ML15261A238

License Amendment Request 259, ISFSI-Only Emergency Plan and Emergency Action Level Scheme

ML15261A238
Person / Time
Site:	Kewaunee
Issue date:	09/14/2015
From:	Clark G Dominion, Dominion Energy Kewaunee
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
15-306
Download: ML15261A238 (118)

Text

Domnio.EnDgoKminieonn 5000 Dominion Boulevard, Glen Allen, VA 23060 Web Address: www.dom.com September 14, 2015 U. S. Nuclear Regulatory Commission Serial No.15-306 Attention: Document Control Desk LIC/JG/R0 Washington, DC 20555-0001 Docket Nos. 50-305, 72-64 License No. DPR-43 DOMINION ENERGY KEWAUNEE, INC.

KEWAUNEE POWER STATION LICENSE AMENDMENT REQUEST 259, ISFSI-ONLY EMERGENCY PLAN AND EMERGENCY ACTION LEVEL SCHEME Pursuant to 10 CFR 50.90, Dominion Energy Kewaunee, Inc. (DEK) requests an amendment to Facility Operating License Number DPR-43 for Kewaunee Power Station (KPS). The proposed amendment would revise the emergency plan and emergency action level (EAL) scheme. The proposed changes are being submitted to the NRC for approval prior to implementation, as required under 10 CFR 50.54(q)(4), 10 CFR 50, Appendix E, Section IV.B.2, and 10 CFR 72.44(f).

By letter dated May 14, 2013, DEK submitted a certification of permanent removal of fuel from the reactor vessel (Reference 1). Consequently, as specified in 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for KPS no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. The Post-Shutdown Decommissioning Activities Report (PSDAR) for KPS dated April 25, 2014 (Reference 2), documented that DEK expects to have all spent fuel transferred to the Independent Spent Fuel Storage Installation (ISFSI) by the end of 2016. To comport to the reduced scope of potential radiological accidents with spent fuel in dry cask storage within the ISFSI, DEK proposes a new EAL scheme and corresponding emergency plan changes.

The proposed emergency plan continues to rely on previously granted exemptions from certain emergency planning requirements (Reference 3) as the basis for these exemptions has not changed and remains in effect.

Attachment I to this letter contains a description, technical analysis, significant hazards determination, and environmental considerations evaluation for the proposed amendment. Enclosures 1, 2, and 3 to this letter provide the proposed emergency plan, the EAL bases document (including EAL scheme), and supporting calculations, respectively.

The KPS Facility Safety Review Committee has reviewed the proposed amendment and a copy of this submittal has been provided to the State of Wisconsin in accordance with 10 CFR 50.91(b).

Serial No.15-306 License Amendment Request 259 Page 2 of 3 DEK requests approval of this proposed amendment by August 31, 2016. Once approved, the amendment will be implemented within 60 days following DEK's submittal of a written certification to the NRC that the final spent nuclear fuel assembly has been transferred out of the spent fuel pool and placed in storage within the ISESI.

Please contact Mr. Jack Gadzala at 920-388-8604 if you have any questions or require additional information.

Sincerely, cLQ~k Gianna C. Clark Vice President - Nuclear Support Services ... Hul Vick1.iNt ' I 4: NOTARY PUBLIC jY *Commonwealth ofVi'rginia Expires.May 31, 20lS y.Comnmission COMMONWEALTH OF VIRGINIA )

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COUNTY OF HENRICO )

The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, Inc. Shetoday has by Gianna C. Clark, who is Vice President - Nuclear Support Services of Dominion Energy Kewaunee, affirmed before me that she is duly authorized to execute and file the foregoing document on behalf of that Company, and that the statements in the document are true to the b )*t of her knowledge and belief.

Acknowledged before me this /L7Lfday of o,4L Jd*,2O ei.,. 201 .//Z/ ., --

My Commission Expires: _________/*:*. 1/ A /.2-- 1 &

Notary Public Attachments:

1. Discussion of Change, Technical Analysis, Significant Hazards Determination and Environmental Considerations

Enclosures:

1. ISFSI-Only Emergency Plan

2. ISFSI-Only Emergency Action Level Bases Document

3. Supporting Evaluations and Calculations

Serial No.15-306 License Amendment Request 259 Page 3 of 3

References:

1. Letter from D. G. Stoddard (DEK) to NRC Document Control Desk, "Certification of Permanent Removal of Fuel from the Reactor Vessel", dated May 14, 2013

[ADAMS Accession No. ML13135A209]

2. Letter from D. G. Stoddard (DEK) to NRC Document Control Desk, "Revision to Post-Shutdown Decommissioning Activities Report", dated April 25, 2014. [ADAMS Accession No. ML14118A382]

3. Letter from Thomas J. Wengert (NRC) to David A. Heacock (DEK), "Kewaunee Power Station - Exemptions from Certain Emergency Planning Requirements and Related Safety Evaluation (TAC No. MF2567)", dated October 27, 2014. [ADAMS Accession No. ML14261A223]

Commitments made in this letter: None.

cc: Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road, Suite 210 Lisle, IL 60532-4352 Ted H. Carter, Senior Project Manager U.S. Nuclear Regulatory Commission Two White Flint North, Mail Stop T-8F5 11545 Rockville Pike Rockville, MD 20852-2738 Public Service Commission of Wisconsin Electric Division P.O. Box 7854 Madison, WI 53707

Serial No.15-306 ATTACHMENT 1 LICENSE AMENDMENT REQUEST 259:

ISFSI-ONLY EMERGENCY PLAN AND EMERGENCY ACTION LEVEL SCHEME DISCUSSION OF CHANGE, TECHNICAL ANALYSIS, SIGNIFICANT HAZARDS DETERMINATION, AND ENVIRONMENTAL CONSIDERATIONS KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No.15-306 Attachment 1 Page 1 of 17 ISFSI-ONLY EMERGENCY PLAN AND EMERGENCY ACTION LEVEL SCHEME DISCUSSION OF CHANGE, TECHNICAL ANALYSIS, SIGNIFICANT HAZARDS DETERMINATION AND ENVIRONMENTAL CONSIDERATIONS 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.90, Dominion Energy Kewaunee, Inc. (DEK) requests an amendment to Facility Operating License Number DPR-43 for Kewaunee Power Station (KPS). The proposed amendment would revise the emergency plan and emergency action level (EAL) scheme. The proposed changes are being submitted to the NRC for approval prior to implementation, as required under 10 CFR 50.54(q)(4), 10 CFR 50, Appendix E, Section IV.B.2, and 10 CFR 72.44(f).

By letter dated May 14, 2013, DEK submitted a certification of permanent removal of fuel from the reactor vessel (Reference 1). Consequently, as specified in 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for KPS no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. The Post-Shutdown Decommissioning Activities Report (PSDAR) for KPS dated April 25, 2014 (Reference 2), documented that DEK expects to have all spent fuel transferred to the Independent Spent Fuel Storage Installation (ISFSI) by the end of 2016. To comport to the reduced scope of potential radiological accidents with spent fuel in dry cask storage within the ISFSI, DEK proposes a new EAL scheme and corresponding emergency plan changes.

The proposed emergency plan is for the operation of the ISFSI. This plan would only be implemented after all spent fuel has been removed from the spent fuel pool and placed in dry storage within the ISFSI. Implementation of the emergency plan would involve DEK establishing administrative controls for radiological source term accumulation limits and methods to control the accidental dispersal of the radiological source.

The proposed emergency plan continues to rely on previously granted exemptions from certain emergency planning requirements (Reference 3) as the basis for these exemptions has not changed and remains in effect.

There are no existing license amendment requests associated with the emergency plan currently docketed for KPS. Therefore, no disposition of other license changes, as they relate to this license amendment request, is needed. Consistent with the condition that the proposed emergency plan may be implemented only after all spent fuel has been removed from the spent fuel pool and placed in dry storage within the ISFSI, KPS License Amendment Request 260 (Reference 11), proposed a revision to the facility operating license to comport to the ISFSI-only condition that includes a prohibition against storage of fuel in the spent fuel pool.

Serial No.15-306 Attachment 1 Page 2 of 17

2.0 PROPOSED CHANGE

The proposed amendment would modify the KPS license by revising the emergency plan and the associated emergency action level (EAL) scheme. The proposed changes reduce the scope of onsite emergency planning requirements to reflect the reduced scope of potential radiological accidents with all spent fuel in dry cask storage within the ISESI. After all spent fuel is in dry cask storage within the ISFSI, the number and severity of potential radiological accidents is significantly less than when spent nuclear fuel was stored in the spent fuel pool. Therefore, the offsite radiological consequences of accidents possible at KPS are substantially lower. There continues to be no need for offsite emergency response plans at KPS because no design basis accident or reasonably conceivable beyond design basis accident can result in radioactive releases that exceed Environmental Protection Agency (EPA) Protective Action Guides (PAGs) beyond the site boundary (Reference 4).

Potential off-site doses were calculated for KPS to verify necessary administrative radiological source term accumulation limits would be adequate during decontamination and dismantling of radioactive systems, structures, and components contained in the non-operational nuclear unit. These administrative radiological source term accumulation limits ensure that if a radiological release were to occur, it would not exceed two times the Offsite Dose Calculation Manual (ODCM) limits (2 x 1500 moremo/year) at the site boundary for 60 minutes (and therefore not result in doses to the public above EPA PAGs beyond the controlled area boundary). In addition to administrative limits on radioactive source term accumulation, administrative controls will be in place to limit the dispersal of radioactive material. These administrative limits and dispersal controls are in addition to the requirements already specified in the ODOM for control of effluent releases.

The current EAL scheme was approved for use at KPS on October 31, 2014 (Reference 5), and is based upon NEI 99-01, Revision 6 (Reference 6). The proposed EAL scheme remains based on NEI 99-01, Revision 6, as appropriate for the ISESI-only condition at KPS.

2.1 Summary of Major Changes The major changes to the KPS Emergency Plan are:

1. Removal of the various emergency actions related to the spent fuel pool.

2. Removal of non-ISFSI related emergency event types.

3. Replacement of the "Shift Manager" title for the "ISESI Shift Supervisor (ISS)" title as the position which assumes the Emergency Director's responsibilities.

4. Revision of the emergency response organization.

The off-normal events and accidents addressed in the KPS ISFSI-Only Emergency Plan are related to the dry storage of spent nuclear fuel at the ISFSI and include only off-

Serial No.15-306 Attachment 1 Page 3 of 17 normal, accident, natural phenomena, and hypothetical events and consequences as presented in the NUHOMS Certificate of Compliance (CoC) 1004 (Amendments 9 and

10) Horizontal Modular Storage System Updated Final Safety Analysis Report (NUHOMS UFSAR), and in the NAC CoC 1031 (Amendment 5) MAGNASTOR System Final Safety Analysis Report (NAC FSAR). After all fuel is removed from the KPS spent fuel pool, there will no longer be any potential for the accidents previously described in the KPS Emergency Plan that would increase risk to the health and safety of the public.

These accidents included events specifically related to the storage of the spent fuel in the spent fuel pool. After the transfer of the spent fuel to the ISESI, the spent fuel storage and handling systems will be removed from operation consistent with the PSDAR for KPS dated April 25, 2014 (Reference 2).

2.2 Elimination of Spent Fuel Pool Initiating Conditions and EALs The initiating conditions (l~s) and EALs associated with emergency classification in the current emergency plan are based on NEI 99-01, Revision 6. Specifically, Appendix C of NEl 99-01 contains a set of ICs and EALs for permanently defueled nuclear power plants that had previously operated under a 10 CER Part 50 license and have permanently ceased operations.

After all spent fuel has been removed from the spent fuel pool and placed in dry storage within the ISFSI, the NEl 99-01, Appendix C, ICs and EALs that are associated with the spent fuel pool are no longer required to be in the emergency plan. Additionally, certain ICs and EALs whose primary function is not associated with the spent fuel pool are also no longer required to be in the emergency plan when administrative controls are established to limit source term accumulation and the offsite consequences of uncontrolled effluent releases.

Therefore, the ICs listed in Table 1 below are being deleted from the currently approved emergency plan for KPS. The I~s being deleted are either associated only with spent fuel pool operation or are ICs for which administrative controls to limit possible effluent releases have been established.

Table I - Emergency Plan Initiating Conditions Being Deleted ALERT UNUSUAL EVENT PD-AA1 (all) PD-AU1I (all)

An uncontrolled release of gaseous or liquid An uncontrolled release of gaseous or liquid radioactivity resulting in detectable levels at the radioactivity for 60 minutes or longer.

site boundary.

PD-AA2 (all) PD-AU2 (all)

UNPLANNED rise in plant radiation levels that UNPLANNED rise in plant radiation levels.

impedes plant access required to maintain spent

• fuel integrity.

Serial No.15-306 Attachment I Page 4 of 17 Table 1 - Emergency Plan Initiating Conditions Being Deleted ALERT UNUSUAL EVENT PD-SU1 (all)

UNPLANNED spent fuel pool temperature rise.

HOSTILE ACTION within the VBS boundary e*

airon atac threa ithin 30#minutcs,,.*

PD-HA1 .2 PD-H UI.3 A validated notification from NRC providing A validated notification from the NRC providing information of an aircraft attack threat within 30 information of an aircraft threat minutes of the site.

PD-HU2 (all)**

Hazardous event affecting SAFETY SYSTEM equipment necessary for spent fuel cooling.

• Only the strike-thru portion is being deleted.

• • For an ISFSI-only facility, the condition addressed by PD-HU2 remains fully addressed by IC EUI.1 (which is being retained in the emergency plan).

The currently existing KPS ICs and EALs not listed in Table I above are being retained.

The EAL ICs being deleted include all ICs associated with the categories of abnormal radioactivity release and system malfunction. These two categories apply only to spent fuel pool operation.

The EAL l~s being retained in the emergency plan are appropriate to address the condition of an ISFS I-only facility (no fuel stored in the spent fuel pool).

2.3 Emergency Response Organization Revision A Resource Manager is provided to assist in assessing the event and obtaining needed resources. The Resource Manager is required to be in contact with the Emergency Director within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of classification of an Unusual Event or ALERT. The Resource Manager augments the Emergency Director by assisting in assessing the emergency condition and coordinating required resources, including public information interface.

Services provided to the Emergency Director by the Resource Manager can be provided remotely and do not necessitate an onsite response of the Resource Manager.

By responding remotely, the actual response time is decreased with no negative impact to services and functional responsibilities provided by the Resource Manager.

Events that require entry into the ISFSI-Only Emergency Plan would be an extreme natural phenomenon (beyond design bases) or a security condition, either of which would negatively impact or restrict access to the site.

Serial No.15-306 Attachment 1 Page 5 of 17 Therefore, the Resource Manager's functional responsibilities would be performed in a timely manner by not requiring the Resource Manager to physically report to KPS during a classified emergency.

3.0 TECHNICAL ANALYSIS

3.1 Radiological Consequences of Design Basis Events Kewaunee Power Station (KPS) is situated in rural northeast Wisconsin, located in Kewaunee County on the west shore of Lake Michigan. The land area within a 20 mile radius is primarily farmland, with a population density of about 60 people per square mile. The current population of the entire county of Kewaunee is about 20,600 people.

Two small cities, Two Rivers (12,000 population) and Manitowoc (34,000 population), lie about 11 and 17 miles, respectively, south of the station. The nearest population center of substance is the city of Green Bay (104,000 population), located about 27 miles west-northwest of the station. The entire 50-mile radius east of the station is occupied by the waters of Lake Michigan (i.e., unpopulated).

Section 14 of the KPS Updated Safety Analysis Report (USAR) previously described the design basis accident (DBA) scenarios and transient scenarios that were applicable to KPS with fuel stored within the spent fuel pool. However, after transfer of all irradiated fuel to storage in the ISFSI, those accident scenarios postulated in the USAR are no longer possible. The ISFSI is a passive system that does not rely on electrical power for heat transfer. After removal of the spent fuel from the spent fuel pool, there are no credible fuel related accidents for which Certified Fuel Handler, Shift Manager, or Non-certified Operator actions are required to prevent occurrence or mitigate the consequences. There is no credible accident resulting in radiological releases requiring offsite protective measures.

The robust design and construction of the spent fuel storage systems selected for use at the ISFSI prevents the release of radioactivity in the event of an off-normal or accident event as described in the ISFSI storage system UFSARs. Leakage of fission products from a canister confinement boundary breach is not considered to be a credible event, given the high integrity nature of the canisters' design and the additional protection afforded the canisters by the storage casks.

DEK submitted a PSDAR for KPS dated April 25, 2014 (Reference 2), which identified that KPS will decommission using a SAFSTOR method in which most fluid systems are drained and the plant is left in a stable condition until final decontamination and dismantlement activities begin. The PSDAR documented that DEK expects to have all spent fuel transferred to the ISFSI by the end of 2016.

After all the spent fuel has been removed from the spent fuel pool, the estimated radiological inventory (non-fuel) that remains at the reactor facility is primarily

Serial No.15-306 Attachment 1 Page 6 of 17 attributable to activated reactor components and structural materials. There are no credible accident sequences that can mobilize a significant portion of this inventory for release. As a result, the potential accidents that could occur during the decommissioning of the reactor facility have negligible offsite and onsite radiological consequences.

With all spent nuclear fuel in dry storage within the ISFSI, the radiological status of the facility required for implementing this proposed revision to the KPS ISFSI-Only Emergency Plan (IOEP) is summarized as follows:

• The remaining radiological source term at KPS is not likely to create an unplanned/unanticipated increase in radiation or in liquid or airborne radioactivity levels outside of the site boundary that would result in doses to the public above ODCM limits at the site boundary.

• Source term accumulation from activities during decontamination and dismantling of radioactive systems, structures, and components are administratively controlled at a level that would preclude the declaration of an Unusual Event (UE).

• Necessary radiological support personnel will be administratively required to be onsite during active decontamination and dismantling of radioactive systems, structures, and components.

The IOEP and certain initiating conditions (I~s) and emergency action levels (EALs), for which administrative controls to limit possible effluent releases will be established, do not apply to decontamination or dismantling of radioactive systems, structures, and components.

NUREG-0586 (Reference 7) supports this conclusion in the following statement.

"The staff has reviewed activities associated with decommissioning and determined that many decommissioning activities not involving spent fuel that are likely to result in radiological accidents are similar to activities conducted during the period of reactor operations. The radiological releases from potential accidents associated with these activities may be detectible. However, work procedures are designed to minimize the likelihood of an accident and the consequences of an accident, should one occur, and procedures will remain in place to protect health and safety while the possibility of significant radiological accident exists."

NUREG-0586 also makes the following supportive statement.

"The staff has considered available information, including comments received on the draft of Supplement 1 of NUREG-0586, concerning the potential impacts of non-spent-fuel-related radiological accidents resulting from decommissioning. This information indicates, that with the mitigation procedures in place, the impacts of radiological accidents are neither detectable nor destabilizing. Therefore, the staff makes the

Serial No.15-306 Attachment 1 Page 7 of 17 generic conclusion that impacts of non-spent-fuel-related radiological accidents are SMALL. The staff has considered mitigation and concludes that no additional measures are likely to be sufficiently beneficial to be warranted."

Accordingly, administrative controls that are designed to minimize the likelihood and consequence of an off-normal or accident event would be implemented when decontamination or dismantling activities of radioactive systems, structures, and components are being performed.

Implementation of the IOEP would involve DEK establishing administrative controls for radiological source term accumulation limits and methods to control the accidental dispersal of the radiological source.

Examples of radiological source term accumulation limits are limits on:

• Radioactive materials collected on filter media and resins (dose rate limit).

• Contaminated materials collected in shipping containers (dose rate limit).

• Surface or fixed contamination on work areas that may create airborne radioactive material (activity limits).

• Radioactive liquid storage tank (activity concentration limit).

An example of a method to control accidental dispersal of the radiological source term is limitation on dispersal mechanisms that may cause a fire (e.g., limits on combustible material loading, use of fire watch to preclude fire, etc.) or placement of a berm around a radioactive liquid storage tank. If the dispersal control fails, the limits on source term would preclude exceeding the site boundary source term limit.

As discussed in the previously granted exemption from various emergency planning requirements contained in 10 CER 50.47 and 10 CFR 50, Appendix E (Reference 3), an analysis of the potential radiological impact of a design basis accident at KPS in a permanently defueled condition indicates that any releases beyond the site boundary are below the EPA PAG exposure levels, as detailed in Reference 4. The basis for these exemptions has not changed and remains in effect for the proposed emergency plan changes.

3.2 Radiological Consequences of Postulated Events Although the limited scope of design basis accidents that remain applicable to the KPS facility justifies a reduction in the necessary scope of emergency response capabilities, DEK also assessed beyond design basis events using past industry precedence, including information contained in Appendix I, "Radiological Accidents," of NUREG-0586 (Reference 7).

Serial No.15-306 Attachment I Page 8 of 17 Under the previous facility condition of fuel stored within the spent fuel pool, the most severe postulated beyond design basis event involved a highly unlikely sequence of events that causes heatup of the spent fuel, postulated to occur without heat transfer, such that the zircaloy fuel cladding reaches ignition temperature. The resultant zircaloy fire could lead to the release of large quantities of fission products to the atmosphere.

However, after removal of the spent fuel from the spent fuel pooi, the configuration of spent fuel stored in dry storage precludes the possibility of such a scenario.

With this previously limiting beyond design basis scenario no longer possible, DEK assessed the following beyond design basis events associated with performance of decommissioning activities with all irradiated fuel stored in the KPS ISFSI. A summary of the assessments is provided below.

1. Cask Drop Event (Fuel Related Accident)

KPS is the holder of a general license for the storage of spent fuel in an ISFSI at power reactor sites in accordance with the provisions of 10 CFR 72.210 and 10 CFR 72.212. The generally licensed ISESI at KPS is used for interim onsite dry storage of spent nuclear fuel assemblies in both the Transnuclear Standardized NUHOMS System (NUHOMS Certificate of Compliance (CoC) 1004 (Amendments 9 and 10))

and in the NAC International MAGNASTOR System (MAGNASTOR CoC 1031 (Amendment 5)).

As documented in the storage system FSARs, analysis of the normal and off-normal events, including drop events, determined that canister drops can be sustained without breaching the confinement boundary, preventing removal of spent fuel assemblies, or causing a criticality accident. There are no evaluated normal conditions or off-normal or accident events that result in damage to the canister producing a breach in the confinement boundary. Neither normal conditions of operation or off-normal events preclude retrieval of the fuel for transport and ultimate disposal.

The dry spent fuel storage casks used at KPS are approved for storage of spent fuel per 10 CFR 72.214; and, as such, are in compliance with the requirements of 10 CFR 72.24 and 10 CFR 72.122 for off-normal and accident events to ensure that they will provide safe storage of spent fuel during all analyzed off-normal and accident events. Therefore, no radiological release would be expected to occur.

The results of the above assessment indicate that the projected radiological doses at the controlled area boundary are less than the EPA PAGs.

2. Radioactive Material Handling Accident (Non-Fuel Related)

The limiting non-fuel related event involves the release of radioactive material from a concentrated source, such as filters, resins, and shipping containers (as discussed

Serial No.15-306 Attachment 1 Page 9 of 17 in NUREG-0586, Appendix I). The initiator for these events could be a fire, explosion, or a handling event (cask drop). During the KPS SAFSTOR dormancy period, after all spent fuel has been moved to the ISESI, there would be no concentrated source of radioactive materials whose release to the environment could exceed two times the 00CM limit at the site boundary (2 x 1500 mrem/year).

During decontamination and dismantlement activities, administrative controls would limit the total amount of activity that can accumulate in a concentrated source.

Calculation RA-0065 (Enclosure 3) develops an activity accumulation limit methodology for decontamination and dismantlement of irradiated stainless steel (e.g., reactor vessel internals) and irradiated concrete (e.g., reactor coolant loop bio-shield walls) based on isotopic mixtures from NUREG/CR-3474, such that a release to the environment from concentrated sources of these radioactive materials would not exceed two times the 0DCM limit at the site boundary.

Representative material samples will be taken and analyzed for actual decommissioning/dismantlement work. The analysis will provide current source term characterization for actual decommissioning/dismantlement work. Using the methodology consistent with RA-0065 calculation, container/filter maximum radioactivity limits will be derived.

The results of the above assessment indicate that the projected radiological doses at the controlled area boundary are less than the EPA PAGs.

3. Accidents Initiated in External Events The effects of external events, such as aircraft crashes, fires, floods, wind (including tornados), earthquakes, lightning, and physical security breaches on the ISFSI remain unchanged from the effects that were considered under the existing emergency plan. Externally initiated events are addressed by the proposed EALs.

In summary, there continues to be a low likelihood of any postulated event resulting in radiological releases requiring offsite protective measures, and there is no credible radioactive material event (non-fuel related) resulting in radiological releases requiring declaration of an emergency.

3.3 ISFSI-Only Emergency Plan The KPS ISFSI-Only Emergency Plan (IOEP) is provided in Enclosure 1 to this submittal for NRC review and approval. This proposed emergency plan is associated with EALs for events related to the ISESI. The IOEP addresses the applicable regulations stipulated in 10 CFR 50.47, "Emergency Plans" (as exempted), 10 CFR 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities" (as exempted), and 10 CFR 72.32, "Emergency Plan," and is consistent with the applicable guidelines established in NUREG-0654/FEMA-REP-1, Revision 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants."

Serial No.15-306 Attachment I Page 10 of 17 The IOEP describes the station's plan for responding to emergencies that may arise at KPS while all spent nuclear fuel is in dry storage within an ISFSI. Currently, irradiated fuel is stored both in the ISFSI and in the spent fuel pool (SFP). After all fuel at KPS is in dry storage within the ISFSI, the number and severity of potential radiological accidents is significantly less than when fuel was also stored in the SEP.

The KPS ISFSI-Only Emergency Plan conservatively provides that the emergency planning zone for the ISFSI is the area within the site boundary. At KPS, the site boundary completely encompasses the controlled area. The controlled area, as defined in 10 CFR 72.3, "Definitions," means the area immediately surrounding an ISFS1 for which DEK exercises authority over its use and within which ISFSI operations are performed.

The controlled area is established to limit dose to the public during normal operations and design basis accidents in accordance with the requirements of 10 CFR 72.104, "Criteria for radioactive materials in effluents and direct radiation from an ISFSI or MRS," and 10 CFR 72.106, "Controlled area of an ISFSI or MRS." DEK's analyses of the radiological impact of potential accidents at the ISFSI conclude that any releases beyond the ISFSI controlled area boundary are expected to be less than the EPA PAGs. The controlled area is completely enclosed within the site boundary. Thus, any radiological releases beyond the site boundary will also be less than the EPA PAGs.

Based on the reduced number and consequences of potential radiological events at KPS with all spent nuclear fuel in dry storage within the ISFSI, there will continue to be no need for offsite emergency response plans for the protection of the public beyond the site boundary. Additionally, the scope of the onsite emergency preparedness organization and corresponding requirements in the emergency plan may be reduced without an undue risk to the public health and safety.

The analysis of the potential radiological impact of an accident in a condition with all irradiated fuel stored in the ISFSI indicates that any releases beyond the site boundary are below the EPA PAG exposure levels, as detailed in Reference 4. Exposure levels, which warrant pre-planned response measures, are limited to onsite areas. For this reason, radiological emergency planning is focused onsite.

3.4 ISFSI-Only Emergency Action Levels provides the site-specific EAL technical bases document, which contains the proposed KPS ISFSI-Only Emergency Action Level (IOEAL) scheme for NRC review and approval. The current KPS EAL scheme was approved by NRC on October 31, 2014 (Reference 5). The new ISFSI EAL scheme is to be implemented by the KPS ISFSI-Only Emergency Plan (provided in Enclosure 1).

Serial No.15-306 Attachment 1 Page 11 of 17 Deletions from the currently approved EAL scheme are listed in Section 2.2, "Elimination of Spent Fuel Pool Initiating Events and EALs," Table 1, "Emergency Plan Initiating Conditions Being Deleted," above.

Related Documents Supporting calculations for establishing appropriate radioactive material administrative control limits are provided in Enclosure 3 to this submittal.

Operatingq Modes and Applicability The proposed EALs are only applicable after the final spent nuclear fuel assembly has been transferred out of the spent fuel pool and placed in storage within the ISFSI.

State and Local Government Review of Proposed Changes State and local emergency management officials are advised of EAL changes that are implemented. Following NRC approval and prior to implementation, KPS will provide an overview of the new classification scheme to State and local emergency management officials in accordance with 10 CFR Part 50, Appendix E, Section IV.B.1.

4.0

SUMMARY

By letter dated May 14, 2013, DEK submitted a certification of permanent removal of fuel from the reactor vessel (Reference 1). Consequently, as specified in 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for KPS no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. The PSDAR for KPS dated April 25, 2014 (Reference 2), documented that DEK expects to have all spent fuel transferred to the ISFSI by the end of 2016. To comport to the reduced scope of potential radiological accidents with spent fuel in dry cask storage within the ISFSI, DEK proposes a new EAL scheme and corresponding emergency plan changes.

This proposed amendment would revise both the emergency plan and the EAL scheme appropriate for the condition of the station wherein all spent nuclear fuel is in dry storage within the ISFSI. The new emergency plan and EAL scheme are being submitted to the NRC for approval prior to implementation, as required under Section IV.B.2 of Appendix E to 10 CFR Part 50. Additionally, 10 CFR 50.54(q)(4) and 10 CFR 72.44(f) require that the proposed changes receive prior approval by the NRC because they are considered to reduce the effectiveness of the plan.

The proposed emergency plan does not meet all standards of 10 CFR 50.47(b) and requirements of 10 CFR 50, Appendix E. However, DEK was granted exemptions from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV, by letter dated October 27, 2014 (Reference 3). The basis for these exemptions has

Serial No.15-306 Attachment 1 Page 12 of 17 not changed and remains in effect for the proposed emergency plan changes. With the granted exemptions, the emergency plan, as revised, will continue to meet the remaining applicable requirements in 10 CFR 50, Appendix E and the planning standards of § 50.47(b).

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Pursuant to 10 CFR 50.90, Dominion Energy Kewaunee, Inc. (DEK) requests an amendment to Facility Operating License Number DPR-43 for Kewaunee Power Station (KPS). The proposed amendment would revise the emergency plan and emergency action level (EAL) scheme. The proposed changes are being submitted to the NRC for approval prior to implementation, as required under 10 CFR 50.54(q)(4), 10 CFR 50, Appendix E, Section IV.B.2, and 10 CFR 72.44(f).

By letter dated May 14, 2013, DEK submitted a certification of permanent removal of fuel from the reactor vessel (Reference 1). Consequently, as specified in 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for KPS no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. The Post-Shutdown Decommissioning Activities Report (PSDAR) for KPS dated April 25, 2014 (Reference 2), documented that DEK expects to have all spent fuel transferred to the Independent Spent Fuel Storage Installation (ISFSI) by the end of 2016. To comport to the reduced scope of potential radiological accidents with spent fuel in dry cask storage within the ISFSI, DEK proposes a new EAL scheme and corresponding emergency plan changes.

DEK has evaluated the proposed amendment to determine if a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed amendment would modify the KPS renewed facility operating license by revising the emergency plan and revising the EAL scheme. KPS has permanently ceased operation and is permanently defueled. The proposed amendment is conditioned on all spent nuclear fuel being removed from wet storage in the spent fuel pool and placed in dry storage within the ISFSI. Occurrence of postulated accidents associated with spent fuel stored in a spent fuel pool is no longer credible in a spent fuel pool devoid of such fuel. The proposed amendment has no effect on plant systems, structures, and components (SSCs) and no effect on the capability of any plant SSC to perform its design function. The proposed

Serial No.15-306 Attachment 1 Page 13 of 17 amendment would not increase the likelihood of the malfunction of any plant SSC.

The proposed amendment would have no effect on any of the previously evaluated accidents in the KPS Updated Safety Analysis Report (USAR).

Since KPS has permanently ceased operation, the generation of fission products has ceased and the remaining source term continues to decay. This continues to significantly reduce the consequences of previously postulated accidents.

Therefore, the proposed amendment does not involve a significant increase in the consequences of a previously evaluated accident.

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 amendment constitutes a revision of the emergency planning function commensurate with the ongoing and anticipated reduction in radiological source term at KPS.

The proposed amendment does not involve a physical alteration of the plant. No new or different types of equipment will be installed and there are no physical modifications to existing equipment as a result of the proposed amendment.

Similarly, the proposed amendment would not physically change any SSCs involved in the mitigation of any postulated accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed amendment does not create the possibility of a new failure mode associated with any equipment or personnel failures. The credible events for the ISFSI remain unchanged.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No Because the 10 CFR Part 50 license for KPS no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation is no longer credible. With all nuclear spent fuel pool transferred out of wet storage from the spent fuel pool and placed in dry storage within the ISFSI, a fuel handling accident is no longer credible. There are no longer credible events that would result in any releases beyond the site boundary exceeding the EPA PAG exposure levels, as detailed in the EPA's "Protective Action Guide and Planning

Serial No.15-306 Attachment 1 Page 14 of 17 Guidance for Radiological Incidents," Draft for Interim Use and Public Comment dated March 2013 (PAG Manual).

