompanied or preceded by a license termination plan to be submitted for NRC approval.
(i) The license termination plan must be a supplement to the FSAR or equivalent and must be submitted at least 2 years before termination of the license date.
(ii) The license termination plan must include -
(A) A site characterization; (B) Identification of remaining dismantlement activities; (C) Plans for site remediation; (D) Detailed plans for the final radiation survey; (E) A description of the end use of the site, if restricted; (F) An updated site-specific estimate of remaining decommissioning costs; (G) A supplement to the environmental report, pursuant to 10 CPR 51.53, describing any new information or significant environmental change associated with the licensee's proposed termination activities; and (H) Identification of parts, if any, of the facility or site that were released for use before approval of the license termination plan.

ZS-2022-0001 Enclosure Page 10 of 13 The initial L TP was submitted to the NRC and incorporated into the ZNPS DSAR as required by 10 CFR 50.82(a)(9)(i). With the proposed changes, the L TP continues to include the information required by 10 CFR 50.82(a)(9)(ii).
4.1.2    10 CFR 50.59, Changes, Tests and Experiments A 10 CFR 50.59 screening was performed to determine if the proposed changes to the L TP should be evaluated against the criteria of 10 CFR 50.59(c)(2). The proposed changes to the L TP screened out; the proposed changes do not meet the requirements for an evaluation in accordance with 10 CFR 50.59(c)(2).
4.1.3    ZNPS, Units 1 and 2, License Numbers DPR-39 and DPR-48, License Condition 2.C.(17), License Termination Plan (LTP)
ZionSolutions shall implement and maintain in effect all provisions of the approved License Termination Plan as approved in License Amendment Nos. 191 and 178 subject to and as amended by the following stipulations:
ZionSolutions may make changes to the L TP without prior approval provided the proposed changes do not meet any of the following criteria:
(A) Require Commission approval pursuant to 10 CFR 50.59.
(B) Result in significant environmental impacts not previously reviewed.
(C) Detract or negate the reasonable assurance that adequate funds will be available for decommissioning.
(D) Decrease a survey unit area classification (i.e., impacted to not impacted; Class 1 to Class 2; Class 2 to Class 3; or Class 1 to Class 3) without providing the NRC a minimum 14 day notification prior to implementing the change in classification.
(E) Increase the derived concentration guideline levels (DCGL) and related minimum detectable concentrations (for both scan and fixed measurement methods).
(F) Increase the radioactivity level, relative to the applicable DCGL, at which an investigation occurs.
(G) Change the statistical test applied other than the Sign test.
(H) Increase the approved Type I decision error above the level stated in the LTP.
(I) Change the approach used to demonstrate compliance with the dose criteria (e.g., change from demonstrating compliance using derived concentration levels to demonstrating compliance using a dose assessment that is based on final concentration data).
(J) Change parameter values or pathway dose conversion used to calculate the dose such that the resultant dose is lower than in the approved L TP and if a dose assessment is being used to demonstrate compliance with the dose criteria.

ZS-2022-0001 Enclosure Page 11 of 13 (K) Reuse concrete from demolished structures, other than from the list of areas specified in Section 2.1.1 ofTSD 17-010, "Final Report - Unconditional Release Surveys at the Zion Station Restoration Project, Revision l ", as backfill.
(L) Assign a dose for reuse concrete other than the dose values provided along with the LTP (as shown in Table 6-53 (Revision 2) of the L TP) and documented in Section 8 and Table 33 ofTSD 14-010, "RESRAD Dose Modeling for Basement Fill Model and Soil DCGL and Calculation of Basement Fill Model Dose Factors and DCGLs, Revision 6."
(M) Use area-specific surrogate ratios that are less than the maximum surrogate ratios (H-3/Cs-137, Ni-63/Co-60, Sr-90/Cs-137) presented in Table 5-15 (Revision 2) of the LTP.
The proposed changes to the LTP meet License Condition 2.C.(17), Criterion I; therefore, the proposed changes to the L TP require prior NRC approval.
4.2    Precedent None.
4.3    Significant Hazards Consideration ZionSolutions is requesting an amendment to the ZNPS L TP. An evaluation to determine whether or not a significant hazards consideration is involved with the proposed amendment was completed by focusing on the standards set forth in 10 CPR 50.92, Issuance of amendment. As discussed below.
4.3.1   Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed changes incorporate, into the LTP, a survey plan for discrete radioactive particles. In addition, the proposed changes provide an estimate of the number of DRPs that may hypothetically remain on site following survey completion and the methods for calculating dose and risk from the hypothetical DRPs. There are no physical changes proposed at ZNPS. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any accident previously evaluated.
4.3.2   Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed changes incorporate, into the L TP, a survey plan for discrete radioactive particles. In addition, the proposed changes provide an estimate of the number of DRPs that may hypothetically remain on site following survey completion and the methods for calculating dose and risk from the hypothetical DRPs. There are no physical changes proposed at ZNPS. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of any accident previously evaluated.

ZS-2022-0001 Enclosure Page 12 of 13 4.3.3    Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No The proposed changes incorporate, into the L TP, a survey plan for discrete radioactive particles. In addition, the proposed changes provide an estimate of the number of DRPs that may hypothetically remain on site following survey completion and the methods for calculating dose and risk from the hypothetical DRPs. There are no physical changes proposed at ZNPS. Therefore, the proposed changes do not involve a significant reduction in a margin of safety.
4.4      Conclusions In conclusion, based on the considerations above: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. Therefore, it is concluded that the requested amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of"no significant hazards consideration" is justified.
5        ENVIRONMENT AL CONSIDERATION The details of the proposed changes are provided in Section 2 of this License Amendment Request. A review has determined the proposed changes require an amendment to the licensing basis. However, ZionSolutions has determined the requested amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c )(9), in that:
(i)   The proposed changes involve no significant hazards consideration.
(ii)  There is no significant change in the types or significant increase in the amounts of any effluents that may be released off site.
(iii) There is no significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed changes meet the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 50.92(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment 6      REFERENCES
: 1)     10 CFR 50.82, "Termination of License"
: 2)    John B. Hickman (U.S. NRC) letter to John A. Christian, (ZionSolutions),
               "Issuance of Conforming Amendments Relating to the Transfer of Licenses for Zion Nuclear Power Station, Units 1 and 2", dated September 1, 2010

ZS-2022-0001 Enclosure Page 13 of 13
: 3)    John B. Hickman (U.S. NRC) letter to John Sauger, (Jl,nergySolutions), "Zion Nuclear Power Station, Units 1 and 2-Issuance of Amendments 191 and 178 for the Licenses to Approve the License Termination Plan", dated September 28, 2018
: 4)    NUREG-1575, Rev. 2 Draft Report, "Multi-Agency Radiation Survey and Site Investigation Manual (MARSSJM)," Revision 1, U.S. Nuclear Regulatory Commission, August 2000.
: 5)    "Risk Assessment Guidance for Superfund, Volume I: Human Health Evaluation Manual, Part E, Supplemental Guidance for Dermal Risk Assessment," U.S.
Environmental Protection Agency, EPA/540/R/99/005, Exhibit 3-5, July 2004.
: 6)    NUREG-1757, Vol. 2, Rev. 2, Draft Report, "Consolidated Decommissioning Guidance - Characterization, Survey, and Determination of Radiological Criteria,"
September 2020.

ZionSolutions, LLC ZS-2022-0001 Enclosure Attachment 1 Zion Nuclear Power Station, Units 1 and 2 License Termination Plan Markup Pages

ZION STATION RESTORATION PROJECT                                                                                           ~
LICENSE TERMlNA TION PLAN                                                                                        =ZIONSOLUTIONSuc=        M.._.....   ~
REVISION 4 TABLE OF CONTENTS
: 5. FINAL STATUS SURVEY PLAN ...... ......... ..... .................................................. ... .... .. ...... 5- l 5.1 . Radionuclides of Concern and Mixture Fractions ..... ... ... ........ ... .... .............. ........... ..... . 5-3 5.2. Release Criteria .................. .............. ............ ...... ............. .. ...... ...................................... 5-7 5 .2.1. Base Case Deri ved Concentrati on Guide] ine Level s for Basement Surfaces .. .. .... 5-7 5.2.2. Operationa l Derived Concentration Guide line Levels for Basement Surfaces ... .. 5-8 5.2.3. Base Case Derived Concentrati on Guideline Levels for Soil.. ................. .... ... ...... 5-9 5.2.4. Operationa l Derived Concentration Guideline Levels for Soil. ... .... .. ..... ...... .. .... . 5- 10 5.2.5. Base Case Derived Concentration Guideline Levels for Buried Piping ..... .. .. ..... 5-11 5.2.6. Operational Derived Concentration Guideline Levels for Buried Piping ..... ... .... 5- 12 5.2. 7. Base Case Derived Concentrati on Guideline Leve ls for Embedded Pipe .. ..... .. .. 5- 12 5.2. 8. Operational Derived Concentration Gu ideline Levels for Embedded Pipe ......... 5-13 5.2 .9. Base Case Derived Concentration Guideline Levels for Penetrations ........... .... .. 5-1 3 5.2.10 .        Operational Derived Concentration Guideline Levels for Penetrations .. ...... 5-14 5 .2. 11 .      Surrogate Radionucl ides ......................................... .... .. .... ... ...... ..... ... ....... ..... 5-1 5 5.2.12.         Sum-of-Fractions .. ..... ... .. .......... ............. ....... .... ....... ...... ..... ......... ... .... ........... 5-16 5.2.13 .        Dose from Groundwater ..... .... ... ..... .. ..... .. .... ... .... ..... ....... ... ... .... .... ............... .. 5- 16 5.2.14.         Demonstrating Compliance with Dose Criterion ..... .. .. .. ....... ...... ..... ... .... .... ... 5-17 5.2.15 .        Soil Area Factors ... ..... ... .. .... .. ...... ..... ... ............ ..... .. .. ... ... .. ....... ........ .. ......... .. .. 5-18 5 .3. Summary of Characterization Survey Results ...... ....... ....... ..... ..... ... .. .... ...... .... .. ......... 5-19 5.3 .1. Survey of Impacted Media ..... ....... ....... ..... .... ........... ......... .. ... .. ..... ........ ..... ...... ... . 5-19 5.3.2. Field Instrumentation and Sensitivities ... ... ....... .. .. ....... ...... ........ ...... .... ...... .... .. .. .. 5-20 5.3 .3. Laboratory Instrument Methods and Sensitivities ......... ...... ..... ... .... ....... ....... .... .. 5-20 5.3.4. Summary of Survey Results ..... .. ........ ...... .. ......... .......... .... .. .. .. ... ... .. ... .. ..... ..... .. .... 5-20
: 5. 3.4.1. Impacted and Non-Impacted Areas ...... ..... .. ....... ... ....... ..... .... ......... .. .... .... .. ... 5-20 5.3.4.2. Justification for Non-Impacted Areas .... .... ... .. .... .. .. ......... .... ... ....... ................ 5-2 1 5.3.4.3. Adequacy of the Characterization ..... ..... ...... ........... ............ ....... .. .. .... ... ...... ... 5-21 5.3.4.4. Inaccessible or Not Readily Access ible Areas .. .... ..... .......... ... .............. ......... 5-22 5.4. Decommission ing Support Surveys ...... ..... .. .... ............... ... .... ......... ... .... .......... .... ...... . 5-25 5.4.1 . Radiological Assessment (RA) ............ .. .. ... ..... .. .... ... ............. .. ... ................ ...... ... 5-25 5.4.2. Remedial Action Support (In-Process) Surveys .. ... .. .... ........ ... .. .... ..... ....... .. ... ..... 5-25 5.4.3. Instrumentation for RA and RASS .... ...... ....... ......... ....... ..... ...... ..... .. ... ...... ...... .... 5-25 5.4.4. Field Screening Methods for RA and RASS ....... ............. .. ... ................. ............. 5-26 5.4.5. Contamination Verification Surveys (CVS) of Basement Structural Surfaces ... 5-26 5.4.6. Post-Demolition Survey ...... .... ..... ... .. .. ...... .. ..... ........ ... ..... .... .. ... .. .... ...... .. ........ ..... 5-27 5.4.7. Survey Considerations for Suspected Discrete Radioactive Particle Areas ........ 5-27 5.5. Final Status Survey of Basement Structures .... ....... ... ...... ..... ... .. ... .. .. ..... .... ..... .. ..... .. ... 5-27 5 .5 .1. Instruments Selected for Performing FSS of Basement Surfaces ....... ...... ..... ... ... 5-27 5.5.2. Basement Surface FSS Units .... ... .. ............ .. ........ .. ......... .. ... .. ..... ..... .. .................. 5-28 5 .5 .2.1. Classification and Areal Coverage for FSS of Basement Surfaces ... ...... .... .. 5-29 5.5.2.2. Samp le Size Determination for FSS of Basement Surfaces ..... ...... ............... 5-30 5.5.3. Survey Approach for FSS of Basement Surfaces ....... .. ........... ..... .......... ......... ... . 5-32 5.5.4. Basement Surface FSS Data Assessment ... ... .... ..... ... .... ..... .. ..... ................... ....... 5-33 5.5.5 . FSS of Embedded Piping and Penetrations ...... ... .. ..... ............... .... .................... .. 5-34 5-i

ZION STATION RESTORATION PROJECT                                     ~
LICENSE TERMINATION PLAN                                         :ZJONSOLUTIONSuc=
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REVISION 4 LIST OF ACRONYMS AND ABBREVIATIONS AF         Area Factor ALARA      As Low As Reasonab ly Achievable AMCG       Average Member of the Critical Group BFM        Basement Fill Model CAQ        Conditions Adverse to Quality CCDD       Clean Concrete Demolition Debris C/LT       Characterization/License Termination CsI        Cesium Iodide CoC        Chain of Custody CVS        Contamination Verification Survey DCGL       Derived Concentration Guideline Levels DQA        Data Quality Assessment DQO        Data Quality Objectives DRP        Discrete Radioactive Particle EMC        E levated Measurement Comparison ETD        Easy to Detect FOY        Field of View FSS        Final Status Survey GPS        Global Positioning System HPGe       High-Purity Germanium HSA        Historical Site Assessment HTD        Hard to Detect IC         Insignificant Contributor ISFSI      Independent Spent Fuel Storage Installation ISOCS      In Situ Object Counting System LBGR       Lower Bound of the Gray Region LTP        License Termination Plan MARSSIM    Multi-Agency Radiation Survey and Site Investigation Manual MDC        Minimum Detectable Concentration MDCR       Minimum Detectable Count Rate NAD        North American Datum Nal        Sodium Iodide NIST       National Institute of Standards and Technology NRC        U .S. Nuclear Regulatory Commission ODCM       Off Site Dose Calculation Manual QA         Quality Assurance 5-v

