LIC-19-0001, License Amendment Request (LAR) 19-01: Independent Spent Fuel Storage Installation (ISFSI) Emergency Plan and Emergency Action Level Scheme

License Amendment Request (LAR) 19-01: Independent Spent Fuel Storage Installation (ISFSI) Emergency Plan and Emergency Action Level Scheme

ML19064A758
Person / Time
Site:	Fort Calhoun
Issue date:	02/28/2019
From:	Fisher M Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards, Office of Nuclear Reactor Regulation
Shared Package
ML19064B016	List: LIC-19-0001, License Amendment Request (LAR) 19-01: Independent Spent Fuel Storage Installation (ISFSI) Emergency Plan and Emergency Action Level Scheme
References
LAR 19-01, LIC-19-0001
Download: ML19064A758 (67)

Text

Omaha Public Power District 10 CFR 50.54(q) 10 CFR 72.212(b)(10) 10 CFR 50.90 10 CFR 72.32 10 CFR 50.47(b)

LIC-19-000 1 February 28, 2019 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Fort Calhoun Station, Unit No. 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285 Fort Calhoun Station Independent Spent Fuel Storage Installation NRC Docket No.72-054

Subject:

License Amendment Request (LAR) 19-01: Independent Spent Fuel Storage Installation (ISFSI) Emergency Plan and Emergency Action Level Scheme

References:

1. OPPD Letter (S. Marik) to USNRC (Document Control Desk) -"License Amendment Request 16-07; Revise the Fort Calhoun Station Emergency Plan to the Permanently Defueled Emergency Plan and Permanently Defueled Emergency Action Level Scheme," dated December 16, 2016 (LIC-16-0108) (ML16351A464)

2. OPPD Letter (T. Burke) to USNRC (Document Control Desk) -

"Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel," dated November 13, 2016 (LIC-16-0074) (ML16319A254)

3. Letter USNRC (J. Kim) to OPPD (M. Fisher)- "Fort Calhoun Station, Unit No. 1, Post-Shutdown Decommissioning Activities Report", dated March 23, 2017 (LIC-17-0033) (CAC No. 9536)(ML18011A687)

4. Letter USNRC (J. Kim) to OPPD (M. Fisher)- "Fort Calhoun Station, Unit No. 1, Exemptions from Certain Emergency Planning Requirements and Related Safety Evaluation", dated December 11, 2017 (LIC-16-01 09) (CAC No. MF9067)(ML17263B198; ML17263B191; ML17278A178)

Pursuant to 10 CFR 50.90, 10 CFR 50.54(q), 10 CFR 50.47(b), and 10 CFR 50, Appendix E, Omaha Public Power District (OPPD) hereby requests an amendment to Renewed Facility License Number DPR-40 for Fort Calhoun Station (FCS). The proposed amendment would replace the FCS Permanently Defueled Emergency Plan (PDEP) (Reference 1) and associated 444 SOUTH 16TH STREET MALL

• OMAHA, NE 68102-2247 EMPLOlMENT WITH EQU!.t OPPORTUNITY

Ll C-19-000 1 Page 2 Emergency Action Level (EAL) technical bases document with the Independent Spent Fuel Storage Installation Only Emergency Plan (IOEP) and its associated Independent Spent Fuel Storage Installation (ISFSI) EAL Technical Bases Document. The IOEP will be used at FCS during the period when all spent fuel is stored in the FCS ISFSI. The proposed changes are being submitted to the NRC for approval prior to implementation, as required under 10 CFR 50.54( q)( 4) and 10 CFR Part 50, Appendix E, Section IV.B.2, and 10 CFR 72.44(f).

By letter dated November 13, 2016 (Reference 2), FCS submitted a certification of permanent cessation of power operations and permanent removal of fuel from the reactor vessel.

Consequently, as specified in 10 CFR 50.82(a)(2), the station's 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel.

This proposed change reflects the complete removal of all fuel from the spent fuel pool (SFP) and permits specific reductions in the size and makeup of the Emergency Response Organization (ERO) due to the elimination of the remaining design basis accident related to spentfuel handling.

The Post-Shutdown Decommissioning Activities Report (PSDAR) (Reference 3) documented OPPD's expectation that all spent fuel would be completely transferred to the ISFSI by the end of 2022. OPPD awarded contracts in the first quarter of 2018 which expedited transferring all spent fuel to the ISFSI by the middle of 2020. To comport to the reduced scope of potential radiological accidents with spent fuel in dry cask storage within the ISFSI, OPPD determined that replacement of the FCS PDEP and EAL Technical Bases Document with the IOEP and the ISFSI EAL Technical Bases Document were warranted.