The proposed amendment does not involve a change in the plant's design, configuration, or operation. The proposed amendment does not affect either the way in which the plant structures, systems, and components perform their safety function or their design margins. Because there is no change to the physical design of the plant, there is no change to any of these margins.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, Dominion Energy Kewaunee, Inc. concludes that the proposed amendment presents no 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.

5.2 Applicable Regulatory Requirements/Criteria The regulatory requirements, as exempted, are discussed below.

Title 10 of the Code of Federal Regulations (10 CFR), Section 50.47, "Emergency plans," sets forth emergency plan requirements for nuclear power plant facilities. The regulations in 10 CFR 50.47(a)(1)(i) state, in part, "No initial~operating license for a nuclear power reactor will be issued unless a finding is made by the NRC that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency."

Section 50.47(b) establishes the standards that emergency response plans must meet for NRC staff to make a positive finding that there is reasonable assurance that the licensee can and will take adequate protective measures in the event of a radiological emergency.

• Planning Standard (1) of Section 50.47(b) states, in part: "[E]ach principal response organization has staff to respond and to augment its initial response on a continuous basis."

• Planning Standard (2) of Section 50.47(b) states, in part: "On-shift facility licensee responsibilities for emergency response are unambiguously defined, adequate staffing to provide initial facility accident response in key functional areas is maintained at all times, timely augmentation of response capabilities is available .... "

Serial No.15-306 Attachment 1 Page 15 of 17

• Planning Standard (4) of Section 50.47(b) requires that a licensee's emergency response plan contain the following: "A standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee."

• Planning Standard (8) of Section 50.47(b) states, in part: "Adequate emergency facilities and equipment to support the emergency response are provided and maintained ."

10 CFR 50.54(q)(4) specifies the process for revising emergency plans where the changes reduce the effectiveness of the plan. This regulation states the following:

"The changes to a licensee's emergency plan that reduce the effectiveness of the plan as defined in paragraph (q)(l1)(iv) of this section may not be implemented without prior approval by the NRC."

Section IV.A of Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities," to 10 CFR Part 50, states, in part: "The organization for coping with radiological emergencies shall be described, including definition of authorities, responsibilities, and duties of individuals assigned to the licensee's emergency organization .... "

Section IV.C.1 of Appendix E requires each emergency plan to define the emergency classification levels that determine the extent of the participation of the emergency response organization.

Section IV.E of Appendix E states, in part: "Adequate provisions shall be made and described for emergency facilities and equipment .... "

As proscribed in 10 CFR 72.13, "Applicability", the applicable emergency plan requirements for an Independent Spent Fuel Storage Installation associated with a general license are specified in 10 CFR 72.32(c) and (d).

The proposed emergency plan continues to rely on previously granted exemptions from certain emergency planning requirements (Reference 3) as the basis for these exemptions has not changed and remains in effect.

In November 2012, NEI published NEI 99-01, Revision 6 (Reference 6). NRC endorsed NEI 99-01, Revision 6, by letter dated March 28, 2013 (Reference 8). The changes being requested herein are based on Revision 6 to NEI 99-01. The proposed changes are conservatively being considered as a change to the EAL scheme development methodology. Pursuant to 10 CFR Part 50, Appendix E, Section IV.B.2, a revision to an entire EAL scheme must be approved by the NRC before implementation.

Serial No.15-306 Attachment 1 Page 16 of 17 5.3 Precedent Similar changes to emergency plans and associated emergency action levels for plants that have transitioned to ISFSI-only status were approved by NRC for the La Crosse Boiling Water Reactor facility on September 8, 2014 (Reference 9) and for the Zion facility on May 14, 2015 (Reference 10).

5.4 Conclusion Based on the considerations discussed 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.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. Letter from D. G. Stoddard (DEK) to NRC Document Control Desk, "Certification of Permanent Removal of Fuel from the Reactor Vessel," dated May 14, 2013.

[ADAMS Accession No. ML13135A209]

2. Letter from D. G. Stoddard (DEK) to NRC Document Control Desk, "Revision to Post-Shutdown Decommissioning Activities Report," dated April 25, 2014.

[ADAMS Accession No. ML14118A382]

3. Letter from Thomas J. Wengert (NRC) to David A. Heacock (DEK), "Kewaunee Power Station - Exemptions from Certain Emergency Planning Requirements and Related Safety Evaluation (TAC No. MF2567)," dated October 27, 2014.

[ADAMS Accession No. ML14261A223]

Serial No.15-306 Attachment 1 Page 17 of 17

4. U.S. Environmental Protection Agency, "Protective Action Guide and Planning Guidance for Radiological Incidents," Draft for Interim Use and Public Comment dated March 2013, (PAG Manual).

5. Letter from Thomas J. Wengert (NRC) to David A. Heacock (DEK), "Kewaunee Power Station - Issuance of Amendment for Changes to the Emergency Plan and Emergency Action Levels (TAC No. MF3411)," dated October 31, 2014.

[ADAMS Accession No. ML14279A482]

6. Nuclear Energy Institute (NEI) 99-01, Revision 6, "Methodology for Development of Emergency Action Levels for Non Passive Reactors," November 2012.

[ADAMS Accession No. ML12326A805]

7. NUREG-0586, "Generic Environmental Impact Statement of Decommissioning of Nuclear Facilities," Supplement 1, Volume 1, November 2002.

8. Letter from Mark Thaggard (NRC) to Susan Perkins-Grew (NEI), "U.S. Nuclear Regulatory Commission Review and Endorsement of NEI 99-01, Revision 6, Dated November, 2012 (TAC No. D92368)," dated March 28, 2013. [ADAMS Accession No. ML12346A463]

9. Letter from U.S. Nuclear Regulatory Commission to Dairyland Power Cooperative (La Crosse Boiling Water Reactor), "Issuance of Amendment Relating to the Dairyland Power Cooperative La Crosse Boiling Water Reactor Request for Changes to the Emergency Planning Requirements (TAC No.

J52956)," dated September 8, 2014. [ADAMS Accession No. ML14155A112]

10.Letter from U.S. Nuclear Regulatory Commission to Zion Solutions LLC (Zion Nuclear Power Station), "Issuance of Amendments Relating to the Emergency Planning Requirements for Zion Nuclear Power Station, Units 1 and 2 (TAC Nos.

J52992 and J52993)," dated May 14, 2015.

11. Letter from Gianna C. Clark (DEK) to NRC Document Control Desk, "License Amendment Request 260, Proposed Changes to License and Technical Specifications to Reflect Permanent Removal of Spent Fuel from Spent Fuel Pool," dated September 14, 2015.

Serial No.15-306 ENCLOSURE 1 LICENSE AMENDMENT REQUEST 259:

ISFSI-ONLY EMERGENCY PLAN AND EMERGENCY ACTION LEVEL SCHEME ISFSI-ONLY EMERGENCY PLAN KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

• Dominion° ISFSI-Only Emergency Plan (IOEP)

Revision Number Effective Date 0ODraft D TDB Revision Summary:

ISFSI-Only Emergency Plan (IOEP) describes the station's plan for responding to emergencies that may arise at the Kewaunee Power Station ISFSI.

1

_ Table of Contents Section Title Page 1.0 Introduction 4 1.1 Purpose 1.2 Scope 2.0 Discussion 5 2.1 Overview of ISFSI-Only Emergency Plan (IOEP) 2.2 Facility Description 3.0 Definitions and Acronyms 6 4.0 References 9 5.0 Assignment of Responsibility 10 5.1 Emergency Response and Responsibilities 5.20Offsite Response Organizations (ORO) 6.0 Onsite Emergency Organizations 11 6.1 On-Shift Positions 6.2 Augmented Organization

__________6.3 Functional Responsibilities 7.0 Emergency Response Support and Resources 13 8.0 Emergency Classification System 14 9.0 Notification Methods and Procedures 15 10.0 Emergency Communications 16 11.0 Public Information 16 12.0 Emergency Equipment and Facilities 16 12.1 Emergency Response Facilities (ERF)

____ 12.2 Emergency Equipment 13.0 Accident Assessment 18 14.0 Protective Actions 18 15.0 Radiological Exposure 19 16.0. Medical and Health Support 20 17.0 Recovery 21 18.0 Exercise and Drills 22 19.0 Emergency Response Training 24 19.1 Emergency Response Personnel Training 19.2 Non-Kewaunee Power Station Emergency Response Support Organizations 20.0 Maintaining Emergency Preparedness 25 20.1 Emergency Preparedness Responsibilities 20.2 Review and Updating of the IOEP

_________20.3 Maintenance and Inventory of Emergency Equipment and Supplies Appendix A Emergency Equipment, Supplies and Reference Materials 27 Appendix B Table B-i 28 Cross Reference IOEP Section to Planning Standards/Requirements/Criteria to Procedures

List of Tables Section Table Title Page 6 6-1 Emergency Response Organization Staffing and Responsibility 13 10 10-1 Communication Systems 16 15 15-1 Response Worker Guidelines 20 Appendix B B-i IOEP Section to Planning Standards/Requirements/Criteria to Procedures 28 3

1.0 INTRODUCTION

The Kewaunee Power Station's (KPS) Independent Spent Fuel Storage Installation (ISFSI) Only Emergency Plan (IGEP) describes the plan for responding to emergencies that may arise at the station's ISFSI. In this condition, no reactor operations can take place and all irradiated fuel is removed from the Spent Fuel Pool (SFP). This IOEP adequately addresses the risks associated with KPS's current conditions.

As provided in the ISFSI storage system UFSARs, the analyses of the potential radiological impacts of postulated off-normal, natural phenomenon, and accident events in an ISFSI-Only condition indicates that any releases beyond the Site Boundary would result in a dose to the public below the radiation limits established in 10 CFR 72.106(b). Exposure levels, which warrant pre-planned response measures, are generally limited to the ISFSI pad and nearby vicinity, and for this reason; radiological emergency planning is focused on this area.

1.1 PURPOSE The purpose of the IOEP is to assure an adequate level of preparedness to cope with the spectrum of emergencies that could be postulated to occur. This plan integrates the necessary elements to provide effective emergency response considering cooperation and coordination of organizations expected to respond to emergencies.

1.2 SCOPE The IOEP is developed to respond to potential radiological emergencies at the KPS ISESI. Because there are no postulated off-normal, natural phenomenon, or accident events that would result in offsite dose consequences large enough to require offsite emergency planning, the overall scope of this plan delineates the actions necessary to safeguard onsite personnel. The concepts presented in this plan address the applicable regulations stipulated in 10 CFR 50.47, "Emergency Plans," and 10 CFR 50 Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities". The plan is consistent with the applicable guidelines established in NUREG-0654!FEMA-REP-1, Revision 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants."

Exemptions from selected portions of 10 CFR 50.47 and 10 CFR 50 Appendix E for Kewaunee Power Station were granted by the Nuclear Regulatory Commission (NRC) on October 27, 2014 (ADAMS Accession Number:

ML14261A223).

The IOEP, revision 0, was approved per NRC Safety Evaluation dated [insert date prior to issuing].

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2.0 DISCUSSION 2.1 OVERVIEW OF ISFSI-ONLY EMERGENCY PLAN (IOEP)

In the event of an emergency at the KPS ISFSI, actions are required to identify and assess the nature of the emergency and to respond in a manner that protects the health and safety of the public and onsite personnel.

This plan is activated by the ISFSI Shift Supervisor (ISS) upon identification of an emergency situation based upon the Emergency Action Level (EAL) criteria. The ISS assumes the position of the Emergency Director (ED). The emergency measures described in the subsequent sections and implementing procedures are implemented in accordance with the classification and nature of the emergency at the direction of the ED.

This emergency plan describes the organization and responsibilities for implementing emergency measures. It describes interfaces with Offsite Agencies (Federal, State and local) which may be notified in the event of an emergency, and may provide assistance. Fire, ambulance, and law enforcement services are provided by local public entities. Medical services are provided by Aurora Medical Center in Two Rivers, Wisconsin.

Because there are no postulated events that would result in offsite dose consequences large enough to require offsite emergency planning, emergencies are divided into two classifications: Unusual Event (UE) and Alert.

KPS is responsible for planning and implementing emergency measures within the Site Boundary. This emergency plan is provided to meet this responsibility.

To carry out specific emergency measures discussed in this plan, detailed implementing procedures are established and maintained.

In addition to the description of activities and steps that-ca-n be im~plemented....

during a potential emergency, this emergency plan also provides a general description of the steps taken to recover from an emergency situation. It also describes the training, drills/exercises, planning, and coordination appropriate to maintain an adequate level of emergency preparedness.

2.2 FACILITY DESCRIPTION KPS has permanently ceased power operations and all irradiated fuel has been removed from the SFP and placed into dry storage within an ISFSI. On May 14, 2013, the station certified permanent removal of fuel from the reactor vessel in accordance with 10 CFR 50.82(a)(1 )(i) and (ii). The 10 CFR 50 license for KPS no longer authorizes operation of the reactor, and emplacement or retention of fuel into the reactor vessel, as specified in 10 CFR 50.82(a)(2).

5

The KPS ISFSI is located in the town of Carlton, Kewaunee County, along the west shore of Lake Michigan in east central Wisconsin. The topography of the region is gently rolling to flat, with elevations varying from 10 to 100 feet above the level of Lake Michigan. The land surrounding the site slopes gradually east towards Lake Michigan from the higher elevations in the west. At the northern and southern perimeters of the site, bluffs form the boundary between the plant site and Lake Michigan.

3.0 DEFINITIONS and ACRONYMS This section provides definitions that are used in this document. Terms capitalized in the text of the definitions indicate that they are defined elsewhere in this section.

Alert - Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the ISFSI or a security event that involves probable life threatening risk to station personnel or damage to ISESI equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Annual - Frequency of occurrence is met if performed within 1.25 times a 12 month interval as measured from the previous performance. This definition does not apply to the term "annual" when it relates to the conduct of the Emergency Preparedness Exercise and off-year Drill. The Exercise and off-year Drill are performed within the calendar year.

Accountability - Discretionary protective action taken for all persons onsite (within the ISFSI PROTECTED AREA) that involves the gathering of personnel into pre-designated areas and subsequent verification that the location of all personnel is known.

Assessment Actions - Those actions taken during or after an incident to obtain and process information necessary to make decisions to implement specific emergency measures.

Corrective Action - Those emergency measures taken to mitigate or terminate an emergency situation at or near the source of the problem in order to prevent an uncontrolled release of radioactive material or to reduce the magnitude of a release (e.g., equipment shutdown, fire fighting, equipment repair, and damage control).

Desigqn Basis Accident (DBA) - Credible accident events as analyzed in the ISFSI Updated Final Safety Analysis Report (UFSAR).

6

Emergqency Action Level (EAL) - A pre-determined, site-specific, observable threshold for an INITIATING CONDITION (IC) that when met or exceeded places the station in a given emergency classification level.

Emergqency Plan Implementingq Procedures (EPIP) - Specific procedures describing actions needed to implement the IOEP.

Emergqency Plan Maintenance Procedures - Specific procedures describing the methods established to maintain and monitor the IOEP.

Emergqency Response Facility (ERF) - The facility containing the communication equipment necessary for emergency conditions. It is operated under the direction of the ED and serves as the primary location for Classification of the incident, Notification of incident to offsite agencies, ASSESSMENT ACTIONS, and CORRECTIVE ACTION direction.

Emerqency Response Orgqanization (ERO) - Individuals who have been assigned an emergency response position within the IOEP.

Environmental Protection Agency (EPA) - An agency of the U.S. federal government which was created for the purpose of protecting human health and the environment by writing and enforcing regulations based on laws passed by Congress.

Hostile Action - An act toward the KPS ISESI or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the KPS ISESI. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals within the SITE BOUNDARY).

Independent Spent Fuel Storage Installation (ISFSI) - A complex designed and constructed for the interim storage of spent nuclear fuel, solid reactor-related Greater Than Class C (GTCC) waste, and other radioactive materials associated with spent fuel and reactor-related GTCC waste storage (10 CFR 72.3).

Initiating Condition (IC)- An event or condition that aligns with the definition of one of the two emergency classification levels by virtue of the potential or actual effects or consequences.

Monthly - Frequency of occurrence is met if performed within 1.25 times a 31 day interval as measured from the previous performance.

7

Personnel Monitoringq Equipment - Radiation exposure measuring devices designed to be worn or carried by an individual for the purpose of measuring the radiation dose received (e.g., direct reading dosimeters and TLDs).

Protected Area (PA) - The area encompassed by physical barriers and to which access is controlled.

Protective Actions - Those measures taken in anticipation of or after an inadvertent release of radioactive material for the purpose of preventing or minimizing radiological exposures to onsite personnel.

Quarterly - Frequency of occurrence is met ifperformed within 1.25 times a 92 day interval as measured from the previous performance.

Radioactive Release - Any radioactive material beyond pre-emergency levels and not attributable to normal operations, either detected or suspected of migrating beyond the PA, while in a classified emergency.

Radiologqical Control Area (RCA) - An area in which radioactive material is present and the potential exists for the spread of radioactive contamination. The area is posted for the purpose of protecting individuals against undue risks from exposure to radiation and radioactive materials.

Security Condition - Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety. A Security Condition does not involve a Hostile Action.

Site Boundary - The perimeter of the land owned by Dominion Energy Kewaunee Inc. The ISFSI Controlled Area, as defined in 10 CFR 72.3, is bounded within the Site Boundary.

Unusual Event (UE) - Events are in progress or have occurred which indicate a potential degradation of the level of safety of the ISFSI or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation OCCU rs.

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4.0 REFERENCES

• 10 CFR 50.47, "Emergency Plans"

• 10 CFR 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities"

• NUREG-0654/FEMA-REP-1, Revision 1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants" (November 1980)

• NUREG-1 140, Final Report published January 1988, "A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Lice nsees"~

• Facility Technical Specifications

• Emergency Preparedness Procedures

• NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors"

• EPA's "Protective Action Guide and Planning Guidance for Radiological Incidents," Draft for Interim Use and Public Comment dated March 2013

• Kewaunee Power Station Exemption from Certain Emergency Planning Requirements and Related Safety Evaluation dated October 27, 2014 (ADAMS Accession Number: ML14261A223)

• NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities

• NRC IN 90-08, KR-85 Hazards from Decayed Fuel

• 10 CFR 72.13, Applicability

• 10 CFR 72.32, Emergency plans

• 10 CFR 72.44, License conditions

• 10 CFR 72.106, Controlled area of an ISFSI or MRS

• ISFSI Storage System Certificates of Compliance, Updated Final Safety Analysis Reports and Technical Specifications 9

5.0 ASSIGNMENT OF RESPONSIBILITY Primary responsibilities for emergency response have been assigned, the emergency responsibilities of the various supporting organizations have been specifically established, and each principal response organization has staff to respond and to augment its initial response on a continuous basis.

5.1 Emergency Response and Responsibilities The ISFSI Shift Supervisor (ISS) is at KPS 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day and is the senior management position during off-hours. This position is responsible for monitoring conditions and managing the activities at the KPS ISFSI.

When an off-normal, natural phenomenon, or accident event becomes apparent, the ISS shall assess the condition and assume the position of Emergency Director (ED). The functions associated within the ED's scope of responsibilities are specified on Table 6-1.

The Emergency Director does not have concurrent duties which conflict with the above responsibilities.

The on-shift staff positions are available 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day. The on-shift staff can perform all required IOEP actions. At the direction of the ED, additional personnel will be activated and augment the on-shift staff.

A Resource Manager assists in assessing the event and obtaining needed resources.

5.2 Offsite Response Organizations (ORO)

The ED coordinates the OROs' response (fire, ambulance and local law enforcement agency (LLEA)), access and radiological controls with the onsite activities. The OROs listed below are capable of 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> emergency response.

State and local government agency response will be in accordance with each agency's plans and procedures, and commensurate with the hazard posed by the emergency. Letters of Agreement are in place for those local agencies that will respond.

City of Kewaunee Fire Department Arrangements have been made with the City of Kewaunee Fire Department to provide the primary response as requested. The City of Kewaunee Fire Department is located about 10 miles from the KPS ISFSI, which allows for a timely response from the initial notification.

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City of Kewaunee Ambulance Arrangements have been made with the City of Kewaunee Ambulance for ambulance services. The agreement includes a commitment for medical transportation of contaminated injured workers.

Aurora Medical Center Arrangements have been made for medical services with Aurora Medical Center, located approximately 14 miles from the KPS ISFSI. The agreement includes a commitment by the hospital to accept and treat personnel with routine industrial injuries as well as injuries complicated by radioactive contamination or radiation exposure. The Aurora Medical Center maintains the capability and facilities to provide radioactive decontamination, first aid, and emergency stabilization medical treatment to injured personnel. These services and facilities are available 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day.

Kewaunee County Sheriff's Department An agreement is maintained with the Kewaunee County Sheriff's Department to provide emergency assistance per the Security Plan.

6.0 EMERGENCY RESPONSE ORGANIZATION (ERO)

ERO responsibilities for emergency response are listed in Table 6-1.

6.1 ON-SHIFT POSITIONS KPS has personnel on-shift at all times that provide the initial response to an off-normal, natural phenomenon, or accident event. Members of the on-shift organization are trained on their responsibilities and duties in the event of a classified emergency and are capable of performing all necessary response actions until the augmenting staff arrives or the event is terminated. The on-shift staffing assignments include the roles and responsibilities for their emergency response functions.

ISFSI Shift Supervisor (ISS)/EMERGENCY DIRECTOR (ED)

The ISS is at KPS 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day and is the senior management position during off-hours. This position is responsible for monitoring conditions and approving all onsite activities.

When an off-normal, natural phenomenon, or accident event becomes apparent, the ISS shall assess the condition and assume the position of Emergency Director (ED).

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The ED assumes overall command and control of the event response. The ED cannot delegate the following responsibilities:

• Classification of event.

• Authorization of radiation exposures in excess of 10 CFR 20 limits.

Other responsibilities assumed by the ED are included in Table 6-1 SECURITY Security is administered by the ISESI Physical Security Plan. Security will perform accountability at the direction of the ED.

6.2 AUGMENTED ORGANIZATION RESOURCE MANAGER The Resource Manager will be in contact with the ED within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of classification. The Resource Manager will augment the ED by assisting in assessing the emergency condition (refer to Table 6-1) and coordinating required resources, including public information interface. The Resource Manager does not need to physically report to KPS to perform their responsibilities.

AUGMENTATION PERSONNEL Additional personnel resources may be directed to report to KPS to provide additional support as needed to assess radiological conditions, support maintenance and repair activities, develop and implement corrective action plans, and assist with recovery actions. The augmentation personnel are available from KPS staff and Dominion facilities, and can be requested from various contractors.

OFFSITE RESPONSE ORGANIZATIONS (ORO)

Additional support is available from OROs, as previously discussed in Section 5.2 of this emergency plan.

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6.3 FUNCTIONAL RESPONSIBILITIES Table 6-1 below lists the functional responsibilities of positions that fulfill emergency staffing capabilities.

TABLE 6-1 Emergency Response Organization Staffing and Responsibility AUGMENTED OFFSITE FUNCTIONAL AREA LOCATION ON-SHIFT STAFF RSOS Assessment of Emergency Emergency Director Resource Manager Condition Response Facility Emergency Direction Emergency Emergency Director--

and Control Response Facility Notifications / Emergency EegnyDrco Communications Response Facility EegnyDrco -

Radiological Accident Emergency Assessment and Response Facility I Emergency Director Resource Manager Protective Actions On Scene Emergency Emergency Director Corrective Actions Response Facility /I-On Scene Per Fire Protection Offsite Response Fie ihtn O ceeProgram Plan Organization Rescue and First Aid On Scene

• Offsite Response Treatment Organization Site Access Control Scrt tto e euiyPa and Accountability Security___tation___erSecurity__Plan_---

• Provided by on-shift personnel who may be assigned other functions 7.0 OFFSITE EMERGENCY RESPONSE SUPPORT AND RESOURCES Arrangements for requesting and effectively using resources have been made and other organizations capable of augmenting the planned response have been identified. Letters of Agreement are in place for those local agencies (fire, ambulance and LLEA) that will respond to an ISFSI emergency condition.

Letters of Agreement for each agency are maintained on file.

The ED coordinates the fire, ambulance and LLEA response as previously discussed in Section 5.2 of this Plan.

The ED is authorized to request Federal assistance as needed. The Nuclear Regulatory Commission (NRC) will act as the lead Federal agency providing coordination and support in response to a nuclear incident.

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8.0 EMERGENCY CLASSIFICATION SYSTEM A standard emergency classification and action level scheme is in use. This section describes emergency classifications, Initiating Conditions, Emergency Action Levels (EAL), and postulated emergency situations.

EMERGENCY CLASSIFICATION SYSTEM The emergency classification system covers an entire spectrum of possible radiological and non-radiological emergencies at the KPS ISESI. The emergency classification system categorizes accidents and/or emergency situations into one of two emergency classification levels depending on emergency conditions at the time of the incident. The emergency classification levels applicable at Kewaunee Power Station ISFSI, in order of increasing severity, are Unusual Event and Alert.

Each of these emergency classes requires notification to the Resource Manager, State and local government agencies, as well as the NRC.

The emergency classification system is based on NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors", revision 6.

Once indications are available that an EAL is met, the event is assessed and classified, and the corresponding emergency classification level is promptly declared as soon as possible. Notification to the State and local government agencies, and the NRC is required within 60 minutes of the event classification.

Incidents may be classified in a lower emergency classification level at first and then escalated to the higher level if the situation deteriorates. The following paragraphs outline the actions at each classification level. Refer to Emergency Action Level Technical Bases for actual parameter values, and status used to classify emergencies.

The Unusual Event status shall be maintained until an escalation in emergency class occurs or the event is terminated. Offsite authorities will be informed of the change in the emergency status and the necessary documentation will be completed as specified in the Emergency Plan Implementing Procedures.

The Alert status shall be maintained until termination of the event or de-escalation in emergency class occurs. The facility may enter recovery operations without de-escalating from an Alert. Offsite authorities will be informed of the change in the emergency status and the necessary documentation shall be completed as specified in the Emergency Plan Implementing Procedures.

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SPECTRUM OF POSTULATED OFF-NORMAL, NATURAL PHENOMENON, AND ACCIDENT EVENTS The ISFSI Updated Final Safety Analysis Report describes the Design Basis Accidents (DBAs) applicable to the KPS ISFSI along with the radiological dose calculation results. Additionally, recovery actions from the DBAs are analyzed for duration and estimated dose to workers.

9.0 NOTIFICATION METHODS AND PROCEDURES Procedures are established for notification to State and local organizations and for notification of KPS emergency personnel; the content of initial and follow-up messages to response organizations has been established.

Notification Process Nuclear Accident Reporting System (NARS) is the communication process used to notify the State and local government agencies of a classified emergency.

The notification contains information that identifies the facility, emergency classification, and EAL. Notification to the State and local government agencies will be made within 60 minutes of event classification, and the process includes a means of message verification. Notification is the responsibility of the ED.

Based upon changing conditions or as requested, follow-up messages will be communicated to the State and local government agencies. The follow-up message will contain the following information as available:

• Identification of facility.

• Identification of caller.

• Date /time of incident.

• Emergency Classification.

• Radiological condition including assessment of any radioactive release.

• Emergency response action.... .

• Request for any needed support by offsite agencies.

• Prognosis for worsening or termination of event based upon available information.

NRC Emergency Notification System (ENS)

The ENS is a dedicated telephone system used to notify the NRC Operations Center. The NRC will be notified as soon as possible after State and local notifications and within 60 minutes of event classification. In the event of failure of the ENS, any telephone will be used to notify the NRC. Notification to the NRC is the responsibility of the ED.

ERO Activation The ERO is activated by an onsite announcement and by the ERO callout system directed by the Emergency Director.

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Support Organizations Medical, LLEA, and fire fighting support services are primarily notified for assistance via the public 911 process. Requests for support services are the responsibility of the ED.

10.0 EMERGENCY COMMUNICATIONS Provisions exist for prompt communications between principal response organizations and emergency response personnel. The communication systems listed in Table 10-1 provide 24-hour onsite and offsite communications capability.

Communication systems are tested to verify proper operation at the testing frequency specified in Table 10-1. Communication systems that are listed with a testing frequency of "Frequent Use" indicates that the associated equipment is normally used at a sufficiently high regularity (e.g., multiple times each day), such that separate additional testing is not needed. Functionality is verified through normal (frequent) use of the system.

TABLE 10-1 Communication Systems Communication System Testing Frequency Commercial / PBX telephone system Frequent Use Portable radios Frequent Use NARS communication equipment/phones Monthly*

NRC ETS Network (ENS) Monthly ERO callout system Semi-annual*

• Performance of drill requirements specified in Section 18 satisfies the Testing Frequency.

11.0 PUBLIC INFORMATION Corporate Communications Department personnel will be notified of a classified emergency. Corporate Communications Department will monitor media activity and coordinate with senior management disseminating public information per communication protocols. As necessary, news conference(s) can be conducted on site or other coordinated location. Corporate Communications Department personnel, or senior KPS or corporate management will represent the facility as the spokesperson.

12.0 EMERGENCY FACILITY AND EQUIPMENT Adequate emergency facilities and equipment to support the emergency response are provided and maintained. This section of the plan identifies and describes the emergency response facility, assessment equipment, the first aid and medical facilities, and protective equipment and supplies that can be utilized during an emergency.

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12.1 EMERGENCY RESPONSE FACILITY (ERF)

The emergency command and control functions are managed within the ERF.

Within the ERF the ED (or other personnel as directed) can assess conditions; evaluate the magnitude and potential consequences of abnormal conditions; initiate preventative and corrective actions; and perform notifications.

The ERF is staffed in accordance with Section 6.0. The facility provides sufficient space to accommodate anticipated response personnel and provides availability of communication systems as specified on Table 10-1.

Radiological conditions as a result of DBAs specified in the ISFSI storage system UFSARs do not inhibit staffing of the ERF.

12.2 EMERGENCY EQUIPMENT This section describes the monitoring instruments used to initiate emergency measures and provide continuing assessment of conditions throughout the course of an emergency.

Specific emergency response equipment and reference materials are listed in Appendix A, Emergency Equipment, Supplies and Reference Materials. The items listed in Appendix A are inspected, inventoried, and operationally checked quarterly and after each use. There are sufficient reserves of instruments/equipment to replace those which are removed for calibration or repair. Equipment in these inventories is checked and calibrated in accordance with approved procedures.

Portable Radiation and Contamination Monitoring Instruments Portable radiation and contamination monitoring instruments normally utilized and maintained by the Radiation Protection group are available for emergency use.

Communication Systems Communication systems are identified and tested as described in Section 10.

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13.0 ACCIDENT ASSESSMENT Adequate methods and equipment are in use for assessing and monitoring consequences of a radiological emergency condition.

The assessment activities required to evaluate a particular emergency depend on the specific nature and classification of the emergency. The ED is responsible for the initial measurement of ISFSI dose rates after an off-normal, natural phenomena, or accident event. The EALs identify the parameter value to determine the emergency condition. Classification of events is performed by the ED in accordance with the EAL scheme.

If the measured ISFSI dose rates exceed the EAL threshold, the ED then performs a radioactive release assessment in the vicinity of the affected storage module or cask. After completing the assessment, the ED contacts the Resource Manager to assist in interpreting the radioactive release assessment results.

Notification of the radiological release assessment is in accordance with Section 9.0.

14.0 PROTECTIVE ACTIONS Protective actions for onsite personnel are provided for their health and safety.

Implementation guidelines for onsite protective actions are provided in EPIPs.

Additionally, the EPIPs provide for a range of protective actions (e.g. relocation of personnel and personnel take cover) to protect onsite personnel during hostile actions.

Accountability Accountability should be considered and used as a protective action whenever a site wide risk to health or safety exists and prudence dictates. -If personnel ---

accountability is required, at the direction of the ED all individuals at the site (including employees without emergency assignments, visitors and contractor personnel) shall be notified of the emergency.

Accountability of all personnel inside the ISESI Protected Area should be accomplished within 60 minutes after event classification and maintained thereafter at the discretion of the ED. If personnel are unaccounted for, teams shall be dispatched to locate the personnel.

Non-ERO personnel, supplemental personnel, and visitors located outside of the ISFSI PA but within the Site Boundary will be directed to report to an assembly area or exit the site as appropriate. The ED is responsible for controlling access to the site when the IOEP is activated.