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The CVS will include extens ive scan surveys on the structural surfaces (walls, floors and miscellaneous equipment) that wi ll be subject to open air demolition , regardless of elevation . The scan coverage is dependent on the contamination potential of the structural surface being surveyed. Class 1 survey units will require 100% scan coverage of all exposed concrete surface areas. Any areas identified in excess of the open air demolition limits will be earmarked for remediation.
For structural surfaces below the 588 foot elevation that will remain and be subject to a PSS (primari ly any basement floor and outer wall s), additional remediation will be performed to ensure that any individual ISOCS measurement will not exceed the Operational DCGLs from Table 5-4 during FSS . Any areas identified that have the potential to exceed the Operational DCGLs by ISOCS measurement during the performance of CVS in these areas will be remediated. Any areas of elevated activity that could potentially approach the Operational DCGLR wil l be identified as a location for a judgmental TSOCS measurement during PSS .
5.4.6. Post-Demolition Survey Following demolition, after al l debris is removed and the floors cleaned, an additional scan survey wi ll be performed to ensure that any individual ISOCS measurement will not exceed the Operational DCGLB from Table 5-4 during FSS. The survey will be performed using hand-held beta-gamma instrumentation as presented in Table 5-27 in typical scanning and measurement modes.
5.4.7. Survey Considerations for Suspected Discrete Radioactive Particle Areas Discrete radioactive particles (DRP) are defined as specks of radioactive material identified usually as either activated corrosion products such as cobalt-60, or an irradiated fuel fragment exhibiting greater than I 0,000 corrected counts per minute (100,000 dpm) . In open land (soil) survey units where the presence of DRPs is suspected, special survey techniques and actions must be considered. The survey techniques for the detection of DRPs are described in the Survey Plan for Discrete Radioactive Particle Identification and Remediation, ZS-LT-07 (the "DRP Survey Plan") (Reference 5-26), which is incorporated into this L TP by reference.
5.5. Final Status Survey of Basement Structures Basement structures are defined as basement surfaces (concrete and steel liner), embedded pipe, and penetrations. As described in section 5.4.5, all remaining floor and wall concrete surfaces will be remediated to levels below the Operational DCGLB as measured by ISOCS . After remediation, a FSS will be conducted to demonstrate that the residual radioactivity in building basements corresponds to a dose below the 25 mrem/yr criteria.
5.5.1. Instruments Selected for Performing FSS of Basement Surfaces The Canberra ISOCS has been selected as the primary instrument that will be used to perform FSS of basement surfaces. Direct beta measurements taken on the concrete surface wi ll not provide the data necessary to determine the residual radioactivity at depth in concrete and therefore, would have to be augmented with core sampling. The ISOCS was selected as the instrument of choice to perform PSS of basement surfaces for the following reasons:
* The surface area covered by a single ISOCS measurement is large (a nominal range of 10-30 m2) which essentially eliminates the need for scan surveys.
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Development" 5-20  Z ionSolulions TSD 14-022, Revisi on 1, "Use ofln-Situ Gamma Spectroscopy for Source Term Survey of End State Structures" 5-21  Z ionSolutions Technical Support Document l 0-002, Revision J, "Technical Basis for Radiological Limits for Structure/Building Open Air Demo lition" 5-22  ZionSolutions ZS-QA-10, Revision 9, "Quality Assurance Project Plan - Zion Station Restoration Project" 5-23  Internatio nal Standard TSO 7503-1 , Part I , "Evalu at ion of Su rface Contamination, Beta-Emitters (maximum beta energy greater than 0.15 Me V) and Alpha-Emitters" - August 1998 5-24  U.S. Nuclear Regulatory Commission Inspection Procedure No. 84750 "Radioactive Waste Treatment, and Effl uent and Environmental Monitoring " - March 1994 5-25  ZionSolutions ZS-AD-08 , "Corrective Action Program" 5-26  ZionSolutions ZS-LT-07, "Survey Plan for Discrete Radioactive Particle Identification and Remediation" 5-86

ZION STATION RESTORATION PROJECT                                                             ~
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REVISION 4 Table 6-50   Penetration Survey Unit Surface Areas .......................... ........... .... .................... .. 6-66 Tab le 6-51  Rati o of Instant Release Maxim um to Diffusion Re lease Maximum for Aux iliary Basement ........................................................ 6-67 Tab le 6-52  Adjusted Penetration DCGLrN (adjusted for insignificant contributor dose) ...... 6-68 Tab le 6-53  Dose Ass igned to C lean Concrete Fil l .................... .......... ........................ ........... 6-69 LIST OF FIGURES Figure  6-1  Z ion Nuclear Power Station Geographica l Location ........................................... 6-73 Figure  6-2  Z ion N uclear Power Station Owner Controlled Area .. ........ ................................ 6-74 Figure  6-3  Z ion Nuclear Power Station Security Restricted Area ........................................ 6-75 Figure  6-4  Backfilled Basement and Structures to Remain Below 588 ' E levation .... .... ...... 6-76 Figure  6-5  Cross Section A-A of Basements/Structures Below ............................................ 6-77 Figure  6-6  Cross Section B-B of Basements/Structures Be low 588' E levation to Remain at License Termination ... .. .............. ............................. .. ...... .................. 6-78 Figure  6-7  Cross Section C-C of Basements/Structures Below 588 ' Elevati on to Rema in at License Termination .......................................................................... 6-79 Figure  6-8  Cross Section D-D of Basements/Structures Below 588' ...................... ...... .... .. .. 6-80 Figure  6-9  Visualization of BFM Conceptual Model .... .. ........... ...... ..................................... 6-81 Figure  6-10 RES RAD Parameter Selection Flow Chart ........................................ .................. 6-82 ATTACHMENT 1           RESRAD Input Parameters for ZSRP BFM Sensitivity Analysis ....... 6-83 ATTACHMENT 2           RESRAD Input Parameters for ZSRP BFM ... ..................... ... .. ... 6-98 ATTACHMENT 3           RESRAD Input Parameters for ZSRP Surface Soil and Subsurface Soil Sensitivity Analysis .................. .... ................. 6-113 ATTACHMENT 4           RESRAD Input Parameters for ZSRP Surface So il and Subsurface Soi l DCGL. ...................................................... 6-124 ATTACHMENT 5           Less Likely But Plausible Scenario for Exposure to Hypothetical Discrete Radioactive Particles 6-iv

ZION STATION RESTORATION PROJECT                                      ~
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REVISIO 4 LIST OF ACRONYMS AND ABBREVIATIONS AF         Area Factor ALARA      As Low As (i s) Reasonab ly Achievable AMSL       Above Mean Sea Level ANL        Argo nne Nati onal Laboratory BFM        Basement Fill Model CRA        Conestoga Rovers & Associates DCGL       Derived Concentration Guidel ine Level DCF        Dose Conversion Factor DRP        Discrete Radioactive Particle DUST-MS    Disposal Unit Source Term - Multiple Spec ies EPA        Environmental Protection Agency FOR        Federal Guidance Report FOY        Field of View FSS        Final Status Survey GW         Groundwater HSA        Historical Site Assessment HTD        Hard-to-Detect JC         Insignificant Contributor ISFSI      Independent Spent Fuel Storage Installation
!SOCS      In-Situ Object Counting System LLBP       Less Likely but Plausible LTP        License Termination Plan MARSSIM    M ul ti-Agency Radiation Survey and Site Investigation Manual MDC        Minimal Detectable Concentration NRC        The U. S. Nuclear Regulatory Commiss ion ODCM       Off-site Dose Calcu lation Manual PRCC       Partial Rank Correlation Coefficient RASS       Remedial Action Support Surveys REMP       Radiological Environmental Monitoring Program RESRAD     RESidual RADioactive materia ls ROC        Radionuclides of Concern SFP        Spent Fuel Pool TEDE       Total Effective Dose Equivalent WWTF       Waste Water Treatment Faci lity ZNPS       Zion Nuclear Power Station ZSRP       Zion Station Restoration Project 6-v

ZION STA TION RESTORATION PROJECT                                               ~
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REVISION 4 A fourth alternate scenario considers the less likely but plausible {LLBP) scenario where a hypothetical discrete radioactive particle {DRP) is ingested, inhaled, or deposited on the skin.
Note that all DRPs identified to date have been remediated and no known DRPs remain. The DRP LLBP scenario assessment is contained in Attachment 5 6.5. Basement Fill Conceptual Model This section describes in detail the BFM conceptual model, including the source term, ROC, future land use and expos ure scenario, AMCG, and expos ure pathways. The BFM is used to calculate dose to the AMCG from residual radioactivity in the backfilled, below ground Basements to remain at the time of license termination. The list of Basements to remain is provided in Table 6-1. The computational model used to implement the conceptual model is described in section 6.6.
6.5.1. Source Term The source term for the BFM is the residual radioactivity, surface plus vo lumetric, remaining in each of the Basements at the time of license termination. The source term includes residua l radioactivity in wall and floor concrete, or steel liner in the case of the Containment Basements, as well as in embedded piping and penetrations that are contained in or interface with a given basement. Embedded pipe and penetrations are treated as separate survey units within the app licable basement that release activity into the basement fill in the same manner as activity from walls and fl oors (see sections 6-1 3 and 6-14) . The embedded pipe and penetration source terms are accounted for by adding the dose from the embedded pipe and penetration survey units to the dose from the app licable basement wall and floor survey unit. The total dose from all three sources within a given basement must be less than 25 mrem/yr. See section 6.17 .1 for discussion of the process for summing the dose from walls/floors, embedded pipe, and penetrations.
L TP Chapter 2 provides detailed characterization data regarding current contamination levels in the Basements. The data is based on concrete core samples obtained at biased locations with high contact dose rates and/or evidence of leaks/spill s. The expected so urce term configuration and radionuc lide distribution expected to remain in each Basement, after remediation is completed, is summarized below.
6.5.1.1.         Unit 1 and Unit 2 Containment Building Basements Both Unit I and Unit 2 Containment Buildings are comprised of concrete walls and floors with all interior surfaces of the containment ' shell' covered by a 0.25 inch steel liner. The liner on the 565 foot elevation floor is covered by a 30 inch thick layer of concrete. The floor of the Under-Vessel area is located at the 541 foot elevation. A 30 inch layer of concrete is present above the liner in the Under-Vessel area and a 15 inch layer of concrete is on the walls in the Under-Vesse l area. The steel liner on walls above the 568 foot elevation and below the 588 foot elevation has surficial contamination with removable contamination levels ranging from less than 1,000 dpm/100cm 2 to approximately 10,000 dpm/ l 00cm 2 as indicated by operational and routine radio logical surveys.
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ATTACHMENT 5 Less Likely But Plausible Scenario for Exposure to Hypothetical Discrete Radioactive Particles 6-135

ZION STA TJON RESTO RA TJON PROJECT LICENSE TERMINATION PLAN REVISION 4
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Less Likely But Plausible Scenario for Exposure to Hypothetical Discrete Radioactive Particles A.5.1 Introduction Discrete radioactive particles (DRP) were identified on the Zion site during decommissioning operations as well as during FSS. As discussed in Chapter 5 of this LTP, after the completion of FSS ZionSolutions prepared the Survey Plan for Discrete Radioactive Particle Identification and Remediation, ZS-LT-07, Revision 1 (DRP Survey Plan). The objective of the DRP Survey Plan is to identify and remediate all of the DRPs identified. Although there is no known DRP source term remaining on the site it cannot be said with absolute certainty that no DRP remain. This attachment provides an estimate of the number of DRPs that may hypothetically remain and the methods for calculating dose and risk from the hypothetical DRPs.
The results of the DRP survey are provided in Technical Support Document 22-001, Discrete Radioactive Particle Survey Report (DRP Survey Report). The dose and risk calculations will be performed using the methods described in this attachment. The activity assumed to be present in hypothetical DRPs and the results of the corresponding dose and risk calculations are provided in the DRP Survey Report. The projection of the number of DRP that could hypothetically remain is made in this attachment based on the number of DRP identified during the DRP survey (i.e.,
one).
AS.2     DRP Exposure Probability and Designation as Less Likely but Plausible Scenario DRP dose and risk is assessed as a LLBP scenario due to the low probability of hypothetical DRP exposure occurring. Treating the low-probability DRP exposure as an LLBP scenario is consistent with the approach used for assessing the low-probability scenario of the well driller contacting the Auxiliary Building drains, which was also designated as an LLBP scenario. In accordance with NUREG-1757, the evaluation of LLBP exposure scenarios ensures that "unacceptably high risks would not result", however, the evaluation of LLBP scenarios is not conducted to demonstrate compliance. Accordingly, the dose from the hypothetical DRP will not be added to the Zion compliance dose.
To justify the designation of the hypothetical DRP exposure pathways as an LLBP scenario, the probabilities of DRP ingestion and inhalation are compared to the probability of drilling into the Auxiliary Building drains, which was accepted by NRC as not likely. The probability of a drill contacting the Auxiliary Building drains is l.5x10-3/y as calculated by Equation A5-l. The lifetime probability of a future site resident ingesting or inhaling a DRP, assuming that one DRP remains, is much lower at l.6xl0- 8 and l.7xl0- 10 , respectively, as calculated using Equation A5-2.
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Equation AS.1 SAdrain Pdrain    =      A CZ where:
P drain = probability of drill contacting Auxiliary Building drain 2 1 SAdrain = projected surface area of Auxiliary Building drains (96 .2 m )
2 2 Ac= = area of contaminated zone (64,500 m )
Equation AS-2 where:
PpRP = lifetime probability of ingesting or inhaling a DRP if one DRP hypothetically remains
!Rs= soil mass ingestion rate {18.3 g/y) or inhalation rate 0.2 {g/y)
Acz = area of Zion contaminated zone for soil DCGL calculations (64,500 m2) le== thickness of soil layer affected by DRP {0.3048 m) 6    3    3 CFcm3lm3 = conversion factor Oxl0- cm /m )
3 ds = density of soil {1 .8 g/cm )
Te = lifetime exposure period of site resident (30 y)
Note: The mass inhalation rate of 0.2 g/y is derived by multiplying the RESRAD mass loading for inhalation value {2.35x10-5 g/m 3) by the breathing rate {8400 m3/y). The assumed 30-year lifetime occupancy period for the site resident is from SECY-97-046A, "Final Rule on Radiological Criteria for License Termination," March 31, 1997 {SECY-97-046A).
The number of unidentified DRPs that could hypothetically remain was calculated by conservatively assuming that all of the DRPs identified during the DRP scan survey were contained in a 1.0 cm layer of soil, regardless of the actual depth at which they were found.
Assuming the DRP are contained in a shallow soil layer provides a qualitative correlation to the depth assumptions in the a posteriori MDC values used in the dose calculations. The first 1 cm layer of soil throughout the entire site is then assumed to contain the number of DRP identified during the scan survey. The DRP{s) are assumed to be randomly distributed in the 1 cm soil layer. In addition, the hypothetical DRPs were assumed to be limited to the first 30.5 cm (I foot) layer of soil. Each successive 1 cm thick soil layer, below the first 1 cm layer, is assumed to have the same DRP density as the first 1 cm layer with the DRP being randomly distributed within each layer. Given these assumptions. the estimate of the number of DRPs that could hypothetically remain was calculated by multiplying the number of DRPs identified during the DRP scan survey by 30.5. One DRP was identified during the DRP scan survey (see DRP Survey 1
ZionSolutions, Zion Station Restoration Project, Final Status Survey Release Record, Revision I, Auxiliary Building 542 ft Embedded Floor Drain Pipe Survey Unit 05119A. 09/01 /2019 2
Zion site conceptual model contaminated area 6-137