The proposed IOEP continues to rely on previously requested exemptions (Reference 4) from certain emergency planning requirements as the basis for these exemptions has not changed and remains in effect. The proposed IOEP changes have been determined to represent changes in both the EAL scheme and the staffing level previously requested to implement the PDEP in accordance with the requirements of 10 CFR 50.54(q) and therefore require NRC approval prior to implementation. to this letter contains a description, technical analysis, significant hazards determination, and environmental considerations evaluation for the proposed amendment. , Attachment 1, contains the supporting evaluations and calculations. Enclosure 1, , contains a comparison matrix of the Proposed FCS Emergency Classification System and ISFSI EALs. Enclosure 1, Attachment 3, contains the ISFSI Only Emergency Plan. , Attachment 4, contains the ISFSI Emergency Action Level Technical Bases Document.

OPPD requests review and approval of the proposed license amendment by January 31, 2020.

Once approved, the Amendment will be implemented within ninety (90) days of OPPD's submittal of a written certification to the NRC that the final spent nuclear fuel assembly has been transferred out of the SFP and placed in storage within the ISFSI.

The proposed changes have been evaluated in accordance with 10 CFR 50.91(a){1) using criteria in 10 CFR 50.92(c), and OPPD has determined that these changes involve no significant hazards.

OPPD has also determined that the proposed changes satisfy the criteria for categorical exclusion in accordance with 10 CFR 51.22(c)(9) and do not require an environmental review. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is required.

LIC-19-0001 Page 3 Pursuant to 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b), OPPD is notifying the State of Nebraska of this application for license amendment by transmitting a copy of this letter and supporting attachments to the designated state official.

If you have any questions regarding this transmittal, please contact Mr. Bradley H. Blome -

Director- Licensing & Regulatory Assurance at (402) 533-6041.

The proposed changes have been reviewed and approved by the Fort Calhoun Station Plant Operations Review Committee (PORC). This letter contains no new regulatory commitments.

I declare under penalty of perjury that the foregoing is true and correct. Executed on February 28, 2019.

Respe..ctfully, ~/ . /

'--#!~o/;Z'-e~- I~

Mary J. Fisher Vice President Energy Production and Nuclear Decommissioning MJF/jef/cac Enclosure 1: Description of Proposed Changes,. Technical and Regulatory Evaluation, Significant Hazards Determination, and Environmental Considerations c: S. A. Morris, NRC Regional Administrator, Region IV M. C. Layton, NRC Director, Division of Spent Fuel Management J. D. Parrott, NRC Senior Project Manager C. D. Steely, NRC Health Physicist, Region IV Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska

OMAHA PUB*IC POWER DISTRICT FORT CALHOUN STATION DOCKET NUMBER 50-285 I LICENSE NUMBER DPR-40 ENCLOSURE 1 DESCRIPTION OF PROPOSED CHANGES, TECHNICAL AND REGULATORY EVALUATION, SIGNIFICANT HAZARDS DETERMINATION, AND ENVIRONMENTAL CONSIDERATIONS

DESCRIPTION OF PROPOSED CHANGES, TECHNICAL AND REGULATORY EVALUATION, SIGNIFICANT HAZARDS DETERMINATION, AND ENVIRONMENTAL CONSIDERATIONS

Subject:

Independent Spent Fuel Storage Installation Only Emergency Plan (IOEP) and Emergency Action Level (EAL) Scheme

3.0 PROPOSED CHANGE

S 3.1 Elimination of SFP Initiating Conditions and EALs 3.2 Emergency Response Organization Revision 3.3 Replacement of the "Shift Manager" with the "ISFSI Shift Supervisor"

4.0 TECHNICAL EVALUATION

4.1 Radiological Consequences of Design Basis Events 4.2 Radiological Consequences of Postulated Events 4.3 ISFSI Only Emergency Plan 4.4 ISFSI Emergency Action Levels

5.0 REGULATORY EVALUATION

5.1 No Significant Hazards Consideration 5.2 Applicable regulatory Requirements/Criteria 5.3 Precedent 5.4 Conclusion

6.0 ENVIRONMENTAL CONSIDERATION

S

7.0 REFERENCES

Attachment 1, Supporting Evaluations and Calculations Attachment 2, Comparison Matrix For ISFSI EALs Based On The Proposed Regulatory Guide DG-1346 "Emergency Planning For Decommissioning Nuclear Reactors" To The Proposed FCS Emergency Classification System And ISFSI EALs Attachment 3, ISFSI Only Emergency Plan Attachment 4, ISFSI Emergency Action Level Technical Bases Document

LIC-19-0001 Page 2

1.0 INTRODUCTION

This evaluation supports a request to amend the Renewed Facility Operating License (OL) DPR-40 for Fort Calhoun Station (FCS).