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15.0 RADIOLOGICAL EXPOSURE Means for controlling radiological exposures, in an emergency, are established for emergency workers. The means for controlling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving Activity Protective Action Guides.

Radiological Control Areas (RCAs) / Access Control During a classified emergency, radiological surveys of the ISFSI pad area will be performed to determine the actual extent of the radiological concern. As necessary, the ED will ensure RCAs and access controls are established to prevent personnel from entering the area. Recovery and corrective actions will be planned and executed in a manner that minimizes exposure to personnel.

Exposure Control and Records Individuals authorized to enter RCAs are required to have in their possession dosimetry capable of measuring a dose received from external sources of ionizing radiation.

Emergency worker dose records are maintained in accordance with Radiation Protection procedures.

All reasonable measures shall be taken to control the radiation exposure to emergency response personnel providing rescue, first aid, decontamination, emergency transportation, medical treatment services, corrective actions or assessment actions within applicable limits specified in 10 CFR 20. The ED is responsible for authorizing emergency response personnel to receive doses in excess of 10 CER 20 limits, if necessary. Table 15-1 contains the guidelines for emergency exposure criteria, which is consistent with the EPA's, "Protective Action Guide and Planning Guidance for Radiological Incidents," Table 2-2, "Response Worker Guidelines."

Personnel Contamination Control All personnel are monitored for radioactive contamination prior to leaving the site.

Portable contamination monitoring instruments are available to frisk personnel for potential contamination.

Documentation of surveys, contamination, and decontamination activities shall be maintained in accordance with Radiation Protection procedures.

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TABLE 15-1 Response Worker Guidelines Guideline Activity :Condition 5 rem All occupational exposures *All reasonably achievable actions have been taken to minimize dose.

10 rein(a) Protecting valuable property Exceeding 5 remn necessary for public welfare unavoidable and all appropriate actions taken to reduce dose. Monitoring available to project or measure dose

25 rem(b) Lifesaving or protection of Exceeding 5 rem large populations unavoidable and all appropriate actions taken to reduce dose. Monitoring available to project or measure dose (a)

(b) For potential In the case ofdoses rem,incident, a very>5large medical may monitoring need toprograms consider should be considered.

raising the property and lifesaving response worker guidelines to prevent further loss of life and massive spread of destruction.

16.0 MEDICAL AND HEALTH SUPPORT Arrangements are made for medical services for injured individuals and/or contaminated injured individuals. KPS maintains on-shift personnel and equipment to provide first aid for personnel working at the site. Medical emergency supplies are located in the ERE.

If immediate professional medical help is required, local ambulance services are available to assist in the transport of seriously injured personnel.

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17.0 RECOVERY The recovery organization will be based on the normal KPS organization and would function with the senior management position being responsible for site activities.

RECOVERY OPERATIONS KPS is responsible for recovery measures and restoring the ISFSI to a stable condition. In an emergency event, immediate response actions are directed towards limiting the consequences of the emergency in a manner that will afford maximum protection to onsite personnel. Once the immediate assessment and protective actions have been implemented, the restoration and recovery measures can be implemented.

The extent and nature of the corrective and protective actions and the extent of recovery will depend on the emergency conditions at hand and the status of ISFSI. The general goals for recovery are:

• An orderly evaluation of the cause and effect of the emergency and implementation of solutions to prevent immediate recurrence of the incident.

• A planned approach for returning the ISFSI to a stable condition by obtaining the appropriate manpower, materials, and equipment.

• A planned approach to coordinate with offsite authorities to identify and resolve situations that may impact the general public.

• An evaluation of the radiation exposure records for all onsite emergency response personnel involved in the incident.

• A planned approach to ensure that radiation exposures and contamination controls are consistent with the ALARA program.

During a classified emergency, a point will be reached where the ISFSI will be placed in a stable condition. Since this condition could be attained even though specific EALs may remain exceeded, the ED will determine that there is no longer a need to keep the emergency organization in effect and to begin recovery. Although de-escalation to a lower emergency level may be performed, it is not necessary to de-escalate prior to initiating recovery.

ISFSI recovery activities shall be in accordance with the Technical Specifications and other license documents. During ISFSI recovery, the radiation exposure limits of 10 CFR 20 shall apply.

If, during recovery, an emergency situation again occurs, the emergency plan would be activated per the implementing procedures. Recovery efforts will be suspended until the emergency condition is resolved. The ED will re-evaluate ISESI conditions prior to resuming recovery.

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STATION RECOVERY TERMINATION The recovery will be terminated by the KPS senior management position after the ISFSI is returned to a stable condition.

18.0 EXERCISE AND DRILLS Periodic exercises are conducted to evaluate major portions of emergency response capabilities. Periodic drijls are conducted to develop and maintain key skills. Deficiencies as a result of exercises or drills are identified and corrected.

Exercise and Drill Kewaunee Power Station conducts a biennial Exercise to test the adequacy of timing and content of implementing procedures and methods; to test emergency equipment and communication networks; and to ensure that emergency personnel are familiar with their duties. Kewaunee Power Station will invite the OROs to participate in the Exercise.

For alternating years, a Drill is conducted for the purpose of testing, developing, and maintaining the proficiency of emergency responders.

Exercise and Drill scenarios will include, at a minimum, the following:

• The basic objective(s) of the exercise / drill.

• The date(s), time period, place(s), and participating organizations.

• A time schedule of real and simulated initiating events.

• A narrative summary describing the conduct of the drill to include such items as simulated casualties, offsite fire assistance, rescue of personnel, and use of protective clothing.

Equipment and Proficiency Drills The following drills are conducted for the purpose of training, developing, and maintaining the proficiency of emergency responders. Equipment and proficiency drills may be performed as part of an exercise, as part of a drill or as an independent drill.

Communication Drills Communications with State and local governments shall be drilled annually. The communication drill includes the aspect of understanding the content of messages.

Performance of the Communication Drill satisfies the testing requirements specified in Section 10.0.

Radiological Monitoring Drills Radiological monitoring drills, which are conducted annually, demonstrate the ability to perform radiological survey and assessment.

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Medical Emergency Drills A medical emergency drill involving a simulated contaminated individual and containing provisions for participation by the Aurora Medical Center shall be conducted at least annually. Both the Kewaunee Power Station and Point Beach Nuclear Plant (PBNP) share the facilities provided by the Aurora Medical Center. To minimize redundant training for the hospital staff, KPS and PBNP will alternate development and conduct of the drill each year.

Augmentation Capability Assessment (ACA) Drills An unannounced off-shift ACA drill shall be conducted semi-annually.

These drills shall involve implementation of the ERO callout system procedure and documentation of the estimated response time for each responder. This drill shall serve to demonstrate the capability to augment the ED after an emergency classification.

Performance of the ACA drill satisfies the ERO callout system testing requirements specified in Section 10.0.

Critique and Evaluation Critiques will evaluate the performance of the organization. The ability of emergency response personnel to self-evaluate weaknesses and identify areas for improvement is the key to successful exercise / drill conduct.

Exercise and drill performance objectives are evaluated against measurable demonstration criteria. As soon as possible following the conclusion of each exercise or drill, a critique, including participants and evaluators, is conducted to evaluate the ability of the ERO to implement the IOEP and associated procedures. Deficiencies as a result of exercises or drills are identified and entered into the corrective action system A written report is prepared following an exercise or drill involving the evaluation of designated objectives. The report evaluates and documents the ability of the ERO to respond to a simulated emergency situation. The report will also contain reference to corrective action and recommendations resulting from the exercise or drill.

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19.0 RADIOLOGICAL EMERGENCY RESPONSE TRAINING Radiological emergency response training is provided to those who may be called on to assist in an emergency.

19.1 EMERGENCY RESPONSE PERSONNEL TRAINING Requirements for emergency preparedness training are specified in the Emergency Preparedness Training Program. This program identifies the level and the depth to which individuals are to be trained.

Emergency Preparedness Training Program The training program for emergency response personnel is based on position specific responsibilities as defined in the IOEP. Emergency response personnel in the following categories receive initial training and annual retraining:

ISFSI Shift Supervisors/Emergency Directors and Resource Managers shall have training conducted such that proficiency is maintained on the topics listed below. These subjects shall be covered as a minimum on an annual basis.

• Emergency Action Level Classification.

• Federal, State and local government notification procedures.

• ERO Activation.

• Dose rate meter operation.

• Radioactive release assessment.

• Emergency exposure control.

• Protective actions for onsite personnel.

• ISFSI DBA Personnel available during classified emergencies to perform emergency response activities as an extension of their normal duties receive duty-specific training. Additional emergency preparedness training is provided as part of annual access training.

First Aid training for personnel assigned to the on-shift responsibility shall include courses equivalent to Red Cross Multi-Media.

Personnel who are badged for unescorted access receive access training annually. Information pertaining to their safety and the safety of visitors under escort during a classified emergency is included in this training.

Access training shall include the following emergency preparedness topics:

• Basic Emergency Plan and implementing procedure information.

• Emergency classification levels.

• Call out of personnel during an emergency.

• Personnel accountability procedures.

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19.2 NON-KEWAUNEE POWER STATION EMERGENCY RESPONSE SUPPORT ORGANIZATIONS Training is offered annually to non-KPS organizations which may provide specialized services during an emergency (e.g., fire-fighting, medical services, transport of injured, etc.). The training shall be structured to meet the needs of that organization with respect to the nature of their support. Training topics such as event notification, basic radiation protection, and interface activities between the offsite organization and KPS shall be made available.

20.0 MAINTAINING EMERGENCY PREPAREDNESS Responsibilities for plan development and review and for distribution of emergency plans are established, and planners are properly trained.

20.1 EMERGENCY PREPAREDNESS RESPONSIBILITIES Kewaunee Power Station Senior Management Position Has overall authority and responsibility for emergency response planning. This responsibility includes ensuring that the emergency preparedness program is maintained and implemented as described in this Plan and applicable requirements and regulations.

Emergency Preparedness Position Responsible for the following tasks:

• Maintaining and updating this IQEP and associated procedures.

• Ensuring Drill/Exercise commitments stated in the plan are met.

• Ensuring material readiness of emergency response facilities.

• Overseeing the Emergency Preparedness Training Program.

• Maintaining Emergency Preparedness interfaces with offsite agencies.

• Performing and documenting appropriate evaluations of program and of classified emergency events.

Individuals assigned the duties of maintaining the IOEP maintain an adequate knowledge of regulations, planning techniques, and the latest applications of emergency equipment and supplies. Training for these individuals includes 50.54(q) and 72.44(f) Evaluation Qualification.

Licensing Responsible for the following tasks:

• Maintaining current knowledge of changes in Federal regulations and other guidance that impact emergency planning activities.

• Submit IQEP and related controlled document revisions to the NRC.

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Oversight Oversight is responsible for performance of independent audit of the emergency preparedness program to meet the requirements of 10 CFR 50.54(t).

20.2 REVIEW AND UPDATING OF THE IOEP It is important that a state of emergency preparedness be maintained at all times.

The IOEP and Emergency Action Level Technical Bases are reviewed annually and updated, as needed. The review shall encompass the need for changes based upon the following aspects:

• Written critiques and evaluations of drills and exercises.

• Changes in the organizational structure.

• Changes in the functions and capabilities of supporting agencies.

• Changes in Federal or State regulations.

• Modifications to the facility which would affect emergency planning.

• Recommendations or agreement changes received from other organizations.

Any needed changes shall be incorporated in the IOEP, Emergency Action Level Technical Bases, and appropriate implementing procedures.

Proposed activities that may impact the IOEP must be evaluated per 10 CFR 50.54(q) and 10 CFR 72.44(f).

Emergency Action Levels (EALs) State and Local Government Agency Review The EALs shall be made available for review with State and local governmental authorities annually.

Emergency Telephone Directory Names and telephone numbers of the ERO and supporting offsite agencies shall be reviewed at least quarterly andl Updated as necessary. -.. . .

Letters of Agreements The letters of agreement with the support agencies shall be reviewed with the support agency at least every two years (biennially). Changes shall be made and the agreements renewed, as necessary.

20.3 MAINTENANCE AND INVENTORY OF EMERGENCY EQUIPMENT AND SUPPLIES Appendix A, "Emergency Equipment, Supplies and Reference Materials," lists each of the emergency response facilities and the required equipment, supplies and reference materials that are to be maintained.

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APPENDIX A Emergency Equipment, Supplies and Reference Materials EMERGENCY RESPONSE FACILITY Procedures / Reference Material ISFSI-Only Emergency Plan ISFSI-Only Emergency Action Level Bases Document Emergency Telephone Directory Emergency Plan Implementing Procedures Equipment Portable radiation monitoring instrument Portable emergency lighting Medical emergency response bag ONSITE LOCATIONS Eqluipment / Supplies Portable radiation and contamination monitoring instruments Contamination control supplies Decontamination control supplies Protective clothing Dosimeters Radiological postings and barricades 27

APPENDIX B Table B-i Cross Reference IOEP Section to Planning Standards/Requirements/Criteria and Procedures Planning Planning NUREG-0654, IOEP Standard Requirement Section II Procedure Section (10 CFR (Appendix Evaluation 50.47)** E.IV)** Criteria ________________

5.0 (b)(1) A.1,2, 4, 7 A TBD 6.0 (b)(2) A.1,2, 4; C.1 B TBD 7.0 (b)(3) A.6, 7 C TBD 8.0 (b)(4) B.1,2; C.1,2 D TBD

9. b) A.6, 7;C.1; D.1, ETBD 9.0__b__5 3E __ _ __E_ _ _ _ _ _ _ _ _ _ _ _ _

10.0 (b)(6) C.1; D.1,3; E F TBD 11.0 (b)(7) Exempt G TBD 12.0 (b)(8) E; G H TBD 13.0 (b)(9) A.4; B.1; C.2; E I____ TBD 14.0 (b)(10) 0.1; E J TBD 15.0 (b)(11) E K TBD 16.0 (b)(12) A.6, 7; E L TBD 17.0 (b)(13) H M TBD 18.0 (b)(14) E9; F N TBD 19.0 (b)(15) F 0 TBD 20.0 (b)(16) G P TBD

• • Refer to the Kewaunee Power Station's exemptions from portions of 10 CFR 50.47 and Appendix E for applicability

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Serial No.15-306 ENCLOSURE 2 LICENSE AMENDMENT REQUEST 259:

ISFSI-ONLY EMERGENCY PLAN AND EMERGENCY ACTION LEVEL SCHEME ISFSI-ONLY EMERGENCY ACTION LEVEL BASIS DOCUMENT (INCLUDES EMERGENCY ACTION LEVEL SCHEME)

KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Revision Summary:

ISFS I-Only Emergency Action Level Basis Document describes the Kewaunee Power Station's classification scheme for emergencies that may arise after all fuel is moved into the IFSFI.

Table of Contents Section Title Page

1. Purpose 1

2. Discussion 1 2.1 Independent Spent Fuel Storage Installation (ISFSI)

3. Key Terminology Used 2 3.1 Emergency Classification Level (ECL) 3.2 Initiating Condition (IC) 3.3 Emergency Action Level (EAL)

4. Guidance on Making Emergency Classification3 4.1 General Considerations 4.2 Classification Methodology 4.3 Classification of Multiple Events and Conditions 4.4 Classification of IMMINENT Conditions 4.5 Emergency Classification Level Upgrading and Downgrading 4.6 Classification of Short-Lived Events 4.7 Classification of Transient Conditions 4.8 After-the-Fact Discovery of an Emergency Event or Condition

__________4.9 Retraction of an Emergency Declaration

5. References 6 Appendix A, Initiating Condition Page Independent Spent Fuel Storage Installation (ISFSl)

EU1, Damage to a loaded cask CONFINEMENT BOUNDARY. 1 Hazards and Other Conditions PD-HU1, Confirmed SECURITY CONDITION or threat. 2 PD-HA1, HOSTILE ACTION within the VBS boundary. 3 PD-HU3, Other conditions exist which in the judgment of the Emergency Director warrant 4 declaration of a UE.

PD-HA3, Other conditions exist which in the judgment of the Emergency Director warrant 5 declaration of an Alert.

Appendix B, Definitions Appendix C, Acronyms and Abbreviations Appendix D, Kewaunee Power Station Emergency Action Levels

1. Purpose This document provides the detailed set of Emergency Action Levels (EALs) applicable to the Kewaunee Power Station (KPS) and the associated Technical Bases using the EAL development methodology found in NEI 99-01 Revision 6.

Personnel responsible for implementation of the Emergency Action Level Matrix may use this document as a technical reference and an aid in EAL implementation. The primary tool for determining the emergency classification level is the Emergency Action Level Matrix. The user of the Emergency Action Level Matrix may (but is not required to) consult the EAL Technical Basis Document in order to obtain additional information concerning the EALs under classification consideration.

2. Discussion 2.1 Independent Spent Fuel Storage Installation (ISFSI)

Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR § 50 emergency plan to fulfill the requirements of 10 CFR § 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFS1 are consistent with the requirements of 10 CFR § 50 and the guidance in NUREG 0654/FEMA-REP-1. The initiating conditions germane to a 10 CFR § 72.32 emergency plan (as described in NUREG-1 567) are subsumed within the classification scheme for a 10 CFR § 50.47 emergency plan.

The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISF5l is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparednessfor Fuel Cycle and Other Radioactive Material Licensees. NUREG-1 140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 rem Effective Dose Equivalent.

Regarding the above information, the expectations for an offsite response to an Alert classified under a 10 CFR § 72.32 emergency plan are generally consistent with those for a Notification of Unusual Event in a 10 CFR § 50.47 emergency plan (e.g., to provide assistance if requested).

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3. Key Terminology Used There are several key terms that appear throughout the NEI 99-01 methodology. These terms are introduced in this section to support understanding of subsequent material. As an aid to the reader, the following table is provided as an overview to illustrate the relationship of the terms to each other.

Emergency Classification Level Unusual Event Alert qJ 4' Initiating Condition ] Initiating Condition (1) - When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition. This includes the Emergency Action Level (EAL), Notes, and the informing Basis information.

3.1 Emergency Classification Level (ECL)

One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

• Unusual Event

• Alert 3.1.1 Unusual Event Events are in progress or have occurred which indicate a potential degradation of the level of safety of the ISFSI or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation occurs.

Purpose:

The purpose of this classification is to assure that the first step in future response has been carried out, to bring the staff to a state of readiness, and to provide systematic handling of unusual event information and decision-making.

3.1.2 Alert Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the ISFSI or a security event that involves probable life threatening risk to station personnel or damage to ISFSI equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

2

Purpose:

The purpose readily available of this classification is to assure that emergency personnel are to respond ifthe situation becomes more serious or to perform confirmatory radiation monitoring if required, and provide offsite authorities current information on status and parameters.

3.2 Initiating Condition (IC)

An event or condition that aligns with the definition of one of the two emergency classification levels by virtue of the potential or actual effects or consequences.

Discussion: An IC describes an event or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed as a measurable parameter (e.g., radiation monitor readings) or an event (e.g., a HOSTILE ACTION).

Appendix 1 of NUREG-0654 does not contain example Emergency Action Levels (EALs) for each ECL, but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency, or events that could lead to a radiological emergency, has occurred). NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific instrumentation readings (as applicable) which, if exceeded, would initiate the emergency classification. Thus, it is the specific instrument readings that would be the EALs.

3.3 Emergency Action Level (EAL)

A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the station in a given emergency classification level.

Discussion: EAL statements may utilize a variety of criteria including instrument readings and status indications; observable events; results of calculations and analyses; entry into particular procedures; and the occurrence of natural phenomena.

4. Guidance on Making Emergency Classification 4.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus Notes and the informing Basis information.

All emergency classification assessments should be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by personnel. The validation of indications should be completed in a manner that supports timely emergency declaration.

A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that 1) the activity proceeds as planned and 2) the station remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the 3

activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10 § CER 50.72.

The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded; the EAL and/or the associated basis discussion will identify the necessary analysis. In these cases, the declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available).

While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to PD-HU3 and PD-HA3). The Emergency Director will need to determine ifthe effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition.

4.2 Classification Methodology To make an emergency classification, the user will compare an event or condition to an EAL(s) and determine if the EAL has been met or exceeded. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with procedures.

4.3 Classification of Multiple Events and Conditions When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:

• If an Unusual Event EAL and an Alert EAL are met, an Alert should be declared.

There is no "additive" effect from multiple EALs meeting the same ECL. For example:

• If two Alert EALs are met, an Alert should be declared.

Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarificationof NRC Guidance for Emergency Notifications During Quickly Changing Events.

4.4 Classification of IMMINENT Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met.

While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification level since it provides additional time for implementation of protective measures.

4

4.5 Emergency Classification Level Upgrading and Downgrading An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate, the new ECL would then be based on a lower applicable IC(s) and EAL(s).

The ECL may also simply be terminated.

The following approach to downgrading or terminating an ECL is recommended.

ECL Action When Condition No Longer Exists Unusual Event Terminate the emergency in accordance with

______________________ procedures.

Alert Downgrade or terminate the emergency in

______________________accordance with procedures.

As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02.

4.6 Classification of Short-Lived Events Event-based l~s and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration (example is a Security event).

4.7 Classification of Transient Conditions It is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.

EAL momentarily met duringq expected station response - In instances where an EAL is briefly met during an expected (normal) response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.

EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required.

It is important to stress that the emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event; emergency classification assessments must be deliberate and timely, with no undue delays.

4.8 After-the-Fact Discovery of an Emergency Event or Condition In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer 5

exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.

In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1 022 is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.

4.9 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1 022.

5. References 5.1 NEI 99-0 1 Rev. 6 Final, Development of Emergency Action Levels for Non-Passive Reactors, November 2012 5.2 10 CFR § 50, Domestic Licensing of Production and Utilization Facilities 5.3 RIS 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007 5.4 NUREG-1 022, Event Reporting Guidelines: 10CFR50.72 and 50.73 5.5 10 CFR § 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors 5.6 10 CFR 50.82, Termination of License 5.7 NUREG-0654/FEMA-REP-1, REV 1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants 5.8 10 CFR § 72.32, Emergency Plan 5.9 NUREG-1567, Spent Fuel Dry Storage Facilities 5.10 10 CFR § 50.47, Emergency Plans 5.11 NUREG-1 140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees 6

Appendix A Independent Spent Fuel Storage Installation ECL: Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY.

Emergency Action Levels: EUl .1 EU1 .1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading greater than two times the ISFSI storage system Technical Specification allowable levels Basis:

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times" is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSIs are covered under l~s PD-HU1 and PD-HAI.

KPS Basis Reference(s):

1. ISESI Storage System Certificates of Compliance, Final Safety Evaluation Reports and Technical Specifications 1

Appendix A Hazards and Other Conditions PD-HUl1 ECL: Unusual Event Initiating Condition: Confirmed SECURITY CONDITION or threat.

Emergency Action Levels: PD-HU1 .1 or PD-HU1 .2 PD-HUI .1 A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by Security Supervision.

PD-HU1 .2 Notification of a credible security threat directed at the site.

Basis:

This IC addresses events that pose a threat to station personnel or spent fuel, and thus represent a potential degradation in the level of safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73.71 or 10 CFR § 50.72. Security events inside of the VEHICLE BARRIER SYSTEM (VBS) boundary that are assessed as HOSTILE ACTIONS are classifiable under IC PD-HAI.

Timely and accurate communications between Security Supervision and the Emergency Director is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to station personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and QualificationPlan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

PD-HU1 .1 references Security Supervision because these are the individuals trained to confirm that a security event is occurring or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.

PD-HU1 .2 addresses the receipt of a credible security threat. The procedure to determine the credibility of a threat is considered security-sensitive information and therefore withheld from the EAL. Credible security threat includes a HOSTILE ACTION within the SITE BOUNDARY outside of the VBS boundary.

Escalation of the emergency classification level would be via IC PD-HAl.

KPS Basis

Reference:

1. Security and Safeguards Contingency Plan 2

Appendix A Hazards and Other Conditions ECL: Alert Initiating Condition: HOSTILE ACTION within the VBS boundary Emergency Action Levels: PD-HAl .1 PD-HAlI.1 A HOSTILE ACTION is occurring or has occurred within the VBS boundary as reported by Security Supervision.

Basis:

This IC addresses the occurrence of a HOSTILE ACTION within the VEHICLE BARRIER SYSTEM (VBS) boundary. This event will require rapid response and assistance due to the possibility of the attack progressing to the ISFSI PROTECTED AREA (PA).

Timely and accurate communications between Security Supervision and the Emergency Director is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering).

The Alert declaration will also heighten the awareness of Offsite Response Organizations, allowing them to be better prepared should it be necessary to consider further actions.

This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc.

Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

PD-HAl1.1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the VBS boundary. This includes any action directed against the ISFSI that is located within the VBS boundary. A HOSTILE ACTION within the SITE BOUNDARY outside of the VBS boundary is considered a credible security threat and should be evaluated under PD-HUI.

KPS Basis

Reference:

1. Security and Safeguards Contingency Plan 3

Appendix A Hazards and Other Conditions IPD-HU3l ECL: Unusual Event Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a UE.

Emergency Action Levels: PD-H U3.1 PD-HU3.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the ISESI or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation occurs.

Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a UE.

KPS Basis

Reference:

None 4

Appendix A Hazards and Other Conditions IPD-HA31 ECL: Alert Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert.

Emergency Action Levels: PD-HA3.1 PD-HA3.1 Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the ISESI or a security event that involves probable life threatening risk to site personnel or damage to ISFSI equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Basis:

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert.

KPS Basis

Reference:

None 5

Appendix B Definitions The following are key terms necessary for overall understanding the NEI 99-01 emergency classification scheme.

Emergqency Action Level (EAL') - A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the station in a given emergency classification level.

Emergqency Classification Level (ECL) - One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The applicable emergency classification levels, in ascending order of severity, are:

• Unusual Event

• Alert Initiatingq Condition (IC) - An event or condition that aligns with the definition of one of the two emergency classification levels by virtue of the potential or actual effects or consequences.

Selected terms used in Initiating Condition, Emergency Action Level, Notes or Basis section are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.

CONFINEMENT BOUNDARY - The barrier(s) between areas containing radioactive substances and the environment.

HOSTAGE. - A person(s) held as leverage against the station to ensure that demands will be met by the station.

HOSTILE ACTION - An act toward the KPS ISFSI or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the KPS ISESI. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the SITE BOUNDARY).

HOSTILE FORCE - One or more individuals who are engaged in a determined assault, overtLy or by stealth and deception, equipped with suitable weapons capabte of kitting, maiming, or causing destruction.

IMMINENT. - The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.

1

INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) - A complex designed and constructed for the interim storage of spent nuclear fuel, solid reactor-related Greater Than Class C (GTCC) waste, and other radioactive materials associated with spent fuel and reactor-related GTCC waste storage (10 CFR 72.3).

PROJECTILE - An object directed toward the ISFSI that could cause concern for its continued operability, reliability, or personnel safety.

PROTECTED AREA (PA) - The area encompassed by physical barriers and to which access is controlled.

SECURITY CONDITION - Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety. A SECURITY CONDITION does not involve a HOSTILE ACTION.

SITE BOUNDARY - The perimeter of the land owned by Dominion Energy Kewaunee Inc. The ISFSI Controlled Area, as defined in 10 CFR 72.3, is bounded within the Site Boundary.

VEHICLE BARRIER SYSTEM (VBS) - A barrier system that is designed, constructed, installed and maintained to protect the facility against the design basis threat.

2

Appendix C Acronyms and Abbreviations CFR Code of Federal Regulations EAL Emergency Action Level ECL Emergency Classification Level EPA Environmental Protection Agency FEMA Federal Emergency Management Agency GTCC Greater Than Class C HSM Horizontal Storage Module IC Initiating Condition ISFSI Independent Spent Fuel Storage Installation NEI Nuclear Energy Institute NRC Nuclear Regulatory Commission ORO Offsite Response Organization PA Protected Area UE Unusual Event VBS Vehicle Barrier System VCC Vertical Concrete Cask 1

Appendix D Kewaunee Power Station Emergency Action Levels None EUI .1 Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by a radiation reading greater than two times the ISFSI storage system Technical Specification allowable levels.

HOSTILE ACTION within the VBS boundary. Confirmed SECURITY CONDITION or threat.

PD-HAI .1 A HOSTILE ACTION is occurring or has occurred within PD-HUI.l A SECURITY CONDITION that does not involve a HOSTILE the VBS boundary as reported by Security Supervision. ACTION as reported by Security Supervision.

PD-HU1.2 Notification of a credible security threat directed at the site.

Other conditions exist which in the judgment of the Emergency Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert. Director warrant declaration of a UE.

Hazards PD-HA3.1 Other conditions exist which in the judgment of the PD-HU3.1 Other conditions exist which in the judgment of the Emergency Emergency Director indicate that events are in progress Director indicate that events are in progress or have occurred or have occurred which involve an actual or potential which indicate a potential degradation of the level of safety of substantial degradation of the level of safety of the ISFSI the ISFSI or indicate a security threat to facility protection has or a security event that involves probable life threatening been initiated. No releases of radioactive material requiring risk to site personnel or damage to ISFSI equipment offsite response or monitoring are expected unless further because of HOSTILE ACTION. Any releases are degradation occurs.

expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

REV 0. Draft D

Serial No.15-306 ENCLOSURE 3 LICENSE AMENDMENT REQUEST 259:

ISFSI-ONLY EMERGENCY PLAN AND EMERGENCY ACTION LEVEL SCHEME SUPPORTING EVALUATIONS AND CALCULATIONS:

1. CALCULATION RA-0065, KEWAUNEE POWER STATION (KPS) NUCLIDE RADIOACTIVITY LIMIT METHODOLOGY TO PRECLUDE RELEASES FROM EXCEEDING TWICE ODCM DOSE LIMIT KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

SCalculation Cover Sheet Paael1 of 49

~Dominion Calculation: RA-0065 Revision: 0 Addendum: N/A Complete the fields with text or an X as required.

Calculation Number: Revision: Addendum: Sub type: Decommissioning Record?

R-050NIA 000 []Yes ElNo Vendor (Ifnot Dominion): Calculation Preparation Risk:

NIA [*Low [-]Medium i--High Vendor Proprietary: El--Yes [S*No Calculation Quality Class: E-lSafety Related [*NSQ E-Non-Safety Related.

Subject (Calculation Title):

Kewaunee Power Station (KPS) Nuclide Radioactivity Limit Methodology to Preclude Releases from Exceeding Twice ODCM Dose Rate Limit Addendum

Title:

NIA Station(s) and Unit(s):

NA [-11 E--]2 [-13 F-ISFSI KW [] r-ISFSI SU [-1 r-]2 [-]ISFSI MP [-11 [-12 [-13 [-1ISFSI Affected System(s), Structure(s), or Component(s):

NIA Purpose (Executive Summary):

Purpose:

Derive and demonstrate a dose rate based nuclide radioactivity limit methodology for a single release that equals two times Offsite Dose Calculation Manual (ODCM) instantaneous dose rate limit (2 x 1,500 mremlyr = 3,000 mremlyr) at the site boundary. Kewaunee Power Station (KPS) will develop and implement administrative controls to limit the radioactivity in shipment containers and other sources of radioactivity (filters, ion exchangers, etc.) to preclude releases from exceeding twice the KPS ODCM instantaneous dose rate limit. This is in an effort to eliminate the need for Emergency Action Levels (EALs) during the SAFSTOR dormancy period, active decontamination and dismantlement.

ResultslConclusions:

Table 6 demonstrates the nuclide radioactivity limit methodology defined in this calculation. Computer derived source terms from NUREGICR-3474 for stainless steel and concrete were used for the demonstration and can be used for initial planning for decommissioning; however, actual stainless steel and concrete samples need to be analyzed prior to undertaking significant work efforts so as to best reflect the appropriate nuclide activity limits for that particular time post-shutdown.

Dale M Flick Reviewer (Qual. Required): Printed Nai Douglas L Gilliatt Approver: Printed Name John R Harrell Note: Physical or electronic signatures are acceptable. Iy Note: (1) Add lines for additional originators or reviewers as necessary. (2) Note if reviews are "Independent', 'Peer", "Subject Matter Expert", =Supervisor" or "Owner's". (3) Enter N/A for Owner's Review of Vendor Calculation.