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Report). Therefore, the estimate of the number of unidentified DRP that hypothetically remain is 31 (30.5 X 1).
The final lifetime probability of ingesting or inhaling a DRP must account for the projection of 31 DRPs hypothetically remaining. The adjustment is made by multiplying the single DRP probability discussed above, i.e., 1.6 xl0-8 and 1.7 x10- 10 for ingestion and inhalation, respectively, by 31 . The lifetime probabilities of a future resident on the Zion site ingesting or inhaling a hypothetical DRP is therefore projected to be 4.8x10-7 and 5.2x10- 9, respectively.
The probability of a DRP being deposited on the skin during the lifetime of a future site resident is l .2xl 0- 1 . The lifetime probability is calculated using Equation A5-3 assuming one DRP remains on the site. The single DRP lifetime probability is multiplied by the 31 DRP projected to hypothetically remain on the site to derive a final lifetime probability of 4xI0-6 . The dermal deposition parameters used to calculate "Mskin" in Equation A5-3 are from the U.S.
Environmental Protection Agency, "Risk Assessment Guidance for Superfund Volume I: Human Health Evaluation Manual (Part E, Supplemental Guidance for Dermal Risk Assessment" (Exhibit 3-5)
Equation AS-3 where:
PoRP = lifetime probability of skin exposure to assuming one DRP present on site Acz = area of Zion contaminated zone (64,500 m2) tcz = thickness of soil layer affected by DRP (0.3048m)
CFcm3/m3 = conversion factor (1E+06 cm 3/m 3) ds = density of soil (1.8 g/cm 3)
Te = time that resident occupies the site (30 y)
Mskin = mass loading of soil on skin (g)
         .:. .:w. .:.ah=e.; . ;re"'"':_ _ _Mskin = fskin mass loading eventEfevent frequencyAfsoitCFg/mgAskin fskin mass loading event = mass loading frequency ( 1 event/d)
Efevent frequency= frequency of event per year (350 d/y)
Afsoil = soil adherence factor for resident gardener (0 .07 mg/cm2)
CF g/mg = conversion factor (1 E-03 g/mg)
Askin= skin surface area (5,700 cm 2)
A.5.3 DRP Ingestion, Inhalation, and Skin Dose Calculations The methods and assumptions to be used in the dose analyses for DRPs are detailed below. Dose analysis methods are detailed for both activated metal and irradiated fuel DRPs. The derivations of the parameters and basis for the assumptions used are documented in TSO 22-001 "Discrete Radioactive Particle Survey Report" .
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The exposure pathways to be evaluated include:
: 1. Potential doses from the ingestion of large non-respirable particles.
: 2. Potential inhalation doses from the fractionation of large particles into l &#xb5;m AMAD respirable particles.
: 3. Complete dissolution of the particles and the resulting localized distributed soil concentrations. Area factors and DCGLEMCS for the DRP radionuclides of concern have been calculated. From these calculated concentrations, the potential ingestion and inhalation doses have been calculated.
: 4. Potential shallow dose and effective dose equivalents (EDE) for the exposure from particles deposited on the skin for a 24-hour exposure period.
Particle Activity Calculations As a part of the DRP survey, gamma scanning was performed using a towed array system. From these measurements a probabilistic model was developed to calculate an a posteriori Minimum Detectable Activity (MDA) for a particle on the ground surface. MDAs were calculated for both Co-60 and Cs-137. These calculations are also contained in TSD 22-01.
The DRP dose analyses have used the activity of a hypothetical particle at an activity equal to the 50 th percentile a posteriori MDA of the towed array. The 50th percentile MDA was chosen because it represents the median and most likely detectable particle activities that would be encountered and thus the most likely doses that would result from exposure to DRPs. This is consistent with the guidance in NUREG-1757 Appendix I, Section 1.3.3.3.7 which states that:
"Analyses of less likely but plausible scenarios are not meant to be 'worst-case' analyses and should not utilize a set of ' worst-case' parameters."
The Co-60 MDA is used to represent activated metal particle activities and the Cs-137 MDA is used to represent irradiated fuel particle activities.
The calculated 50 th percentile a posteriori MD As are:
* 0.12 &#xb5;Ci Co-60 and
* 0.41 &#xb5;Ci Cs-137.
Activated Metal Radionuclide Fractions Characterization of activated metal particle SO 124 was performed by ORISE and is documented in ORISE report 5271-SR-09-0. The nuclide activities reported in the characterization of particle S0124 have been scaled to the Co-60 activity. Scaling factors to Co-60 for the remainder of the activation radionuclides were derived from the activation activity calculations of the reactor internal components. These scaling factors have been applied to the Co-60 activities at the 50th percentile of the DRP scan MDA distribution.
The radionuclide scaling factors (i.e., activity ratios) and particle activities are included in Table AS-1.
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ZION STATION RESTORATION PROJECT LICENSE TERMINATION PLAN REVISION 4 Table AS-1      Activated Metal DRP 50 th Percentile Scan MDA Activity
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Activation                           Particle Activitv Activitv Activity                            Scaled to 50 th Nuclide                             Ratio to Calculation for                        Percentile MDA Co-60 Ul (Ci}                                   1l!Q}
H-3           2.53E+02           2.53E-03             3.04E+02 C-14           3.59E+02           3.59E-03             4.31E+02 Mn-54           2.85E+0l           2.85E-04             3.42E+0l Fe-55          7.15E+03           7.14E-02             8.57E+03 Co-60          l.00E+0S           l.00E+00             l.20E+0S Ni-59          l.66E+03           l .66E-02            l .99E+03 Ni-63          2.27E+0S           2.27E+00             2.72E+0S Nb-94           5.54E+00           5.53E-05             6.64E+o0 Tc-99          l.18E+00           l . l 7E-05          l .40E+00 Irradiated Fuel Particle Radionuclide Fractions Characterization of irradiated fuel particle SO 126 was performed by ORISE and is documented in ORISE report 5271-SR-09-0. The nuclide activities analyzed in the characterization of particle S0126 have been scaled to its Cs-137 activity. These scaling factors have then been applied to the Cs-137 activity at the 50 th percentile of the DRP scan MDA distribution.
The radionuclide scaling factors and particle activities are included in Table AS-2.
Table AS-2      Irradiated Fuel Particle at DRP 50 th Percentile Scan MDA Activity Particle Activity Activity Ratio           Scaled to 50 th Radionuclide to Cs-137             Percentile MDA (oCi)
Eu-155           8.47E-03                3.47E+03 Am-241           8.08E-0l                3.31E+0S Cm-244            l .S0E-01              6.ISE+04 Cs-137           l.00E+00                4.lOE+0S NQ-237           3.94E-05                1.62E+0l Pu-238           2.65E-0l                l .09E+0S Pu-239           7.62E-02                3. 12E+04 Sr-90           1.59E+00                6.52E+0S Particle Size Estimates Particle sizes have been estimated for both activated steel and irradiated fuel. The particle sizes estimated for both types of particles at the 50th percentile MD As are substantially larger than the respirable size limit of 10 &#xb5;m .
The minimum activated metal DRP diameter has been calculated by using the highest activated Co-60 specific activity from decommissioning activities, which is represented by the activation calculations of the reactor vessel internals. The Unit 1 baffle plates were used to estimate a 6-140