By letter dated August 25, 2016, OPPD informed the NRC that FCS will permanently cease power operations in accordance with 10 CFR 50.82(a)(1)(i), specifying a shutdown date of October 24, 2016 (Reference 7.1 ). By letter dated November 13, 2016, FCS submitted a certification of permanent cessation of power operations and permanent removal of fuel from the reactor vessel (Reference 7.2). Consequently, as specified in 10 CFR 50.82(a)(2), the station's 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel.

The proposed IOEP continues to rely on previously requested exemptions from certain emergency planning requirements (Reference 7.3), as the basis for these exemptions has not changed and remains in effect. The proposed IOEP has been determined to represent changes in both the EAL scheme and the staffing level previously requested to implement the Permanently Defueled Emergency Plan (PDEP) (Reference 7.4) in accordance with the requirements of 10 CFR 50.54(q) and therefore, require NRC approval prior to implementation. Additional changes to the FCS PDEP and EAL Technical Bases Document are warranted to reflect the storage of all fuel in the Independent Spent Fuel Storage Installation (ISFSI) facility.

The Post-Shutdown Decommissioning Activities Report (PSDAR) (Reference 7.5) documented OPPD's expectation that all spent fuel would be completely transferred to the ISFSI by the end of 2022. OPPD awarded contracts in the first quarter of 2018 which expedited transferring all spent fuel to the ISFSI by the middle of 2020. To comport to the reduced scope of potential radiological accidents with spent fuel in dry cask storage within the ISFSI, OPPD determined that implementation of the IOEP and the ISFSI EAL Technical Bases Document will be warranted.

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

The NRC approved AREVA TN Americas' Amendment 14, to the standardized NUHOMS Certificate of Compliance (CoC) No. 1004 for Spent Fuel Storage Casks, on April 25, 2017 (Reference 7.6). This revision deleted the License Condition requiring a return to the SFP for inspection. With the approval of the CoC, there is no longer a requirement to return spent fuel to the SFP.

Consistent with the condition that the proposed emergency plan may be implemented ninety (90) days after all spent fuel has been certified to have been removed from the SFP, FCS has submitted a LAR to revise the FCS Facility Operating License to comport to the ISFSI-Only condition that there is no longer a requirement to return spent fuel to the SFP.

LIC-19-0001 Page 3

2.0 DESCRIPTION

The proposed amendment would modify the FCS license by replacing the existing FCS PDEP and the associated EAL scheme with the IOEP and the ISFSI EAL scheme to reflect the storage of all fuei in the ISFSI. The proposed changes reduce the scope of onsite emergency pianning requirements to reflect the reduced scope of potential radiological accidents with all spent fuel in dry cask storage within the ISFSI. After all spent fuel is in dry cask storage within the ISFSI, the number and severity of potential radiological accidents possible at FCS are substantially lower.

There continues to be no need for offsite emergency response plans at FCS because no postulated design basis accident or reasonably conceivable beyond design basis accident can result in a radioactive release that exceeds Environmental Protection Agency (EPA) Protective Action Guides (PAGs) beyond the "site boundary", as described in EPA's PAG Manual "Protective Action Guides and Planning Guidance for Radiological Incidents" dated January 2017 (EPA PAG Manual) (Reference 7.7).

The robust nature and high integrity of the spent fuel storage system selected for use at the ISFSI is designed to prevent the release of radioactivity in the event of an accident, including environmental phenomena (e.g., earthquake and flooding). As a result of the high integrity dry shielded canister's design and the substantial protection afforded the canisters, leakage of fission products from a canister is not considered to be a credible event.

The radioactive source term for an accidental release at the defueled reactor site is reduced by radioactive decay and transfer of spent fuel from the SFP to the ISFSI. Potential offsite doses were calculated at FCS to verify that the 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 (two (2) times 1500 millirem/year) at the site boundary for sixty (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 ODCM for control of effluent releases.

The PDEP EAL scheme used at FCS in is based on NEI 99-01, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6 (Reference 7.8). The proposed IOEP EAL scheme format is based on NEI 99-01, Revision 6, as appropriate after the transfer of the spent fuel from the SFP to the ISFSI. The proposed revisions constitute a change in the emergency planning function commensurate with the ongoing and anticipated reduction in radiological source term at FCS.