CM-AA-CLC-301 Rev 9, Attachment 3 38 (Sep 731189 (e204 2014)

• *Calculation Worksheet Pae2of 49 i*JDr O il1nliBn Calculation: RA-0065 Revision: 0 Addendum: N/A Table of Contents

1. Record of Revision and Addenda..................................................................... 3

2. Cumulative Effects Review (Required for Revisions and Addenda) ............................. 3

3. References............................................................................................... 3

4. Computer Codes Used................................................................................. 3

5. Identification of Computer Inputs and Outputs..................................................... 4

6. Purpose (Optional) ..................................................................................... 4

7. Background (Optional)................................................................................. 4

8. Design Inputs............................................................................................ 4

9. Assumptions ............................................................................................ 4

10. Methodology............................................................................................. 4

11. Calculations ............................................................................................. 6

12. Acceptance Criteria (Optional) ...................................................................... 18

13. Results and/or Conclusions ......................................................................... 18

14. Precautions and Limitations ......................................................................... 18

15. Recommendations (Optional)........................................................................ 18

16. Calculation Review Checklist........................................................................ 18

17. Attachments............................................................................................ 18 CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292 (Sep 2014)

• *Calculation Worksheet Pae3of 49 D*ominitonl1 Calculation: RA-0065 Revision: 0 Addendum: N/A

1. Record of Revision and Addenda RA-O065-O-O (Rev 0, Add N/A) has no revisions or addenda.

2. Cumulative Effects Review (Required for Revisions and Addenda)

RA-0065-O-O (Rev 0, Add N/A) has no cumulative effects.

3. References

1. KW-MANUAL-000-ODCM "Dominion Energy Kewaunee, Inc. Kewaunee Power Station Offsite Dose Calculation Manual (00CM)", Rev 1_7 (09/25/2014). [ODCM Instantaneous dose rate &

DCF methodology, X/Q basis]

2. NUREG-0133 "Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants" (October 1978). [ODCM BCF methodology]

3. NUREG/BR-01 50 NRC RTM-96 "Response Technical Manual", Vol 1 Rev 4 (March 1996).

4. NUREG-0172 "Age-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake" (November 1977).

5. NUREG-1140 "A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees" Final Report (January 1988).

6. NUREG-1 940 "RASCAL 4: Description of Models and Methods" (December 2012).

7. NUREG/CR-3474 "Long-Lived Activation Products in Reactor Materials" (August 1984).

[Stainless steel and concrete source terms]

8. NUREG/CR-1276 "User's Manual for LADTAP II -- A Computer Program for Calculating Radiation Exposure to Man from Routine Release of Nuclear Reactor Liquid Effluents" (May 1980). [Ground dcf basis]

9. RG 1.109 Regulatory Guide 1.109 "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix r", Rev 1 (October 1977).

10. FGR 12 Federal Guidance Report 12 (EPA-402-R-93-081) "External Exposure to Radionuclides in Air, Water and Soil" with erratum (September 1993). [Halflives]

11. en.wikipedia.org/wikilisotopes of niobium. [Nb-92 & Nb-92m haiflives, Attachment A screen print].

12. www.webelements.comlniobiumlisotopes.html. [Nb-92 halflife = 3.7e7 yrs, Attachment A screen print].

13. "Kewaunee Power Station Meteorological and Atmospheric Dispersion Report" by Chesapeake Nuclear Services, Rev 1 (October 2011 ). [D/Q basis] Copy in INDEX with calculation.

14. "Evaluation of Kewaunee Power Station's Meteorological and Atmospheric Dispersion Report and 2011 Land Use Census" by KPS, Rev 3. Copy in INDEX with calculation.

15. NEI 99-01 "Development of Emergency Action Levels for Non-Passive Reactors", Rev 6 (November 2012).

4. Computer Codes Used None CM-AA-CLC-301 Rev 9, Attachment 4 729292 (Sep 2014)

Calculation Worksheet Pn Pa*e4 off449 SDomi nion-Calculation: RA-0065 Revision: 0 Addendum: N/A

5. Identification of Computer Inputs and Outputs None

6. Purpose (OPTIONAL)

See Calculation Cover Sheet

7. Background (Optional)

None

8. Design Inputs

1. KPS Permanent Shutdown Date/Time 05/07/2013 11:15 Attachment C eSoms Log Entry (Reactor Shutdown) 3

2. Site Boundary Atmospheric Dispersionl X/Q = 7.44e-7 sec/in ODCM (Ref 1) Table 2.3 [See Attachment B]

Deposition Factors D/Q = 6.2e-9 m2 (Ref 13) Table 10 [See Attachment B]

3. Nuclide Haiflives (except Nb-92) [See Tables 2A & 2B] FGR 12 (Ref 10), Table A. 1

4. New Ground ODCM Dose Conversion [See Attachment D Table D-3] RG 1.109 (Ref 9) Table E-6 or NUREG/CR-1276 (Ref Factor dcf0 , 8) DF Table 1

5. New Inhalation Dose Conversion Factor [See Attachment D Table D-2] NUREG-0 172 (Ref 4) Table 5-Table 8 dcfao,

6. Breathing Rates BRa [See Attachment D Table D-2] RG 1.109 Table E-5

7. Fire Release Fractions [See Table 3] NUREG-1940 (Ref 6), NUREG-1140 (Ref 5)

9. Assumptions tel Paramee~s hterw ixs[eablue Vuld 1]NRG/R37 BRf7 able5s

1. StanlresStee Shutdown Nuclide Mixes [See Table 1] NUREG/CR-3474(Rf7 Table 5.1*

3. Nb-92 Halflife 3.7e7 yrs Ref 12 [See Attachment A]

• Absolute magnitudes in NUREGICR-3474 Tables 5.1l and 5.4 are not directly applicable to this analysis because the nominal neutron flux used to irradiatethe materials would be a function of the reactor design and the power level; however, the spectra (relative isotopic concentrations)will be representative of PWR vessel internals and ex-vessel concrete.

10. Methodology Overview Based on NUREG-3474 (Ref 7) computer-generated decommissioning/dismantlement stainless steel and concrete sample analysis results, derive and demonstrate the methodology for determining container/filter maximum radioactivity limits to obtain twice ODCM (Ref 1) limits at the Site Boundary using the ODCM / NUREG-O0133 (Ref 2) instantaneous dose rate methodology. In planning!

preparation for actual decommissioning/dismantlement work, representative material samples must be taken and analyzed to provide source term characterization for use with the methodology in this calculation.

CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292(Sep 2014)

• • Calculation Worksheet Pa~e 5 of 49 S Dominion- Calculation: RA-0065 Revision: 0 Addendum: N/A Discussion NUREG-0 133 §5.2.1 is the basis for the instantaneous dose rate methodology used in the ODCM

§2.3.3 SITE BOUNDARY Dose Rate - Radioiodine and Particulates, i.e. the dose rate and dose conversion factor algorithms. The ODCM Normal Condition 13.2.1.b limits the dose rate to < 1500 mrem/yr to any organ for 1-131, 1-133, tritium and particulates with half-lives greater than 8 days.

The lowest EAL (AU1 / PD-AU 1 Unusual Event per NEI 99-01 (Ref 15)) limit is twice the ODCM limit of 1,500 mrem/yr or 3,000 mrem/yr. By ensuring that an instantaneous dose rate cannot exceed twice the ODCM through administrative controls, the current EALs are no longer necessary and can be eliminated.

ODCM Table 2.3 specifies the following for ODCM Normal Condition 13.2.1 .b:

Site Boundary (0.81 mi, NNW)

Pathways Inhalation, Ground Plane (hereafter called Ground)

X/Q 7.44e-7 (sec/rn3)

D/Q N/A (1/rn 2)

ODCM Table 2.3 specifies that the Ground pathway be included; however, ODCM §2.3.2 Equation 2.6 does not address the Ground pathway which requires a D/Q. Therefore, the following were used to accommodate the Ground pathway:

• NUREG-O0133 Ground dose rate (§5.2.1) and dose conversion factor (§5.2.1.2) algorithms

• Largest D/Q (6.2e-9 m2 ) from 2011 Chesapeake Nuclear Services Report (Ref 13) Table 10 The ODCM Equation 2.6 dose rate (DR) algorithm can be expanded to accommodate the Ground pathway as follows:

Total DR DR =DRG + DR1 (mrern/yr)

Ground DR DRG D/Q *.(Qi " DCF~1 ) (mremn/yr)

Inhalation DR DRI = X/Q Maxlao[*i(Qi " DCFlaoi)] (mnremn/yr) where: DCFGi = 8.766e9 . dcftai* (1 - e-;it)/2Li (m~rnremn/yrperlli/sec)

DCFiaot = le6 . BRa* dcfiaoi (mrem/yrperptCi/m3)

D/Q = Ground deposition factor from plume (1/mn2)

X/Q = Atmospheric plume dispersion factor (sec/rn3)

- ~ - -Q -- -=Release Rate for ithnu~clide (lpCi/sec~) __

BRa = Age-dependent Breathing Rate (m3/yr) [RG 1.109 (Ref 9) Table E-5]

dcfci = Ground Dose Conversion Factor for ith nuclide (mrem/hrperpCi/rn2) dcfiaoi = Inhalation age-organ Dose Conversion Factor for ith nuclide (rnremn/pCi)

.i = Decay Constant for ith nuclide (sec-1 ) [T1/2 from FGR-12 (Ref 10) Table A.))

t = 1 yr x (365.25 x 24 x 60 x 60 seclyr) = 3.15576e7 (sec)

The ODCM Inhalation age-organ Dose Conversion Factors DCFiaoicome from ODCM Tables 2.4-2.7 or need to be derived using Breathing Rates BRa from RG 1.109 Table E-5 and Inhalation age-organ Dose Conversion Factors dCf~aoi from NUREG-01 72 Table 5-Table 8.

The ODCM Ground Dose Conversion Factors DCFGI need to be derived using Ground Dose Conversion Factors dcfG, from NUREG/CR-1276 (Ref 8) DF Table 1.

Due to some discrepancies with halflives (T112) in NUREG-0 133, halflives need to be based on FGR 12 Table A. 1; however, a Nb-92 halflife of 3.7e7 yrs (Ref 12 Attachment A) should be used because it does not appear in FGR 12. (Note: Ai = Ln(2) / T112)

The Total Instantaneous Dose Rate DR must calculated using the equations previously provided for Inhalation DR1 and Ground DRG for each unique material-nuclide mix combination. Dividing the twice ODCM limit (3,000 mrem/yr) by the Total Instantaneous Dose Rate DR yields the inverse of the fraction of the limit (FinlvL) the input source term provides. Multiplying each input nuclide activity CM-AA-CLC-301 Rev 9, Attachment 4 729292 (Sep 2014)

SCalculation Worksheet Pa~e 6 of 49

~i~ Dominion Calculation: RA-0065 Revision: 0 Addendum: N/A (Q,) by FfnvL yields the input nuclide activities (QLI) that would yield an instantaneous dose rate equivalent to twice the ODCM limit, i.e. Qi FInV, QuI.

NUREG-1140 (Ref 5) states: "Aside from fires or accidents that lead to fires, UF6 releases, or criticality accidents, no other significant accidents were identified. Explosions were not seen to yield as large a release unless they were followed by a fire. Earthquakes also were not identified as leading to significant releases unless they were followed by a fire." [pp ill-iv]..... "The significant accidents were determined to be UF6 releases, fires, and criticality accidents. Aside from the special cases of UF6 releases and criticalities, the release fractions for fires were considered to be larger than the release fractions for other types of accidents. Thus, release fractions for fires, as described in Section 2.3.1.2, are used to determine the need for emergency preparedness." [p 9] Therefore, for a permanently shutdown plant, fire appears to be the worst case scenario for transporting radioactivity to the Site Boundary. The fire release fractions used in this calculation are consistent with NUREG-1 140 Table 13, NUREG/BR-0 150 (Ref 3) Tables F-2 & F-3 and NUREG-1940 (Ref 6)

Tables 3-11 & 3-12. To find the activity of each nuclide that must be present before a fire_(QFI), such that when burned meets twice the ODCM limit at the Site Boundary, divide QLI by FF1 = QFi, where FF1

=Fire Release Fraction. See Table 3 for Fire Release Fractions. So if a source term consists of QEi and there is a fire, then QFEI FFj = QLi and this source term will meet twice the ODCM limit at the Site Boundary.

Thus, QLI represents the source term released from a container as a result of a fire; whereas, QEi represents the source in the container prior to a fire:

QFi = QL / FF1 = Q"( FinvL / FF1 ) = (3000 / (DR" FFi ))

The following is a summary of the methodology:

1. Lacking actual radioactive sample results for stainless steel and concrete, use characteristic stainless steel and concrete source terms from NUREG/CR-3474 Table 5.1 and Table 5.4 (which are as of shutdown) to derive respective source terms that have been decayed for an appropriate time period post-shutdown for work being done. Use actual sample results prior to commencing work.

2. Using these decayed nuclide activities, determine the associated maximum dose rate (Ground +

Inhalation). Divide the two times ODCM limit (3,000 mrem/yr = 2 x 1,500 mrem/yr) by the Smaximum-dose rate to~yield the inverse of-the-fraction of the limit (FqnlvL) . ................. ....

3. Multiply each decayed nuclide activity by FinlvL to derive the nuclide activities required to meet the limit.

4. Divide each limiting nuclide activity by its fire release fraction to obtain the maximum container loading which if burned will cause the instantaneous dose rate to be twice the ODCM limit.

5. In order to be cautious, a limiting fire composite stainless and a limiting composite concrete nuclide activity mix will be derived using the minimum non-zero activity for each nuclide.

6. The activity limits are in units of Cl/hr. If these activities are released over one hour, then these values become the Container Activity (Ci) Limits. Note NEI 99-01 AU1/PD-AU1 specifies the release must be for 60 minutes or longer.

11. Calculations New DCFs Inhalation: Attachment D Table D-2 contains newly generated DCFiaoi values, as well as the Breathing Rates BRa and dcfiaoi values used in generating them. These newly generated Inhalation DCFiaoi values are in addition to those already in ODCM (Ref 1) Tables 2.4-2.7.

CM-AA-CLC-301 Rev 9, Attachment 4 729292 (Sep 2014)

SCalculation Worksheet Pane 7 of 49

~D ominion, Calculation: RA-0065 Revision: 0 Addendum: N/A Ground: Attachment D Table D-3 contains the newly generated Ground DCF0 i values, as well as the dCfGi values used in generating them. These newly generated Ground DCFGI values are new instantaneous dose rate dose conversion factors to be used with ODCM §2.3.2 dose rates, as opposed to the dose conversion factors in ODCM Table 2.15, which are used with ODCM §2.4.2 doses.

See Attachment D for newly-derived Inhalation and Ground ODCM DCFs.

Decayed Source Term Table I contains nuclides and computer-generated activities from NUREG/CR-3474 (Ref 7) Table 5.1 304 Stainless Steel (Shroud, Core Barrel, Thermal Pads, Vessel Cladding) and Table 5.4 BioShield Concrete (Inner Edge, 10 cm, 24 cm, 55 cm).

Table 2A contains Stainless Steel (Shroud, Core Barrel, Thermal Pads, Vessel Cladding) nuclides, halflives (yr), decay constants (sec1 ) and activities decayed for two different times from permanent shutdown (Now = 3.57 yrs (11/30/16) and 30 yrs) which are displayed below the main section of this table. FGR-12 (Ref 10) Table A.1 halflives (T112) and decay constants (A) derived from these halflives were used in this table and calculation.

Table 2B contains Concrete (Inner Edge, 10 cm, 24 cm, 55 cm) nuclides, halflives, decay constants and activities decayed for two different times from permanent shutdown (Now = 3.57 yrs (11/30/16) and 30 yrs) which are displayed below the main section of this table. The halflives and decay constants are the same as those in Table 2A.

Note that NUREG/CR-3474 Table 1.1 lists Nb-92m halflife of 2.7e+7 yrs; however, web based sources have a Nb-92 halflife of 3.47e+7 yrs (Ref 11, Attachment A) and days to fractions of seconds for Nb-92m. This calculation used Nb-92 (with a 3.7e7 yr halflife (Ref 12, Attachment A) instead of Nb-92m to be consistent with the halflife instead of the nuclide designation. This affects activities but has no effect on dose rate because Nb-92 or Nb-92m dose conversion factors (DCFs) are unavailable.

Nuclide Activity Limits Based on the aforementioned equations and inputs, Attachment E provides the Nuclide Activity Limits Summary for the contents of a container prior to a fire (Pre-Fire) and for what is released from a container duelto _a fire (Fire)._ This attachmenthas results for~theJ~lUREGLCLR-347+/-4StainlessSteel_

(Shroud, Core Barrel, Thermal Pads, Vessel Cladding) and Concrete (Inner Edge, 10 cm, 24 cm, 55 cm) scenarios for Now (11/30/2016) and 30 years. These two time periods shows the effects of two primary factors, decay and having to maintain the 3000 mrem/yr. Ergo, the need to ensure that any Nuclide Activity Limits used must be time and material representative.

Table 4A (Stainless Steel) and Table 4B (Concrete) show the various scenario Container Activity Limits based on results from Attachment E. In order to be cautious, the Minimum Nuclide Activity Limit for each nuclide across the four stainless steel scenarios was selected to use as the Stainless Steel Nuclide Activity Limits to preclude KPS performing numerous dose rate calculations to ensure the limit was not exceeded and similarly for the concrete scenarios. A minimum non-zero activity criterion was used because Eu-I152 for the Stainless Steel-Shroud scenario has a zero value which would erroneously drive its limit to zero. Note these limits were calculated as Ci/hr; however, the values can be considered as Ci because they were released over one hour.

The Stainless Steel Now and 30-yr and Concrete Now and 30-yr Nuclide Activity Limits were used as the input source terms to provide dose rate statistics used in Table 5A Stainless Steel Summary, Table 5B Concrete Summary and Table 6 ContainerNuclide Activity Limits Summary. Tables 5A and SB provides the Pre-Fire and Fire Nuclide Activity Limits, Fire Nuclide Activity Limit percentage, Ground and Inhalation overall and nuclide percentages. Also included below Table 5B are the total dose rates at the Site Boundary for the four material-time Nuclide Activity Limits to demonstrate their CM-AA-CLC-301 Rev 9, Attachment 4 729292 (Sop 2014)

• Calculation Worksheet Pae8of 49

'Dominion Calculation: RA-0065 Revision: 0 Addendum: N/A dose rates were less than or equal to 3,000 mrem/yr. Percentages were provided to easily identify the nuclides with the most significant activity and dose rate contributions. Table 6 provides the Nuclide Activity Limits for the four material-time Nuclide Activity Limits formatted for easier reading.

These NUREG/CR-3474 computer generated stainless steel and concrete scenarios were used to demonstrate this methodology. However, because this methodology is decay sensitive, these stainless steel and concrete nuclide activity limits must be recomputed for the specific periods of stainless steel and concrete decommissioning work. Recomputed stainless steel and concrete limits can be used for initial planning purposes but need to be updated as soon as feasible using actual stainless steel and concrete lab sample results. The specific work period lengths for a given set of nuclide activity limits will be dictated by the actual nuclide mixes and period start and end times, such that over the period length the limits are appropriate indicators for ensuring dose limits are not exceeded.

This calculation estimated the total activity to meet the twice ODCM dose rate limit in a single release. Administrative controls should consider things such as intermediate lower limits to provide defense in depth, individual shipment container nuclide activity limits, number of separate releases to be accommodated, etc.

CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292 (Sep 2014)

Calculation Worksheet Pa~e 9of 49 SDominion Calculation: RA-0065 Revision: 0 Addendum: N/A Table 1 NUREG/CR-3474 Shutdown Mixes Table 5.1304StainlessSteel Table 5.4 Concrete Nulie hrud Core Thermal Vessel Inner 0o 24c 55m Nde Shod Barrel Pads Cladding Edge 1c 4m 5c

____ (0C/

gm) ( 0/gm) (Cl/gm) (Cl/gm) (Ci/gm) (Ci/gm) (Ci/gm) (C'/gm )

H-3 1.00E-05 9.60E-06 2.20E-06 3.30E-07 5.00E-06 8.30E-06 4.40E-06 2.50E-07 C-I4 2.50E-05 3.00E-06 4 60E-07 6.60E-08 1.70E-09 2.80E-09 1i50E-09 8.30E-11 cl-36 5.10E-07 6.40E-08 1.00E-08 1.40E-09 9.80E-11 1.50E-10 7.90E-11 4.40E-12 Ar-39 1.50E-07 8.O0E-09 4.20E-10 6.00E-10 2,70E-08 1.20E-08 3.10E-09 1.60E-10 Ca-41 4.70E-09 5.60E-10 8.60E-11 1.20E-11 1.20E-08 1.90E-08 1.00E-08 5.60E-10 Mn-53 3.20E-09 2.10E-10 1.00E-11 6.90E-12 1,70E-14 7.40E-15 1.90E-15 1.00E-16 Mn-54 6.50E-03 3.60E-04 1.80E-05 1.20E-05 2.80E-08 1.20E-08 3.10E-09 1.70E-10 Fe-SS 2.10E-01 2.40E-02 3.70E-03 5.40E-04 2.90E-06 4.70E-06 2.50E-06 1.40E-07 NI-59 1.10E-04 1,90E-05 3.00E-06 4.30E-07 1.60E-11 2.60E-11 1.40E-11 7.90E-13 Co-S0 1.30E-01 1.40E-02 2.00E-03 3.30E-04 2,20E-07 3.60E-07 1.70E-07 9.20E-09 Ni-63 1.80E-02 2.40E-03 3.80E-04 5.40E-05 2.00E-09 3.30E-09 1.70E-09 9.80E-11 Zn-6S 6.4OE-04 4.50E-05 5.90E-06 9.90E-07 2.20E-08 2.50E-09 1.2E-08 6.40E-10 5e-79 6.10E-10 4.60E-11 4.70E-12 1.20E-12 2.30E-15 2.80E-15 1.40E-15 7.90E-17 Kr-81 7.60E-10 3.40E-12 3.50E-14 2.10E-15 1.10E-16 5.50E-17 7.60E-18 1.70E-20 Kr-S5 8.50E-07 1.90E-08 9.50E-10 1.30E-10 4.30E-11 4.20E-11 1.70E-11 9.40E-13 Sr-90 2.00E-06 5.00E-09 2.60E-09 3,60E-10 4.50E-11 7.30E-11 3.80E-11 2,10E-12 Nb-g2m 1.20E-12 6.50E-14 3.20E-15 2.00E-15 1,40E-18 1.20E-19 1.90E-19 1.00E-20 Zr-93 1.10E-10 3.90E-12 2.90E-13 8.60E-14 3.00E-14 2.90E-14 1.30E-14 7.00E-16 Mo-93 9.40E-07 3.90E-08 2.10E-09 1.10E-09 1.20E-12 5.60E-13 1.40E-13 4.40E-15 Nb-94 4.OOE-07 2.90E-08 3.00E-09 7.50E-10 7,50E-12 5.30E-12 2.OOE-12 9.40E-14 Tc-99 1.30E-07 8.20E-09 4.20E-10 2.40E-10 2.80E-13 1.40E-13 3.70E-14 1.30E-15 Ag-lOSm 1.00E-07 9.00E-09 1.10E-09 2.20E-10 3.60E-12 3.20E-12 1.40E-12 7.30E-14 Sn-l2lm 4.80E-09 3.40E-10 1.70E-11 1,10E-11 6.50E-13 5.20E-13 1.80E-13 1.20E-14 I-129 6.00E-13 1.40E-14 7.90E-16 1.10E-16 1.40E-17 2.20E-17 1.20E-17 6.60E-19 Sa-133 3.00E-06 3.00E-07 4.50E-08 7.10E-09 1.30E-O9 2.00E-09 1.00E-09 5.70E-h1 Cs-134 7.O0E-06 1.00E-06 1.70E-07 2.30E-08 8.30E-09 1.60E-08 8.70E-09 4.90E-10 Cs-135 4.00E-11 9.00E-13 5.10E-14 7.80E-15 9.50E-16 1.40E-15 7.50E-16 4.20E-17 Cs-137 2.00E-06 S.OOE-08 2.70E-09 4.20E-10 5.1OE-11 7.60E-11 4.00E-11 2.20E-12 Pm-145 8.90E-10 1.40E-10 2.20E-11 3.10E-12 5.90E-12 9.80E-12 5.20E-12 2.90E-13 Sm-146 1.00E-16 1.30E-17 6.70E-19 4.30E-19 3.90E-19 1.70E-l9 4.40E-20 2.30E-21 Sm-151 4.60E-09 4.50E-09 2.90E-09 5.90E-10 1.40E-09 l.40E-09 6.10E-10 3.10E-11 Eu-1S2 0.00E+00 9.10E-10 1.70E-07 7.60E-08 2.40E-07 3.90E-07 2.10E-07 1.20E-08 Eu-154 5.60E-07 6.00E-07 6.20E-08 1.20E-08 5.60E-08 4.80E-08 2.00E-08 1.00E-09 Eu-155 4.10E-07 1.40E-07 S.00E-09 5.20E-10 1.70E-09 1.30E-09 5.00E-10 2.40E-11 TI-1S8 1,90E-09 1.70E-10 8.50E-12 5,60E-12 2.20E-13 9.60E-14 2.50E-14 1,30E-15 Ho-166m 1.60E-07 1.30E-08 1.10E-09 3.00E-10 3.30E-11 3.90E-11 1.30E-11 5.60E-13 Hf-175m 4.30E-08 2.80E-08 3.10E-09 1.20E-09 3.40E-10 1.B0E-10 5.30E-11 2.10E-12 pb-205 1.80E-12 1.30E-13 1.30E-14 3.40E-15 2.80E-16 2,70E-17 1.20E-16 6.60E-18 U.-233 3.60E-10 1.00E-10 1.10E-11 3.4OE-12 2.80E-12 1.20E-11 5.70E-13 2.50E-14 PU-239 7.OOE-09 2.30E-09 1.40E-10 9.60E-11 4.OOE-11 1.70E-11 3,70E-12 9.30E-14 Total 3.65E-01 4.08E-02 6.1E-03 9.38E-04 8.52E-06 1.39E-05 7.34E-06 4.15E-07 CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292 (Sep 2014)

Calculation Worksheet Page 10 of 49 SDominion-Calculation: RA-0065 Revision: 0 Addendum: N/A Table 2ZA Decayed Stainless Steel Mixes 143 1.2350E+01 1.7785E-O9 8.19E-06 7.86E-06 1.80E-06 2.70E-07 1.86E-06 1.78E-06 4.08E-07 6.13E-08 C-A 5.7300E+03 3.8332E-12 2.50E-05 3.00E-06 4.60E-07 6.60E-08 2.49E-05 2.99E-06 4.58E-07 6.SBE-08 Q# 3.0100E+05 7.2972E-14 5.I0E-07 6.40E-08 1.00E-08 1. 40E-09 5.1 0E-07 6.40E-08 1.00E-08 1.40E-09 A-9 2.6900E+02 8.1652E-11 1.49E-07 7.93E-09 4,16E-10 5.95E-10 1.39E-07 7.40E-09 3.89E-10 5.55E-10 cs4.1.4000E+05 1.5689E-13 4.70E-09 5.60E-10 8.60E-11 1.20E-11 4.70E-09 5.60E-10 8.60E-11 1.20E-11 M-3 3.7000E+06 5.9364E-15 3.20E-09 2.10E-40 1.0OE-11 6.90E-12 3.20E-09 2.10E-10 1.00E-11 6.90E-12 M-4 8,SSS8E-01 2.5672E-08 3.62E-04 2.006-05 1.00E-06 6.686-07 1.81 E-13 1.006-14 5.01E-16 3.346-16 F-S 2.7000E+00 8.1350E-09 8.416-02 9.61E-03 1.486-03 2.16E-04 9.49E-05 1.09E-05 1.67E-06 2.446-07 N-I 7.50O0E+04 2.9286E-13 1.10E-04 1.90E-05 3.006-06 4.30E-07 1.106-04 1.906-05 3.00E-06 4.306-07 coO 5.2710E+00 4.1670E-09 8.13E-02 8.766-03 1.25E-03 2.06E-04 2.526-03 2.71E-04 3.87E-05 6.396-06 N13 9.6000E+01 2.2880E-10 1.756-02 2.34E6-03 3.70E-04 5.266-05 1.45E-02 1.936-03 3.06E-04 4.35E-05 ZaS 6.6776E-01 3.2893E-08 1.58E-05 1.I1E-06 1.466-07 2.446-08 1.91E6-17 1.356-18 1.766-19 2.96E-20 e-S 6.5000E+04 3.3792E-13 6.106-10 4.60E-11 4.706-12 1.206-12 6.106-10 4.60E-11 4.706-12 1.206-12 r-I 2.1000E+05 1.0459E-13 7.606-10 3.40E-12 3.506-14 2.106-15 7.60E-1O 3.406-12 3.50E-14 2.10E-15 K-S 1.0720E+01 2.0489E-09 6.75E-07 1i5lE-O8 7.546-10 1.03E6-10 1.226-07 2.736-09 1.376-10 1.87E-11 S-D 2.9120E+01 7.5428E-10 1.846-06 4.596-09 2.396-09 3.316-10 9.796-07 2.456-09 1.276-09 1.76E-10 Nb9m3.7000E+07 5.9364E-16 1.206-12 6.506-14 3.206-15 2.00E-15 1.20E-12 6.50E-14 3.20E-l5 2.006-15 r-I 1.5300E+06 1.4356E-14 1.106-10 3.906-12 2.906-13 8.60E-14 1.106-10 3.906-12 2.906-13 8.60E-14 td-33.5000E+03 6.2756E-12 9.396-07 3.906-08 2.106-09 1.106-09 9.346-07 3.88E-08 2.096-09 1.09E-09 b-4 2.0300E+04 1.0820E-12 4.006-07 2.906-08 3.006-09 7.506-10 4.006-07 2.90E-08 3.O0E-09 7.496-10

1. -S 2.1300E+0S 1.0312E-13 1.306-07 8.206-09 4.206-10 2.406-10 1.30E-07 8.20E-09 4.206-10 2.406-10 AK408n 1.2700E+02 1.7295E-10 9.816-08 8.836-09 1.08E-09 2.16E-10 8.49E-08 7Z64E-09 9.34E-10 1.87E-10 Se-him S.5(00OE+01 3.9935E-10 4.596-09 3.256-10 1.636-11 1.056-11 3.296-09 2.336-10 1.16E-11 7.54E-12 li 1.5700E+07 1.3990E-15 6.006-13 1.40E-14 7.90E-16 1.106-16 6.006-13 1.406-14 7.90E-16 1.10E-16 fa151.0740E+01 2.0451E-09 2.38E-06 2.38E-07 3.57E-08 5.64E-09 4.33E-07 4.33E-08 6.49E-09 1.02E-09 c44 2.0620E+00 1.0652E-08 2.11E-06 3.02E-07 5.136-08 6.946-09 2.92 E-10 4.176-11 7.096-12 9.60E-13 Cs,43S 2.3000E+06 9.5498E-15 4.006-11 9.006-13 5.106-14 7.806-15 4.006-11 9.006-13 5.10E-14 7.80E-15 C47 3.0000E+O1 7.3215E-10 1.846-06 4.606-08 2.496-09 3.876-10 1.006-06 2.50E-08 1.356-09 2.IOE-10 Pm131.7700E+01 1.2409E-09 7746-10 1.226-10 1.91E6-11 2.70E-12 2.75E-10 4.32E-11 6,80E-12 9.58E-13 Su161.0300E+08 2.1325E-16 1.00E-16 1.306-17 6.706-19 4.306-19 1.006-16 1.306-17 6.70E-19 4.30E-19 5419.0000E+O1 2.4405E-10 4.486-09 4.38E-09 2.82E-09 5. 74E-10 3.65E-09 3.5 7E-09 2.30E-09 4.68E-10 Eu121.3330E+01 1.6478E-O9 O.00E+O0 7.56E-10 1.416-07 6.31E-08 0.00E+00 1.91E-10 3.57E-08 1.60E-08 iuiS8.8000E+00 2.4960E-09 4.23E-07 4.53E-07 4.68E-08 9.06E-09 5.27E-08 5.65E-08 5.84E-09 1.136-09 E-.S4.9600E+00 4.4283E-09 2.496-07 8.516-08 3.046-09 3.166-10 6.19E-09 2.126-09 7.556-11 7.866-12 ibiS 1.5000E,02 1.464E-1.O 1,876-09 1.67E-10 8.36-12 5.516-12 1.656-09 1.46-10 74*06-12 4.88-12 14-65 .2000E+03 1.80E-11 1.606-07 1.30E-08 1.106-09 2.99-10 1.576-07 1.286-08 1.086-0 2.956-10 H-7m3.1000E+01 7.0853E-10 3.97E-08 2.59E-08 2.86E-09 1.11E-09 2.20E-08 1.43E-08 1.59E-09 6.14E-10 Pb2S1,4300E+07 1.5360E-15 1.80E-12 1.306-13 1.306-14 3.406-15 1.806-12 1.30E-13 1.306-14 3.40E-15 U-23S 1.5850E+05 1,3858E-13 3.60E-10 1.OOE-10 1.10E-11 3.40E-12 3.606-10 1.006-10 1.106-11 3.406-12 PaZS2.4065E+04 9.1272E-13 7.006-09 2.30E-09 1.406-10 9.606-11 6.996-09 2.306-09 1.40E-10 9.596-11 Shutdowni~~i!