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minimum DRP volume for activities corresponding to the 50th percentile of the a posteriori Co-60 MDA. Using the Unit 1 baffle concentration of 4.06x10-2 Ci/cm 3 , the volume for 50 th percentile scan MDA activity (0 .12 &#xb5;Ci Co-60) DRP is 2.90xl0-6 cm3 . This volume corresponds to a particle diameter of 124 &#xb5;m.
The equation to calculate aerodynamic equivalent diameter (dae) equivalent to AMAD for a distribution of particle sizes weighted by activity, from the physical equivalent volume diameter, is taken from Equation 2-1 on page 2-2 in a report prepared for the NRC. 3 Therefore, the intact particle aerodynamic equivalent diameters (AED) for 8 g/cm 3 density steel sphere with a 1.5 shape factor for the 50th percentile is 287 &#xb5;m, respectively. Thus, activated metal DRPs detected at the Co-60 MDA far exceed the 10 um respirable particle threshold. Also, since the particle would not likely be truly spherical the median aerodynamic diameter could be substantially larger.
The irradiated fuel particle mass was estimated using the activity and the specific activity (Bq/kg) of Pu-239 in sample S0126. Therefore, the mass of Pu-239 in particle S0126 (1.22xl0-7 g) is equivalent to 2.43x 10-5 g of spent fuel as follows . Spent fuel contains 941 kg/tonne of U-238 or 0.941 grams of U-238 per gram of spent fuel. Based on this, the U-238 or uranium oxide mass is 2.29x10-5 gram.
Converting the U-238 mass to volume using a Uranium Dioxide density of 10.97 g/cm 3 results in a particle volume of 2.09xl0-6 cm 3 . This fuel mass corresponds to a physical spherical diameter of 155 &#xb5;m and an aerodynamic equivalent diameter, dae, of 420 um AMAD using a density of 10.97 g/cm 3 and a shape factor of 1.5, well above the size considered respirable.
When the S0126 radionuclide mix is scaled to the Cs-137 Scan a posteriori MDAs, the Pu-239 activity is 3.15xl 04 pCi. The corresponding physical parameters are : 5.08x10-7 grams of Pu-239, 9.02x10- 5 grams ofU-238, a U-238 volume of 8.20xl0-6 cm3 and a physical diameter of l.76xl0 2 fil!h Particle Solubility For each particle type, it has been assumed that each is best represented by the most insoluble form since these have been exposed to weathering and have likely become stable since the particle's creation.
There appears to be no available data within the literature for absorption of radionuclides within the GI tract for activated steel particles. but it is reasonable to expect that this absorption will be very low. There is some data available for irradiated fuel fragments. Therefore, to represent these insoluble states. we have selected forms with the lowest value of the parameter that represents absorption from the GI tract (alimentary tract) to the blood stream following a hypothetical ingestion event. In each case, the calculated internal dose follows the methodology from ICRP-30, the basis of the methods represented in 10 CFR 20, including the tissue weighting factors.
The ingestion internal doses are calculated using IMBA Professional Plus Version 4.1.11 using 3
Airborne Particle Resuspension and Inhalation Radiological Dose Estimation Following Volcanic Events.
prepared for U.S. Nuclear Regulatory Commission. Contract NRC--02--07--006. September 2011 .
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parameters from ICRP-26/30 or directly from Federal Guidance Report No. 11 (FGRl 1) as noted below.
Ingestion Dose Calculations from Activated Metal Particles The ingestion dose from an activated metal DRP has been based on the activity of a hypothetical particle at an activity equal to the towed array a posteriori 50th percentile Co-60 MDA.
The lowest fi value and its corresponding dose conversion factor from FGRl 1 have been used.
Irradiated Fuel Particle Ingestion Dose The ingestion dose from this particle has been calculated using IMBA. ICRP-137 provides a source of estimating absorption from the GI (i.e., alimentary) tract for this type of particle (irradiated fuel). It adopts a {A value of 5x 10-4 for all chemical forms. However, ICRP-137 does not provide direct guidance for ingestion exposure to irradiated fuel particles for the actinides. It does provide an in-depth discussion on Cs-137 in irradiated fuel particles. This states that for ingestion, of all forms of Cs, except irradiated fuel, the [A value (same as fi for ICRP-30) is 1.0 but for irradiated fuel, the value is 0.1 or a factor of 10 reduction.
ICRP-137 does not include [A values for actinides in irradiated fuel particles. However, it does include a value for Cs-137, a very soluble element. Using this factor of 10 reduction is applicable to the {A value of 5xl 0-4 for a final value of Sx10- 5 . Therefore, the value for fi for the actinides used in this internal dose analysis is Sxl0- 5
* For Sr-90 and Eu-155 the lowest fi value from FGRl 1 is used. A factor of 10 reduction in the lowest fi values from FGRl 1 is used for the other radionuclides identified in this particle.
This approach is considered appropriate and conservative for the hypothetical doses from this type of particle since an fi value of zero may actually apply to such an exposure. This value is consistent with the work reported in Environment Health Perspectives4 with a value of 3x 10-5 for the fractional absorption by ingestion of radionuclides within irradiated fuel fragments.
Additionally, this reference states, in regard to ingestion absorption of elements within a fuel fragment: " ... fission products in the fused particulate form renders them virtually inert in metabolic terms and the radionuclides are not metabolized along biological pathways characteristic for the elementary form. "
The  fi values are shown in Table AS-3 along with the source of the             fi values used .
4 Environmental Health Perspectives, Review, Volume 103, No. 10, October 1995, p. 920 - 934: "Biokinetics of Nuclear Fuel Compounds and Biological Effects of Nonuniform Radiation," Sakari Lang {Department of Environmental Sciences. University ofKuopio. Kuopio. Finland). Kristina Servomaa (Department of Research.
Finnish Centre for Radiation and Nuclear Safety, Helsinki, Finland), Veli-Matte Kosma (Department of Pathology.
University ofKuopio. Kuopio. Finland). and Tapio Rytomaa (Finn ish Centre for Radiation and Nuclear Safety).
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Irradiated Fuel Particle at Scan MDA Ingestion Dose Radionuclide       /.J.        Source of u Eu-155       l.00xI0-3        FGRll Am-241       5.00xI0-5  ICRP 137 modified Cm-244       5.00xI0-5  ICRP 137 modified Cs-137       l.00xI0- 1      ICRP 137 Np-237       5.00xI0-5  ICRP 137 modified Pu-238       5.00xI0-5  ICRP 137 modified Pu-239       5.00xIQ-5  ICRP 137 modified Sr-90       l.00xJ0-2         FGRII Inhalation Doses from 1 &#xb5;m AMAD DRPs The potential for a particle to be reduced in size over time has also been considered. The anticipated particles sizes accounting for changes in size over the 1,000-year compliance period have also been evaluated. Calculations have been performed to assess the inhalation dose from respirable 1 um AMAD activated metal and irradiated fuel particles.
The doses from inhalation of these particles created by fractionation of larger particles would be reduced in proportion to this decrease in activity. Thus, even if over time the size of activated metal particles is reduced to a respirable particle size, the dose from inhalation would be insignificant.
The inhalation dose from DRPs that have been reduced in size to 1 &#xb5;m AMAD over time has been calculated. The ICRP-30 lung model is based upon 1 &#xb5;m AMAD particles.
The spherical volume of a 1 &#xb5;m AMAD activated steel particle is 4.25xl0- 14 cm 3 . Thus, the activity of a 1 &#xb5;m AMAD particle derived from a particle at the 50 th percentile Co-60 MDA would be reduced by a factor of 8.63x10-9 . This would result in a Co-60 particle activity of l.727x10- 3 pCi for an equivalent volume diameter of 1 &#xb5;m AMAD.
The spherical physical volume of the 1 &#xb5;m AMAD irradiated fuel particle is 2.65x10-14 cm 3
* Thus, the activity of a 1 &#xb5;m AMAD particle derived from a particle at the 50th percentile Cs-137 MDA would be reduced by a factor of 1.90xl0-9 . This reduction in size would result in a Cs-137 activity of l.33x10- 3 pCi for a 1 &#xb5;m particle.
These calculations for the activity of 1 &#xb5;m AMAD irradiated fuel particle and activated steel particles have then been used to calculate inhalation doses using the FGRl I dose conversion factors.
Doses from Complete Dissolution of Particles Ingestion and inhalation doses have also been evaluated resulting from the complete dissolution of particles over time into distributed residual radioactivity. Both activated metal and irradiated fuel particles have been considered in this evaluation.
The particle activities in pCi were divided by the soil masses associated with areas of 0.0 l, 0.1, and 1 m2 and at soil depths of 15 cm to calculate the distributed contamination soil concentration in pCi/g.
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Surface Area Factors for all the activated metal and irradiated fuel nuclides were then used to calculate the dilution mass of the soil within these areas . A soil density of 1.8 g/ cm3 was used.
The Surface Area Factors and DCGLEMCS calculations from the LTP Chapter 6.11 were used if available. Area Factors for nuclides not listed in the L TP were calculated using the RESRAD computer model.
The analysis does not account for radioactive decay and is thus very conservative since complete dissolution would likely occur over many years, if at all.
Skin Doses from DRPs External skin doses have been calculated for exposure to both activated metal and irradiated fuel particles. The doses were evaluated using the 50 th percentile scan a posteriori MDAs.
The Varskin Version 6.2. l computer model has been used to calculate Shallow Dose Equivalent (SOE) from DRPs . For each of the type of particle, the equivalent volume diameter, de, was calculated. This diameter for a spherical particle was used in Varskin as the variable that accounts for the self-attenuation of beta particles within each DRP .
For activated metal DRPs, only the Co-60 skin dose was analyzed since this nuclide clearly dominates the activity profile. For irradiated fuel DRPs, only Sr-90 and Cs-137 have been included in the calculations since the alpha emitters would make an insignificant contribution to SDE. Varskin includes the doses from daughter nuclides. Therefore, the calculated dose includes the contribution from Y-90.
To allow comparison of SDE to the 25 mrem/year dose limit, the SDE is multiplied by a risk factor to calculate fatal cancer risk and compared to the risk corresponding to the 25 mrem/year TEDE criterion as provided in SECY-97-046A, which is 4xl0-4/y.
The fatal cancer risk factor applied to the SDE dose is from the 2002 Final Rule for Revision of the Skin Dose Limit5
* The risk factor from DRP SDE exposure is given as 6.6x10- 10/rem.
The calculated dose rates are converted to a total dose for a 24-hour exposure period.
To better evaluate the dose relative to the 25 mrern/year site release criteria for DRPs on the skin, the EPRI guidance for calculating a DRP skin dose has been used. 6 This approach, authorized by the NRC in RIS-2003-04. 7 has been used to calculate the Effective Dose Equivalents for Co-60 and Cs-137 at the 50 th percentile MDAs.
Using the Cs-137 and Co-60 EDE dose conversion factors, the EDE for a 24-hour exposure period from particles at the 50th percentile MDAs have been calculated.
5 Final Rule, Revision ofthe Skin Dose Limit, 67 FR 16298 , April 5, 2002.
6 Implementing the EPRI Effective Dose Equivalent (EDE) Methodology for Discrete Radioactive Particles on the Skin. Electric Power Research Institute, EPRJ I 002823. October 2004.
7 Use Ofthe Effective Dose Equivalent in Place of The Deep Dose Equivalent in Dose Assessments. U.S. Nuclear Regulatory Commission, Regulatory Issue Summary (RJS) 2003-04, February 13 , 2003.
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Risk and Dose from Hypothetical DRP Ingestion, Inhalation, and Skin Exposure As stated above, in accordance with NUREG-1757, the evaluation of LLBP scenarios such as the hypothetical DRP exposure ensures that "unacceptably high risks would not result" . The LLBP scenarios are not considered compliance scenarios. SECY-97-046A addresses the dose and risk from widespread residual radioactivity but does not contemplate the dose or risk from DRPs.
However, there are two sites where NRC has approved DRP exposure assessments in support of unrestricted use license termination, i.e., the Shelwell Site in Hebron Ohio and the Sacramento Municipal Utility District (SMUD) Ranch Seco Nuclear Generating Station (Rancho Seco). The methods approved by the NRC at these two sites will be applied to the Zion DRP assessment of ingestion and inhalation exposures.
In addition to the two NRC approved assessment approaches, ZionSolutions has calculated the lifetime fatal cancer risk from ingestion and inhalation of the hypothetical DRPs that are projected to remain in order to demonstrate that the fatal cancer risk from DRP exposure is less than the fatal cancer risk corresponding to the 25 mrem/y criterion as provided in SECY-97-04A.
The fatal cancer risk from the shallow dose equivalent (SDE) due to DRP skin exposure is also calculated as well as the effective dose equivalent (EDE) from skin exposure.
In summary, because there is no NRC guidance on the dose and/or risk assessment of DRPs, five separate assessments are performed to demonstrate that "unacceptably high risks would not result" from hypothetical exposure to DRP. The dose and risk assessments include the following:
: 1. expectation dose (rnrem/yr) from ingestion and inhalation using the Shelwell method
: 2. dose (rnrem/yr) from ingestion and inhalation using the Rancho Seco method
: 3. lifetime fatal cancer risk from ingestion and inhalation
: 4. EDE (mrem) from skin exposure.
: 5. lifetime fatal cancer risk from SDE due to skin exposure.
The criteria to be used to demonstrate that "unacceptably high risks would not result" are listed in Table A5-4.
Expectation Dose The DRP assessment method for the Shelwell site is described in SECY-98-117, "Shelwell Services, Inc., Risk Assessment", May 27, 1998. The method applied a risk informed approach that entailed calculating an "expectation dose" by multiplying the CEDE from DRP exposure by the annual probability of DRP exposure occurring. In a Staff Requirements Memorandum dated June 30, 1998, the NRC staff was directed by the Commission to apply the SECY-98-117 DRP assessment method to the review of Shelwell license termination request. The NRC terminated the license on July 14, 1999, using the SECY-98-117 risk informed approach as justification.
The expectation dose is calculated using Equation A5-4.
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Equation AS-4 where:
ED = expectation dose (mrem/yr) p, = lifetime probability of DRP ingestion or inhalation (4.8x10-7 and 5 .2x 10-9 for ingestion and inhalation, respectively, as calculated in section AS.2) to = resident lifetime site occupancy time (30 y)
Ddrp = dose from ingestion or inhalation ofDRP (mrem)
TEDE from DRP Activity Distributed in Soil The DRP assessment method at Rancho Seco was described in Attachment 2 of a {{letter dated|date=June 8, 2009|text=June 8, 2009 letter}} from Einar Ronnington, SMUD, to John Hickman, USNRC, "Phased Release of the Ranch Seco Site" . The method was approved in a {{letter dated|date=September 25, 2009|text=September 25, 2009, letter}} from Keith McConnell, USNRC, to Einar Ronnington, SMUD, which states that "the NRC staff has reviewed your proposed partial site release, as described in your {{letter dated|date=June 8, 2009|text=June 8, 2009, letter}}, and finds the proposed release to be acceptable". The Rancho Seco DRP assessment method assumed that the activity in the DRP is uniformly distributed within a volume of soil that is 1 m 2 in area and 0.15 m deep.
The dose from the radionuclide concentrations uniformly distributed in the soil volume is then calculated using soil DCGLs and the unity rule. The result represents a DRP dose that can be compared to the TEDE dose units that are used to express the unrestricted use criteria in 10 CFR 20 Subpart E. The Rancho Seco approach was applied but modified to be more comprehensive by also assessing soil areas less than 1 m2 and more realistic by using soil area factors in conjunction with soil DCGLs to properly account for the dose from small areas. See section AS .3 for details on the dose calculation using the Rancho Seco method.
Fatal Cancer Risk from DRP Ingestion and Inhalation The third method that will be used to assess the risk from DRP, in a manner that can be compared to a dose expressed in units of TEDE, is to convert DRP ingestion and inhalation dose to a corresponding fatal cancer risk. In support of the promulgation of the L TR, SECY-97-046A states that the lifetime fatal cancer risk corresponding to the 25 mrem/yr unrestricted use criterion is 4xl 0-4 and that this risk is estimated assuming a risk coefficient of Sxl 0-4 rem- 1 and a 30-year lifetime exposure. As shown in Section AS.I, exposure to a DRP is a low probability, once in a lifetime event, as compared to the assumed continuous exposure over a 30-year lifetime from uniformly distributed source terms such as soil. Therefore, the ingestion and inhalation doses (CEDE) from the hypothetical DRP will be multiplied by the risk coefficient applied in SECY-97-046A, i.e., Sxl0-4 rem- 1, and compared to the risk of 4xl0-4 that corresponds to the 25 rnrem/year unrestricted use criterion. If the lifetime fatal cancer risk from the LLBP DRP ingestion and inhalation exposure scenarios is less than 4xl0-4, then the risk is not considered unacceptably high.
Fatal Cancer Risk from SDE Due to Skin Exposure As described in section A .5.3, Varskin Version 6.2.1 will be used to calculate shallow dose equivalent (SDE) to the skin. However, the dose criterion in 10 CFR 20 Subpart E applies to 6-146

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TEDE, not SDE to the skin, and therefore direct comparison of SDE to the TEDE criterion is not appropriate. To allow comparison of SDE to a dose in units ofTEDE, the SDE was multiplied by a risk factor to calculate fatal cancer risk and compared to the fatal cancer risk corresponding to the 25 mrem/yr TEDE criterion as provided in SECY-97-046A, i.e., 4x10-4 yr- 1* The fatal cancer risk factor applied to the SDE dose is from the Final Rule, "Revision of the Skin Dose Limit", 67 FR 16298, April 5, 2002, which recommends a value of 6.6x10- 10 rem-1*
EDE from Skin Exposure In addition to fatal cancer risk from SDE, the EDE from DRP will be calculated using the method authorized in NRC Regulatory Issue Summary (RIS) 2003-04, "Use of the Effective Dose Equivalent in Place of The Deep Dose Equivalent in Dose Assessments". February 13.
2003 (see section A5 .3 for details).
Demonstration That "Unacceptably High Risks Would Not Result" from DRP Exposure The final dose and risk from DRP ingestion. inhalation. and skin exposure will be calculated and provided in the DRP Survey Report. The acceptance criteria provided in Table A5-4. column 4, will be used to demonstrate that "unacceptably high risks would not result" from the LLBP exposure to hypothetical DRPs. Note that the ingestion and inhalation fatal cancer risks. and the skin doses in Table A5-4 assume that the exposure occurs notwithstanding the very low lifetime probability of such an exposure actually occurring. See column 2 of Table A5-4 for the lifetime probabilities of DRP exposure. The acceptance criteria in Table A5-4 are not intended to be the absolute limits at which the LLBP DRP scenario is assessed but are intended as criteria that if met, do demonstrate that unacceptably high risks do not result. Higher values may also be acceptable on a case-by-case basis if justified in the DRP Survey Report.
Table AS-4        Acceptance Criteria for Demonstrating that "unacceptably high risks would no t resu It" f rom th e LLBP DRP E xoosure S cenano          .
DRP Exposure           Lifetime Probabilitv of Assessment Criteria         Acceptance Criteria Pathway           DRP Exposure Occurrini:
Ingestion and                    NA 1 Exgectation Dose              <25 mrem/yr Inhalation Ingestion and            Ingestion: 4.8xI0-7 Fatal Cancer Risk2               <4xl0 4 Inhalation             Inhalation: 5.2xI0-9 NA3                     TEDE from DRP Ingestion and Activitv Distributed in          <25 mrem/yr Inhalation Soil Skin                      3.7xl0-6                  Skin Dose (EDE)                 <25 mrem 3.7xI0-6               Fatal Cancer Risk from Skin                                                                                 <4x104 SOE
: 1)  The expectation dose calculation includes annual probability of exposure occurring
: 2)  The 4x 104 fatal cancer risk is the risk corresponding to the 25 mrem/yr unrestricted use criterion as stated in SECY-97-046A
: 3) The exposure to a small soil area of distributed activity is assumed to occur annually in the same manner as other small soil elevated areas 6-147