LIC-19-0001 Page 4

3.0 PROPOSED CHANGE

S Replacement of the FCS PDEP and associated EAL Technical Bases Document with the IOEP and the ISFSI EAL Technical Bases Document involves the following major changes to the FCS PDEP:

1) Removal of the various emergency actions related to the SFP, 2} Removal of non-ISFSI-related emergency event types,

3) Removal of the judgment EAL's

4) Clarifying definitions for security EALs

5) Revision of the Emergency Response Organization (ERO), and

6) Identification of the "ISFSI Shift Supervisor (ISS) title as the position that assumes the Emergency Director (ED) responsibilities following an emergency declaration

7) Removal of requirement to perform accountability after declaration of an emergency.

The off-normal events and accidents addressed in the IOEP are related to the dry storage of spent nuclear fuel within the ISFSI and include only the off-normal, accident, natural phenomena, and hypothetical events and consequences presented in the Updated Final Safety Analysis Report (UFSAR), NUH-003, "Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel", for the AREVA TN Americas. After all fuel is removed from the FCS SFP, there will no longer be any potential for the accidents previously described in the FCS 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 SFP. After the transfer of the spent fuel from the SFP to the ISFSI, the spent fuel storage and handling systems will be removed from operation.

The proposed revisions to the FCS emergency plan and associated EAL scheme are commensurate with the reduction in radiological hazards associated with the transfer of the spent fuel from the SFP to the ISFSI and will allow the facility to transition to an emergency plan and EAL scheme specifically related to the storage of the spent fuel in the ISFSI. The proposed changes are necessary to properly reflect the conditions of the facility and to maintain the effectiveness of the emergency plan.

3.1 Elimination of SFP Initiating Conditions and EALs and Alert Classification The Initiating Conditions (ICs) and EALs associated with emergency classification in the PDEP are based on NEI 99-01, Revision 6. Specifically, Appendix C of NEI 99-01 contains a set of ICs and EALs for permanently defueled nuclear power plants that had previously operated under a 10 CFR Part 50 license and have permanently ceased power operations.

After all spent fuel has been transferred from the SFP to dry storage within the ISFSI, the NEI 99-01, Appendix C ICs and EALs that are specifically associated with the SFP are no longer required to be in the emergency plan. Additionally, certain ICs and EALs, the primary function of which is not associated with the SFP, 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 proposed for elimination and are not included in the IOEP and EAL scheme.

With respect to the aircraft-related EALs; Interim Compensatory Measures (ICM) Order EA-02-026, Section B.5.b mitigation strategies (dated February 25, 2002) (Reference 7.9) was

LIC-19-0001 Page 5 issued and subsequent security-based ICs and EALs were provided to licensees in NRC Bulletin (BL) 2005-02, "Emergency Preparedness and Response Actions for Security Based Events," dated July 18,2005 (Reference 7.10). BL 2005-02 was addressed to all holders of operating licenses for nuclear power reactors, except those who had permanently ceased operation and had certified that fuei has been removed from the reactor vessel.

In 2009, the NRC amended its security regulations adding new security requirements pertaining to nuclear power reactors. This rulemaking established and updated generically applicable security requirements similar to those previously imposed by Commission orders issued after the terrorist attacks of September 11, 2001. In the Statements of Consideration (SOC) for the Final Rule for Power Reactor Security Requirements (7 4 Federal Register (FR) 13926; March 27, 2009), the Commission stated, in part:

"Current reactor licensees comply with these requirements through the use of the following 14 strategies that have been required through an operating license condition. These strategies fall into the three general areas identified by

§§ 50.54(hh)(2)(i), (ii), and (iii). The firefighting response strategy reflected in

§ 50. 54(hh)(2)(i) encompasses the following elements: ....

7. Spent fuel pool mitigation measures" As such, the staff maintained EALs for potential or actual aircraft threats for facilities transitioning into decommissioning with spent fuel stored in a SFP, in addition to maintaining the mitigative strategies license conditions required by NRC Order, EA-02-026, "Interim Compensatory Measures (ICM) Order," issued February 25, 2002 (67 FR 9792; March 4, 2002).

The SOC further stated, in part:

'The NRC believes that it is inappropriate that § 50. 54(hh) should apply to a permanently shutdown defueled reactor where the fuel was removed from the site or moved to an ISFSI. The Commission notes that the § 50. 54(hh) do not apply to any current decommissioning facilities that have already satisfied the§ 50.82(a) requirements."

On November 28, 2011, the NRC issued a letter that rescinded Item B.5.b of the ICM Order EA-02-26 (Reference 7.18). The rulemaking codified generically applicable security requirements previously issued by orders and updated the existing power reactor security requirements.

Neither the ICM Order nor 10 CFR 50.54(hh) continue to apply to FCS. Therefore, the ICs deleted also include those associated with the mitigative strategies and response procedures for potential or actual aircraft attack procedures as the spent fuel has been removed from the SFP and is stored in the ISFSI.