Dateii~iiiii 05/07/201 12:00 05/07/201 12:00

. at 130/2016 0:00

• /ry 05/08/2043 0:00 Ocylm .56 yrs 30.000 yrs CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292 (Sep 2014)

Calculation Worksheet Pe11of 49 SDominion-Calculation: RA-O065 Revision: 0 Addendum: N/A Table 28 Decayed Concrete Mixes iiii!i*!il iii*!iY*rsiiiiil 14 1.2350E+01 1.7785E-09 4.09E-06 6.79E-06 3.GOE-06 2.05E-07 9.28E-07 1.54E-O6 8.17E-07 4.64E-08 5730oE+0 s-4 3.832E-12 1.70E-09 2.80E-09 1.50E-09 8.30E-11 1.69E-09 279E-09 1.49E-09 8.27E-11 0-0 3.0100E+05 7.2972E-14 9.80E-11 I.50E-10 7.90E-11 4,40E-12 9.80E-11 1.50E-lO 7.90E-11 4.40E-12 A33 2.6900E+02 8.1652E-11 2.68E-08 1.19E-08 3.07E-09 1.59E-1O 2.50E-OB 1.11E-08 2.87E-09 1.48E-10 4S1 1.4000E+0S 1.S689E-13 1.20E-08 1.90E-08 1.OOE-08 5.60E-10 1.20E-08 1.90E-08 1.00E-08 5.GOE-10 M-3 3.7000E+06 5.9364E-15 1.70E-14 740E-15 1.90E-15 1.00E-16 1.70E-14 7.40E-15 1.90E-15 1.00E-16 MI-4 8.5558E-01 2.$672E-08 1.56E-09 6.68E-10 1.72E-10 9.46E-12 7.80E-19 3.34E-19 8.63E-20 4.73E-21

• 55 2.7000E+00 8.1350E-09 1.16E-06 1.88E-06 1.O0E-06 5.60E-08 1.31E-09 2.12E-09 1.13E-09 6.33E-11 N-$ 7.5000E+04 2.9286E-13 1.60E-11 2.60E-11 1.40E-11 7.90E-13 1.60E-11 2.60E-11 1.40E-11 7.90E-13 C9 5.2710E+00 4.1670E-09 1.38E-07 2.25E-07 1.06E-07 5.76E-09 4.26E-09 6.97E-09 3.29E-09 1.78E-10 t43 9.6000E+01 2,2880E-10 1.95E-09 3.22E-09 1.66E-09 9.55E-11 1.61E-09 2.66E-09 1.37E-09 7.89E-11 5 &000E+04 64 3.3792E-13 2.30E-15 2.80E-15 1.40E-15 7.90E-17 2.30E-15 2.80E-15 I.40E-15 7.90E-17 E~I 2.1000E+05 1.0459E-13 1.I0E-16 5.50E-17 7.60E-18 1.70E-20 1.10OE-16 5.50E-17 7.60E-18 1.70E-20 IrS 1.0720E+01 2.0489E-09 3.41E-11 3.34E-11 1.35E-1I 7.46E-13 6.18E-12 6.04E-12 2.44E-12 1.35E-13 SrO 2.9120E+01 7.5428E-10 4.13E-11 6.71 E-11 3.49E-11 1.93E-12 2.20E-11 3.57E-11 1.86E-11 1.03E-12 Nb492. 3.7000E+07 5.9364E-16 1.40E-18 1.20E-19 1.90E-19 1.00E-20 1.40E-18 I.20E-19 1.90E-19 1.OOE-20 Z93 1.5300E+06 1.4356E-14 3.00E-14 2.90E-14 1.30E-14 7.00E-16 3.OOE-14 2.90E-14 1.30E-14 7.OOE-16 Mc3 3.5000E+03 6.2756E-12 1.20E-12 5.60E-13 1.40E-13 4.40E-15 1.19E-12 5.57E-13 1.39E-13 4.37E-15 N-4 2.0300E+04 1.0820E-12 7.50E-12 5.30E-12 2.OOE-12 9.40E-14 7.49E-12 5.29E-12 2.00E-12 9.39E-14 T3 2.1300E+05 1.0312E-13 2.80E-13 1.40E-13 3.70E-14 1.30E-15 2.80E-13 1.40E-13 3. 70E-14 1.30E-15 AgOSM 1.2700E+02 1.7295E-10 3.53E-12 3.14E-12 1.3 7E-12 7.16E-14 3.OGE-12 2.72E-12 1.19E-12 6.20E-14 Sn-I2lu 5.5000E+01 3.9935E-10 6.21E-13 4.97Eo13 1.72E-13 1.15E-14 4.45E-13 3.56E-13 1.23E-13 8.22E-15 Ii 1.57O0E+07 1.3990E-15 1.40E-17 2.20E-17 1.20E-17 6.60E-19 1.40E-17 2.20E-17 1.20E-17 6.60E-19 S.33 1.0740E+O1 2.0451E-09 1.03E-09 1.59E-09 7.94E-10 4.53E-11 1I88E-I0 2.89E-1O 1.44E-10 8.22E-12

£3# 2.0620E+00 1.0652E-08 2.50E-09 4.83E-09 2.62E-09 1.48E-10 3.46E-13 6&67E-13 3.63E-13 2.04E-14 c35 2.3000E+06 9.5498E-15 9.50E-16 1.40E-15 7.50E-16 4.20E-17 9.50E-16 1.40E-15 7.50E-16 4.20E-17

.OOO+01

• 007 7.3215E-1O4.0Eno-1 7OOE-11 3.68-1 2.03E-12 2.55E-1 3.80E-11 2.O-1l1. 0E-12 P l 1.7700E+01 1.2409E-09 5.13E-12 8.52E-12 4.52E-12 2.52E-13 1.82E-12 3.03E-12 1.61E-12 8.96E-14 m461.0300E+08 2.1325E-16 3.90E-19 1.70E-19 4.40E-20 2.30E-21 3.90E-19 1.70E-19 4.40E-20 2.30E-21 Sn-S19.OO00E+01 2.4405E-10 1.36E-09 1,36E-09 5.93E-1O 3.02E-11 1.IIE-09 1.I1E-09 4.84E-1O 2.46E-11 i4 1.3330E+01 1.6478E-09 1.99E-07 3.24E-07 1.74E-07 9.9 7E-09 5.04E-08 8.20E-08 4.41E-O8 2.52E-09 u448.B000E+00 2.4960E-09 4.23E-O8 3.52E-08 1.51 E-08 7.55E-l0 5.27E-09 4.52E-09 1.88E-09 9.41E-11 E4S 4.9600E+00 4.4283E-O9 1.03E-09 7.9OE-1O 3.04E-10 1.46E-11 2.57E-11 1.96SE-11 7.55E-12 3.63E-13 lss 1.5(00OE+02 1.4643E-10 2.16E-13 9.44E-14 2.46E-14 1.28E-15 1.92E-13 8.36E-14 2.18E-14 1.13E-15 tH04* 1.2000E+03 1.8304E-11 3.29E-1I 3.89E-11 1.30E-11 5.59E-13 3.24E-11 3.83E-11 1.28E-11 5.50E-13 H$4!71m* 3.1000E+01 7.0853E-10 3.14E-10 1.66E-10 4.89E-1l l.94E-12 1.74 E-10 9.20E-11 2. 71 E-11 1.07E-12 P4S 1.4300E+07 1.5360E-15 2.80E-16 2.70E-17 1.20E-16 6.60E-18 2.80E-16 2.70E-17 1.20E-16 6.60E-18 U2 1.58SOE+05 1.3858E-13 2.80E-12 1.20E-11 5.70E-13 2.50E-14 2.80E-12 1.20E-11 5.70E-13 2.50E-14 Pu239 24065E+04 9,1272E-13 4.O0E-11 1.70E-1l 3.70E-12 9.30E-14 4.00E-11 1.70E-11 3.70E-12 9.29E-14 t F~~i!iii;i~i*!i* T 1 30 Yam * ! Ii~ii~ii~i~i i!~~~iiiiiiiiii~ii!i*'!i!!*!~iiiiii*!!!~

'Dyo'De 11/30/20160:00 05/08/2043 0:00 05/07/2013 12:00

$butdown Outnii~iiiiiiiiiiil 05/07/20132:00 CM-AA-CLC-301 Rev 9, Attachment 47292(p204 729292 (Sep 2014)

• • II Calculation Worksheet Pace 12 of 49 Do inion"l Calculation: RA-0065 Revision: 0 Addendum: N/A Table 3 Nudlide Fire Release Fraction H-3 0.5 o,14 0.01 Ca-3 0.5 1

At-39 1 Ca-4i 0.01 Mn-S3 0.01 Mn-54 0.01 Fe-55 0.01 2

Ni-59 0.01 co-60 0.001 Ni'-63 2 0.01 Zn-65 0.01 se-79 0.01 Kr-S1i 1 Kr-8S

• 1 Sr-S0 0.01 Nb-92n, 0.01 Zr-93 0.01 Mo-93 0.01 Nb-94 0.01 Tc-99 0.01 Ag-*lOm 0.01 Sn-121m 0.01 1-129 0.5 Ba-133 0.01 cs-13 0.01 cs-135 0.01 Cs-137 0.01 pm-14s 0.01 Sm-146 0.01 Sm-151 0.01 Eu-152 0.01 Eu-154 0.01 Eu-IS5 0.01 Tb-lS8 0.01 Ho-166m 0.01 Hf-178m 0.01 Pb-205 0.01 u-233 0.001 Pu-239 0.001 Source: NUREG-1 940 Table 3.12 Except:'J NUREG-1940 Table 3.11 2 NUREG-1140 Table 13 CM-AA-CLC-301 Rev 9, Attachment 4 729292 (Sep 2014)

Calculation Worksheet Page 13 of 49 SDominion-Calculation: RA-0065 Revision: 0 Addendum: N/A Table 4A Stainless Steel Container Activityo Limits*

3@Yazs i~3 6,40E-05 5.70E-04 9.15E-04 8.31E-04 6.401-05 0.0000% 4,54E-04 4.01E--03 6,39E-03 5.85E-03 4.541-04 o.0oooi C-A 9.76E-03 1.09E-02 1.17E-02 1.02E-02 9.761-03 0.0027% 3.04E-01 3.36E-01 3.59E-01 3.14E-01 3.041-01 0.0630%

• 4 3.98E-06 4,64E-06 5.08E-06 4.31E-06 3.981-06 0.o0000 1.251-04 1,44E-04 1.57E-04 1.34E-04 1.25E-04 0.0000o%

k- S .81E-07 2.88E-07 1.061-07 9.15E-07 1.061E-07 o.oooo% 1.70E-05 8.32E-06 3.04E-06 2.65E-05 3.04E-06 0.00oo0%

C.1 1.84E-06 2.03E-06 2.18E-06 1.85E-06 1.841-06 o.oooo% 5.741-05 6.29E-05 6.73E-05 5.731-05 5.73E-05 0.00o0%

Mn31.25E-06 7.62E-07 2.54E-07 1.06E-06 2.541-07 0.o00o% 3.911-05 2.361-05 7.83E-06 3.29E-05 7.83E-06 0.0000%

M-41,41E-01 7.26E-02 2.54E-02 1.03E-01 2.541-02 0.0071% 2.21E-09 1.13E-09 3.92E-10 1.59E-09 3.92E-10 0.00o00%

e-53.28E+01 3.48E+01 3.76E+01 3.33E+01 3.281401 9.1858%A 1.161+(00 1.22E+00 1.31E+00 1.16E+00 1.161400 0.2402%

1. 0!4.30E-02 6.89E-02 7.62E-02 6.621-02 4.301-02 0.0120% 1.34E+00 2.13E+00 2.351+(00 2.05E+-00 1.341400 0.2782%

C.03.18E+02 3,18E+02 3.181+02 3.18E+02 3.181402 88.87z33% 3.07E+02 3.04E+02 3.031+02 3.05E+02 3.0314O2 62.7441%;

1. 03 6.85E+00 8.49E+00 9.40E+00 8.10E+00 6.85E+040 1.9174% 1.77E+02 2.17E+02 2.40E+02 2.07E+02 1.77E+02 36.6714%

l-S6.16E-03 4.021-03 3.69E-03 3.76E-03 3.69E-03 0.0010"% 2,34E-13 1.51E-13 1.381-13 1.41E-13 1.381-13 0.o00oo 5-9 2.38E-07 1.67E-07 1.19E-07 1,85E-07 1.19E-07 0.0000% 7.45E-06 5.17E-06 3.681-06 5.72E-06 3.68E-06 o.o0ooo M-I2.97E-09 1.23E-10 8.89E-12 3.23E-12 3.231-12 o.oooo% 9.28E-08 3.821-09 2.74E-10 1.001-10 1.001-10 o.o~ooo%

%-5 2.64E-06 5.47E-07 1.92E-07 1.59E-07 1.591-07 0.0000% 1.49E-05 3.07E-06 1.07E-06 8.92E-07 8.92E-07 o.o~ooo%

Sr!7.18E-04 1.67E-05 6.06E-05 5.09E-05 1.67E-05 0.0000% 1.20E-02 2.75E-04 9.97E-04 8.411-04 2.7SE-04 0.0001%

1. b0m4.69E-10 2.36E-10 8.13E-11 3,08E-10 8.131-11 o.o~ooo 1.47E-08 7.31E-09 2.50E-09 9.541-09 2.501-09 0.00oo%

r-E 4.30E-08 1.41E-08 7.36E-09 1,32E-08 7.361-09 o.0000% 1.34E-06 4.38E-07 2.27E-07 4.10E-07 2.271-07 0.00ooo%

M-E3.67E-04 1 41E-04 5.33E-05 1.69E-04 S,33E-05 o.0oooo 1.14E-02 4.36E-03 1.631-03 5.22E-03 1.631-03 0.0o03%

N-E 1.56E-04 1.05E-04 7.62E-05 1.1SE-04 7.62E-05 o.oooo% 4.88E-03 3.26E-03 2.351-03 3.58E-03 2.35E-03 0.0oo5%

T- 5.08E-05 2.97E-05 1.071-05 3.69E-05 1.071-05 o.o~ooo% 1.$9E-03 9.22E-04 3.29E-04 1.15E-03 3.291-04 0.0001%

Ag*--O* 3.83E-05 3.20E-05 2.74E-05 3.32E-05 2.741-05 o.oooo% 1.04E-03 8.59E-04 7.31E-04 8.91E-04 7.31E-04 0.00o2%

S4* 1.79E-06 1.18E-06 4.13E-07 1.62E-06 4.131-07 o.oooo% 4.02E-05 2.62E-05 9.12E-06 3.60E-05 9.12E-06 0.0000%

-Zt4.69E-12 1.02E-12 4.011-13 3.39E-13 3.391-13 o.oooo% 1.47E-10 3.15E-11 1.24E-11 1.05E-11 1.05E-11 o.oooo%

• a139.31E-04 8.65E-04 9.08E-04 8.68E-04 8.65E-04 o0.002% 5.29E-03 4.86E-03 5.08E-03 4.89E-03 4.86E-03 0.0010%

C448.241-04 1.09E-03 1.30E-03 1.07E-03 8.241-04 0.0002% 3.57E-06 4.69E-06 5.551-06 4.58E-06 3.57E-06 0.0oo0o CelE 1.561-08 3.26E-09 1.29E-09 1.20E-09 1.201-09 o.oo~oo 4,89E-07 1.01E-07 3.99E-08 3.72E-08 3.72E-08 0.0000%

Cs-137 7.20E-04 1.67E-04 6.31E-05 5.951-05 5.951-05 o.0oooo% 1.22E-02 2.81E-03 1.06E-03 1.00E-03 1.001E-03 0.0002%

P-153.02E-07 4.42E-07 4.86E-07 4.15E-07 3.02E-07 o.o~ooo% 3.36E-06 4.86E-06 5.32E-06 4.57E-06 3.361E-06 0.0000%

Se-E .1-447E1 .0-466E1 .0-4 ooo%12E1 .6-2524-1 2.05-1 S.241-1 oooo Sdi1.75E-06 1.59E-05 7.16E-05 8.83E-05 1.75E-06 o.o~oo% 4.46E-05 4.01E-04 1.80E-03 2,23E-03 4.46E-05 0.0000%

Eu120.00E+00 2.74E-06 3,59E-03 9.71E-03 2.741-06oooo0.000%0OE+00 2.15E-05 2,80E-02 7.62E-02 2.15E-05 0.0000%

lE-I 1.65E-04 1.64E-03 1.19E-03 1.39E-03 1.65E-04 0.000o% 6.44E-04 6.35E-03 4.57E-03 5.39E-03 6.44E-04 0.0001%*

fu159.73E-05 3.09E-04 7.71E-05 4.86E-05 4.86E-05 o,oooo% 7.57E-05 2.38E-04 5.91E-05 3.751-05 3.751-05 0.0oo0%

Tbl87.30E-07 6.07E-07 2.12E-07 8.48E-07 2.12E-07 o.oooo%* 2.02E-05 1.66E-05 5.79E-06 2.33E-05 5.79E-06 o,oojo%

IHo46mn 6.24E-09 4.71E-05 2.79E-05 4.61E-05 2,79E-05 o.oooo% 1.92E-03 1.44E-03 8.46E-04 1.41E-03 8.46E-04 0.0002%

hf-hw 1.S5E-05 9.38E-05 7.27E-05 1.71E-04 1.55E-OS o.oooo% 2.69E-04 1.61E-03 1.24E-03 2.93E-03 2.69E-04 0.0001%

PblS7.03E-10 4.72E-10 3.30E-10 5.23E-10 3.301-10 o.0oooo% 2.20E-08 1.46E-08 1.02E-08 1.62E-08 1.02E-08 0.0o000%

1.431.41E-06 3.63E-06 2.79E-06 5.23E-06 1-.41E-06 o,oooo% 4.40E-05 1.12E-0,4 8.61E-05 1.62E-04 4.40E-05 0.0oooo Pa2,9273E-05 8.34E-05 3,55E-05 1.48E-04 2.73E-05 o,000o% 8.54E-04 2.58E-03 1.09E-03 4.58E-03 8.54E-04 0.0002%

• Container Activity needed to meet 3,000 mrem/yr instantaneous dose rate limit (twice OOCM)at Site Boundaryvia 1-hr fire release

• • Minimum is the 'non-zero' minimum because SS-SH Eu-152 has a value of zero CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292 (Sep 2014)

Calculation Worksheet Pa~e 14, of 49 SDom inion- Calculation: RA-0065 Revision: 0 Addendum: N/A Table 48 Concrete Container Activity Limits*

30~ae~

5i.52E.00 5.85E+O0 6.09E+OO 6.23E+00 5.52E'400 2.9457% 2.40E+00 2.53E+0O 2.56E+0O 2.59E+00 2.40E+00 14.6105%

04 1.15E-01 1.21E-01 1.27E-01 1.26E-01 1.15IE-01 0.o612% 2.19E-01 2.29E-01 2.34E-01 2.31E-01 2.29E-01 1.3329%

• 5 1.32E-04 1.29E-04 1.34E-04 1.3E-04 1.291-04 0.001%, 2.3E-04 2,45-04 2.47E-04 2.,*E-o4 2.,61.04 0.0015%

.80E-02 5.12E-03 2a. 2.60E-03 2.41E-03 2.41E-03 0.0013% 3.22E-02 9.13E-03 4.49E-03 4.14E-03 4.141E-03 0.0252%

Ca4 8.09-Ol 8.18-Ol 84,E-o1 &53E-o1 8.01-01 0.431,,%1.SE0 1.5oo 6Es+oo 1s.5+Oo 1s.5Eoo 1.5ss0 9.4413%,

M-I1.15E-06 3.19E-07 1.61E-07 1.52E-07 1.52E-07 0.0000% 2.19E-06 6.08E-07 2.97E-07 2.79E-07 2.791-07 0.0000%

t441.05E-01 2.87E-02 1.46E-02 1.44E-02 1.441E-02 0.0o77%* 1.01E-10 2.75E-11 1.35E-11 1.32E-11 1.32E-11 o.oooo%

5$37.83E+O1 8.10E+01 8.46E.O1 8.53E+01 7.83E+01 41.7640% 1.69E-02. 1.75E-01 1.77E-01 1.77E-01 1.69E-01 1.0317%

81-9 1.OSE-03 1.12E-03 1.18E-03 1.20E-03 .061E-03 0.0006% 2.06E-03 2.14E-03 2.19E-03 2.21E-03 2,061-03 0.0126%;

~Q09.28E+O1 9.70E.01 9.00E+0O1 8.76E*O1 8.761E+01 46.7595% 5.49E+00 5.72E+0O 5.15E+0O 4.97E+OO 4.971E400 30.3353%

814 .31E-01 1.38E-01 1.40-01 1.4sE-o1 1.3110 .0 70o~1,, 2OSE-0l 2.8-01 2.14E-01 2.21E-01 2.01-01 1.2573%

1n5 3.66E-02 2.65E-03 2.50E-02 2.40E-02 2.651-03 0.oo14% 8.49E-14 6,14E-1S 5.62E-14 5.35E-14 6.141-15 0.000o%

S.9 1.SSE-07 1.21E-07 1.18E-07 1.20E-07 1.1.81E-07 0.000o% 2.97E-07 2.30E-07 2.19E-07 2.21E-07 2.191-07 o.0000%

Ivl7.42E-11 2.37E-11 6.43E-12 2.59E-13 2.591-13 o.ooo0% 1.42E-10 4.52E-11 1.19E-11 4.75E-13 4.75E-13 0.o0o0%

1. -5 2.30E-05 1.44E-05 1.14E-05 1.14E-05 1.141-05 o.oooo% 7.98E-06 4.96E-06 3.83E-06 3.7BE-06 3.78E-06 0.0000%

S-O 2.79E-03 2.89E-03 2.95E-03 2.94E-03 2.79E-03 0.0015% 2.84E-03 2.94E-03 2.91E-03 2.87E-03 2.841-03 0.01 73%

-gm9.44E-11 5.17E-12 1.61E-11 1.52E-11 5.17E-12 o.oooo% 1.81E-10 9.86E-12 2.97E-11 2.79E-11 9.86E-12 0.0000%;

b-S 2.02E-06 1.25E-06 1.10E-06 1.07E-06 1.071E-06 0.0000% 3.87E-06 2.38E-06 2.04E-06 1.96E-06 1.961-06 0.00oo0%

M-I8.09E-05 2.41E-05 1.18E-0S 6.69E-06 6.691-06 o.oooo% 1.54E-04 4.57E-05 2.18E-05 1.22E-0S 1.22E-05 0~o1 b-45.06E-04 2.28E-04 1.69E-04 1.43E-04 1.431-04 0.0o01% 9.67E-04 4.35E-04 3.13E-04 2.62E-04 2.621-04 0.0016%

Y-S 1.89E-05 6.03E-06 3.13E-06 1.98E-06 1.98E-06 o~oooo% 3.61E-05 1.15E-05 5.79E-06 3.63E-06 3.63E-06 0.00oo%

A.ISe 2.38-04 13sE-0 1.16-04 109-04 1.091-04 oool, 3.94-04 2.23E-04 16-04 1.73E-04 1.3-04 0.0011,,

Sr12u 4.19E-O5 2.14E-05 1.46E-OS 1.75E-05 1.461-05 o, oooo,, 5.75E-05 2.93E-05 1.93E-05 2.30E-05 1.93E-05 0.0001%

145 1.89E-11 1.89E-11 2.03E-11 2.01E-11 1.891-11 o~oooo* 3.61E-11 3.62E-11 3.76E-11 3.69E-11 3.61E-11 0.000o%

9406.96E-02 6.84E-02 6.72E-02 6.89E-02 6.721-02 0.0359% 2.42E-02 2.37E-02 2.26E-02 2.30E-02 2.261-02 0.1377%

(i3 .9E-01 2.08-01 2.22E-01 2.2sE-01 2.69-01 0.0900% 4.47-05 5.,8-os s.68-o5 s.71-o 4.471-0 0.0oo3,,

a336.41E-08 6.03E-08 6.34E-08 6.39E-08 6.03E-08 0.0000% 1.23E-07 1.15E-07 1.17E-07 1.17E-07 1.15E-07 0.0000%

(sIl3.17E-03 3.01E-03 3.12E-03 3.08E-03 3.01E-03 0.0016% 3.29E-03 3.12E-03 3.13E-03 3.07E-03 3.071-03 0.0187%

P453,46E-04 3.67E-04 3.83E-04 3.84E-04 3.461-04 0.0002% 2.35E-04 2.49E-04 2.SIE-04 2.50E-04 2.351-04 0.0014%

au162.63E-11 7.32E-12 3.72E-12 3.50E-12 3.501-1.2 0.oooo% S.O3E-11 1.40E-11 6.89E-12 6.43E-12 6.43E-12 0.00oo%

SiiI9.19E-02 S.86E-02 5.02E-02 4.59E-02 4.591-02 0.0245% 1.43E-01 9.13E-02 7.58E-02 6.88E-02 6.88E-02 0.4193%

Eu121.34E+01 1.39E+01 1.48E+O1 1.$2E+01 2,341E+01 7.1745% 6.51E+00 6.73E+00 6.91E+00 7.0SE+00 6.S1E+00 39,*6o60 1442.85E+00 1.56E+0O0 1.28E+00 1.15E+00 1.15E+00 0.6135% 6.80E:01 3.71E-01 2.95E-01 2.63E-01 2.63E-01 1.6o41%

Eu156.96E-02 3.40E-02 2,57E-02 2.22E-02 2.22E-02 0.0118%*3.31E-03 1.61E-03 1.18E-03 1.O1E-03 1.01E-03 0.0062%

TbIa1,46E-OS 4.07E-06 2.08E-06 1.95E-06 1.95E-06 0.0000% 2.47E-05 6.87E-06 3.41E-06 3.16E-06 3.16E-06 0.oooo%

f*-1.* 2.22E-03 1.68E-03 1.10E-03 8.51E-04 8.S1E-04 O.ooos% 4.19E-03 3.1SE-03 2.00E-03 1.54E-03 1.541E-03 0.0094%

111478 2.12E-02 7.16E-03 4.14E-03 2.9SE-03 2.951-03 0.o016% 2.24E-02 7.561-03 4.24E-03 3.00E-03 3.001-03 0.0183%

Pb2 19-08

, 1.16-09 1.02E-08 100.E08 1.16-09 ooooo 31-08 2.22E-09 1.8-08 1.84-08 2.22-09 0.000 ii23 1.89E-03 5.17E-03 4.82E-04 3.81E-04 3.81E-04 0.o002% 3.61E-03 9.86E-03 8.92E-04 6.99E-04 6.99E-04 0.0043%

Pu252.70E-02 7.32E-03 3.13E-03 1.42E-03 1.42E-03 0.0008% 5.16E-02 1.40E-02 5.79E-03 2.60E-03 2.60E-03 0.0158%

• ContainerActivity needed to meet 3,O000mrem/yr instantaneous dose rate limit(twice ODCM)at Site Boundaryvia 1-hr fire release

• • Minimum is the 'non-zero' minimum because SS-SH Eu-IS2 has a value of zero CM-AA-CLC-301 Rev 9, Attachment 47292(e214 729292 (Sep 2014)

Calculation Worksheet Pace 15 of 49 SDominion Calculation: RA-0065 Revision: 0 Addendum: N/A Table 5A Stainless Steel Summary CJ J 80.47% 19.53%  :*;*:7js!:*,::~.*::' 10 1) 77.71% 22.29%

6.408-OS 3.208-05 0% 0% 0% 4.548-04 2.27E-04 0.0108% 0% 0%

S9.76E-03 9.76E-05 0.0136% 0% 0% 3.048-01 3.04E-03 0.1448% 0% 0%

S3.98E-06 1.99E-06 0% 0% 0% 1.25E-04 6.23E-05 O% 0% 0%

i1.068-07 1.06E-07 0% 0% 0% 3.04E-06 3.048-06 0% 0% 0%

S1.848-06 1.84E-08 0% 0% 0% 5.73E-OS 5.73E-07 0% 0% 0.0145%

i2.548-07 2..548-09 0% 0% 0% 7.83E-06 7.83E-08 0% 0% 0%

i2.54E-02 2.548-04 0.0355% 0.0199% 0.0178% 3.92E-10 3.92E-12 0% 0% 0%

S3.28E+01 3.28E-01 45.8944% 0% 1.4374% 1.16E+00 1.16E-02 0.5517% 0% 0.0451%

4.308E-02 4.30E-04 0.0601% 0% 0% 1.3148+00 1.34E-02 0.6391% 0% 0.0475%

3.138+02 3.18E-01 44.4033% 99.9762% 97.79S0% 3.03E+02 3.03E-01 14.4120% 99.9965% 82.7886%

S6.858+00 6,85E-02 9.5800% 0% 0.7428% 1.77E+02 1.77E+00 84.2322% 0% 17.0352%

!3.698-03 3,69E-05 0.0052% 0% 0% 1.388-13 1.38E-15 0% 0% 0%

i1.191E-07 1.19E-09 0% 0% 0% 3.688-06 3.68E-08 0% 0% 0%

i3.23E-12 3.238-12 0% 0% 0% 1.008-10 1.O0E-10 0% 0% 0%

i1.598-07 1.59E-07 0% 0% 0% 8.92E-07 8.92E-07 0% 0% 0%

i1.67E-05 1.67E-07 0% 0% 0% 2.758-04 2.75E-06 0% 0% 0%

8.13E-11 8.13E-13 0% 0% 0% 2.508-09 2.50E-11 0% 0% 0%

S7.368-09 7.368-11 0% 0% 0% 2.27E-07 2.27E-09 0% 0% 0%

i5.33E-05 5.33E-07 0% 0% 0% 1.63E-03 1.638-05 0% 0% 0%

i7.62E-05 7.62E-07 0% 0% 0% 2.35E-03 2.35E-05 0% 0% 0%

1.07E-05 1.07E-07 0% 0% 0% 3.29E-04 3.298-06 0% 0% 0%

2.74E-05 2.74E-07 0% 0% 0% 7.31E-04 7.31E-06 0% 0% 0%

4.13E-07 4,13E-09 0% 0% 0% 9.121E-06 9.128-08 0% 0% 0%

3.39E-13 1.69E-13 0% 0% 0% 1.05E-11 5.25E-12 0%" 0% o%

8.65E-04 8.658-06 0% 0% 0% 4.868-03 4.86E-05 0% 0% 0%

8.248-04 &24E-06 0% 0% 0%i 3.57E-06 3.57E-08 0% 0% 0%

1.201E-09 1.20E-11 0% 0% 0%j 3.72E-08 3.72E-10 0% 0% 0%

5.95E-05 5.958-07 0% 0% 0%1 1.008-03 1.00E-05 0% 0% 0%

3.02E-07 3.02E-09 0% 0% 0%* 3.36E-06 3,36E-08 0% 0% 0%

1.70E-14 1.708-16 0% 0% 0%* 5.248-fl 5.24E- 15 0% 0% 0%

1.758-06 1.758-08 0% 0% 0% 4.468-05 4.468-07 0% 0% 0%

2.748-06 2.74E-08 0% 0% 0% 2.1SE-05 2.15SE-07 0% 0% 0%

1.658-04 1,65E-06 0% 0% 0% 6.448-04 6.448-06 0% 0% 0%

4.86E-05 4.86E-07 0% 0% 0% 3.75E-05 3,75E-07 0% 0% 0%

2.128-07 2.12E-09 0% 0% 0% 5.79E-06 5.79E-08 0% 0% 0%

2.798-05 2.79E-07 0% 0% 0% 8.468-04 8.46E-06 0% 0% 0%

1.55E-05 1.SSE-07 0% 0% 0% 2.691E-04 2.69E-06 0% 0% 0%

3.308-10 3.308-12 0% 0% 0% 1.028-08 1.02E-10 0% 0% 0%

1.41E-06 1.41E-09 0% 0% 0% 4.408-05 4.40E-08 0% 0% 0%

• Container activity pirior to afire

• • Activity Released (from container via fire) = Activity Limit x Fire Release Fraction CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292 (Sep 2014)

Calculation Worksheet Page 16 of 49 Calculation: RA-O065 Revision: 0 Addendum: N/A

~Dominion-Table 5B Concrete Summary (Ci, (ci,

• !*

  f

  ;::i f 4.7%

41.71% 5.2%

58.29% :i~i)*rij i i*i*)i);*

(0) f%,:

-10.01% 89.99%;::*00*

I8999 5.52E+00' 2.76E+00O 72.7742% 0% 0.0436% 2.40E+00 1.208+00 92.3349%( 0% 0.0118%(

l.i5E-O1 1.15E-03 0.0302% 0% 0%( 2.198-01 2.19E-03 0.1685%( 0% 0%(

1.29E-04 6.46E-05 0% 0% 0%( 2.468-04 1.23E-04 0.0095%( 0% 0%A 2.41E-03 2.41E-03 0.0636% 0% 0%( 4.14E-03 4.14E-03 0.3190%( 0% 0%A 8.09E-01 8.09E-03 0.2134% 1.1067% 81.3865% 1.55E+00 1.55E-02 1.1933%( 8.5228% 97.4034%(

1.52E-07 1.528-09 0% 0% 0%( 2.79E-07 2.79E-09 0%( 0% 0%

1.448-02 1.44E-04 0% 0.0229% 0%( 1.32E-11 1.32E-13 0%{ 0% 0%(

7.83E+01 7.83E-01 20.6357% 0% 1.2079% 1.698-01 1.69E-03 0.1304%( 0% 0%i 1.08E-03 1.08E-05 0% 0% 0%( 2.068-03 2.06E-05 0%( 0% 0%A 8.768+01 8.76E-02 2.3104% 55.9824% 9.5106% 4.978+00 4.97E-03 0.3834%( 12.7924% 0.3378%

1.31E-01 1.31E-03 0.0347%1 0% 0.0050% 2.088-01 2.08E-03 0.1602%( 0% 0%A 2.65E-03 2.65E-05 0% 0% 0% 6.148-15 6.14E-17 0% 0% 0%(

1.188-07 1.18E-09 0% 0% 0% 2.19E-07 2.19E-09 0% 0% 0%(

2.598-13 2.598-13 0% 0% 0%( 4.758- 13 4.75E- 13 0% 0% 0%

1.148-05 1.14E-05 0% 0% 0% 3.788-06 3.78E-06 0%( 0% 0%

2.798-03 2.79E-05 0%( 0% 0.0057% 2.848-03 2.84E-05 0%( 0% 0%(

S5.17E-12 5.17E-14 0%A 0% 0% 9.86E-12 9.86E-14 o0% 0% 0%

1.07E-06 1.07E-08 0% 0% 0%( 1.968-06 1.96E-08 0%( 0% 0%

6.69E-06 6.69E-08 0%i 0% 0% 1.f2E-05 1.22E-07 0%( 0% O0%

1.4.38-04 1.43E-06 0% 0% 0% 2.628-04 2.62E-06 0%( 0% O0%

1.98E-06 1.98E-08 0% 0% 0% 3.638-06 3.63E-08 0% 0% 0%

i1.09E-04 1.09E-06 0%A 0% 0% 1.738-04 1.73E-06 0%( 0% 0%

S1.468-06 1.46E-07 0%A 0% 0% 1.938-05 1.938-07 0%( 0% 0%*

1.89E-11 9.448-12 0%( 0% 0% 3.61E-11 1.81E-l1 0%( 0% 0%

6.728-02 6.72E-04 0.0177% 0% 0% 2.268-02 2.26E-04 0.0174% 0% 0%

1.698-01_ 1.69E-03 0.0445%( 0.6896% ,0% 4.47E-05 4.47E-07 0% 0% 0%(

6.038-08 6.03E-10 0% 0% 0% 1.15E-07 1.15E-09 0% 0% 0%

S3.01E-03 3.01E-05 0%( 0.0050% 0% 3.078-03 3.07E-05 0% 0.0206% 0%(

3.468-04 3.46E-06 0% 0% 0% 2.35E-04 2.35E-06 0% 0% 0%(

3.508-12 3.50E-14 0% 0% 0% 6.438-12 6.43E-14 0% 0% 0%(

4.59E-02 4.598-04 0.0121%( 0% 0% 6.888-02 6.88E-04 0.0530% 0.0053% 0%(

1.348+01 1.348-01 3.5450%( 38.7252% 6.7072% 6.518+00 6.51E-02 5.0161% 75.4500% 2.0309%(

1.15E+00 1.158-02 0.30319( 3.4583% 1.0440% -03 0.2028% 3.1852% 0.1494%(

0.00599( 0% 0.0334% -05 0% [ 0% 0%(

0% 0% 0% -08 0% 0% 0%

0% 0% 0% -05 0% 0.0221% 0%(

0% 0% 0% -05 0% 0% 0%(

0% 0% 0% -11 0% 0% 0%(

0% 0% 0% -07 0%1 0% 0%(

nod no'

• Activity Released (from container via fire) = Activity i~mitx Fire Release Fraction Now I30 rsI Stlees 2,997 mrem/yr

/t:et 2,8438mrem/yr mrem/yr 2,959 mrem/yr Qan 2,949 CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292 (Sep 2014)

Calculation Worksheet Pafe 17 of 49 oDminlow- Calculation: RA-0065 Revision: 0 Addendum: N/A Table 6 Container Activity Limits Summary

• Stainless Steel Concrete H1-3 6 40E-05 4.54E-04 5.520 2.396 C-14 9.76E-03 3.04E-01 1.15E-01 2.1!9E-01 CI-36 3.98E-06 1.25E-04 1.29E-04 2.46E-04 Ar-39 1.06E-07 3,04E-06 2.4.1E-03 4.14E-03 Ca-41 1.84E-06 5.73E-05 &09E-01 1.548 Mn-53 2.54E-07 7,83E-06 1.52E-07 2.79E-07 Mn-54 2.54E-02 3.92E-1O 1.44E-02 1.32E-11 Fe-55 32.833 1.160 78.263 1.69E-01 Ni-59 4.30E-02 1.34 1.08E-03 2.06E-03 Co-60 317.661 302.932 87.624 4.975 Ni-63 6.854 177.051 1.31E-01 2.08E-01 Zn-6S 3.69E-03 1.38E-13 2.65E-03 6.14E-i5 Se-79 1.19E-07 3,68E-06 1.18E-07 2.19E-07 Kr-81 3.23E-12 1.00E-IO 2.59E-13 4.75E-13 Kr-SS 1.59E-07 8.92E-07 1.14E-05 3.78E-06 Sr-90 1.67E-05 2.75E-04 2.79E-03 2.84E-03 Nb-92m 8.13E-11 2.50E-09 5.17E-12 9.86E-12 Zr-93 7,36E-09 2.27E-07 1.07E-06 1.96E-06 Mo-93 5.33E-05 1.63E-03 6.69E-06 1.22E-05 Nb-94 7.62E-05 2.35E-03 1.43E-04 2.62E-04 Tc-99 1.07E-05 3.29E-04 1.98E-06 3.63E-06 Ag-108m 2.74E-05 7.31E-04 1.09E-04 1.73E-04 Sn-121m 4.13E-07 9.12E-06 1.46E-05 1,93E-05 1-129 3.39E-13 1.05E-11 1.89E-11 3.61E-i1 Ba-133 8.65E-04 4.86E-03 6.72E-02 2.26E-02 Cs-134 8.24E-04 3,57E-06 1.69E-01 4.47E-05 cs-13s 1.20E-09 3.72E-081 6.03E-08 1.15E-07 Cs-137 5.95E-05 1.O0E-03 3.01E-03 3.07E-03 Pm-145 3.02E-07 3.36E-06 3.46E-04 2,35E-04 Sm-146 l.70E-14 5.24E-13 3.50E-12 6.43E-12 Sm-l51l1.75E-06 4.46E-05 4.59E-02 6.88E-02 Eu-152 2.74E-06 2.15E-05 13.445 6.508 Eu-154 1.65E-04 6.44E-04 1.150 2.63E-01 Eu-155 4.86E-05 3.75E-05 2.22E-02 1.01 E-03 Th-1S8 2.12E-07 5,79E-06 l.95E-06 3.16E-06 H1o-66m 2.79E-05 8.46E-04 8.51E-04 1.54E-03 Hf-178m l.55E-05 2.69E-04 2.95E-03 3.00E-03 Pb-205 3.30E-I0 1,02E-08 1.16E-09 2.22E-09 U-233 1.41E-06 4.40E-05 3.81E-04 6.99E-04 Pu-239 2.73E-05 8.54E-04 1.42E-03 2.60E-03

• Pre-fi re Activity Ž1 mCi is bolded CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292 (Sep 2014)

oPD minion-

• DCalculation Worksheet Calculation: RA-0065 Page 18 N/A Revision: 0 Addendum: of 49

12. Acceptance Criteria (Optional)

Instantaneous dose rates must not exceed twice the ODCM instantaneous dose rate limit (2 x 1,500 mrem/yr =3,000 mremn/yr) at the site boundary.

13. Results and/or Conclusions Table 6 provides stainless steel and concrete nuclide activity limits to preclude site boundary dose rates exceeding twice the ODCM instantaneous dose rate limit (2 x 1,500 mrem/yr = 3,000 mrem/yr).

These are container limits based on a worst case fire release to the Site Boundary. Table 5A (Stainless Steel) and Table 5B (Concrete) show the nuclide activity and dose rate impacts and illustrate the impact of nuclide decay while maintaining a 3,000 mrem/yr dose rate limit.

These limits can be used for initial planning until actual stainless steel and concrete lab sample results are available at which time refined nuclide activity limits should be derived using the methodology used in this calculation.

14. Precautions and Limitations If this method is used to determine dose rate (mrem/yr), then the nuclide source term input units must be in pCi/sec. To derive IpCi/sec dose rate units assume the nuclide activity is released over one hour and then convert to pCi/sec accordingly.

If this method is used to determine Nuclide Activity Limits (NALs), then nuclide sample mix proportionality is fundamental in determining its associated NAL--not the absolute activity of each nuclide. Thus, container contents must be consistent with the basis of the container's NALs.

15. Recommendations (Optional)

• New Inhalation instantaneous dose rate DCFs in Attachment D Table D-2 (Inhalation) be added to the appropriate ODCM Tables 2.4-2.7.

• New Ground instantaneous dose rate DCFs in Attachment D Table D-3 (Ground) be added to a new OCDM Table 2.16 and titled for Site Boundary Instantaneous Dose Rates (10 CFR 20).

Existing ODCM Table 2.15 be titled for Unrestricted Area Doses (10 CFR 50).

• ODCM §2.3.2 add the Ground dose rate formulation in similar fashion to ODCM §2.4.2.

• D/Q be added to ODCM Table 2.30ODCM Normal Condition 13.2.1.b.

• Consider adding Table 3 Fire Release Fractions to ODCM as new table.

16. Calculation Review Checklist (See page 20)

17. Attachments A Nb-92/Nb-92m Halflives B )(IQ and DIQ Basis C eSOMS Log Permanent Shutdown Date Time D NewOD0CM DCFs CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292 (Sep 2014)

Calculation WorksheetPae1of4 PaKe 19 of 49

~Domini.n- Calculation: RA-0065 Revision: 0 Addendum: N/A E ODCM Methodology Results F Source Term vs Activity Limit CM-AA-CLC-301 Rev 9, Attachment 47292(e204 729292 (Sep 2014)

• Calculation Review Checklist Paee 20 of 49

~D ominion" Calculation: RA-0065 Revision: 0 Addendum: N/A NOTE: If "Yes" is not answered, an explanation may be provided below. Reference may be made to explanations contained in the calculation or addendum.

1. Have the sources of design inputs been correctly selected and referenced in the calculation? [X ] []

2. Are the sources of design inputs up-to-date and retrievablelattached to the calculation? [X ] []

3. Where appropriate, have the other disciplines reviewed or provided the design inputs for which they [X ] []

are responsible?

4. Have design inputs been confirmed by analysis, test, measurement, field walkdown, or other pertinent means as appropriate for the configuration analyzed?[X] []

5. Have the bases for assumptions been adequately and clearly presented and are they bounded by the Station Design Basis?[X] []

6. Were appropriate calculation/analyticmethods used and are outputs reasonable when compared to [ X] []

inputs?

7. Are computations technically accurate? [ XJ []

8. Has the calculation made appropriate allowances for instrument errorsand ca/ibrationequipment [X] []

errors?

9. Have those computer codes used in the analysis been referenced in the calculation? [ ] [ X]

10. Have all exceptions to station design basis criteriaand regulatoryrequirements been identified and ~X justified in accordance with NQA-1-1 994?[] [X]

11. Has the design authority/original/preparer for this calculation been informed of its revision or ~ ~ X addendum, if required?[] [X]

12. Was the pre-job brief completed without any identified HU error precursors/compensating actions?

(If HU error precursors/compensating actions were identified, then mark N/A and provide [X ] [ ]

explanation/summary below or attach pre-job brief form to calculation.)

Comments: (Attach additional pages if needed)

9. No computer codes were used in this calculation

10. No exception taken

11. RevO0 Sign(rperatu re: Dale M Flickk*Da te:*/ 5 Signature: Douglas L GilliattDae (Reviewer) * t "2- /' ,

Signature: N/A Date:

(Owner's Review, if applicable)

Note: Physical or electronic signatures are acceptable.

CM-AA-CLC-301 Rev 7, Attachment 57310(e214 731190 (Sep 2014)

Pe21of 49

• Dominion- Calculation: RA-0065 Revision: 0 Addendum: N./A Attachment A - Nb-92/Nb-92m Haiflives representative i i range of natural nuclide iZ(p) iN~n) i sot*opi mass (u) decay daughe half-lie (mole...acion (mole fraction) excitation energy i ,p i'%Z UNb 41t 40 &0.94003{151)# <44 rns i Nb i 41 i0+

UNb i41 4*2i82,.93,071(34) 4.1(3) s

+)*

i p .*)i {> =Zr i'Nb 41 43 83.93357(32~W 103(19) ns

• "Nb 7.9.0(10 key "'zr (9,2+)

• Nb 41 45 85.92504{9) "zr i(0+)

3.75s(9) mm i'Zr * (I2-)

! =Nb 2540(1410) key i2.0(1) min =Zr ,*"

L~Nb 41 4 0 180 920e341(2T i14 55(0) ri s--- 3 34(14)keY

• Nb 41 47187}918338511) i2,03(7) h, "Zr i(9:2+1

] '*N 40(142302)keV 11 (o3) h i1450{() h i *Nb 12.0(35) keY 56+

i18,81(0) s "Nb i4-

"KmlNb 122372.0(25) keV IT

...... . ... T ;7......

......................... ;1 + ....

472(13) fis iEC (99.98%} 31Zr 080 (13{ ) (.........................-.....r.. G~2+

rr (93%)} Nbz

,! Nb! 104.00(5) keV 3,75(12) 'is

~~nt~Nb: 2034.30(19)keV i(17V2-)

'3Nb 41 4t 51* 91,907*194(3) i3.47(24)X10* al' P~ (99.95%)

'3eZr

~ (.05%)

s b 22574 keV i5.9(2) 'is e*bt 2203.3(4) keV . 107(4) r'*s (11-)

en.wikipedia.org/wiki/isotopes of niobium

Pe22of 49 SDomninion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment A - Nb-92/Nb-92m Haiflives Radiosotope data Further data for naturally occuring isotopes of niobium are listed above. This table gives information about some radiosotopes of niobium, their masses, their hafflii es, their modes of diecay, their nuc ear spins, and their nucl~ear magnetic momernts.

Isotope Mass I Da Half-life Mode of decay Nuclear spin Nuclear magnetic moment

• Nb 88.91349 1.10 h EC toaQZr 112

• Nb 89.911263 14.6 h EC to *Zr 8 4.961 1/2

• Nb 90.906989 700 y EC to91 Zr 6.114 I Nb 91.907192 3.7x107y jECto*Zr, 13-to*Mo 7

• Nb 93.907282 24000 y f3- to 94Mo 6 g*Nb 94.906834 34.97 d f3- to 9-*Mo 6.141

• Nb 95.908099 23.4 h f3- to *Mo 6 4.976 97 Nb 96.908096 1.23 hi f3"to Q7Mo 6.15 www. webelements. corn/niobium/isotopes.html Isotopes

[Isotope Half Life

[Nb-90 I14.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

[Nb-91 700.0 years Nb-91m 162.0 days Nb-92 {3.6E7 years Nb-92m 10.13 days Nb-93 Stable Nb-93m 16.1 years Nb-94 20000.0 years Nb-94m 6.26 minutes Nb-95 34.97 days Nb-95m 3.61 days

!Nb-9 I]123.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Nb-97 1.23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> Nb-97m 58.1 seconds www.chemicalelements.com/elements/nb.htmi

Pa~e 23 of 49 SDominion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment B X/Q and D/Q Basis KPS ODCM excerpt for Site Boundary xIQ Basis KEWAUNEE PONER STATION ODOM 2.0 OFFSITE DOSE CALCULATION MANUAL Revision 17 Sept. 25, 2014 Table 2.3 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations Atmospheric Dispersion ODCM Normal Condition Location Pathways xtQ (sec/rn3 ) I)/Q (i'm 2 )

3,Aa Site Boundary Noble gases 7A4E07 N/A

__3__2___a (0.81 mile. NI*/) Direct exposure ______

13,2A.b Sit Bonay naain 7.44F-07 N/A

___________(0.81 mile, Nf*/} Ground Plane______

13.22 Site Boundary (0.81 mile, NNW?) Beta AirAirE07N/ 7A4___-__7 Gammaw 13.2_____2 _______

13.2.3 Resience/dairy inhalation F18F0 13,2,3Resience/dairy Vegetation, Mi 391:-8 .6E0 (1.3 mile SW and Ground Plane______

KPS Mereoroloqical and Atmospheric Dispersion Report excerpt for Site Boundary DIQ Basis Table 10 Meteorological Dispersion Parameters for Site Boundary Locations X/Q (s/rn+)

No Decay 2.26 Day 8 Day Decay Decay Sector Diht (fI) Dist (m) Dist (mi) Undepleted Undepleted Depleted D/Q (m-)

SSW 4100 1250 0.78 3.3E-07 3,3E-07 2.9E-07 6.1E-09 SW 4010 1222 0.76 3.2E-07 3.2E-07 2.8E-07 4.2E-09 WSW 4328 1319 0.82 2.1E-07 2,1E-07 1,8E-07 2.5E-09 W 4271 1302 0.81 2.3E-07 2.2E-07 2.0E-07 2.4E-09 W~N"W 4391 1338 0.83 ,.7*E-07, 2.7E-07 2,4E-07 2.7E-09 NW 4851 1479 0.92 2.7E-07 2.7E-07 2.4E-07 2.8E-09 N'NW 4827 1471 0.91 6.0E-07 6,0E-07 5,3B-07 [ 6.2E-09 N2819 .95,3E-07 5.3E-07 4.6E-07 4.4E-09

• D Pa~e 24 of 49 OD~ in~iliu~n Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment C - oSOMS Log PermanentShutdown Date Time Coatls Room Log AHRENS, GARY 7sL 5:7/201312:05: Tl)AFWTcaisLCO is metEsitActiooStttm*-js I*iE ycopeodfr*omTehSpec eacr~k n-DaySl - 05:0?:O13byAHIBRNS, GARYM.,ŽdHSCl AHREN-S, iARY.M.

5/'7,2013 11:43 T *usioe to KW-GOP-203. Slu~astdsmo :Mode 3 to RHl.

5/712013 11:40 Sh.*dowa* Coto Rod Odv Syste pcmNOP-CRD-002, Costeol Room Log GIESE. IR.NDAll SCOTT Trassoe to KW-GOP-204. Slaitdwnm Mod 2. to Mod 3 (Remztor Shutdovos).

517;/2013 11:10 Completed OSP-RCS-006. Shatdowa Mxipo to eam* Mode 2. Slutdtwa Mmgi is Met C~onto Room Log AHEENS. GARY M

"/7/2013 10:-49 Shtdown M.gs caclto lAW OSP-RCS-006 is met far eaea Modes 2 emd 3.

C~onto Room Log SCHERWINSKI. DOUGLAS 51712013 10:38 Secure the A Heaer Drau Pimp jAW NOP-HD-00lI Cotted Room Log SCHERWINSKI, DOUGL.AS 5-'712013 10:403: Secozedtle B Heatti Drain* o, IAW NQP-HD.001 SCH*ERWINSKI, DOUGLAS Sf712013 16:00: C'opleted Shutdowva of -ester Drum S'ytem perOP-KW-NOP-HD- 001sertcts- 5. SIODOLA, C~A.DWICK 517/2013 10:33 Shstdown Tixisie Traps an Drain pm NOP-TD-00l.

Costol Room Log F*ICTUAM. JOHN JOSEPH 5/712013 09:34 Trasloe fo*maKW-GOP-206 to KW-GOP-205 Shutdowna froam35% Power to Mode 2.

Conerol Room Log SCHERWIINSKI. DOUGLAS 5/7,/2013 07:57, Serad Chemiojeci for slattw

• D Paxe 25 of 49 ODlmini~inr Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment D - New ODCM DCFs Table D-1 shows the status of the Ground and Inhalation ODCM DCFs. This attachment is concerned with the new 0DCM DCFs (New) generated in support of this calculation. Existing Inhalation 0DCM DCFs (X)were also used in this calculation.

Table D-2 shows the new Inhalation 0DCM DCFs derived using the equations in §10 Methodology DCF6 i with dc!6 i from NUREG-0172.

Table D-3 shows the new Ground 0DCM DCFs derived using the equations in §10 Methodology DCF,0 o, with dcf,0 oi from either RG 1.109 Table E-6 or GASPAR II.

Table D-1 Nuclide ODCM DCFs Irun Inhalation H-3 0 X C:-14 0 X a-36 n/i n/I Ae-39 n/I New Ca-41 New New Mn-53 n/I n/I Mn-54 New X Fe-SS 0 X Ni-59 0 New co-6o New X Ni-63 0 X Zn-65 New X Se-79 0 New Kr-81 n/I n/il Kr-S5 n/I New Sr-90 0 X Nb-9Zn. n/i n/i Zr-93 0 New Mo-93 New New Nb-94 n/I n/I Tc-99 0 New Ag-108m n/I n/I Sn-121m n/I n/I I-129 New New Ba-133 n/il n/I cs-134 New X Cs-135 0 New C~s-137New X Pm.145 n/I n/I Sm-146 n/il n/I sm-151 New New Eu-152 New New Eu-iS4 New New Eu-1SS New New Tb-iSa n/I n/I Ho-lG66mNew New Hf-178mi n/I n/I Pb-o0 n/I n/I U-233 New New Pu-239 New New x =ExistingODcM DCFs New =Newly derived DCF(s) 0 =DCF o~fzero n/i = Not listed in source Note: ODCM DCFs con be updoted using the NUREG-0133 approoch.

Page 26 of 49 SDominion- Calculation: RA-0065 Revision: 0 Addendum: N__L Affachment D New ODCM DCFs -

Table D-2 New Inhalation ODCM DCFs (DCF,.o,,

dot lo b fn R~

ODM 41805 .380 0.08 08.8 .4L08E405 1 804 1.080 3.M8728407 0.08 0 0 2.03EE0 3008+E0O 7 4.26183180 .38+06 39.606E 441.12 431.4601 0 3.0E0 1.7201,0 4.13-05 67751-07 0.0084.00 O.0OE8.00 0.008+00 0.00E400 31208401 O.0OE+O0 O.OOE+00 4.05E-05 0 0 8 1,011-01 3.031-07 4.331-01 3,.248405 0.oo0.,0o 0.00E+00 o0ooE-4o 8.068408 2.42840 3.50840 5.441-06 2,021,,6 0 0 1.411-05 6.4.81-07 5.240-87 4.358+04 1.62E'04 0.008403 0.O08400 1.13-405 S.18E4*03 7.39E+03 0 1,431-07 0 81236.07 7.71E1,05 3.53E-46 8,711E-08 0.008400 4,34i0 0.0(8400 6.508403 0.17E.4)05 2.82E404 6.971E402 6,03-05 3.38105 0 3161-04 3.67E-04 1.600E-5 1.841-05 5.468E-3 2..708405 0.006400 9.23E805 2.948406 1.288405 1.4784*05

'21 0 1860.-06 0 51061-07 8 811-05 3,991-06 45$26-01 21 I3.56E4.02 4.481-88 8-.26.+32 4.581-08 0.OtX-4-0 0 8.31E-07 6.668E43 1.39E406 1,741-04 6.398.404 7159E-04 1.43E.4T2 1 79E1,08 3.531-06 2.841-06 3.661E-83 5.2614 000.0 2.291-07 480.90E-2.82E8+)4 2.358+04 2.93E-4)7 4.218404 0.0038.00 1.83E+03 3.92E404 77 I] 1.668,,.5 2.081-45 1.46E-415 1.821-05 0.O.+00,O 0 58844+04 7,30E-01 2.168404 3,70106 1.7684.03 2.231*-07 3.588404 41471-06 1.7-4 3.0-5 0 2.27E-OS 7.60E-01 3.531-06 4861-06 8.568405 1.68E405 0.00E-40O 182E405 6.148405 2.82E404 3.898404 2.966-04 7 19E-05 0 3,3,41-04 5.81E-64 1.351-05 0.301-05 2.371E406 5.758405 0.00E400 2.678406 4.018406 1.0084E05 5.048405 9.411-04 1.231-04 8 5.44E-04 9.121E-0.4 3.341-05 8.601-05 l 7.548406 9.848405 0.0084*00 4,35E406 7.308406 2.67E405 6.868405 7.001-04 1.961-05 0 7.656-05 1.511-83l 5.97E-05 1.311-0 1..608406 1.578+,05 O.00840.0 0.12E405 1.218407* 4.788405 9.688404 4.4,01-04 1.361-04 0 2.006-04 6,246-04 1.601-05 9.871-05 3.52E406 1.098406 O.00E8400 1608406 4.998406 1.348405) 7.908405 3.55r102 0 0 3.63E-03 9.181-,6. 4.121.05 9.421-04 1.24844808 .OE8403 0.008400 2.908+07 7.348406 3.308405 7.54840I6 o]3.311e00 2.658+10 3.608409 4.5E01-0 0.008400 8 2.758409 3,446-01 2.348409 2,931-,01 3.508405 4.371-05 6.448408 8,051-02

Page 27 of 49 Calculation: RA-0065 Revision: 0 Addendum: N/A

~D@mini.n Attachment D New ODCM DCFs Table D-2 (Cant)

New Inhalation ODCM DCFs (DCF,0 )

0 4 0 0 4.49449 o 0 O.00(+00 O.O0(+O0 0.00(400 0.00E0 1.O 81E401 0.00(.00 0.00(+00 7440 0 0 0 7.214..02 2.966'07 7.706-04 2.61E405 0.00O400 0.00(400 O.00(400 2.6"(4+8 1.09E+03 2.85E404 1.060 4.671-06 0 0 2,736-04 4,29647 2.036-06 6.14E404 1.73E+04 O.00(+00 0.00(4O0 1.01E405 2.33(+03 1.05E(404 0 t2144*6 0 1.716.06 1.49604 3.476-04 24060-7 0.O(400] 4.54(+03 0.0(.o00-H 6_33(4"03 S.51(+05 1.27(4434 9.67E'402 2.076-04 7.BE4060 0 3.00E.44 7.106,04 1.4744'5 135601I 7.66(-OS 2.89(405 O.O0E+O0 1.113406 2.63(406 5.44(404 2.05E+05 0 3,76E-06 0 1.046E-0 1.70604 3.74E-06 1.35E07 0.00(400 1.39(404 0.00(400 3.92(-48 6.2g(+05 1.40(404 5.00(+02 1246-07 1.49E-47 0 1.756-44 3.37644 71364,6 4.36-*09 4.96(+02 5.51E402 0.00(400 6.48E400 1. 25E+06 2.87(+04 1.98E,.02 1.444405 4.40E606 4.14643 101-.0564 040.+04 7.1464*7 4.714E'.0 3.89E,4)4 2.371(+04 1.58(407 4.00(404 O.00(400 7.96(40]2 2.11E404 63-05. 4.11E444 0 1.136*05 4.224-06 2.176-07 444446 2.31(405 1.53(.05 O.0O(400 5.66(404 1.93(404 8.031E-402 1.65E404 3.146-44 4.74445 0 4,.09605 1.40E-44 3.41*64 14.49,5i 1.16(406 1.76(405 O.00(4O0 1.81(0 5.48E405 1.27(404 5.51(<40 7.426,46 1 3764,4 0 1.736,46 9.44644 1.14649O 1616-44 2.79E.4] 507.07E5 0.846+00 2.123406 3.33E+06 4.22(s04 5.96E-(

7.746-03 2.49E,44 0 1.0944.3 1.4644,0 234645 2.27,044 1.01(407 9.21(.05 0.00(400 4.03(406 6.14(406 1.10(405 6.40(+O5 50E.,064 4.056-05 0 1.11644 2.79E-04 5.094415 3.18145 2.07E"406 1.50(405 0.00(400 5.59(405 1.05E+06 1.99(+05 1.18E.4)5 1.34443 2.01E-04 0 4.01644 1.13643 1.63E-45 2.37-044 4.96(406 1.04(406 0.00(+00 1.48(406 4.18(406 6.0(4E04 6.77(405 4.64642 0 0 7.2624.3 1.776E1 4.06E,0O 2.02E-43 1.72E.05 O.00(+00 O.00E+O0 2.82E+07 6.55(+08 1.48(+05 1.04(407 5.140.00 6.44E'01 0 4.701641 5.726-01 4.24E41 1.206-0 0.00(400 O.00(+O0 0.00(+O0 0.00(400 1.40(401 O.00(-00 0.00(4000 7.4E445 0 4 0 6.94442 2,96647 014606 1.05(+06 OO0('00 0.00E-400 0.00(400 9,72E(407 4.14(402 1.14(404 1.0161 4.44E-44 0 0 1.40441 4.344-47 3.0*046 2.$3E+04 7.62(403 0.00E400 0.0OE400 7.67E404 8.881E+02 4.3,(403 0 2.25E-04 0 7,474446 2.996,04 14,64-6 4.20E-47 0.00(40 3.15(400 0.00(-00 3146(403 4.19(4+05 4.84(403 5.88(402 2,246-46 9,516-04 0 3.196-04 1.3764D3 1.,4*044 6.14441 3.14(40]5 1.33E'05 0.00(-400 4,47(*405 1.92(406 2.07(404 8.65E+04 0 544614,6 0 1454646 3*.DE046 3.76E,04 2.2264'7 0.00(406 9.04(4,06 0.846.092 2.16(4,206 4.76(405* 5.26(4+03 3.11E'402 2.040,.7 2,60E,47 0 2.49646 6.7760*4 7.204E4 0.0514 2.98(+02 3.75(402 O.00(+00 3.49(+03 9.48(405 1.03E404 1.24(402 2.14640 1.594,41 1.0464-0 1,.4044 0.O046,0 2,1264,7 1.166.46 3.02(4.104 2.23(404 1.46(-107 2.63(404 0.00E(400 2.97E402 1.62(+04 1.41064 4 .4644-5 0 2.19641. 3.144140 6,14607 6.74446*

1.40(s0 1.21E'4]5 0,00E(400 3.61E404 1.41(404 3.05E'02 6.62(403

$3BE,3O4. 6.45E.44 0 .424604 2.9E-04 0,46,E-6 3.63645 4.73(4,00 9.03(-4)4 0.00(+00 7.34(4E04 4.17E405 4.84(403 2.28E-104 7.034-04 1 77,044 0 1.946,04 3.406403 9.0E064 1.726*44 1.10(406 2.48(*405 0.00(+O0 6.321(405 2.07(+06 1.38(404 2.41(+05 4.14(406 4.84(405 0.00(+00 1.60(4016 4,27E406 3.98(40H4 3.48(405 4.97E,44 4,726405 0 1.40604 4,20644 4.196-01 3,446(0 6.36(4056 .0LOIE404 0.00-400 2.-21(405 7.28(40D5 7,27E(404 4.84(4044 1.44643 30744O4 0 4.2204,4 70464.3 1.4654~5 2.516*44 2.03(406 4.30(405 O.00~r(400 5.91E+4]5 2.87E+06 2.31(404 3.51E(405 0 0 3.03442 3,446-01 4.04641* 3.036-43 O.00(400 O00OE4O0 1.58(407 4.98(408 5.64(4*04 5.36(406 4.10E+O0 4,72E041 0 4.91641 3,47441 4.20E401 1,341419 7.70(409 9.411E+05 0.0(30£00 6,93E+05 1.19E409 5.99(404 1,88(408

Page 28 of 49 SDomninion- Calculation: RA-0065 Revision: 0 Addendum:N/

Attachment D New ODCM DCFs Table 0-3 New Ground 0DCM DCFs (DCF 6J) nNUREG-0133 §5.2.1.2

Page 29 of 49 9 Dom in ion- Calculation: RA-0065 Revision: 0 Addendum:NA Attachment E - ODCM Methodology Results Stainless Steel - Shroud - Now Results Input IShroud Su~° yr ost I °'00.