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* Attachment .2
. Zion N)iclear Power Station, U D,~ts 1 *and 2 License Terinjnation Plan Clean Pages
* ZION STATION RESTORATION PROJECT                                                                           -~
LICENSE TERMINATION PLAN                                                                                 ~ZIONSOLui:JoNsuc=   1#,8   I      Os;m REVISION 4 TABLE* OF CONTENTS
: 5.      FINAL STATUS SURVEY PL.AN .......... , ......................... .'.............. :................*................. 5-l 5.1. Radionuclides of Concern and Mixture Fractions .......................... ,............................. : 5-3 5.2. Release Criteria .... :....................... :........ :........ :............................ 7.............. : ........... : ........ 5-7 5.2.1. Base Case Derived Concentration Guideline Levels for Basement Surfaces ........ 5-7 **: _
      . 5.2.2. Operational Derived Conce&#xb5;tration *Guideline Levels for Basement Surfaces ..... 5-8 5 .2.3. Base Case Derived Coneentratiori Guidelin*e Levels for Soil ................................ 5-9 5 .2.4.. Operational Derived Concentration Guideline Levels for Soil............................. 5-10 5.2.5. Base Case Derived Concentration Guideline Levels for Buried Piping .............. 5-11 5.2.6. Operational Derived Concentration Guideline Levels for Buried Piping ............ 5-12
* 5 .2. 7. *Base Case Derived Concentration Guideline Levels for Embedded *Pipe ........... 5-12 5.2.8. . Operational Derived Concentration Guideline Levels for Embedded Pipe ......... 5-13 5.2.9. . Base Case Derived Concentration Guideline Levels for.Penetrations ..'..-............. 5-13 5 .2.10.        Operational Derived Concentration Guideline Levels for Penetrations ......... 5-14 5 .2.11.        Surrogate Radionuclides ..................... .'.. :........................................................ 5-15
* 5.2.12.         Sum-of-Fractions ................................... :............................... :................ :....... 5-16 5.2.13.         Dose from Groundwater:... :.: .....*............................ ,
                                                                                                               .       -................... 5-16 5.2.14.         Demonstrating Complian~ with Dose Griterion ....................,. ....... ,............. 7 5-17
    *.. 5.2.15.         Soil Area Factors ............................... :..... :...................................................... :. 5-18 5.3. Summary of Characterization Sur:vey Results .......................... :... .'...... ,...................... 5-19 5.3.1. Survey of ltnpacted Me~ia ................................................... ,...-....... :........... ,........ 5-19 5.3.2. Field ln~mentation and Sensitivities ....................... ;..... .-.: ....... i, ....... :................. 5-20
       *5.3.3. Laboratory Instrument Methods and Sensitivities; .............................................. 5-20
        ~ .3 .4. Summary of Survey Results ...... - ................. ;......... :..................... :.................... :.. 5-20 5.3.4.1. Impacted and Non-Impacted Areas ........................ :...... :.:.:.................... :..... : 5-20 5.3.4.2.* Justifi'?ation for Non-Impacted Areas .... :................................. :....................... 5-21 5.3.4.3. Adequacy of the Characterization .. :.*.............. ,..._. ... :....... .-.... :.. ::.'..............._...... 5.-21 5.3.4.4. Inaccessible.or Not Readily Accessible Ar~as ........,. .......................,. .............. 5-22 5.4.
* Decommissioning Support Surveys ............................................ :................... - .... :...... 5-25.
5.4J. Radiological Assessment (RA) .... :............ .'............ :....................................... :...... 5-25 5.42. Remedial Action Support. (In-Process) Surveys ........... ,...*........... :.................,. ..... 5-25 5.4.3. Instrumentation for RA and RASS ........ :.... :........ :........... .'....................... .'.. :........ 5-25 5.4.4. *Field Screening Methods for RA and RASS *..................... ,................................. 5-26 5.4.5. Contamination Verification Surveys (CVS) of Basement Structural Surfaces ... 5-26
        ;:!:~: -:  ~~~-~e;o 0
                               ~~11:ra~::=~~;*s~~~*;d*Di~~~;~*i~<li~~~~i~~:P~i~i~*k~~*::::::*:: ;:;;
5.5. Final Status Survey of Basement Structures .................. :.............................._. .............. 5-:-27 I*
5.5. l. Instruments Selected for Performing FSS of Basement Surfaces ........................ 5-27 5.5.2. Basement Surface FSS Units ............................... :....... :....................................... 5-2&deg;8 5.5.2.l_.
* Classification.and Areal Coverage for FSS of Basement Surfaces ............... 5-29 5.5.2.2. Sample Size Determination for FSS of Basement Surfaces .................... :..... 5-30 5.5.3. Survey Approach for FSS of Basement Surfaces ................................................ 5-32 5.5.4. Basement Surface FSS Data Assessment .........................................-................... 5-33 5:5.5. FSS of Embedded Piping and Penetrations .................................................... :... : 5-34 5-i
* ZION STATION RESTORATION PROJECT                                          ~-.
LICENSE TERMINATION PLAN REVISION 4
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                                                                            -   M~O....,
LIST OF ACRONYMS~ ABBREVIATIONS AF           Area Factor ALARA*       As..Low As Reasonably Achievable AMCG         Average Member of  . the Critical Group BFM          Basement Fill Model CAQ          Conditions* A4verse to Quality
   *econ::      Clean Concrete Demo'tition Debris C/LT         Characterization/License Termir\ation Csl          Cesium Iodide CoC         :chain of Custody CVS                            y Contamination eri.fic~tion *Survey DCGL         Derived Concentration Guideline Levels DQA          Data Quality Assessment DQO         .Data Quality' Objective~
. DRP                . Radioactive Discrete        ,
Particle ,
EMC          ~levated Measurement Comparison .
ETD          Easy to De~ec~
FOV          Field of View FSS          Final Status Surv(?y GPS          Global Positioning System
* HPGe         High-Purity Germanium BSA        . Historical Site Assessment*
HTD          Hard to.Dete~t
* IC           lnsignifiqant Contributor ISFSI        Indep~ndent Spent Fuel Storage Installation
   !SOCS        In Situ Object Counting System LBGR         Lower Bound of the Gray Region LTP          License _Termination Plan MARSSIM      Multi-Agency Radiation *Survey' and ~ite fuvestigation Manu~l MDC          Minimum Detectable Concentrati.o,n MDCR'        Minimum betectable Count Rate NAD          N9rih American'Datuni Nal          Sodium Iodide
  *NIST         National Institute of Standards and Technology NRC          U.S., Nuclear Regulatory Commission' ODCM        Off Site Dose Calculation *Manual QA          Quality Assurance
                                                  . 5-v

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A-tar* I C-,...,
REVISION 4 The CVS will include extensive scan surveys on the structural surfaces (walls, floors.and miscellaneous equipment) that will be subject to open air demolition, regardless of elevation. The s*can coverag*e is.
dependent on the contamination potential of the structural surface being surveyed. Class 1 survey units will require 100% s.can COV(?rage of all exposed concrete surface areas. Any.areas identified in excess of the open air demolition limits'will be earmarked for remediation .
. For structur~l surfaces below the'588 foot elevation that win remain and be subject to a*FsS (primariiy ariy 1:>asement floor: and outer wails), additional remediation will be performed to *ensure that any individual !SOCS measurement will not e~ceed the Operational DCGLB from Table 5-4 .during FSS .. AI1Y areas identified that have the potential to exceed the Operational DCGLB by ISOCS measurement during the performance of CVS in* these areas will be rem'ediated. Any areas o( elev~ted activ.ity that could .
potentially approach the Operational DCGLB will be identified as a location for a judgmental ISOCS measureme_nt during FSS.
* 5.4.6. Post-Demolition Survey Following d.emolition, after all debris is ren{oved and :the *flo.ors cleaned, an additional scan survey will be performed to ensure that any individual ISOCS measurement will not exceed the Operational DCGLB from Table 5-4 during FSS; The survey'will be performed using hand-held beta-gamma instrumentation as presented in Table 5-27 'in fyJ?ical scanning and me~uremerit modes.
5.4.7: Survey Consi~erations for Suspected Discrete Radioactive Particle Areas Discrete ra_dioactive particles (DRP) are defined as specks *of radioactive material identified usually as either activated corrosion products such as cobalt~0, or an irradiated fuel fragment exhibiting greater than 10,000 *corrected counts. per minute (100,000 dpm). In open land (soil) survey units where the pr:esence of DRPs is suspected,* special survey techniques and actions must be con*sidered." The survey
 -techniques for the detection of DRPs are described in the *survey Plan for Discrete* Radioactive Particle Ide_ntification arid _Remediation, ZS-LT-07 (the "bRP Survey Plani;) (Reference 5-26), which *is
* incorporated irtto this L TP by reference. *       *
* 5.5.      Final Status Survey   of Basement St~ctures Basement s~c&#xb5;ires are* defined as b~ement surfaces (concrete and steel liner), embedded pipe, an~
penetrations. As described in section 5.4.5, all remaining floor and wall conc,rete surfaces. will be remediated to levels below the Operational DCGLB as measured by. !SOCS. After remediatio'ri,. a *pss
  ':"ill be conducted to demonstrate that the residual radioactivity in building basements corresponds to-a dose below the 25. mrem/yr criteria.
* 5.5.1. Instruments Selected-for Perfon&#xb5;ing FSS o'fBasement Surfaces The Canberra ISOCS has beeO: selected as-the primary instnunent that will be.used to perform FSS of basement su~ces.
* Direct beta measurements taken on the concrete surface will not provtde the data necessary to* determine the residual radioactivity at depth in *concrete and therefore, would have to be augmented with *core s~mpling. The ISOCS was selected as the instrument of choice to perform FSS of basement surfaces (or the:following reasons:
* The surface area*covered by a single ISOCS measurement is large (a nominal.range'of 10-30 m2) which essentially eliminates* the need for scan suryeys.
27

ZION *sTAnoN RESTORAnoN PROJECT                                                      ~-*
                                                                                    ~,r, LICENSE TERMINATION PLAN REVISION 4                                                                    ;ZIONSoLlJTIONSuz-:-.
                                                                                  .    , _,.._,. s;r   Clialllot Development" 5-20     ZionSolutions TSD *14-022, Revision l; "Use* of In-Situ Gamm~ Spectrosc~py for Sourcetenn
: Survey of End State Structures" * *              *            *    *
* 5-21     Zio~olutions Technical S~pport Document -1 Q-002, Revision 1, "Technicai Basis f~*r Radiological Limits for Structure/Building Open Air pemolition" 5-22 . *zionSolutions ZS-QA~l 0, Revision 9, "Quality As.surance Project Plan*- Zion Station Restoration Project"                                                  **
5-2:3   Tnternational S~ndard ISO 7503-1, Part 1, ''Evaluation of Surface Contamination, -Beta-Emitters (maximum beta energy greater than 0.15 MeV) and Alpha-Erriitters''-August 1998
* 5-24    U.S. 'Nuclear Regulatory Commission Inspection Proced&#xb5;re No. 84750 ''Radioactive Waste
* Treatment, and *Effluent and. Enviromn*ental Monit9ring" - March 1994
.5-25    ZionSolutions ZS-AD~08, ~Co.rrective Action Program" 5-26    ZionSolutions ZS-LT-07, "Survey Pian*for Discrete Radioactive Particle Identification and Remediation"                             *                                            *
* 5-86

ZION STATION RESTORATION PROJECT
    .LICENSE TERMINATION PLAN REVISION 4
.. Table 6-50      Penetration Survey* Uni~ Surface Areas: ............................... :....... :....................... 6-66 Table 6-51    Ratio of Instant Release Maximum to Diffusion .
* Release Maximum for Auxiliary Bas~me~t .... .'.... ;..................... :.: .......................6-67 Table 6-52     Adjusted Penetration DCGLPN (adjusted for insignificant contributor dose).: .... ~68
* Table 6-53    Dose As-signed to Clean Concrete Fill ........*.................. :...... :........... :................... 6-6~
LIST OF FIGURES Figure 6-1    Zion Nuclear Pow.er Station Geographical Location .............. :............................ 6-73 Figure 6-2    Zion. Nuclear:
                              . Power. Station Owner Controlled Area .................................... :..... 6-74
* Figure 6-3    Zion Nuclear Power Station Security Restricted Area ........... ::...................... :.... 6-75 Figure 6-4    Backfilled Basement and* Structures to Remain Below 588' Elevation .............. 6-76 Figure 6-5    Cros.s. Section A-A ~f B~ements/~tructures Below.'........................................... ~- 7.7 Figure.6-6    Cross Section B-B of Basements/Structures Below 588' Elevation to
                  *Remain at.License Termination ... :............. :....................................... :.............. :.6-78.
Figure.6-7 Cross Section C-C of Basements/Structures Below 588' Elevation to
* Remain at License Termination ... :............. :: ................. :...... .'... :..... :........ :........... 6-79
   . Figure 6-8 . Cross Section D.-D ofBasenie~ts/Stru9tures Below 588' .................... :.. :.......... :, 6-80 Figure 6-.9 Visu3:lizati9n of BFM Conceptual Mqdel.. .. *: ........................................ ......... :.... 6-81 Figure 6-10 ' RESRAD
                       .        Parameter Selection Flow Chart .................................
                                                                              .                           .'' ............... :....... 6-82 ATTACHMENT l** RE SUD Input Parameters for ZSRP BFM Sensitivity Analysis ..... :.6-83 A1TACHMENT 2 RESRAD lnput P~eters for ZSRP BFM'. ......... : ..................... 6,-98 ATIACHMENT3 RESRAD lnput Parameters for.ZSRP Swface Soil and
* Subsurface Soil Sensit.ivity Analysis ... *... .'.............. *... *............... 6-i 13 .
: A TTACHfyffiNT 4 RESRAD Iriput Parameters for ZSRP Surface Soil and Subsurface Soil DCGL ............. .'., ....................................... 6-124 ATTACHMENT 5 Less Likely But Plf!;usible Scenario for E:x~sure to Hypoth_etical Discrete
                         *
* Radioactive Particles*
6-iv