10 CFR Part 50, Appendix E (IV)(A)(7) defines "hostile action" as an act directed toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end, as it applies to the capability of implementing the emergency plan during such events. However, in the Statement of Considerations for the 2011 Emergency Plan Final Rule, the NRC excluded non-power reactors from the definition of "hostile action" because a non-power reactor as

LIC-19-0001 Page 6 defined in 10 CFR 50.2, "Definitions," is not a nuclear power plant, and presently a regulatory basis had not been developed to support the inclusion of non-power reactors in the definition of "hostile action."

Even though FCS will continue to maintain a facility license under the auspices of 10 CFR 50, the FCS ISFSI is licensed in accordance with the requirements of 10 CFR 72.212, "Conditions of General License Issued Under 10 CFR 72.210". As such, the radiological consequences to the public from the FCS ISFSI have been developed 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." The use of these regulations to develop the FCS ISFSI Technical Specifications provides corollary alignment for development of an ISFSI EAL scheme that meets the historical purpose of an Emergency Plan, protecting the public from radiological exposure in the event of a design basis accident, using the regulatory technical bases for ISFSI facilities. This technical basis also provides the foundation for development of a radiological EAL that is more in line with the standardized risk from an ISFSI.

FCS recognizes that the practice of using Emergency Planning requirements set forth under 10 CFR 50.47 for Independent Spent Storage Facilities located at operating Nuclear Power Reactors is prudent, and that prudency extends through the period that used fuel is stored in the Spent Fuel Pool for a facility that has submitted the certifications required under 10 CFR 50.82. During these periods, having an Emergency Plan and EAL scheme that is familiar to the Certified Fuel Handlers and operating staff allows for a manageable transition from power operations to removal of all fuel from the Spent Fuel Pool. Once all fuel is placed in dry storage in the ISFSI, the makeup of the facility staff can shift dramatically. This shift, concurrent with the significant reduction in risk to the public, predicates the use of an emergency plan that more closely aligns to that of an emergency plan developed under 10 CFR 72.32. The most significant difference between the proposed FCS EAL scheme and that of a decommissioning power reactor using a 10 CFR 72.32 Emergency Plan is the required use of the ALERT classification for 10 CFR 72.32 emergency plans. All other terminology is essentially the same.

This rationale justifies the exclusion of facilities with permanent removal of fuel from the reactor vessel from the definition for a "hostile action" and its related requirements (including conducting hostile action exercises) as they apply to the Emergency Plan.

Elements for security-based events should be maintained for facilities, including ISFSI-only facilities with a 10 CFR Part 50 license to help ensure assistance can be made available during these events. As such, the Alert security classification based on a "hostile action" is being redefined for the FCS IOEP.

Even though a Hostile Action-Based program is not necessary for an ISFSI-only site, precedence from other utilities and regulatory guidance provides that consideration of actions by an adversary for EAL purposes is still applicable. Therefore, the use of the term "ADVERSARIAL ACTION" and the revised definition is included, to reflect those aspects associated with an ISFSI-only site and is utilized in the EALs.

Judgements EALs are being eliminated as part of this submittal to align with Draft Regulatory Guide 1346. The draft does not include the Judgement EALs as part of the IOEP scheme.

L1 C-19-000 1 Page 7 Draft Regulatory Guide 1346 also proposes an alternate EAL for determining the occurrence of damage to a loaded storage cask following an event that may cause damage to the loaded casks . This proposed methodology identifies any change in radiation levels above normal background as the initiating condition for the EAL. FCS is proposing to base this EAL on a change in radiation ieveis significant enough to warrant concern for exceeding the limit to dose to the general public as defined in 10 CFR 20.1301(a)(2) of 0.002 Rem (2 mRem) in any one hour. Establishing an EAL threshold of >2 mRem/hr within the ISFSI protected area or on a Horizontal Storage Module (HSM) concrete surface provides a level of margin to maintain protection of the public, while providing an easily identifiable set point for ISFSI personnel. This level of radiation is high enough to minimize instrument error and operational differences while still providing positive indication of an emergency condition.

The ICs listed in Table 1 are not included in the proposed ISFSI EAL scheme for FCS. The ICs in Table 1 are either associated only with SFP operation or are ICs for which administrative controls to limit possible effluent releases have been established.