"Decay Mix To" Date XIQ 11/30/2016 12:00 7.44E-07 3.57 yrs I1os/o7/2o1 12:00111130/2016 12:001 05/08/2043 0:00 I D/Q 6.20E-09 Nuclide 11-3 H1-3 0.5 0%

O-14 C-i4 0.01 0.003%

I  %*.I2. 76E+00mrem/yrI3.00E-O3mrem/yrI1.39E.tO3IuCi/sec C1-36 Cl-36 0.5 0%

Ar-39 3.125E-06 31498E-05 Minimum 3.91E400 ci/hr Ar-39 1 0%

Ca-4i 2.404E-052 *6E0 Ca-4i 0.01 0%

Mn-53 5.100E-07 4.29E-04 Uimit Reduction Factor =1 Mn-53 0.01 0%

Mno54 81348E-072 i07- Mn-54 0.01 0.039%

4.700E-09 6Z*Q2.*';"*

Fe-55 Fe-55 0.01 9.101%

NI.59 3.200E-10 Ni-59 0.01 0.0o12%

Co-SO 0.001 08.845%

6.612E-04 2.437E-0$ Ni-63 0.01 1.917%

Zn-65 18340E-02 17E0 Zn-65 0.01 0.002%

Se-79 1.200E-14 *427-* Se-79 0.01 0%

Kr-Si 81320E-102 Kr-Si 1 0%

Kr-B5 9 753E-02 .7E0 Kr-SS 1 O%

Sr-SO 4.578E-OS *1.6306 Sr-SO 0.01 0%

Nb-92m 61300E-07IO 7E0 NB-92m 0.01 0%

Zr-93 Zr-93 0.01 0%

Mo-93 Mo-93 0.01 0%

Nb-94 Nb-94 0.01 0%

2.3837E-06 7.317E-06 TC-SS Tc-99 0.01 0%

Ag-1AI~m Ag-i08m 0.01 0%

9.393E-07 3.44-0 Sn-121m Sn-121m 0.01 0%

1-129 1-129 0.5 0%

4 OO7E-11 Ba-133 13,-133 0.01 0%

Cs-i34 Cs-134 0.01 0%

Cs-135 6,00OE-13 Cs-13S 0.01 0%

Cs-137 Cs-137 0.01 0%

Pm-145 Pro-145 0.01 0%

Sn-2A6 Sm-146 0.01 0%

Sm-151 Sm-iB1 0.01 0%

Eu-1S2 Eu-152 0.01 0%

Eu-154 Eu-154 0.01 0%

Eu-155 Eu-1SS 0.01 0%

Tb-158 Tb-158 0.01 0%

Ho-166m Ho-i66m 0.01 0%

Hf-i7Sm Hf-17Sm 0.01 0%

Pb-2O5 Pb-205i 0.01 0%

U-233 U-233 0.001 0%

Pu-239 Pu-239 0.001 0%

Total Total 3.57bEOHZ Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire RsI Fraction.) the activity shown in 3rd column (multi-colored) from the left.

Pagle 30 of 49 SDominion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E - ODCM Methodology Results Stainless Steel - Corel Barrel - Now Results Input ICoreBarrel Sht w Shutdown sDecayMixTo"~ Date 11/30/20161:0 3.57 yrs 1o5/o7/2o13 12:001 11/30/2016 12:001 05/08/2043 0:00 X/Q 7.44E-0 D/Q 6.20E-0 Nudildei Activiity jjci/sec) 11-3 7.858E-06 p~Ci/sec o%

C-14 2.999E-06 C-14 0.01 o 003%

Cl-36 6.400E-08 0-36 0.5 Ar-49 7.927E-09 M~inimuni 3.63E401 c/i/h Ar-39 1 5.600E-10 Ca-41 0.01 Ca-41 Mn-53 2.100E-1O Limit Reduction Factor= 1 Mn-53 0.01 Mn-54 2,000E-05 Mn-54 0.01 Fe-S5 9.604E-03 Fe-55 0.01 N1-59 1,900E-05 Ni-59 0.01 8.758E-03 0i%

Co-SO Co-60 0.001 NI-63 2.339E-03 Ni-SI 0.01 Zn-ES 1.1OSE-06 Zn-6S 0.01 0%

4.600E-11 Se-79 0.01 5e-79 o02%

Kr-Si 3.400E-12 Kr-8i 1 Kr-S5 1.509E-08 Kr-85 1 009%

4.593E-09 Sr-90 Sr-9O 0.01 6790%

Nb-92m 6.500E-14 NB-92m 0.01 Zr-93 3 900E-12 Zr-93 0.01 Mo-93 3.897E-08 MO-93 0.01 000%

Nb-94 2 900E-08 Nb-94 0.01 0%

Tc-99 8.200E-09 Tc-99 0.01 Ag- lOem 8,826E-09 3.251E-10 Ag-lO~m 0.01 0%

Sn-121m Sn-121m 0.01 0%

1-129 1,400E-14 1-129 0.5 0%

Ba-133 2.383E-07 Ba-133 0.01 Cs-134 3.014E-07 Cs-134 0.01 0%

cs-135 9.000E-13 i Cs-135 0.01 Cs-137 4.604E-08 Cs-137 0.01 0%

1.217E-10 Pm-14S 0.01 Pm-145 0%

Sm-146 1.300E-17 Sm-14S 0.01 4.378E-09 Sm-T11 0.01 0%

Eu-152 7.$59E-10 Eu-152 0.01 Eu-154 4.530E-07 Eu-154 0.01 Eu-155 8.504E-08 Eu-:155 0.01 Tb-i5O 1.672E-10 Tb-15a 0.01 Hlo-166m 1.297E-08 Ho-156m 0.01 2.585E-08 Hf-170m 0.01 Pb.205 1.30E-13 i Pb-205 0.01 1.000E-10 I U-l33 0.001 Pu-239 2.300E-09 Pu-239 0.001 Total 2.08E-02 Total s*.61Zt'*uz Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction) the activity shown in 3r column (multi-colored) from the left.

Pe31of 49 Calculation: RA-0065 Revision: 0 Addendum:NA 9 D@mini.n-Attachment E - ODCM Methodology Results Stainless Steel - Thermal Pads - Now Results 30 VyrsPost-Input Thermal~ads Shutdown Now Shutown 05/07/2013 12:00 11/30/2016 12:00 05/08/2043 0:00 "DecayMixTo" DoteI 11/30/201612:00 3.57 yrs x/(1 7.44E-07 D/Q 6.20E-09 Nudide Activity 1.801E-06 0%

C-14 4.598E-07 E ~~3.42E-0omrem/yrI. 00E+03 mrem/vrI8.77E.04I c;-o C-14 0.01 0.003%

Cl-36 1.OOOE-08 CI-36 0.5 0%

At-39 4.162E-10 Minimum 2.54E402 cifrr Ar-39 1 0%

Ca-41 8.600E-11 Ca-41 0.01 0%

Mn-53 1.OOOE-11 Limit Reduction Factor = 1 Mn-53 0.01 0%

Mn-54 1.OO0E-06 Mn-54 0.01 0.00 7%

Fe-55 1.481F-03 Fe-SS 0.01 l0.306%

Ni-59 3.O00E-06 Ni-59 0.01 0o.021%

Co-SO 1.251E-03 Co-6O 0.001 87.082%

Ni-63 3.703E-04 Ni-63 0.01 2.570%

1.454E-07 Zn-65 0.01 0.00 1%

Se-79 4.700E-12 Se-79 0.01 0%

Kr-S1 3.500E-14 Kr-81 1 0%

7.$43E-10 Kr-85 1 0%

Sr-SO 2.388E-09 Sr-SO 0.01 0%

Nb-92m 3.200E-15 NB-92m 0.01 0%

Zr-93 2.900E-13 Zr-93 0.01 0%

Mo-93 2.099E-09 Mo-93 0.01 0%

Nb-94 3.000E-09 Nb-94 0.01 0%

Tc-SS 4.200E-10 Tc-99 0.01 0%

1.079E-09 Ag-l0Sm 0.01 0%

Sn-121m 1.625E-11 Sn-121m 0.01 0%

7.900E-16 1-129 0.5 0%

Ba-133 3.575E-08 Ba-133 0.01 0%

Cs- 134 5.124E-08 Cs-134 0.01 0%

Cs-US 5,100E-14 Cs-13S 0.01 0%

Cs-137 2.486E-09 Cs-137 0.01 0%

Pm-145 1.91.3E-11 Pm-145 0.01 0%

Sm-146 6.700E-19 Sm-146 0.01 0%

Sm-151 2.821E-09 Sm-151 0.01 0%

Eu-152 1.412E-07 Eu-152 0.01 0%

Eu-3M4 4.681E-08 Eu-154 0.01 0%

Eu-ISS 3.037E-09 Eu-15S 0.01 0%

Tb-1S8 8.3616-12 Tb-1S8 0.01 0%

H4o-166n 1.098E-09 Ho-1E6m 0.01 0%

Hf-173m 2.86216-09 Hlf-178m 0.01 0%

Pb-205 1.3006-14 Pb-205 0.01 0%

U-2r3 1.100E-11 U-233 0.001 0%

Pu-239 1.400E-10 Pu-r3n 0.001 0%

Total 3.11E-03 Total Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction) the activity shown in 3 rd column (multi-colored) from the left.

Pa~e 32 of 49 SDominion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E - ODCM Methodology Results Stainless Steel - Vessel Cladding - Now Results I 30Shutdown Input jVessel~ladding Shutdown Now yrs Post-

"Decay Mix To' Date 11/30/201612:00 3.57 yrs 1o5/o7/2o13 12:001 11/30/2016 12:001 05/0812003 0:00 1 X/Q 7.44E-07

/Q D___ 6.20E-09 Nuclide Actvity h*Cilsec)

'4-3 2.701E-07 0%

C-14 6.597E-08 C-14 0.01 0.003%

CI-36 1,4ooE-09 CI-36 0.5 0%

Ar-39 5.945E-10 Minimum 1.54E+03 Ciihr 1 0%

1.200E-II1 Ca-41 Ca-41 0.01 0%

Mn-53 6.90OE-12 Limit Reduction Factor= 1 Mn-53 0.01 0%

Mn-54 6.668E-07 Mn-54 0.01 0.029%

Fe-55 2.161E-004 Fe-5S 0.01 9.257%

Ni-59 4.300E-07 Ni-59 0.01 0.018%

Ca-E0 2.064E-00 Ca-EQ 0.001 88.433%

Ni-E3 5.263E-o5 Ni-El 0.01 2.254%

Zn-E5 2,440E-08 Zn-E5 0.01 o0001%

1.200E-12 Se-79 Se-79 0.01 0%

Kr"-Si 2.100E-15 Kr-81 1 0%

Kr-US 1.032E-10 Kr-U5 1 3.307E-10 i 0%

St-90 Sr-90 0.01 Nb-92m 2.OO0E-15 NB-g2m 0.01 0%

Zr-93 8.600E-14 Zr-93 0.01 0%

Mo-93 1.099E-09 Mo-93 0.01 0%

Nb-94 7.499E-10 Nb-94 0.01 0%

Tc-99 2.400E-10 Tc-S9 0.01 0%

Ag-lOmn 2.158E-10 Ag-l00m 0.01 0%

1.052E-11 i Sn-31m* Sn-221m 0.01 0%

1-129 1.100E-16 I-129 0.5 0%

Ba-133 5.640E-09 Ba-133 0.01 0%

Ci-134 6.933E-09 Cs- 134 0.01 0%

Cs-135 7,800E-15 Cs-135 0.01 0%

Cs-137 3.868E-10 Cs-137 0.01 0%

Pm-145 2.696E-12 i Pm-145 0.01 0%

Smi-14E 4.300E-19 Sm-I4E 0.01 0%

Sm-151 5.740E- 10  ; Sm-151 0.01 0%

Eu-152 6.313E-008 Eu-152 0.01 0%

Eu-154 9.060E-09 Eu-354 0.01 0.0%

Eu-155 3.15gE-zo Eu-155 0.01 0%

Tb-158 5.500E-12

• Tn-i58 0.01 0%

Ho-lEGnm 2.994E-10 Ho-IE~m 0.01 0%

Hf-178m 1.108E-09 Hf-178m 0.01 3.400E-15 0%

Pb-20S Pb-20S 0.01 0%

U-233 3.400E-12 U-233 0.001 Pu-239 9.599E-11 Pu-239 0.001 Total 4.77E-04 Total 3.592E402 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction) the activity shown in 3r column (multi-colored) from the left.

Page 33 of 49 Calculation: RA-0065 Revision: 0 Addendum: N/A oDminion-Attachment E - ODCM Methodology Results Stainless Steel - Shroud - 30 Year Results Input Shroud Shutdon Shutdwn Now 05/07/2013 12:00 11/30/2016 12:00 I05/08/2043 I 30Shutdown yrs Post-0:00 "Decav Mix To" Datel 05/08/204300 3000 yrs X/Q 7.44E-0 D/Q 6.20E-0 Nudide Activity H-3 1.857E-06 2,26&04 C-14 2.491E-05 3.043E-03 IiU 6.87E-0*nmrem/yI3.O0E+03 mrem1/yrI 4.36E044 0 C/,ec C-14 0.01 Cl-3B 5.100E-07 &*219E-05 CI-36 0.5 0%

Ar-39 1.38E-07 1.9E0 Minimum 1.22E+0)2 cihr Ar-39 1 Ca'41 4.699E-09 574*47 Ca-41 0.01 0%

Mn-53 3.200E-09 $Q0 Limit Reduction Factor = 1 Mn-53 0.01 . N 0%

Mn-54 1.810E-13 i Mn-54 0.01 o0*%

Fe-55 9.494E-05 tIEOE02 Fe-55 0.01 Ni-59 1.100E-04 1.34*-0 N1-59 0.01 Co-EQ 2.516E-03 3Y3-~ Co-EQ 0.,01 Ni-US 0.01 Zn-65 1.914E-17 '*. Zn-U5 0.01 se-79 6.o98E-10 i~ Se-iS 0.01 Kr-81 1 Kr-US 1.222E-07 1 49E-0 Kr-85 1 Sr-S0 9.793E-07 t~EO Sr-SO 0.01 Nb,,., 1.2ooE-12 i NB-92m 0.01 zr-93 1.100E-o 0%

Zr-93 0.01 MO-93 9.344E-07 1.141E-04 MO-93 0.01 0%

Nb-.94 3.996E-07 4.88"E-05 Nb-94 0.01 Tc-99 1.300E-07 1.58&-0 TC-99 0.01 A.g-108m 8.490E-08 1.037E-0 Ag-10*m 0.01 Sn-inrm 3.289E-09 W.8O Sn-121m 0.01

• i 0%

1-129 0.5 o%

Ba-133 4.328E-07 58E0 Ba-133 0.01 Cs-134I 2.920E-10 Cs--134 0.01 Cs-135 0.01 0%

Cs-137 1.OO0E-06 t*222E-04 Cs-137 0.01 Pm-US5 2.749E-10 Pm-145 0.01 0%

Sm-IAE 1.O00E-16 Sm-146 0.01 0%

Sm-i51 3.651E-09 444 Sm-iS15 0.01 Eu-15. 0 It! ti* Eu-152 0.01 0:

EU-154 5.272E-08 6439£-06 Eu-154 0.01 0!i Eu-155 6.195E-09 7.567E*07 Eu-155 0.01 Th-15 1.654E-0

• Tb-1LS8 0.01 Ho-lE6m 1.573E-07 t9ZlE-05 Ho-166m 0.01 ff-178m 2.199E-08 2.68E-06 Hf-178m 0.01 Pb-205 0.01 U-233 0.01 Pu-239 6.994E-09 8.543E-0 Pu-239 Total 1.72E-02 2.107E.O Total 4.872E-402 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction) the activity shown in 3 rd column (multi-colored) from the left.

Page 34 of 49 SDominion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E ODCM Methodology Results Stainless Steel - Core Barrel- 30 Year Results Input ICore~arrel I I sro Post "Decaly Mix To",Oate 05/08/20430 30 00yrs 105/07/2013 12:001 1V30/ 2016 12:00l 05/08/2043 0:00 1 x/oQ 7.44E-0

____o/Q. 6.20E0-Nucilde Activity 11-3 1.782E-06 0%

C-14 2.989E-06 C-14 0.01 0.064%

CI1-36 6.400E-08 0.5 0%

At-39 7.405E-09 Minimum 1.12E+03 ci/hr Ar-39 1 0%

Ca-41 5.599E-10 Ca-41 0.01 0%

Mn-53 2'.100E-10 Limit Reduction Factor= 1 Mn-53 0.01 0%

Mn-54 1.002E-14 Mn-54 0.01 0%

Fe-55 1.085E-05 Fe-55 0.01 0.232%

NI-59 1.899E-05 Ni-59 0.01 0.406%

Co-6O 2.709E-04 Co-6O 0.001 57.00 1,933E-03 Ni-63 Ni-63 0.01 41.31 Zn-65 1.346E-18 Zn-O5 0.01 0%

Se-79 4.599E-11 Se-79 0.01 0%

Kr-Si 3.400E-12 Kr-S1 1 0%

Kr-85 2.731E-09 Kr-SS 1 0%

Sr-SO 2.448E-09 Sr-SO 0.01 0%

Nb-92.m 6.500E-14 NB-92m 0.01 0%

Zr-93 3.900E-12 Zr-93 0.01 0%

Mo-93 3.877E-08 Mo-93 0.01 0%

Nb-94 2.897E-08 Nb-94 0.01 0%

Tc-99 8.199E-09 TC-S9 0.01 0%

Ag-lOem 7.641E-09 Ag-00m 0.01 0%

2.330E-10 Sn-121m 0.01 Sn-121rn 0%

1-129 1.400E-14 1-129 0.5 0%

Ba-iSS 4.328E-08 Ba-ISS 0.01 0%

Cs-134 4.172E-11 Cs-134 0.01 0%

Cs.135 9.000E-13 Cs-135 0.01 0%

Cs-137 2.500E-08 Cs-137 0.01 0%

Pm-145 4.324E-11 2 Pm-145 0.01 0%

Sm-146 1.300E-17 2 Sm-140 0.01 0%

Sm-151 3.572E-09 Sm-151 0.01 0%

Eu-i52 1.912E-10 Eu-152 0.01 0%

Eu-154 5.648E-08 Eu-154 0.01 0o.o01%

Eu-155 2.115E-09 Eu-i55 0.01 0%

in-isa 1.480E-10 2 Tb-iss 0.01 0%

Ho.-16Or 1.278E-08 Ho-1O6m 0.01 0%

Kf-178m 1.432E-08 Hf-178m 0.01 0%

Pb-205 1.300E-13 2 Pb-205 0.01 0%

U-2.33 9.999E-11 U-233 0.001 0%

Pu-239 2.298E-09 Pu-239 0.001 0%

Total Z.24E-O3 Total 5.ZS4t44JZ Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction.) the activity shown in 3r column (multi-colored) from the left.

Page 35 of 49 SDom in ion Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E - ODCM Methodology Results Stainless Steel - Thermal Pads - 30 Year Results Input IThermal~ads Sht~n N~w NWShutdown I30 yrs POSt-

"Decay MixTo" Date X/Q 05/08/2043:0 7.44E.0 30.00 yrs 105/07/2013 12:00111/30/2016 12:001 05/08/2043 0:00 I

______D/Q 6.20E-0 Nuclide Activity L Ij*Ci/sec) 14.3 4.085E-07 *~cI/sec 1"1-3 0.5 0.001%

C-i4 4.583E-07 C-14 0.01 0,066%

CI-36 9.999E-09 Cl-36 0.5 0%

Ar-39 3.888E-10 Minimum 7.83E.03 Cl/hr Ar-39 1 0%

ca-41 8.s99E-11 Ca-41 0.01 0%

Mn-53 1.000E-11 limit Reduction Factor = 1 Mn-53 0.01 0%

Mn-54 5.011E-16 Mn-S4 0.01 0%

Fe-SS 1.673E-06 Fe-55 0.01 0.240%

Ni-59 2.999E-06 Ni-S9 0.01 0.430%

Co-Go 3.870E-05 Co-GO 0.001 Ni-63 3.060E-04 Ni-63 0.01 Zn-65 1.765E-19 Zn-65 0.01 0%

Se-79 4,698E-12 Se-79 0.01 0%

Kr-SiJ 3.500E-14 Kr-81 1 Kr-85 1.365E-1O0: Kr-B5 1 0%

0%

Sr-S0 1.273E-09 Sr-9O 0.01 Nb-S,,w 3.200E-15 0%

NB-92m 0.01 0%

Zr-93 2.900E-13 Zr-93 0.01 0%

MO-93 2.088E-09 MO-93 0.01 0%

Nb-94 2.997E-09 Nb-94 0.01 0%

Tc-99 4.200E-10 TC-99 0.01 9.339E-10 0%

Ag-lO~m Ag-l08m 0.01 1.16SE-11 3..o87o 0%

Sn-121m Sn-121m 0.01 0%

1-129 7.900E-16 1-129 0.5 6.492E-09 O%

Ba-133 Ba-133 0.01 7.092E-12 0%

Cs-134 Cs-134 0.01 0%

Cs-135 5. 1OE- 14 Cs-135 0.01 0%

Cs-137 1 350E-09 Cs-137 0.01 6.79SE-12 i 0%

Pm-45 Pm-145 0.01 0%

Sm-146 6.700E-19 Sm-3M6 0.01 0%

Srn151 2.302E-09 Sm-lS1 0.01 0.0302 Eu-152 3.572E-08 Eu-152 0.01 0.005%

Eu-154 5.837E-09 Eu-154 0.01 0%

7,SSE-11  !

Eu-5S5 Eu-155 0.01 0%

Tb-158 7.400E-12 Tn-1.58 0.01 0%

Ho-lO6im 1.081E-09 Ho-166m 0.01 0%

Hf-17Snm 1.585E-09 Hf-178m 0.01 0%

Pb-205 1,300E-14 Pb-205 0.01 0%

i0.001 i*

U-233 1.100E-11 U-233 0.001 0%

Pu-239 1.399E-10 Pu-239 0,001 0%

Total 3.SOE-04 Total 5.465Eo02 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fractionl the activity shown in 3d column (multi-colored) from the left.

Page 36 of 49 SDomninion Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E ODCM Methodology Results Stainless Steel - Vessel cladding - 30 Year Results Input Vessel[fladding ShutdSwntdoNo "Decay Mix To" Date X/Q 05/08/2043, 7.44E-0 C 30.00 yrs 105/0712013 12:00113130/2016 12:001 05/08/2043 0:00 I

_____ /Q 6.20E-0 Nuclide IActivity 11-3 6.127E-08 2.: 0.00 1%

C-14 6.576E-08 3.1 C-14 0.01 Cl-36 1.400E-09 6.E Cl-SO 0.5 Ar-39 S.S54E-10 2.1 Minimum 4.77E+04 CiAhr Ar-39 1 1.200E-11 Ca-41 Ca-41 0.01 o%

Mn-S3 6.9005-12 Mn-53 0.01 Limit Reduction Factor = 1 Mn-54 3.341E-16 , Mn-54 0.01 Fe-55 2.441E-07 1. Fe-55 0.01 Ni-59 4.299E-07 . Ni-59 0.01 0%

Ca-60 6.386E-06 *( Co-60 0.001 0%

Ni-63 Ni-63 0.01 0%

Zn-OS Zn-6S 0.01 Se-79 1.200E-12 Se-79 0.01 0.001%

2.100E-15 i4 Kr-81 1 Kr-8S1 Kr-8S Kr-85 1 Sr-9O 1.763E-10 a Sr-S0 0.01 0%

Nb-92m 2.000E-15 NB-92m 0.01 0%

0%

Zr-93 8.600E-14 Zr-93 0.01 Mo-93 1.093E-09 . Mo-93 0.01 Nb-94 7.492E-10 3.:. Nb-94 0.01 0%

TC-99 2.400E-10 1.1 Tc-99 0.01 0%

Ag-10*m 1.868E-10 8. Ag-lOgm 0.01 7.537E-12 0%

Sin-U2tm Sn-121m 0.01 1-129 1.100E-16 1-129 0.5 Oa-133 1.024E-09 4. Ba-133 0.01 Cs-134 9.595E-13 Cs- 134 0.01 0%

Cs-135 7.800E-15 Cs-135 0.01 Cs-*17 2.100E-10 34 Cs-137 0.01 9.575E-13 Pm-14S 0.01 Po-14S Sm-140 4.300E-19 4 Sm-14O 0.01 Sm-S15 4.6835-10 . Sm-1S1 0.01 Eu-152 1,597E-08 7!1 Eu-152 0.01 0%

Eu-1*54 1.1305-09 5S. Eu-154 0.01 0%o Eu-155 7.857E-12 1* Eu-155 0.01 o0 Tb.-I.S 4.875E-12

• Tb-i58 0.01 Ho-lO6m 2,948E-10 1,* Ho-lO6m 0.01 H4f-173m 6.1365-10 34 Hf-178m 0.01 Pb-l0S 3.400E-15 Pb-l0S 0.01 U-233 3.400E-12 U-233 0.001 Pu-239 9.5925-11 4.! Pu-239 0.01 Totul 5.07E-OS 7.4 Total 5.1585+02 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction.) the activity shown in 3 rd column (multi-colored) from the left.

Page 37 of 49 SDomninion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E ODCM Methodology Results Concrete - Inner Edge - Now Results Input Inneri~dge I I I 30 yr, P.o.st-.o..

"Decay Mix To" Datel 11/30/2016 20 3.57 yrs 0o5/07/2013 12:00!111/30/2016 12:001 05/08/2043 0:00 X/Q 7.44E-0 0/, 6.20E-0 Nuclide Activity MCI/sec rlI H-3 4.093E-06 H-3 0.5 C-14 1.699E-09 1.1468-03 C-14 0.01 0.115 0.059%

CI-36 9.800E-11 6:.609-0 CIl-3 0.5 Ar-IS 2.67sE-o8 1.8080 Minimum 6.74E+05 ci/hr Ar-39 1 C-1 1.200E-08 8.092E-03 Ca-41 0.01 Mn-SI 1.700E-14 Limit Reduction Factor= 1 Mn-SI 0.01 0.3 40.00S%

Mn-54 1.556E-09 1.0198-3 Mn-54 0.01 Fe-55 1.161E-06 2.41 Fe-55 0.01 Ni-59 16OOE-11 I* 8-0 CO-60 1.376E-07 9.2811-02 Ni-5S Co-60 0.01 0.001

-* 47 701%

Ni-63 1.949E-09 13148-03 Ni-63 0.01 0.133 0 068%

Zn-ES* 5.423E-10 3 6578-04 Zn-ES 0.01 Se-f9 2.300E-15 **** Se-f9 0.01 Kr-SI. 1.1IOOE- 16 ........... Kr-81 1 Kr-ES 3.414E-11 2.302E-O5 Kr-ES 1 Sr-SO 4.134E-11 2.788E-0S Sr-SO 0.01 o oo10%

NB-92m 0.01 Mb-92m 1.000E-14 8

• Zr-93 0.01 0%

o Mo-S3 0.01 Nb-54 7.499E-12 i

• Nb-94 0.01 Tc-SS 2.800E-13

• i! Tc-99 0.01 Ag-108m 0.01

• n-I2m 6.214E-13 i!*I Sn-121m 0.01 o~o*0%

1-125 0.5 Ba-133 1.033E-09 6.9648-04 Ba-133 0.01 Cs-1.34 2.502E-09 1.687E..03 Cs-134 0.01 o~oo0%

CS-135 9.500E-16 Cs-135 0.01 cs-137 4.6968-11 3 178O Cs-137 0.01 Pm-145 0.01 Sm-146 3.900E-19 Sm-146 0.01 Sni-151 1.362E-09 9.185E-04 Sm-151 0.01 Eu,-152' 1.9948-07 134-1 Eu-152 0.01 Eu-1,4 4.228E-08 2.851-02 EU-154 0.01 Eu-155 1.033E-09 6.94804 Eu-155 0.01 0.02 047 1"b-158 0.01 Ho-lE6m 3.2938-11 .18O Ho-166m 0.01 If-17Sni 3.139E-10 2 11780 Hf-178m 0.01 Ub-233 2.800E-12 Pb-lO5 0.01

~0%

U-233 0.001 Pu-235 4.000E-11 2.6978-05 Pu-239 0.001 Total 5,.611E..6 3.833E.00 Total 1.946E-H}2 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction) the activity shown in 3d column (multi-colored) from the left.

Page 38 of 49 9 Calculation: RA-0065 Revision: 0 Addendum: N/A Dminion-Attachment E ODCM Methodology Results Concrete - 10 cm - Now Results Input 10 cm Shutown NoW j I 30 yrs Post-Shutdown "Decay Mix To" Date 11130/2016 20 3.57 yis 105/07/2013 12:00111/30/2016 12:001 05/08/2043 0:00 I X/Q 7.44E-0 D/0 6.20E-0 Nuclide Activty 11-3 6.794E-06 p Ci/se.c 2.913%

C:-14 2,799E-09 C-14 0.01 0.060%

Cl-36 1.500E-10 Cl-36 0.5 0%

1.189E-08 Minimum 4.31E+05 CiA*r At-39 1 0.003%

Ca-41 1,900E-O8 Ca-41 0.01 0.407%

Mn-53 7.400E-15 Mn-53 0.01 0%

Limit Reduction Factor= 1 Mn-54 6.668E-10 Mn-54 0.01 0,014%

Fe-55 1.881E-06 i Fe-55 0.01 Ni-59 2.600E-11 Ni-59 0.01 0%

Co-60 2.252E-07 Co-SO 0.001 40,284%

NI-63 3.216E-09 Ni-63 0.01 0.069%

6.162E-11 Zn-65 0.01 0.00 1%

Se-79 2.800E-15 Se-79 0.01 0%

Kr-S1 5.500E-17 Kr-8i 1 0%

Kr-8S 3 335E-11 Kr-SS 1 0%

Sr-90 6 706E-11 St-90 0.01 0.00 1%

Nb-92m 1 200E-19 NS-92w, 0.01 0%

Zr-93 2 900E-14 Zr-93 0.01 0%

Mo-93 5.596E-13 MO-93 0.01 0%

Nb-94 5.299E-12 Nb-94 0.01 0%

Tc-g9 1.400E-13 Tc-99 0.01 0%

3.138E-12 Ag-lOn 0.01 0%

Sn-12im 4.971E-13 Sn-121m 0.01 0%

1-129 2.200E-17 1-129 0.5 8a-133 1.589E-09 Ba-133 0.01 0.034%

Cs-134 4.823E-09 Cs-134 0.01 0.103%

Cs-135 1.400E-15 Cs-135 0.01 0%

Cs-137 6.999E-11 Cs-137 0.01 0.002%

8.522E-12 Pm-145 0.01 Pm-145 0%

Sm-146 1.700E-19 Sm-146 0.01 0%

Sm-151 1.362E-09 Sm-151 0.01 0.029%

Eu-352 3.240E-07 Eu-152 0.01 0.946%

Eu-1S4 3.624E-08 Eu-1.54 0.01 0,777%

Eu-155 7.896E-10 Eu-155 0.01 0.0 17%

"rb-i58 9.443E-14 Tb-is8 0.01 0%

Ho--166mi 3.892E-11 Ho-166m 0.01 0%

Hf-175m 1.662E-10 H~f-178m 0.01 0.004%

2.700E-17 Pb-205 0.01 Pb-205 0%

U-233 1.200E-11 U-233 0.001 0.003%

PU-239 1.700E-11 Pu-239 0.001 0.004%

Total 9.31E-.6 Total 2.OO8EO2 Note: Fire Source Term is the container Activity Limit if burned will release (based on Fire Rsl Fraction.) the activity shown in 3d column (multi-colored) from the left.