ZION STATION RESTORATION PROJECT LICENSE TERMINATION PLAN .
REVISION 4 LI.ST OF ACRONYMS AND ABBREVIATIONS .
AF            Area Factor.
ALAR.A:       As Low As (is) Reasonably Achi~vable AMSL          Above Mean Sea Level .
:ANL          Argonne National Laboratory BFM           Basement Fill Model
* CRA           Conystoga Rovers & Associates DCGL.         Derived Concentration Guideline I;evel DCF           Dose Conversion Factor DRP           Discrete Radioactive Particle
* DUST-MS        Disposal Unit:Sou:rce Term - Multiple Species
   . EPA*             Environmental Protection ,Agency FGR           Federal G~idance Report
     .FOV          . Field of View FSS           Final Status Survey GW            Groundwater HSA*        'Historical Site Assessment HTD           Hard-to-Detect .
IC      * : Insignificant Contributor
* ISFSI        Indepen~ent Spent Fuel Storage Installation
        !SOCS         In-Situ Object Counting System LLBP         Less Likely .but Plausible LTP          License Termination Plan MARSSIM      Multi-Agency Radiation Survey and Site Investigation Manual MDC          Minimal Detectable Concen,tration NRC           The U.S. Nuclear:Regulatory Commission ODCM          Off-site Dose Calculation Manual PRCC          Partial Rank.Correlation Coefficient
. . . RASS            ,Remedial ~ction ~upport Survey~
REMP.         Radiologic~ Environmental Monitoring Program
      . RESRAD.   *
* RESidual RADioactive materials**
       *ROC
* Radionuclides of Concei:n SFP           Spent Fuel Pool TEDE          Total Effective Pose Equivalent WWTF          Waste Water Treatment Facility ZNPS          Zion Nuclear Power.Station.
ZSRP          Zion Station Restoration Project 6-v

                                                                                 .. ~ :
   ,ZION STATIQN RESTORATION PROJECT LICENSE TERMINATION PLAN                                                    :ZJONSOLUTJONS~
M~O:.,-,{
REVISION 4 A fourth. alternate scenario considers the less likely but plausible (LLBP) scenario where a*.
hypothetical discrete radioactive particle (DRP) is ingested, inhaled, or deposited on .the skin ..
* Note that all DRPs identified to date have been remediated and no known DRPs remain. The DRP LLBP scenario assessment is contain~d in Attachmen( 5 6.5. Basement Fill Conceptual Model This section *describes in detail the BFM ,conceptual model, including the *source term, ROC, futu~e lanq use and exposure scenario, AMCG; and* exposure pathways.* The BFM* is used to calculate do.se to the AMCG from residual radioactivity in the backfilled, below grqund B~sements to remain at the .fone of license termination. The list of Basements to remain is*
provided in Table 6-1. The computational mod'el
* used. to implement the conceptual' ~odel is described in section 6.6.
6.5.1.
* Source Term
* The source term for the BFM is the residual radioactivity, surface plus volumetric, retn~ining in each of the Basements at th~ time of license termination. The source term includes residual.
radioactivity .in.wall.and floor concrete, or steel* liner in the cas~ of the *Containment Basements; as well as in embedded piping and penetrations that are contained in or interface with a given basement. Embedded pipe and penetrations are treated as separate survey units within the*
* applicable basement that release activity &#xb5;1to tl).e basement fill in the same: manner as activity:
from walls and floors (see sections 6-13 and 6-14). The embedded pipe and penetration source*
terms are accounted for by adding the dose from the embedded pipe and penetration survey units.
to the dose from the applicable basement' wall and floor survey unit. The total dose from all three sources within a* given basement must be less than 2~ mrem/yr. See section *6.17.1 for
    ~iscussion .of the process for. summing the dose from walls/floors, embedded pipe,. and' penetrations.
L TP Chapter. 2 provides detailed* characterization data regarding current contamination levels in
* the Basements. ~e data is based on *concrete core samples obtained .at biased locations with high contact-dose rates and/or evidence 9f leaks/spills .. The expected* source. term configuration and* radionuclide distribution* expected' to remain in each.. Basement, after remediation is compfoted, is summarized below. .
6.5.1.1.        Unit 1 and Unit 2 Containme~t Building Basements Both Unit 1 and Unit 2 Containment J3uildings are comprised of concrete walls and. floors with
.. all interior surfaces of the containment 'shell' co:vered by a 0.25 inch steel liner. The liner on the 565 foot elevation floor *is covered by a 30 inch thick layer of concrete .. The floor of the Under-Vessel area is located at the 541 foot elevation. A 30 inch layer.of.concrete is present al;>ove the liner in the Under-Vessel area and a 15 inch layer of concrete is on the walls .in the. Under.,Vessel area. The steel liner on walls above the 568 foot elevation and below the 588 foot elevation has
* surficial contamina(ion .with removable contamination levels ranging from less than 1,000 dpm/100cm 2 to appro~iinately 10,00Q dpm/IOOcmtas indicated by operational and routine.*
radiological surveys. ..
* 6-9

ZION STATION, RESTORATION PROJECT LICENSE TERMINATION PLAN REVISION 4
                                *ATTACHMENTS Less'Likely But Plausible Scenario for Exposure to Hypothetical
                *
* Discr.ete Radioactive Particles     *
* 6-135

ZION STATION RESTORATION PRO,WCT                                            .,.~.:
LICENSE TERMINATION PLAN
  .REVISION 4                                                                   =ZIONSOWTioNs~
                                                                                            .." :wi**  *ir-r-r.
tess* Likely But Plausibl~ Scenario' for Exposur~ to* Hypothefical Discr:ete R!idioac~e hrt~es                                              .
A.5.1 Introduction*
Discrete radioactive particles (DRP) were id'entified on the Zion site* during decommissioning operations as well as during FSS . As ~iscussed in Chapter 5 of.this L TP, after the. completio~ of FSS'ZionSolutions prepared the Survey Plan for Discrete Radioactive Particle Identification and.
Remediation, ZS-L T-07, Revision 1 (DRP Survey Plan). The objective of the DRP Survey* Plan is to identify'.and remed{ate all ofthe DRPs identified. Although there is no known DRP 'sour~e
. term remaining on the site it cannot be said with absolute certainty that no DRP *remain. This*
attachmept provides an estimate of the number of :PRP.s that may hypothetically remain and the methods for calculating dose and risk from the hypothetical DRPs.                                 .
The results 'of the DRP survey are provided in Tec~cal Support Documerit 22-001, Discrete Radioactive Particle-Survey Report (DRP Survey Report). Th~ dose and risk calculations will be perform~ using tl).e methods described in this attachment. Tp.e activity as sum&#xa2; to be *presen~ &#xb5;i hypothetical .DRPs and the results of the corresponding:dose and risk*calculati.ons are provided in the DRP Survey Report. The projection of the number of DRP that could hypothetically reI11ain is made in this attachment based on the number of PRP identifi_ed during the DRP sui:vey (i.e.,
one).          *
* A5.2     DRP Exposure Probability and Designa~on as Less* Likely but Plausible Scenario as DRP dose and risk is assessed a LLBP scenariq_due_to ~e i~w prob~bil~ty of hypothetical D~ exposure occurring. Treating the*low-probability DRP exp<;>sure as an LLBP scenari<;:> is consistent with the approach used fo~ assessing the !_ow-probability scenario of the well driller contacting the Auxiliary Building drains, whicb was also designated as*an LLBP scenario. In :
accordance.with NUREG-1757, tb.e ev~luation ofLLBP exposure scenarios _ensures that *
    ''unacceptably high risks would not result", *however, the evaluation ofLLBP scen~os is not conducted to demonstrate compliance. Accordingly, the dose from the hypothetical DRP \.viii not be added to* the Zion compliance* dose.                                            .     .
To justify the designation of the hypothetical DRP exposure pathways as an _LLBP scenario, the
*
* probabilities of DRP ingestion and inhalation are compared to the probability of drilling into the Auxiliary Building drains, which was accepted by NRC as not'likely.* The probability of 1:1 drill contacting the Auxiliary Building drains is l.Sxl0-3/y as calculated by Equation AS-1. The*
lifetime* probability of a future site residet?-t ingesting or inhaling a DRP, assuming that one DRP remains, is much lower at 1.6x10-8 and 1.7x10- 10 , respe~tively, as_ calculated using Equation AS-.
2..
6-136

ZION STATION RESTORATION fROJECT                                                     .:    ~~.
                                                                                                 .~                   '
LICENSE TERMINATION PLAN
 .REVISION 4                                                                            ~ZJ.ONSowTJONSui=
Equation AS.1
                                                 .      .       SAdrain Pdratn     =    A CZ where:
Pdiam= probability of drill contacting Auxiliary Building drain SAmam =:projected surface area*of Auxiliary Building drains (96.f m 2) 1 Ac::.= area of contaminated zone (64,500 m.2)2                   .
: Equa~on AS-2 where:
PDRP     = lifetime probability of ingesting or inhaling a DRP if one DRP hypothetically remains
  *IJls = soil mass ingestion rate (18.3, g/y) or inhalation rate 0.2 (g/y)
AC%.= area of Zion contaminated zone for s~il DCGL calc4latio.p.s (64,500 .m2)"
tc:: = thickness of soil layer affecte~ by D~ (0.3048 m)                          .
3 CFcm3lm3 = conversion factor (lxl0-6 cm /in 3) d/l = density of soil (1.8 g/cm 3) .
 .Te= lifetime *expos*u:re period of site resident (30 y)
Note: The mass inhalation rate of 0.2 g/y is. derived. by multiplying the.RESRAD mass loading for inhalation value (2.35~10-5 *g/m3) by the breathing rate (8400 m 3/y). The
* assumed 30-year lifetime occupancy :period for the site resident is from SECY-97-046A,
              ''Final R&#xb5;le on Radiological Criteria for License Termination," March 31,' 1997 (SECY-97-046A).
The number of unidentified DRPs that c,ould hypothetically remain.was calculated by' co.nservatively assuming that all of the DRPs identified during the DRP scan survey were contained in a 1.0 cm layer of soil, regardless of the actual depth at w_hich they were found.
Assumip.g the DRP are contained in a shallow soil layer provides a qualitative correlation to the depth' assumptions in the a posteriori MDC values used iii the do*se calculations. The first 1 cm layer of soil throughout the entire site is then &-SSUmed to contain the number of DRP identified during the SC~ survey. The DRP(s) are 'assumed to be randomly .distributed in the 1 cm soil .
layer. In addition, the hypothetical DRPs were assumed to be limited to the first 30.5 cm (.1 foot) layer of soil. Each successive 1 cm thick soil layer, below the first 1 cm fayer, is assumed to have the same DRP densjty. as the first. I cm layer with.the DRP being ranqomly distributed within e~h layer. Gi~eh these assumptions*; the estimate of the number of DRPs that co.uld                          *
. hypothetically remain was calculated by multiplying the number of DRPs identified during the DRP scan survey: by 30_.5. One DRP: was identified during the DJ;U> scan survey (see DRP Survey.
1 ZionSolutions, Zion Station Restoration Project, Final Status Survey Release Record, Revision 1, Auxiliary Building 542 ft Embed<led* Floor Drain Pipe Survey"Unit 05119A, 09/01/2019                 .
2 Zion site conceptual model contaminated  area        .
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ZION STATION RESTORATION PROJECT LICENSE TERMINATION PLAN REVISION 4.
.Report). Therefore, the estimate of the number o.f unidentified bRP that hypothetically'remain is 31 (30.5x 1).
The final lifetime probability of ingesting or inhaling a. DRP must account for the projection of 31 DRP~ rypothetically rerri~ining. The adj~s.trrient is made by'multiplying the single DRP probability discussed above, i.e., 1.6 xl0- 8 and 1.7 x10- 10 for ingestion and inhalation,*
respectively, by 31. The lifetime pro6abilities of a future re;iqent op the Zion site ingesting or inhaling a hypothetical DRP is *therefore projected to be 4.8x10-7 and 5.2:xI0- 9, respectively.
Th~ probability of a DRP being deposited on the skin during. th_e lifetime of a. future site' resi~ent is l.2xl0-7 . The lifetime probability is calculated using Equation A5-3 assuming one DRP remains on the site. The single DRP lifetime probability is multiplied by the 31 DRP projected to hypothetic<!,lly remain on the site to derive a final ljfetime probability of 4x10-6. The dermal deposition parameters used to calculat;e "Mskm" in Equation A5.-3 are from the U.S.
* Environmental'Protection Agency, "Risk Assessment Guidance fo~ Superfund Volume I: Human Health Ev'aluation Manual (Part E; Supplemental.Guidance.for Dermal Risk Assessment" *
(Exhibit 3-5)                 .
Equation AS-3 where:
PoRP := lifetime probability of skin exposure to' a~suming one            DRP present on site Acz = ~ea of Zion contaminated zone (64,500 m 2) tcz = thickness of soil layer affected by DRP (0.3048m) 3 CFcm3/m3 = conve~sion factor (1E+06 cm /m~}
3 ds = density of soil (1.8 g/cm )
T~ = time that resident occupies the site (30 y)
Msian = mass loading of soil on skin (g) *
         . where:              Msktrt .= fsktn ~ass loading eventE f ev~rtt frequency Afsotl CFg /mgAskin:
fs1an mass Ioadmg event= mass loading frequency (1 event/d) * .
Efev~t frequency = frequency. of eve~t per Yc:ar (3 50 d/y}
2 Afso11 = soil adherence factor for residen~ gardener (0.07 mg/cm )
CFgtmg = conversion factor (IE-03 g/mg)
Asian=:= skin s~rface area (5,700 cni)
A.5.3 DRP Ingestion, Inhalation~ and Skin Dose Calculations
.The metJ:iods and assumptions to *be used in the dose analyses f<;>r DRPs .are detailed below.' Dose analysis methods are detailed for both activated metal and irradiated fuel DRPs. The derivations of the parameters and basis for the assumptions used are documented in T~D 22-001 ''Discrete Radioactive Particle Survey Report''.                            *
* 6-138