Ll C-19-000 1 Page 8 Table 1- Emergency Plan Initiating Conditions Being Deleted ALERT UNUSUAL EVENT PD-RA1 Release of gaseous or liquid PD-RU1 Release of gaseous or liquid radioactivity resulting in offsite dose greater radioactivity greater than 2 times the ODCM than 10 mRem TEDE or 50 mRem thyroid limits for 60 minutes or longer.( 1l CDE.( 1l PD-RA2 UNPLANNED rise in facility PD-RU2 UNPLANNED rise in facility radiation levels that impedes facility access radiation levels.( 1l required to maintain spent fuel integrity.( 1l PD MA~ ~GS+Ibe AG+IGN 'llitRiR tRe PD MU~ [EU1] Confirmed SECURITY GWNeR GGN+RGbbeQ AReA eF aiFB9FRe CONDITION or threat.(2) attack tRFeat witRiR 30 miRutes. [EA1

1. A SECURITY CONDITION tRat dees Ret ADVERSARIAL ACTION is occurring or has iRvelve a HGS+Ibe AG+IGN as reported occurred.] (2l by [the security supeFVisieR force and

1. [An ADVERSARIAL ACTION is occurring impacting the ISFSI].

or has occurred as reported by the security force.] A HGS+Ibe AG+IGN is 2. Notification of a credible security threat 9CCUFFiR§ 9F Aas 9CCUFFed witRiR tRe directed at the site [ISFSI].

GWNeR GGN+RGbbeQ AReA as Feperted ey secuFity supeFVisieR. 3. A validated RetificatieR fmm tRe NRG pmvidiR§ iRfeFmatieR ef aR aiFcmft tRFeat.

~- A 1.talidated RetificatieR fFem NRG ef aR aiFcFaft attack tRFeat witRiR 30 miRutes ef tRe site.

PD-HU2 Hazardous event affecting SAFETY SYSTEM equipment necessary for spent fuel cooling.( 1l PD-HA3 Other conditions exist which in the PD-HU3 Other considerations exist which in judgment of the Emergency Director warrant the judgment of the Emergency Director declaration of Alert. (1l warrant declaration of an Unusual Event. (1l PD-SU1 UNPLANNED spent fuel pool temperature rise.( 1l E MU~ [EU2]: Damage to a loaded cask CONFINEMENT BOUNDARY.

1. Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an 9R-ceRtact [abnormal] radiation reading [of >2 mRem/hr (gamma) within the ISFSI Protected Area or on a Horizontal Storage Module (HSM) concrete surface.]

LIC-19-0001 Page 9

• > 1600 mRem/hr (gamma

• neutron) on the Horizontal Storage Module (HSM) front surfaoe OR

• > 400 mRem/hr (gamma* neutron) on the HSM door oenterline OR

• > 16 mRem/hr (gamma

• neutron) on the end shield Viall exterior

( 1) Indicates the IC and the associated EALs are being deleted in their entirety.

(Z) Indicates only the portion of the IC orEAL shown in strikethrough text is being deleted. Text included with brackets []will be added in the proposed ISFSI EAL Scheme.

The ICs being deleted include all ICs associated with the categories of abnormal radioactive release and system malfunction associated with the SFP as well as security conditions associated with aircraft. These categories apply only to SFP operation and are not appropriate given the minimized risk of having all spent fuel stored within the ISFSI.

The ICs listed in Table 2, below, are being retained. The ICs being retained in the ISFSI Only Emergency Plan are appropriate to address the condition of a facility in which all spent fuel is stored in the ISFSI.

Table 2 - ISFSI Emergency Plan Initiating Conditions UNUSUAL EVENT SECURITY EU1 (formally PD-HU1) Confirmed SECURITY CONDITION, or threat, at the independent spent storage installation (ISFSI).

Independent Spent Fuel Storage Installation (ISFSI)

EU2 (formally E-HU1): Damage to a loaded cask CONFINEMENT BOUNDARY.

ALERT SECURITY EA1 (formally PD-HA1) ADVERSARIAL ACTION is occurring or has occurred.

3.2 Emergency Response Organization Revision The FCS PDEP provides for two (2) ERO augmented positions- a Technical Coordinator and a Radiation Protection Coordinator. The proposed FCS IOEP replaces these positions with a Resource Manager and an individual trained in radiological monitoring and assessment.

LIC-19-0001 Page 10 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 (ED) within two (2) hours of declaration of an Unusual Event or an Alert. Entry into the IOEP would result from an extreme natural phenomenon (beyond design basis) or a security condition, either of which wouid negativeiy impact or restrict access to the site.

The Resource Manager augments the ED by assisting in assessing the emergency condition and coordinating the required resources, including serving as the public information interface. Services provided to the ED by the Resource Manager can be provided remotely and do not necessitate an onsite response by 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. The Resource Manager's functional responsibilities could be performed in a timely manner either by reporting to the site or performing the function remotely in the specified timeframe.