Page 39 of 49 SDominion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E - ODCM Methodology Results Concrete - 24 cm - Now Results Input 24 cm Shto*

Shtdw I NW Shutdown 30

,O yrs Post-

"Decay Mix To" DateI 11130/201612:00 3.57 vrs 05/07/2013 12:00111130/2016 l2:00 05/08/2043 0:00 X/Q 7.44E-07 D/Q 6.20E-09 Nuclide Activity H-3 3.602E-06 ,i i 0.12 .0064%

C-14 1.499E-09 1.268E-03 C-14 0.01 Cl-3N 7.900E-11 6.683E-05 *',**12E0 rmy .0+3me/r23E0 I4 Ci/sec 0.5 Ar-39 3.072E-09 2.598E-03 Minimum 8.46E+05 Cl/hr Ar-39 1 003 0.001%

Ca-41 1.000E-08 8.459E-03 Ca-41 0.01 Mn-53 1.00E-15 N Limit Reduction Factor= 1 Mn-53 0.01 Mn-54 1.723E-10 L45TE-04 Mn-54 0.01 ~0.007%

Fe-55 0.01

• 42 608%

Ni-S9 1.400E-11 1.184E-05 Ni-59 0,01 S0%

Co-60 1.063E-07 8.996E-02 Co-60 0.01 * ~45.375%

Ni-63 1.657E-09 I!402E-03 Ni-63 0.01 7Jn-6S 2.958E-10 2.50804 Zn-65 0.01 Kr-S 7.600E-18 Se-79 0.01 Kr-81 1 Kr-US 1.350E-11 1.1428-O Kr-US 1 Sr-90 3.491E-11 2953E-0S Sr-9O 0.01 NS-92m 0.01 Zr-93 1.300E-14 " ..

r~MO-93 1.399E-13 ,* Zr-93 0.01 Mo-93 0.01 *i*! 0%

NI-94 0.01 Tc-99 3.700E-14 Tc-99 0.01 Ag-l0*m 0.01 Sn-l2lni 1.721E-13 i:

Sn-121m 0.01 1-129 0.5 Ba-133 7.943E-10 6.7208-04 Ba- 133 0.01 007 0.034%

Cs-134 2.622E-09 2 2198-03 Cs-L..34 0.01 Cs- 135 0.01 Cs-13 3.6848-11 3 116-05 Cs-137 0.01 003 0.002%

Sin-t46 4.400E-20 Pii-145 0.01 ~0%

Sm-1,6 0.01 Sin-151 5.935E-10 5.021.E-04 Sm-151 0.01 Eu-1.52 1.744E-07 14768-OX Eu-352 0.01 Eu-154 1.510E-08 1.2778-02 Eu-154 0.01 17 0.644%

eu-ISS 3.037E-10 2.5680 EU-iS5 0.01 Th-158 2.459E-14 Tb-iSa 0.01 lo-.lO=m 1.297E-11 1.0978 05 Ho-166m 0.01 if-178m 4.894E-11 4;140805 Hf-178m 0.01 004 0.002%

U-133 5.700E-13 4 Pb-2Z05 U-233 0.01

~0%

Pu-239 3.700E-12 3~- Pu-7.39 Total 4.92E-06 4.1615+00 Total 1.983E+02 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction) the activity shown in 3rd column (multi-colored) from the left.

Page 40 of 49 SDom~inion Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E ODCM Methodology Results Concrete - 55 cm - Now Results Input I55 cm I ,,.*o.I .o. 3Oyr ost-,,**

"Decay Mix Ta" Date 11/30/2016 20 X/Q 7.44E-0 3,7yrs Ios/o7/2o13 12:0o111/30/201612:0o1 05/08/2043 0:00 I

_____ O/ 6.20E-0 Nuclide Act~tvy 2.046E-07 C-14

• ,*3 2.96-07mrem/y 3.OE.+03mrem/yrI

, 1.01E÷1 i.c,/.o 3.163%

8.296E-11 (C-14 0.01 0.054%

CI-36 4.400E-12

• 7.9E-~mrm/yr3.0E+0 mrm/vr4.2E+0 I6 ci/sec Cl-36 0.5 0%

Ar-39 1.585E-10 Minimum 1.52E+07 Ci/h Ar-39 1 0.O01%

Ca-41 5.600E-10 Ca-41 0.01 0,433%

Mn-53 1.0O0E-16 Limit Reduction Factor = 1 Mn-53 0.01 0%

Mn-54 9.447E-12 Mn-54 0.01 0,007%

Fe-S5 5.603E-08 Fe-5S 0.01 43.29S%

Ni-59 7.S00E- 3. Ni-59 0.01 0%

Co-E0 5.755E-09 Co-Eo 0.001 44.474%

Ni-6U 9.551E-11 Ni-63 0.01 0.074%

Zn-ES 1.578E-11 Zn-ES 0.01 0 .012%

Se-79 7.90OE-17 Se-79 0.01 0%

Kr-81 1.70OE-20 Kr-81 1 0%

K~r-85 7.464E-13 Kr-ES 1 0%

Sr-SO 1.929E-12 Sr-SO 0.01 0.001%

Nb-92m 1.OOE-20 NB.92m 0.01 0%

Zr-93 7.OOE-16 Zr-93 0.01 0%

Mo-93 4.397E-15 Mo-93 0.01 0%

Nb-94 9.399E-14 Nb-94 0.01 0%

Tc-g9 1.30OE-15 TC-99 0.01 0%

7.159E-14 Ag-10Orm 0.01 0%

Sn-121m 1.147E-14 Sn-Ifl1m 0.01 0%

1-129 6.600E-19 1-129 0.5 Ba-133 4.528E-11 Ba-U33 0.01 0.030%

Cs-134 1.477E-10 Cs-U34 0.01 0.114%

Cs-US5 4.20OE-17 Cs-US5 0.01 0%

Cs-U?7 2.026E-12 Cs-U17 0.01 0.002%

Pm-145 2.522E-13 Pm-14S 0.01 0%

Sm-146 2.3OOE-21 Sm-14E 0.01 0%

Sm-151 3.016E-11 Sm-U11 0.01 0.02 3%

Eu-252 9.968E-09 Eu-352 0.01 7. 703%

Eu-15*4 7.550E-10 Eu-3M* 0.01 0.,563%

eu-15s 1.458E-11 Eu-155 0.01 0.0 11%

"rb-158 1,279E-15 TbIS15 0.01 0%

Hlo-166nm 5.588E-13 Ho-166m 0.01 0%

Hf-178m 1.939E-12 Hf-178m 0.01 0.00 1%

Pb-205 6.6OOE-18 Pb-205 0.01 0%

U-233 2.50OE-14 U-233 0.0O1 0%

Pu-23S 9.299E-14 Pu-239 0.0O1 0%

Total 2.78E-O7 Total 1.970E.4O2 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction.) the activity shown in 3td column (multi-colored) from the left.

Page 41 of 49 Calculation: RA-0065 Revision: 0 Addendum: N/A 9 Domini.n-Attachment E ODCM Methodology Results Concrete - Inner Edge - 30 Year Results Input jInner~dge I Shutdown j I NOW I 30yrs Post" Shutdown "De aMixTo" Date 05/08/2043 30.00 yrs 105/07/2013 12:00 112130/2016 12:001 05/08/2043 0:00 XqQj 7.44E-0 0/04 6.20E-C Nuclide I.I-3 1t-3 0.5 *.36,8%

C-14 C-14 0.01 CI-36 8I*Ci/ser Cl-36 0.5 Ar-3S Minimum l.29E+06 Ar-39 1 0.032 0.184%

9.281E-07 1.620 CIA.r Ca-41 0.01 Ca-41 Mn-53 Limit Reduction Factor =1 Mn-S3 0.01 1.694E-09 2,0186-03 Mn-54 Mn-54 0.01 Fe-55 Fe-55 0.01 Ni-59 0.01 002 Ni-59 Co-60 2.499E-18 2.203E-08 7.76-0 2*543E-0S Co-60 0.001 - 0.012%

31.360%

N1l-63 Ni-63 0.01 Zn-65 Zn-65 0.01 1.703E- 14

• Se-79 5e-79 0.01 7.795E-129 .0 -6 KCr-4 Kr-81 1 Kr-US Kr-85 1 3.060E-12 3S444E!05 Sr-90 0.01 4.254E-09 59330 Nb-92M NB-92.w 0.01 Zr-93 1.870E-10 2.407E-03 Zr-93 0.01 Mo-93 Mo-93 0.01 2.299E-15 3.9E0 Nb-94 Nb-94 0.01 Tc-99 Tc-99 0.01 6.l81E-22 1.34-0 Agioem 0.01 0.003 0.016%

Sn-2mm Sn-121m 0.01 I.200E-01 6 80... .

1-129 1-129 0.5 8.-433 2.203E-11 3.831E-05 0.01 Cs-134 1.243E-11 .8 5~ Cs- 134 0.01 Cs-135 Cs-13S 0.01 Cs-iS7 1.938E-102 .4-0 Cs-137 0.01 0.124 0 810%.

Pm-145 Pm- 145 0.01 Sm-146 Sm-146 0.01 Sm-151 3.056E-12 .5E0 Sm-151 0.01 4.4354E- 7 1 " .31.1E- 0 Eu-152 0.01 Eu-152 Eu-154 Eu-154 0.01 Eu-i55 Eu-155 0.01 Tb-158 *

• 0% .

Tn-isa 0.01 Ho-166m Ho-166ni 0.01 Hf-178m tff-178m 0.01 0.004 8.024%

Pb-20S Pb-205 0.01 U-233 U-233 0.01 Pu-239 Pu-239 0.01 Total Total 1.752E5.)Z Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction.) the activity shown in 3 rd column (multi-colored) from the left.

Page 42 of 49 SDomin ion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E - ODCM Methodology Results Concrete - 10 cm - 30 Year Results Input 10 cm I Sht~n I Nw I ° TsP$Shutdown "Decay Mix To" Datel 05/08/20430(00 30.00 yrs 105/07/2013 12:00111/30/2016 12:001 05/08/2043 0:00 I o

X/Q 7.44E-07 0/Q 6.208-09 Nducide Activity

• 4-3 1,541E-06 1.47E-06 mrem/yrj 3.100E+03 mrem/yrI 2,128+09 Ioci/ec 1H-3 0.5 C-14 2.790E-09 2.292E-03 C-14 0.01 o~o1426 CI-36 1.500E-10 1.232E-04 Ei1,31E-06 mrem/yrl3.00E.03 mrem/yrI2.28E.O08 6 ci/se CI-36 0.5 At-39 1.111E-08 9.26E-0 Minimum 8.22E+05 ci,,nr At-39 1 Ca-41 0.01 Limit Reduction Factor =1 Mn-53 0.01 0.00 o0o92%

Mn-54 3.341E-19 Mn-54 0.01 Fe-5S 2.125E-09 t.746E-03 Fe-55 0.01 Ni-59 2.599E-11 2,136E-05 Ni-59 0.01 Co-SO 6.966E-09 5.724E--03 Co-E0 0.001 Ni-U3 2.657E-09 2.83-03 Ni-63 0.01 Se-79 2,799E-15 £ Zn-U5 0.01 0.002 o0o12%

Se-79 0.01 Kr-al 5.499E-17 K~r-Si 1 Kr-85 6.037E- 12 4,96006 Kr-US 1 Sr-90 3.574E-11 2.937E-0S Sr-9O 0.01 Nb-92fi1 1.200E-19 i* NB-92.w 0.01 Zr-93 0.01 Mo-93 S.S67E-13 ..... Mo-93 0.01 Nb-94 5.29SE-12 4350*:-0i Nb-94 0.01 Tc-99 1.400E- 13 Tc-99 0.01 Ag-lO~m 0.01 Sn-l2lm 3.S63E-13 '

43S0-O 0.002%

Sn-121m 0.01 1-129 2,200E-17 *' 1-129 0.5 Sa-1.33 2.88SE-10 2.371E-04 Ba-133 0.01 C$-134 6.675E-13 , Cs-134 0.01 Cs-135 0.01 Cs-137 0.01 Pm-145 3.027E-12 Pm-145 0.01 005 0.010%

Sm-146 1.700E-19 Sm-146 0a01 Sm-151 0.01 Eu-1S2 8.196E-08 6.734E-02; Eu-152 0.01 * ~36014%

EU-154 4.519E-09 3.713-03 Eu-154 0.01 Eu-155 1.964E-11 L614E-05 Eu-155 0.01 Tb-1SS 8.357E-14 Tb-158 0.01 lo-lS~m 3.833E-11 3.149E-05 Ho-1S6m 0.01 0.003 0.010%

Hf-175m 9.204E-11 7562E05 Hf-178m 0.01 Pb-205 0.01 U-233 1.200E-11 9.859E-,6 U-233 0.001 0.010 0.0s6%

Pu-239 1.699E-11 1.3968-05 Pu-239 0.001 0.014 0.079%

Total 1.67E-06 1.3768+00 1.771E'401 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction.) the activity shown in 3r column (multi-colored) from the left.

Pane 43 of 49 SDominion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E - ODCM Methodology Results Concrete - 24 cm - 30 Year Results IShutdown Input I24 cm NO NoIhUO~

30 yrs Post-

"Deci iy Mix To" Date 05/08/2043 3.0.0 yrs 105/07/201.3 12:001 11/30/2016 12:001 05/08/2043 0:00 1 X/Q 7.44E-0 0/0 6.20E-0 Nuclide Activty (iCifsec) ifi H4-3 8.170E-07 I H4-3 0.5 7,17E-O7nemfrI30em/r3+O3mrem/yrI 4.06E+09Iocilsee C-14 1.495E-09 2.3408-'03 C-14 0.01 14oo8

• 6,97E07-06m/rem/ 3.0E+03 mrem/yrI4.35E408 eci/se*

CI-36 7.899E-11 12371E..4 Cl-36 0.5 Ar'39 2.869E-09 4.492E-O3 Minimum 1.57E+06 ci/r, Ar-39 1 Ca-41 9.999E-09 1.S6SE-02 Ca-41 0.01 Mn-53 Umit Reduction Factor = 1 Mn-53 0.01 8.631E-20  : i 074 1.244%

Mn-54 Mn-54 0.01 Fe-55 1.130E-09 1.769E-03 Fe-55 0.01 Ni-59 Ni-,59 0.01 Co-60 3.290E-09 5.1508-03 Co-60 0,001 NI-63 1.369E-09 2.143E0 Ni-63 0.01 1247E-04 o.0o01%

Zni-6 Zn-65 0.01 7,599E-18 5e-79 Se-79 0.01 Kr-81 Kr-81 1

~0%

Kr-f5 Kr-85 1 St-90 1,861E-11 2.913c-05 Sr-90 0.01 0.077 1.027%

Nb-92m N8-92m 0.01 Zr-93 1.300E- 14 Zr-93 0.01 Mo-95 Mo-93 0.01 Nb-94 1.998E-12 3J.28E-06 Nb-94 0.01 Tc-g9 3.700E-14 Tc-99 0.01 Ag-lSm Ag-l0im 0.01 Sn-121rn Sn-121m 0.01 1.233E-17 1-129 1-129 0.5 Ba-133 1.4,43E-10 2.25880 Ba-*133 0.01 Cs-134 Cs-134 0.01 Cs-135 7.500E-16 Cs-135 0.01 Cs-137 Cs-137 0.01 0.7 40.131%

Pm-14S Pm-145 0.01 Sm-146 Sm-146 0.01 * " *!*:i*

• • • 0%

5m-151 4.842E-10 7.57S9-O Sm-151 0.01 Eu-152 Eu-152 0.01 ~0,018%

Eu-1S4 1.883E-09 2.947*E-3 Eu-154 0.01 Eu-155 Eu-15S 0.01 Tb-i38 2.176E-14 Tb-158 0.01 0.0256 Ho-166m 1.278E-11 2.000E-0 Ho-166m 0.01 Hf-178m 2.710E-11 4.242E-05 lI1-17hm 0.01 Pb-205 16 S.699E-131,0E R* Pb-205 0.01 U-233 U-233 0.001 &922E-O4 0.00s%

Pu-239 3.697E-12 5.787E-06 Pu-239 0.001 o.0

.034%

Total SJI4E-07 1,384E,*00 Total 1.723E+01 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction) the activity shown in 3 rd column (multi-colored) from the left.

Paaee of 49 SDom in ion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E ODCM Methodology Results Concrete - 55 cm - 30 Year Results Input I55 cm I ShtowShtdw Nwo I30°sPost-Shutdown "Decay Mix To" Date0SO20 30.00yrs S05/07/2013 2:oo1 11/30/2016 12:00) 05/08/20430:o0oo X/Q 7.44E-0 O/0 6.0E Nuclide Activity 11-3 4.642E-08 p~ci/ser 11-3 0.5 C-14 8.270E-11 2.31.1E-03 C-14 0.01 CI-36 4.400E-12 "1229E-04 d-36 0.5 Z45.9E-E o00o1%

Ar-3 1.481E-1o 418E0 Minimum 2.79E+07 ciA*r Ar-3S 1 0.1004 0.024%

Ca-41 5.599E- 10 1.565c-02 Ca-41 0.01 Limit Reduction Factor = 1 Mn-53 0.01 Mn-54 4.733E-21 Mn-54 0.01 Fe-S5 6.329E-11 1.769E-03 Fe-55 0.01 Nt-59 7.898E-13 Z2O27E,05 Ni-59 0.01 Co-E0 1.780E-10 4.975E-03 Co-EO 0.001 Ni-63 7,891E-11 2.205E-0 Ni-63 0.01 Zn-ES 1.914E-23  !

Zn-E5 0.01 0.177 o1020

$e-79 7.897E-17 t* Se-79 0.01 IKr-Si 1.700E-20 Kr-81 1 Kr-85 1.35IE-13 3.776E-06 Kr-85 1 Sr-90 1.028E-12 2.873E-05 Sr-SO 0.01 Nb-92m 1.000E-20 NB-g2m 0.01 7.r-93 7.O00E-16 Zr-93 0.01 MO-93 4.374E-15 Mo-Si 0.01 N~b-94 9.390E-14 2,624E-06 Nb-94 0.01 Tc-99 0.01 Ag-10Im 6.197E-14 1.3t0 Ag-lOSm 0.01 Sn-121m 8.222E-15 Sn-I2lm 0.01 o~z0%

1-129 0.5 Ba-3.33 8.223E-12 2.298E-04 Ba-133 0.01 CS-134 2.044E-14 ................. Cs-134 0.01 Cs-135 0.01 Cs-137 1.100E-12 3.074E-O5 Cs-137 0.01 Pm-145 8.gS7E-14 2.503E-06 Pm-145 0.01 Sm-146 0.01 Sm-IS1 2.460E-11 6.876E-04 Sm-151 0.01 Eu-152* 2.522E-09 704E2 Eu-1.2 0.01 - 41.003%

EU-154 9.414E-11 2.63/E-0 Eu-154 0.01 Eu-155 3.626E-13 1.013E-05 Eu-155 0.01 0.001 0006%

Tb-I58 1.132E-15 Tb-158 0.01 Ho-lE6m 5.504E-13 t538E-05 I-o-l66m 0.01 002 0.009%

Hf-178m 1.074E-12 3.0011-05 Hf-178m 0.01 Pb-205 6.600E-18 Pb-205 0.01 U-233 2.500E-14 E*:*,- 7 U-233 0.ool 6.965E-04 0.004%

Pu-235 9.292E-14 2.597E-06 Pu-239 0.ool 0.003" 0.011%

Total S.OZE-.08 1A4O2E.00 Total 1.719E*01 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction.) the activity shown in 3 rd column (multi-colored) from the left.

Page 45 of 49 S---Domninion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E ODCM Methodology Results Stainless Steel - Minimum - Now Results Input SS5-Minimum-Now Shutdown Now 30htonYrSPost-

"Decay Mix To" Date1/320_120 05/07/2013 12:00 11/30/2016 12:00 05/03/2043 0:.00 3.57 yrs X/Q 7.44E-07 OIQ 6.0E j Nuclide Activty 1H-3 .883E-03 3.2018-05 O-14 2.712E-02 9.773E-05 0-14 0.01 0.010 0.003%.

Cl-SO 5.535E-04 1.995E-06 0.5 Ar-59 2.935E-05 100 Minimum 3.60E-03 CiAr Ar-39 1 Ca4 5~l.10 O0 ........ Ca-41 0.01 Mn-S3 7.053E-07 Uimit Reduction Factor= 1 Mn-55 0.01 Mn-54 7.055E-02 2.542E-04 Mn-54 0.01 Fe-55 9.1208+01 Fe-55 0.01 Ni-59 1.194E-01 432804 Ni-59 0.01 Co-6O &824E+01 Co-60 0.001 0.005 00o12%

Ni-OS Zn-OS 1.904E+01 1.0268-02

.O-0 3.6978-05 Ni-OS Zn-OS 0.01 0.01 -* So 03 Se-79 3.315E-07 " Se-79 0.01 6.861 1o17 Kr-Si 8.976E-10 Kr-81 1 Kr-ESi 4.412E-05 1.5908-07 Kr-85 1

• .*E.O* 0%

Sr-SO 4.628E-05 1.6688-07 Sr-SO 0.01 Nb-92m 2.257E-10 i*** ":

Zr-93 2.045E-08 NB-92m 0.01 Zr-93 0.01 Mo-93 1.4808-04 i334-07 Mo-93 0.01 Nb-SO 2.116E-04 7.625E-07 NI-94 0.01 Tc-9 2.962E-05 680 Tc-S9 0.01 2.7488-0 0%

Ag-1O~m 7.609E-05 2.7428-07 Ag-1O8m 0.01 Sn-I1*m 1.146E-06 *;** **""

1-129 4.702E-11 Sn-121.m 0.01 1-129 0.5 Ba-453 0.01 &6.548-04 0%

Cs- 134 0.01 Cs15 3.33 E-09 Cs-iS5 0.01 Cs-137 1.653E-04 5.9588-07 Cs-137 0.01 Pm-l4S 8.399E-07 " Pm-145 0.01 Sm-146 0.01 Sm-151 0.01 Eu-152 0.01 Eu-154 0.01 i 654E-0% 0%

Eu-155 0.01 o0%

Tb-158 0.01 Ho-lO6m 0.01 Hf-l7Sm 0.01 Pb-205 0.01 U-233 0.001 Pu-239 0.001 2.737£:-05 0%

Total 3.S78E+02 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction) the activity shown in 3 rd column (multi-colored) from the left.

Page 46 of 49 SDom in ion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E QDCM Methodology Results Stainless Steel - Minimum - 30 Year Results Input JSS-Minimum-30yr II 30yIot "Decay MixTo" Date 05/08/2043 0.00 I30.00 yrs 105/07/2013 12:00111/30/201612:0ol 05/08/2043 0:00 j

- - D/Ot 6.0E J Nucild. Ac'tivity (iC/sec) 8.452E-01 3.0851-03 0.01 .03o39 O-14 ZU 2.3OE+03nmrem/yrI3,0E+03 mrem/yrI1.3OE +00IeaCi/sec C-14 1.730E-02 6.3171-OS Cl-3O 0.5 oAo0%

At-39 8.4s3E-04 3.0861-06 Minimum 3.65E-03 cl/hr At-39 1 Ca-41 1.590E-04 5.806107 Ca-41 0.01 Mn-53 2.174E-05 Limit Reduction Factor = 1 Mn-53 0.01 ii *".oi:

!**'>** 0%i Mn-S4 1.090E-09 Mn-54 0.01 Fe-55 3.221E+00 1.16102 Fe-SS 0.01 NI-59 Ni-59 0.01 Co-60 8.41SE+01 3.04 !1,01 Co-60 0.0o1 Ni-OS3 4.918E+02  ! Ni-OS 0.01 Zn-ES Zn-OS Se-79 1.022E-05 l Se-79 Kr-81 Kr-81 1 Kr-S5 2.477E-04 9.0411-07 Kr-85 1 Sr-S0 7.643E-04 2.7901-06 0.01 Nb-92m 6.958E-09 * . NB-92m 0.01 Zr-93 Zr-93 0.01 MO-93 4.539E-03 1.657E-05 Mo-93 0.01 Nb-94 6,516E-03 2.3791-OS Nb-94 0.01 0.002 0%

Tc-g9 9. 1311-04 3.3331-06 Tc-99 0.01 3.333E-O4 0%

At -O0m 2.031E-03 7.121-06 Ag-10*m 0.01 Sn-fl21m Sn-121m 0.01 l-129 1.458E-09 1-129 0.5 a-.433 1,351E-02 4.9321-05 Ba-isS 0.01 Cs-134 Cs-134 0.01 Cs-135 1.034E-07 0.01 Cs-135 Cs-137 2.784E-03 1.016-05 0.01 Pm-45S Pm-14S 0.01 *

• 0%

1,457E-12 Sm-146 5m-140 0.01 Sm-151 1.239E-04 4S27E-*07 Sm-1S1 0.01 Eu-152 5.9701-05 ;4i  ? Eu-152 0.01 6.5221-4 0%

Eu-I.SO 1.789E-03 6.$301-06 Eu-154 0.01 Eu-3.55 Eu-155 0.01 Tb-iS8 Tb-158 0.01 Ho-lO6m Ho-lO6m 0.01 8.581E-04 o%

Hf-178m, Hf-178m 0.01 pb-205 Pb-205i 0.01 ~0%

U-233 U-233 0.001 4.458E-O5 0%

Pu-239 Pu-239 0.oo1 0 o83F04 Total Total 4.896E.02 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction.) the activity shown in 3d column (multi-colored) from the left.

Paxe 47 of 49 SDominion- Calculation: RA-0065 Revision: 0 Addendum: N/A Attachment E ODCM Methodology Results Concrete - Minimum - Now Results Input ICo-Minimum-Now ShutdowShuowdown "Decaw Mix To" Date 11/30/ 2062:0 3.57 yrS 105/07/2013 12:00111/30/2016 12:001 05/08/2043 0:00 ]

X/Q 7.*44E-07 D/Q 6.:

N~uclide 11-3 7.667E+02 I C-14 3.183E-01 1.207E, C-14 0.01 CI-36 1.794E-o2 6.002E -03 CI-36 0.5 0.003 001 Ar-39 6,705E-01 2..542E, Minimum 3.791-03 ci/Sr Ar-39 1 Ca-41 2.248E+00 8&$23E:-03 ca-41 0.01 Mn-53 4.229E-o7 Limit Reduction Factor = 1 Mn-53 0.01 Mn-54 3.995E-02 1.515, Mn-54 0.01 Fe-55 2.174E+02

• Fe-55 0.01

• 41 764%

Ni-SO 2.997E-03 1.1361 Ni-SO 0.01 Co-60 2.434E+01 9.229 -02 Co-RO - 46.759%

Ni-63 3.651E-01 1.3841, Ni-63

-03 Zn-OS Zn-OS 0.01 Se-79 Se-79 0.01 Kr-81 Kr-81 1 Kr-OS 3.157E-03 1.1971 Kr-OS 1 Sr-.9O 0.01 Nb-Olin NB-9l2n 0.01 Zr-03 0.01 M-O-9 Mo-93 0.01 Nb-94 Nb-94 0.01 Tc-OO 1,743E-11 TC-OO 0.01 Ag-lOam 7.743E-03 3.1741 Ag-lOSm 0.01 003 0.002%

Sn-lim Sn-l2.1m 0.01 1-129 2.9611E-06 ~ 4 1-129 0.5 Ba-J33 1.760E-05 ".837 Ba-1*33 0.01 Cs-134 3.975E-01 1416 C.s-134 0.01 Cs-135 Cs-135 0.01 5.198E-06 .38 Cs-137 Cs-137 0.01 0.040 02%

34028E-04 Piz-145 Pm-145 0.01 Sm-146 2.623E-03 9 6. Sm-140 0.01 1.1 0 013%

Sm-l11 Sm-1Sl1 0.01 Eu-1S2 Eu-152 0.01 83701E-03 3.17OE Eu-154 Eu-3M4 0.01 Eu-1,55 .9.61E-04 EU-1L55 0.01 T1b.1SO Tb-iSO 0.01 Hoc-1O6m Ho-lO6m 0.01 H.f-178m Hf-170m 0.01 0.00 0o1%

Pb-2OS Pb-lOS 0.01 U-23n 8.201E-03 3.995E, U-233 0.001 Pu-239 Pu-239 0.001 Total Total 1.9745402 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction.) the activity shown in 3 rd column (multi-colored) from the left.

Paag~e 48 of 49 Calculation: RA-0065 Revision: 0 Addendum:NA SDominion-Attachment E - ODCM Methodology Results Concrete - Minimum - 30 Year Results I

Input ICo-Minimum-3Oyr Shutown l Now -

30 yro Post-Shutdow 05/07/2013 12:00111/30/201612:01 05/08/2043 0:001 "Decay Mix To" Date 05/08/204300 30.00 vrs X/C[ 7.44 Nuldge Activity 11-3 3. 328E+02 I 14-S 0.5 *1460196 O14 6.072E-01 2.223E-03 mrmremyrI3.0E+03 mrem/yrI1.02E401I *cI/se Z95E+02 O-14 0.01 0.22 .333%

3.415E-02 1.l251E-04 CI-36 0.5 1.350E+00 t209t-03 Minimum 3.661-03 Cmvh Ar-39 1 004 0.025%

Ca-41 4,301E+00 1.57SE-02 Ca-41 0.01 Mn-53 Umit Reduction Factor= 1 Mn-53 0.01 Mn-54 Mn-S4 0.01 4.699E-01 172-03 0.01 Fe-55 Fe-55 o.0%

Ni-59 Ni-59 0.01 Co-GO 1.382E+00 5.0fiOE-03 Co-6O 0.001 Ndi-63 5.773E-01 2,114E-03 Ni-63 0.01 Zn-65 Zn-65 0.01 Se.79 1.707E-14 * "* - 6 Se-79 0.01 Kr-Si Kr-81 1 1.o0%

Kr-85 1.319E-10 Kr-US 1 Sr-90 Sr-90 0.01 6.089E-07 2.69.0 Nb-92M NB-92m 0.01 0.2 130.3%

Zr-93 Zr-93 0.01 4.849E-03 3.784E-06 Mo-93 S.36E-o05 2EO Mo-93 0.01 Nb-94 Nb-94 0.01 62.739E-11 229~0 Tc-B9 Tc-99 0.01 Ag -0Gm 12434E-06 4 Ag-1O0in 0.01 * .*i 0%

Sn-12lm Sn-121m 0.01 3.395E-05 .2E0 1-129 1-129 0.5 7.529E-04 2.39106 Ba-133 Ba-133 0.01 Cs-134 1.900E-05 .910 Cs-134 0.01 Cs-135 Cs-135 0.01 Cs-137 5.80E018 662E-09 Co-137 0.01 Pm-145 7.273E-02 2.97E-.03 Pm-145 0.01 002 0r009%

Sm-146 Sm-146 0.01 Sm-151 Sm-T11 0.01 Eu-1S2 Eu-152 0.01 Eu-154 Eu-154 0.01 Eu-155 Eu-155 0.01 Tb-158 Tb-i58 0.01 Ho-166mn 11o-166m 0.01 IHf-178m Hf-178m 0.01 Pb-20S Pb-205 0.01 U-233 U-rn3 0.oo1 7, O5E-O* 0.004%

Pu-239 Pu-239 0.ooi 003 0.016%

Total Total 1.668E'+01 Note: Fire Source Term is the Container Activity Limit if burned will release (based on Fire Rsl Fraction) the activity shown in 3rd column (multi-colored) from the left.

Page 49 of 49 Calculation: RA-0065 Revision: 0 Addendum: N/A.

~Dominion-Attachment F - Source Term Input vs Activity Limit Through experimentation it was discovered that both fractional and non-fractional source term input provided the same nuclide activity limits.

Based on this, the following derivation was generated to support this observation:

Non-Fractional nuclide source term for ith nuclide - 0 Fractional nuclide source term for ith nuclide = f, = 04 / ( *0Q ) or 04 = fi 0 Using Non-Fractional nuclide source term:

Non-Fractional Dose Rate for the ith nuclide: DR 01 = ai. 04 where a1 = Nuclide Constant = ( D/Q. DCF6 i + X/Q. DCFiaoi )

Total of Fractional Dose Rates (DRQ): DR 0 = *DR 0 1= [ ai" 4]

Factor to Scale up/down to 3000 mrem/yr (Fs0): FsQ = ( 3000 / DR 0 )

Nuclide Activity Limit (Q~ai): 04_aI = Fs 0Q" 1= (3000 /DRc 0)"4= ( 3000 / [ a. 0] )" 04 . ... using substitution Using Fractional nuclide source term:

Fractional Dose Rate for the ith nuclide: DR* = a1 ' fi where a, = Nuclide Constant = ( D/Q. DCFG, + X/Q. DCFiaoi )

Total of Fractional Dose Rates (DRf): DRf = *DRfi='[ai'f,]

Factor to Scale up/down to 3000 mrem/yr (Fsf): Fsf = ( 3000 / DRf )

Nuclide Activity Limit (O~fi): 0.i= Fsf"fi=( 3000 /DRf ) *f 1 = (3000 / *[ ai*f,] )"f .f...using substitution

= ( 3000 / [ a" (0 4 / ( *0,) ) ] )' (Q04/( ) ).O... -,terms willicancel out

=( 3000 / I[ai" 04] )" 0 By observation: CQ.no~ = Ctf, = ( 3000 / }j[a,'0] )" 0 = ( 3000 / *[a,'f 1 ] )" f Therefore: Whether the nuclide source term is fractional or non-fractional, the same Nuclide Activity Limit will be generated. Because the *0 constant term cancelled out and any units of measure can be converted by multiplying by a conversion constant, any conversion constants will cancel in a similar manner. The underlying, fundamental principle is proportionality of source term nuclide activities determines the Nuclide Activity Limit. Note that the above applies only to determining the Nuclide Activity Limit. If the instantaneous dose rate is desired, then the nuclide source term activity inputs must be in the proper units oflpCi/sec!