ZION STATION RESTORATION PROJECT LICENSE TERMINATION PLAN
                                                                                --~7" REVISION 4 ..
                                                                               ;ZIONSoLUTJONS~ *-s = e c.....,.,. ,
Exposure* Pathways to be Evaluated The exposure J?athways to be evaluat~d include:
: 1. Potential doses from the *ingestion ~f large non-re~pirable particles.
: 2. Potential inhalation doses from the fractionation of large particles iAto 1 &#xb5;m AMAD
* respirable particles.                                    .           .
: 3. Complete dissolution of the particles and the resulting localized .distributeq soil*
concentrations. Area factors and DCGLEMcS for the DRP radionuclides of concern.have been calculated.* From these c~lculated: concentrations, the potential ingestion and                 *
           *inhalation doses have been calculated.
: 4. Potential shallow dose and effective dose equivalents (EDE) for the e~posure from
           'particles deposited on the skin for a 44-hour exposure period.
Particle Activity Calculations As a part or'the PRP survey, gamma ~canning V.:as performed using a towed array system. From these measurements a proba~ilistic model was develo~d to calculate an q posteriori Minimum .
Detectable Activity (MDA) for a p~icle on the.ground surface. MPAs were calculated for QOth
* .Co-60 and Cs-13 7. These calculations are also contained in TSD 22..:01.
The DRP dose analyses have used the activity of a-hypothetical particle at_an activity equal to the 50~ percentile a posteriori MDA of the towed array. The 50th percentile MDA was chosen because it represents the* median and most likely detectable particle actiyities that would be encountered and thus the most likely doses that would result from eX:po_sure to DRPs. This is consistent with the guidf&#xb5;lce inNUREG-1757 Appendix I, Section 1.3.3.3.7 which states that:
  "A.l;lalyses of less likely but plausible scenarios are not meant to be* 'worst-case' analyses and should no't utilize a set o_f 'worst-case' :Parain_eters."                  *
* The Co-60 MDA i~ used to represent activated metii.l particle activities and the Cs-137 MDA is used to represent irradiated fuel particle activities. *
* Th~ calculatea'SOth percentile.a-~ster.iori MDAs are:
       *
* 0.12 &#xb5;Ci Co-60 and
* 0.41 &#xb5;Ci Cs-137.
Activated Metal .Radionuclide 'Fradti~ns Characterization of activated metal particle S0124 wa_s:perfonned by ORISE and is docqmented in ORISE report 5271:-SR-09-0. The nuclide activities reported in the characterization of particle
  *s.0124 have been scaled to the Co-60 activity. Scaling factors to Co-6o' for tp.e remainder of the activation'radionuclides were*derived from the activation activity calculations o(the reactor
* th int~mal components. These scaling factors have ,been applied to the Co-60 activit~es *at the 5_0 p~rcentile of the DRP scan*MDA distribution.
The radionuclide s_caling factors (i.e., activity ratios) ~d particle activities are included in Table A5-l.                           . .
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ZION STATION RESTORATION PROJECT                                                                           ...
LICENSE TERMINATION PLAN REVISION 4
                                                                                                          *z    ~ '
                                                                                                          ;**IO.NS..'
VliUTIONS~
FA(:  llf, b C ; j Table ~-1            A~ated l\'Ietal DRP 50 th Percentile S~n MDA Activity
_. Activation                            .. '    -    Particle Activity .
                         .                                               Activity
* Scaled to.SO th ; .
                                         . *. Acttyity: .
                   . ..* Nuclide'.                                   .
* Ratio fo           .. Pe.rcen.tile :MD.A Calculation for* .. Co-6()':
                                      ..        U1 (C,i) :: .                 ,,*                   .*,(pCi) ..
H-3                2.53E+o2               . 2.53E-03                   3.04E+o2 C-14.               3:59E;i-02*               3.59E-03                  4.3 lE+-02.
Mn-54                 2.85E+0l                  2.85E:.04                 3.42E+ol Fe-55               7.15E+o3                  7.14E-02                  8.57E+03,.
c~o                 l.00E+oS                  l.00E+00                   1.20E+o5 Ni-59               l.66E+o3                  l.~E-02                   '1.99E+o3 Ni-63               2.27E+o5' ..             2.27E+00                   2.72E+o5 Nb-94               5.54E+o0                  5.53E-05                  6.64E+o0 Tc-99            :* 1:.18E+00
* l.l?E-05                   l.40E+o0 Irradiated Fuel Particle Radionuclide Fractions Characterization of irradiated fuel particle S0126 was performed by ORISE and is documented in ORISE report 5271-SR-0~-0. The nuclide act1vitie*s analyzed in the characterization of particle S0126 have been scaled to its Cs-137 activity.'These scaling factors have'then*been applied.to the Cs-137 activity at the 50th percentile of the                   PRP        scan MD.A distribution. :                     .
The radionuclide scaling factors and particle. activities are.in~luded in Table A5-2.
Table A,5-2          lrradiat~ I?uel Particl~ at DRP 50 th -Percentile Scan MDA A~tivity
                                                    .       .* ,.,                 *Particle* Adivity Activity Ratio                   'Scaled*to*5ot1a ** . '
                             .~dionuclide*    .. - *: to CS-:1j7
                                                                .'                  Percentile MDA .
                                                                                       *'    (pCi.)''.       '
Eu-155                8.47E-03                          3.47E+o3
                                . Am-241                8.08E-01                          3.31E+o5 Cm-244                 l.S0E--01                        6.15E+o4 Cs-137               l.00E+o0                           4.l0E+-05 Np-,237               3.94E-05                          l.62E=+0l Pu-238                2.65E-01                          l.09E+05 Pu-239                7.62E-02                          3.12E+o4 Sr-90               l.59E+oo                           6.52E+o5 Particle* Size Estimates Particle. sizes have* been estimated for both activated steel and *irra4iate'd fuel. The partfole sizes estimated for }?oth types of particles at the 50th percentile MDAs are substantially l~ger than the respirable size limit of 10 &#xb5;m. .
* The minimmn activated metal DRP diameter has been calculated by using the highest activated.
Co-60 specific activity from decommissioning activities, which is represented by the activation calculations o(the reactor vessel internals. _The Unit 1 baf.fle plates were *used to estimate a 6-140 .