In addition, FCS proposes that, for a classified event involving radiological consequences, a minimum of one person trained in radiological monitoring and assessment will report to the ISFSI within four hours of the emergency declaration.

The proposed FCS IOEP also provides that additional personnel resources may be directed to report to FCS 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 FCS staff, OPPD, and from various contractors.

3.3 Replacement of the "Shift Manager" with the "ISFSI Shift Supervisor" The FCS PDEP assigns the authority and responsibility for control and mitigation of emergencies to the Shift Manager (SM). If an emergency condition develops, the SM would assume the role of ED. The proposed FCS IOEP proposes replacing the SM position with an ISS within the IOEP.

The ISS will be at FCS on a continuous, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day basis, and is the senior management position during off-hours. This position is responsible for monitoring ISFSI conditions and managing the activities at the FCS ISFSI. This position assumes overall command and control of the response as the ED and is responsible for monitoring conditions and approving all onsite activities. The IOEP identifies non-delegable responsibilities, along with other designated tasks. OPPD considers this an administrative change which will not impact the timing or performance of existing emergency response duties.

3.4 Removal of requirement to conduct accountability following declaration of an emergency.

The specification for accountability from section J.5 of revision 1 of NUREG-0654 (Reference 7.13) reads as follows.

"Each licensee shall provide for a capability to account for all individuals on site at the time of the emergency and ascertain the names of missing individuals within 30 minutes of the start of an emergency and account for all onsite individuals continuously thereafter."

Ll C-19-000 1 Page 11 The previously approved exemptions and PDEP for FCS removed the requirements for Site Area Emergencies and General Emergencies. Accountability of personnel is a process required for the Protected Area at most nuciear piants when a Site Area Emergency or General Emergency has been declared. Accountability is necessitated at these classifications due to the potential for significant radiological exposure or other health hazards to site personnel. As the facility transitions to an ISFSI only site, the need for accountability diminishes as a result of the following:

- significantly smaller staff (less than 5% of an operational facility)

-the facility only has one building

-entry into the Protected Area is intermittent, with no permanent occupation other than that required for the security plan

-the staff at the ISFSI will be in continuous communication with each other

-the entire facility is under video surveillance, and monitored 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a day The proposed IOEP for FCS does not contain an emergency classification higher than ALERT, and considering the factors specified previously, the requirement to conduct accountability following an emergency declaration is no longer warranted.

3.5 Removal of emergency notification to the State of Iowa.

The State of Iowa Department of Homeland Security formally requested to be removed from any emergency notifications associated with FCS.

4.0 TECHNICAL EVALUATION

4.1 Radiological Consequences of Design Basis Events FCS is located midway between Fort Calhoun and Blair, Nebraska, on the west bank of the Missouri River. The site is located approximately 19 miles North of Omaha, Nebraska and four (4) miles South of Blair, Nebraska. The ISFSI is located within a Protected Area on the site. Except for the city of Blair and the villages of Fort Calhoun and Kennard, the area within a ten mile radius is predominantly rural and land use is primarily devoted to general farming. There are no private businesses or public recreational facilities on the plant property.

Chapter 14 of the FCS Final Safety Analysis Report, as Updated described the Abnormal Operational Transients and Design Basis Accident (DBA) scenarios applicable to FCS during power operations. However, after permanent cessation of power operations and transfer of all irradiated fuel from the SFP to dry storage within the ISFSI, the remaining accident scenarios postulated in the Defueled Safety Analysis Report (DSAR) are no longer possible. The ISFSI is a passive storage system that does not rely on electric power for heat transfer. After removal of the spent fuel from the SFP, there are no credible fuel-related accidents for which actions of a Certified Fuel Handler, SM, or Non-Certified Operator are required to prevent occurrence or to mitigate the consequences. There is no credible accident resulting in radioactive releases requiring offsite protective measures.

LIC-19-0001 Page 12 The robust design and construction of the spent fuel storage system 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 NUHOMS UFSAR. 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 canister's design and the additional protection afforded by the storage casks.

FCS PSDAR documents the decommissioning strategy selected for FCS. Systems that are not required to support the spent fuel, HVAC, Emergency Plan, or site security will be drained, de-energized, and secured and the plant will remain in a stable condition until final decontamination and dismantlement activities begin. The PSDAR documents the time period that OPPD expects to have all spent fuel transferred to the ISFSI. After the fuel transfer is completed, the SFP and associated systems will be drained and de-energized .