ZION STATION RESTORATION PROJECT                                                                          -
                                                                                         .-... .~~
LICENSE TERMINATION PLAN REVISION 4                                                                          ?ZIONSOlJ!.l'IO!j&sect;'f;;:;.
ii         , '
to minimum DRP volume for activities *corresponding the 50th. percentil~ 0f the a posteriori Co-60 MDA. Using the.Unit 1 ball~ concentration of 4.06x10-2 Ci/cm3, the volume for 50 th
* percentile scan MDA activity (0.12 &#xb5;Ci Cp-60) DRP-is 2.90x10-6 cm3
* This volume corresponds
 . -to a p~icle diameter of 124 &#xb5;m.
* The equation to' calc~late aerodynamic ~uivalent diameter (dae) equivale~t to AMAD for a ':
distribution of particle sizes weighte4 by activity, from the physical equivalent volume diameter, is taken from.Equation 2-1 on page 2-2 ma report prepared for the NR~. 3                              .
                                '                '   I                            '       '
_Therefore, the iii.tact partic.le aerodynamic equivalent diameters (AED) for.8 g/cm3*dens.ity steel sphere with a 1.5 shape. factor for the: 50th percentile is 287 &#xb5;m, respectively. Thus, activated* . :
metal DRPs detected at the Co-60 MDA far*exceed the 10 &#xb5;m respirable particle threshold. Also, -
since .the particle would not liJcely be truly ~pherical the median aerodynamic diameter could be
* substantially larger. . *           *
* The irradiated fuel particle :mass was estimated using the activity and the specific activity (Bq/kg) of Pu-239 in.sample S,0126. Therefore, the mass of Pu-239 in particle S0I.26 (1.22xl0-7
.. g) is equivalent to 2.43x10- 5 *g of spent fuel as follows, Spent fuel contams 941 kg/tonne ofU-
* 238 or p.941 grams of U-238 per.gram of spent fuel. 'Based on this, the U-238 or uranium oxide*
mass is 2.29xl o-5 gram.                                                    *                                 *
:Converting the U-238 mass to volume using a Uranium Dioxide.density:of 10.97 g/cm3 results in a particle volume of2.Q9xl0-6 cm3
* This fuel mass corresponds to a physical spherical diameter of 155* &#xb5;m and ap aerodynamic equivalent diameter, dae, of 420 &#xb5;m AMAb using a density of 10.9i g/cm3 and a*shape_factor of 1.5, well above the size considered respirable.                             .
Wpen the S0126 radionuclide mix.is scaled.to.~l).e Cs-137 Scan a po,steriori MDAs, the Pti-239 activ~ty is 3.15xl04 pCi. The corresponding physical parameters are: 5.08x10-7 grams of Pu-239, 9.02xl0- 5_grams ofU-238, a U-238 v~lume of 8.20x10-6 cm3 anq a physical-diameter ofl.76xla2
  . &#xb5;m.
Particle Solubility For each particle type, it has been assumed that ea~h is be~t represented by the most insoluble form since these*have b.een exposed to weathering and have likely become stable since the particle '*s creation.
There appears to be no available data wit:Wn the literature for absorption of radionuclides within the GI tract for actiyated steel particles, but it is reasonable to expect that :fuis absorption will be very low. There.is.some data available for irradiated fuel fragments. Therefore, to represent these insoluble-states, we have selected forms with the.lo_west value.of the parameter.that represents*
absorption from th.e* GI tract (alimentary tract) to ~e blood stream following a hypoth.etici:i,l               .
ing.estion event. rri each case, the calculated internal dose foll~ws the methodology from ICRP-30, the basis of the methods represented in 10 CFR 20, including the tissue w:eighting factors.
The ingestion internal doses are calculated using IMBA Professional Plus Version 4.1.11 using 3
Airborne Particle Resuspension and Inhalation Radioiogical_Dose &timatton Following. Vo/came Events, prepared for U.S. N~clear Regulatory Commission, Contract NRC---02---07---006, September 2011.
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parameters from I~RP-26/30 or directly fro~ Fede~l Guid~ce Report No. 11 CfGRI 1) as noted below.
* Ingestion Dose Calculations from Activated Metal Particles The ingestion dose from an activated metal DRP has-been based on t~e activity. of a hypothetical .*
particle at an activity equal t9 the towed array a posteriori 50th percentile Co-60 MDA. .
The lowe~fi valu~ and its corresp0nding dose conversion:faotor from *FGRl 1 have been us~d.
* . Irradiated Fuel Particle Ingestion Dose_*
* The ingestion dose from this particle has been calculated using lJ'vffiA. ICRP-137 provides a source *of estimating absorption from the GI (i.e., alimentary) tract for this type of particle .
   -(irradiated fuel). It adopts a fa value of 5x10-4for all chemical forms. However, ICRP-137 does not*provide direct guid_an~ for ingestion exposure to _irradiated_ fuel particles f9r the actinides. It
* does provide an in-depth discussion on Cs-.137 in irradiated fuel particle_s: This states that for ingestion, of all forms of Cs, except irradiated fuel, the fA value (same as fi for ICRP-30) is 1.0 but for irradiate'd fuel, the value is 0.1 or. a factor of 10 reduction.
* ICRP-137 doe~ not include./A values for actinide; in &#xb5;-radiated fueJ:particl~s. However, it does include a.value for Cs*-137, a very soluble element. Using this factor of 10 reduction is applicable to the/A value of Sxl()-4,for 'a final value of 5x10*5 _ Ther~fore, the.value forjj for the actinides used in this internal dose analysi~ is 5x 10*5 : For Sr-90 and Eu-15 5 t:he: lowest fi value from FGRl 1 is used. A factor of 10 reduction in the lowestfi values from FGR.11 is used for the other radionuolides identified in this:particle ...
This approach is considered appropriate and conservat~ve for the hypothetical doses from this type of part~cle since anfi value of zero. may actually apply to such an exposure. This value is
* consistent with the work reported in Environment Health Perspectives4 ~ith a 'value, of 3xl 0-5 for the fractional absorption by ingestion of i;aqionuclides within irradiated fuel fragments.
Additionally, this refe.rence state~, in regard to ingestion absorption of elemei:its within i;t fuel
  . fragment: " ... fission products in the fused particulate form renders _them virtually inert in
   . metabolic. temis and 'the radionuclides are not metabolized along biological pathways characteristic for the elementary form."
Theji val~~s.are shown      in Table AS-3. along with the so~ce ofthefi values used.
4
    *   &vzronmental flea/th Perspecttv_es, Review, Volume 103, No. 10, October 1995, p. 920 - 934: "Biokinetics of Nuclear Fuel Compounds and Biological Effects of Nonuniform Radiationt Sakari Larig (Department of
* Environmental Sciences, University ofKuopio, Kuopio, Finland), Kpstina Servomaa (Department of Research,
* . Finnish Centre for Radiati<;m and Nuclear Safety, Helsinki, Finland), Veli-M.atte Kosma (Department of'Pathology, University ofKuopio*,.Kuopio, Finland), and Tapio.Rytomaa (Finnish Centre for Radiation and Nuclear Safety).
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Table AS-3       Irradiated Fuel Particle at Scan MDA.lnges,tion o*ose Ra~omiclide    : . *Ji'.   .* .*.: . So&#xb5;rce offi' Eu-155       l.00x10* 3
* FGRll Am-241'      5.00x10* 5      ICRP 137 modified Cm-244       5.00xl.o-5 ;ICRP 137 modified Cs-137       1.oox10-1             JCRP.137 .
Np-237       5.00x10*5       icRP 137 modified Pu-238       5.00xl'0: 5 ICRP 137 modified Pu-239       s:oox10* 5 icRP 13.7 modified Sr~0        1.oox1a:2               FGRll Inhalation Doses from* I &#xb5;m AMAD DRPs The potential foi- a particle to be reduced in size over time.has also beeil'considered. The anticipated particles sizes accounting for changes in size over the 1,000-year coinpl_iance period .
have also be(?n evaluated. Calculations have been performed to assess the inhalation dose from respiral;>le 1 um AMAD activated nietal and irradiated fuel par:ticles,                         *
* The doses froni inhalation *of thes_e particles created *by fractionation of larger particles would: be reduced in proportion ~o this decrease. in activity. Thus, even if over time the size of activated metal particles .is redu~d to a.respirable particle size, the dose from inhalation would be insignificant.
* The inhalation dose.from DRPs that have been reduced in size to 1 &#xb5;m.AMAD over *time has been calculated. The ICRP-30 lung model is based upon 1 &#xb5;m AMAD*particles.
Th~ spherical volume of a 1 &#xb5;m AMAD activated ~teel particle is 4.25xl 0- 14 cm3
* Thus, the activity of a I &#xb5;m AMAD particle derived from a particle at the 50 th percentile ~o-60 MDA would be reduced by a factor of 8.63xl o*9 . This would result in a Co-60 particle activity of l.727x10~3 pCi for an equivalent volume diameter of 1 &#xb5;m AMAD.
* The spherical physi<;al volume of the 1 &#xb5;m AMAD ~diated fuel particle is 2.65xI0* 14 cm3*
Thus, the *activity of a 1 &#xb5;m AMAD part~cle derived from a particle at the 50th perce~tile Cs-137 MDA would be reduced by a factor of 1-.90xI0*9* Tbis reduction in. size would result in:a Cs-137 activity of l .33xl 0- 3 pCi for a 1 &#xb5;m particle.
These*calculations for the activity of 1 &#xb5;m A.MAD irradiated fuel particle         and activated steel*
.particles have then been used to calculate inhalation doses using the FG;Rl 1 dose conversion factors. *
*noses from Complete.Dissolution of Particles Ingestion and inhalation doses have also been evaluated resulting from the complete dissolution of parti.cles over time into distributed residual radioactivity. Both activated metal and irradiated
* fuel particles have been considered in this evaluation.                 *           *       **
The partioie.activitie~ in pCi were ,divided by the soi~ masses associated with areas of 0.01,*o.1, and 1 m 2 and*at soil depths of 15 cm to calculate the distributed contamination soil concentration .
fu~~                                                                                              .
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Surface Area Factors for all the activated m~tal and 'irradiated fuel nuclides were then used to calculate the dilution mass of the s~il within these areas. A soil density of 1.8 _g/ cm3 w.as used.
The Surface Area Factors ~nd D.CGLEMcS calculations from the L TP Chapter 6.11 were used if ayailable. Area Factors for nuclides not listed in 'the L TP were calculated using the RESRAD CO!l}puter model.                                           *                                                      *
* The analysis does not accm,mt for radioactive decay and is thus very coqservative since complete dissolution would likely o~ur o:ver many years, if at all.
Skin Doses from DRPs External skin doses have been calculated for exposure to hoth activated metal and irradiated fuel
  . particles .. The doses were evaluated using tlie 50th percentile scan a posteriori MD As.
The'varskin Version 6.2.1 computer model has-been used&deg;to calculate Sh_allowDose Equivalent (SDE) from DRPs. For each of.the type of p~cle, the equivalent volum~ diameter, de, was calculated. This diameter for a spherical particle was us*ed in V arsk:in as the variable that
* accounts for the self-attenuation of beta partie:les within each DRP.
For activated metal DRPs,         only  the Co;-60 skin dose was analyzed since this nuclide' ~!early dominates f:he activity profile. For irradiated fuel DRPs, only Sr-90 and Cs-137 have been included in the c;alculations since the alpha emitters would make*an insignificant contribution to SDE. Vars.kin includes the doses from daughter nuclides. Therefore 1 the calcul~ted dose includes the contribution from Y-90.                                                           *
* To allow comparison of SDE to the* 25 rhrem/year dose limit, the SDE is multiplied by a risk
* factor to: calculate fatal cancer risk and compared to the risk corresponding to the 25 mrem/year TEDE criterion a*s provided in ~ECY-97-046A,'which is 4x10-4/y.
The fatal c~cer risk factor. applied to the SDE dose is from the 2002 Final Rule for Revision of
 . the. Skin D.ose Lin&#xb5;t5.      The  risk factor from DRP SD~ e:xposure is gi_ven as 6.6x10- 10/rem. :
The calculated do~e rates are converted to~ totaf dose fo~ a 2.4-hour expo~re.p~ri<?d. *
* To better evaluate the dose relative to the 25 mrem/year site release criteria for DRPs o&#xb5; the skin, the EPRI guidance 'for calculating a* DRP skin dose has been used. 6 This approach, authorized by the NRC in RIS-2003-04, 7 has ~n used to calculate the Effective Dose Equivalents for Co-60 and Cs-137 at the.50 th percentile.MDAs.                             *
* U~ing the Cs-137 and Co-60 EDE dos*e conversion factors, 'the EDE for a 24~hour exposure period*from p~cles at the 50th percentiie MDAs have been calculated. . . .
* 5 Final Rule, Revision of the Skin pose Limit, 67 FR 16298, April 5, 2002.
* 6 Implementing the EPRI Effective Dose Equivalent (EDE) Methodology/or Discrete Radioactive Particles 'or,z      the
* *skin, Electric Power Research lrishtute, EPRI 1002823, dctober 2004. * *
* 7
* Use Of the Effectivff Dose Equiv.a/en/ in Place of The Deep Dose Equivalent in Dose Assessments, U.S. Nuclear Regulatory Commission,. Regulatory Issue Summary, (RIS) 2003.,04, February
                                                                             . 13, 2003.
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As stated above, in accordance with,NUREG-1757, the evaluation o_fLLBP scenarios .such 'as the hypothetical DRP exposure ens~res that "unacceptably high risks would not result''. The LLBP ,
scen,arios are not considered compliance scenarios. SECY~97-046A addresses the dose apd risk from widespread residual rfi:dioactivity but does* not contemplate the dose or risk from pRPs.'
However, there are two sites where NRC has approved DRP exposure assessments in support of unrestricteq use license termination, i.e., the Shelwell Site in Hebron Ohio and the Sacramento Municipal Utility District (SMUD) Ranch Seco Nucle~r Generating Station (Rancho Seco). The methods approved by the NRC at these two sites wi~l be applied to the Zion DRP'assessment of
* ingestion and inhalation e~posures.,              .       .       ,                             ,
In ~dditfon t6 the two NR<;: approv~d ass~ssment *approaches, ZionSolutions has calculated th~
lifetime fatal cancer risk from ingestion and inhalation of the hypothetical DRPs that are projected ~oTemain in order to demonstrate that the fatal cancer. risk from DRP e~posure is l,ess than the fatal cancer risk corresponding to the 25 miem/y c~terion as provided in SECY-97-04A.
The fatal cancer risk fr9m the shallow dose equivalent (SDE) due to DRP skin exp9sur~ is also calculated as well as the effective dose equivalent (EDE) from skin exposure ..
In summary, because there is-no NRC guidance on the-dose and/or risk assessment of DRPs, five separate assessments are performed to, demons~te that ''un~cceptably hfgh risks would not result" from hypothetical exposure to DRP: The dose and risk assessments include,the
, following:
       . 1. expectatio11 dose. (mrem/yr) from ingestion and inhalation using tb_e* Shelwell method
        . 2. dose (mrem/yr) from ingestion and inhalation using the Rancho -Seco method
: 3. ,lifetime fatal cancer risk from ingestion and inhalation
: 4. EDE (mre.m) from skin exposure.                ,.
: 5. lifetime fatal' cancer risk from SDE*due* to skin exposure .
. The criteria to be used to demonstrate that "unaccep:tably high risks would not result" are listed in Table A5-4.                              ,
Expectation Dose The DRP assessment method for the Shelwell .site is described in SECY-98-1 i 7, :'Shelwell Services, Inc., Risk Assessment", May 27, 1998. The method applied a,risk* informed approa~h that entaiJed calculating an ''expectation dose" by multiplying.the CEDE from DRP exposure by the_ annual *probability of DRP exposure occurring. In a Staff Requirements Memorandum dated June 30, 1998, the NRC staff was directed by the Commission to apply the SECY-98-117 DRP assessment method to the,rev1ew of Shelwell license termination request: Tue NRC terminated the license on July 14, 1999, using the SECY-98-117 risk informed approach as justification.
The expectation dose is calcul~ted using Equation A5-4.'
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                                                                                             , MC' lli,f I rt C.-V,,,*
                                                                                           *.Equation            AS-4 where: .*
* ED= expectation dose (mrem/yr)                               .       .
p1 = lifetime probability of DRJ;> ingesticm or inhalatio'n (4.8,xlo-7 and
                           *S.2xl 0-9 for ingestion and inhalation, respectively, as calculated in section A5.2)
                           'to= resident lif~time site o~upancy time (30 y)
Ddrp ~ dose. from . ingestion or inhalation ofDRP (mrem) '
TEDE from DRP Activity Distributed in Soil The DRP assessment method ~t :Rancho Sec;o was described in Attachment 2 of a June 8, 2009
 .letter from Einar Ronnington, SMUD, to John Hickman, USNRC; ''Phased Release of the Ranch Seco Site". Toe method was approved in a September 25, 2009, lett~r from Keith Mc~onnell,.
USNRC, to Einar Ronnington, SMUD;which states that "the l'qRC staffras reviewed you*r proposed partial site release, as described in your {{letter dated|date=June 8, 2009|text=June 8, 2009, letter}}, and finds the proposed*
release to be.acceptable". The Rancho Seco DRP assessment method assumed that the activity-in the _DRP is unifo~ly d~stributed within*~ volume of.soil that is 1 m 2 in. area *and 0.1_5 m deep.
  ~e dose from the radionuclide conceptrations uniformly distributed in the soil volume i_s then*
calculated using soil DCGLs and.the unity rul~: The result represents-a DRP. dose that can be compared to 'the .TEDE dose units that are.used to express the unrestricted _use criteria in 10 CFR 20 *Subpart E. The Rancho Seco approach was applied but modified to be more comprehensive by also assessing soil ;:treas less than 1 m 2 and more reafo;tic by using soil area factors in conjuncti<:-m with soil DCGLs to properly account for the dose from small areas. See section AS .3 for details .on the dose. calculation using the Rancho Seco meth9d.
Fatal Cancer Risk from. DRP Ingestion and Inhalation The third method that will be used to assess the risk from DRJ>, in a manner that can be compared to a dose express~d in units of *TEDE, i~ to convert DJU> ingestion and inhalation dose to a corresponding fatal cancer risk. In support of the promulgation of the -L TR, SECY-97-046A states that the lifetime fatal cancer risk corresponding to the 25 mrem/yr unrestricted use criterion is 4xl 0-4 and. that this risk is estimated assuming a risk coefficient of Sxl 0-4 rem.~ 1 and a
* 30-year lifetime exposure. As shown in Section A5.l, exposure to a DRP is a low probability, once in. a lifetime event, as compared to'the assumed continuous exposure over a 30-year lifetime from uniformly di~trib~d source terms s.uch as soil. Therefore1 the ingestion and inhalation
* doses (CEDE) fro*m the hypothetical DRP will be multiplied by the risk coefficient applied in SECY-97-046A, i:e., 5xl 0-4 rem- 1, and compared to the risk of 4xl 0-4 that corresponds to the 25 mrem/year unrestricted use criterion. If the lifetime fatal cancer risk from the LLBP DRP ingestion.and inhalation exposure scenarios is less than 4~10-4, then the risk is not considered unacceptably high. . *         *
* Fatal Cancer ~sk fr?m SDE Due to Skin Exposure As described in section A.5.3, Varskin Version     .
6.2.1
                                                           . will be. used to*calculate shaUow dose equivalent (SDE) to the skin. However, the dose criterion in 10 CFR 20 Subpart E applies to 6-146

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TEDE, not SDE to the skin, and therefore direct comparison of SDE to the TEDE crit~rion is not appropriate. To allow comparison of SDE to a dose in units of TEDE, the SDE was multiplied by .
arisk factor to calculate fatal cancer risk and compared to the fatal cancer risk corresponding to the 25 mrem/yr TEDE criterion as provided in SECY-97-046A, i.e., '4x10_-4 yr- 1. The fatal*cance_r
    .risk'factor applied to the SDE do.se is from the'J:"inal Rule, "Revision of the Skin Dose Limit;', 67
* FR 16298, April 5; 2002, which recommends a*value of 6.6x1Q- 10 rem- 1*
:EDE from Skin Exposure :
In addition to fatal cancer risk from SDE, the EDE. from DRP will be calculated using the method authorized in NRC Regulatory Issue Summary (RIS) 2003-04, '~ se of the Effective Dose Equivalent in P)ace of The Deep Dose Equivalent in Dose A.ssessments", February 13,
* 2003 (se*e section A5.3 for details).                         *                           *             *           *
* Demonstration That ''Unacceptably High Risks Would Not Result" from DRP Exposure The final do-~e and ris.15: from DRP 'ing~stion, inhalation, and skin exposure will be .calculated and provided in the DRP Survey Report The acceptance 'criteria provided iri Table A5-4) column. 4,
* wil.J be used to demonstrate that "unacceptably high risks would not r~sult'' from the LLBP *
 * *exposure to hypothetical DRPs.* Note that the ingestion and inhalation fatal cancer risks, and .the
. skin dose_s in Table _A5-4 assume that the exposure occurs notwithstanding the very* low lifetime
.. pro~_ability .of such an exposure ~ctua1ly occurring. S~e column 2 :of Table A5-4 for* the lifet~e probabilities ofDRP exposure. The acceptance criteria in Table.A5-4 are*not intended to be*the ..
absolute. limits *at which the LLBP DRP scenario is *assesseci but are intended as criteria that if .
met, do c;l~monstrate that unacceptably_ high risks ~o not ie.sult. _ffi.gher *values may also lie acceptable on~ case-by-case basis if justified in the DRP Survey Report ..
  . -Table A5-4 Acceptance Criteria for Demonstrating that "unacceptably high risks would no t resu It" f rom th.e LLBP DRP Exposure         .      S cenano
_DRP Exposure_.          Life.time Probability of
                                                                      -.i&#xb5;sessme.nt Criteria        Acceptince Criteria* :
P~thway
* nitj> E~pos~r~ Occ~rring                                                     ..          ..
* Ingestion and                       NA1                      .            .
Expectation Dose.              <25 .mrein/yr Inhalation Ingestion and             Ingestion: 4.8x10*7 Fatal Cancer Risk2               <4x1CJI Inhalation             Inhalation: 5.2x10-9 NA3                       TEDE from DRP Ingestion and Activity Distributed in          . <25 mremlyr Inhala&#xb5;on
                                                                                 'Soil Skin                         3.7xlo-6                   Skin D.os~ (EDE)         ' .    <25 mrem
          '.                                . 3.7xlo-6               Fatal Cartcer Rj.sk from Skin                                                                                     <4x10-4 SDE
: 1)    Toe expectation dose calculation includes annual probability of exposure occurring             .           .   .
: 2)    The 4x10__. fatal cancer risk: is the risk corresponding to the 25 mrem/yr unrestricted use criterion as stated in
* SECY-97-046A                                                                .
: 3) The exposure to a small soil area of distributed activity is assumed to occur annually in the saJ&#xa5;e manne~ as
          *other small soil elevated areas               .
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