After all the spent fuel has been removed from the SFP, the estimated radiological inventory (non-fuel) that remains at the reactor facility is primarily attributable to activated reactor components and structural materials. There are no credible accident scenarios that can mobilize a significant portion of this inventory for release. As a result, the potential accidents that could occur during 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 IOEP is summarized as follows:

• The remaining radiological source term at FCS will not create an unplanned/unanticipated increase in radiation or in liquid or airborne radioactivity levels that would result in doses to the public above EPA PAG limits at the site boundary.

• Source term accumulation from activities during decontamination and dismantlement of radioactive systems, structures, and components are administratively controlled at a level that would preclude declaring an Unusual Event.

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

• The IOEP, and certain ICs and EALs for which administrative controls to limit possible effluent releases will be established, do not apply to the decontamination and dismantlement of radioactive systems, structures, and components.

NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," (NUREG-0586) (Reference 7.11) 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

Ll C-19-000 1 Page 13 while the possibility of significant radiological accident exists."

NUREG-0586 also includes the following 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 detectible nor destabilizing. Therefore, the staff makes the 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 off-normal or accident events would be implemented when decontamination or dismantling activities involving radioactive systems, structures, or components are being performed.

Implementation of the IOEP would involve FCS 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 based 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(s) (activity concentration limits)

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 requested exemptions from various emergency planning requirements contained in 10 CFR 50.47 and 10 CFR 50, Appendix E, an analysis of the potential radiological impact of a design basis accident at FCS in a permanently defueled condition indicates that any releases beyond the site boundary are below EPA PAG exposure levels. The basis for these exemptions has not changed and remains in effect for the proposed IOEP.

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

LIC-19-0001 Page 14 With spent fuel stored within the SFP, 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 any heat transfer, such that the zircaloy fuel cladding reaches ignition temperature (adiabatic heat up). The resultant zircaloy fire could lead to the release of large quantities of fission products to the atmosphere. However, after removai of the spent fuel from the SFP, the configuration of the spent fuel stored in dry storage precludes the possibility of such a scenario.

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

1. Cask Drop Event (Fuel-Related Event)

FCS is the holder of a general license for the storage of spent fuel in an ISFSI at power sites in accordance with the provisions of 10 CFR 72.210 and 10 CFR 72.212. The generally licensed ISFSI at FCS is used for interim onsite dry storage of spent nuclear fuel assemblies in the Standardized NUHOMS Horizontal Modular Storage System for Irradiated Nuclear Fuel, (Certificate of Compliance (CoC) 1004).

As documented in the NUHOMS UFSAR, NUH-003, analysis of the normal events, including drop events, determined that canister drops can be sustained without breaching the confinement boundary, preventing removal of spent fuel assemblies, or creating 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 FCS 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 beyond the site boundary would be expected to occur.

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

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 in NUREG-0586, Appendix 1). The initiator to these events could be a fire, explosion, or a handling event (cask drop). After all spent fuel has been moved to the ISFSI, there would be no concentrated source of radioactive material available to be released to the environment in an amount that could exceed two (2) times the ODCM limit at the site boundary (2 times 1500 millirem/year). During decontamination and dismantlement activities, administrative controls would limit the total amount of activity that could accumulate in a concentrated source. FCS Calculation FC08566 (Attachment 1) details 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, "Long-Lived Activation Products in Reactor Materials," (Reference 7.12) such that a release to the environment from concentrated sources of these radioactive materials would not

LIC-19-0001 Page 15 exceed two times the ODCM at the site boundary.

It is expected that representative material samples will be taken and analyzed prior to actual decontamination/dismantlement work. Using the methodology consistent with this 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 by External Events The effects of external events, such as fires, floods, wind (including tornados), earthquakes, lightning, and physical security breaches on the ISFSI remain unchanged from the effects that were considered under the proposed PDEP. Externally initiated events are addressed by the proposed ISFSI 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.

4.3 ISFSI ONLY EMERGENCY PLAN The FCS IOEP is provided in Enclosure 1, Attachment 3 to this submittal for NRC review and approval. This proposed emergency plan is associated with EALs for events related to the ISFSI. The IOEP addresses the applicable regulations stipulated in 10 CFR 50.47, "Emergency Plans;" 10 CFR 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities" (considering the exemptions requested in Reference 7.3); 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" (Reference 7.13 ).

The IOEP describes FCS's plan for responding to emergencies while all spent fuel is in dry storage within an ISFSI. After all spent fuel at FCS is in dry storage within the ISFSI, the number and severity of potential radiological accidents is significantly less than when fuel is stored in the SFP.

The FCS IOEP conservatively provides that the emergency planning zone for the ISFSI is the area within the site boundary. At FCS, 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 ISFSI for which FCS 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." FCS's analysis of the radiological impact of potential accidents at the ISFSI conclude that any releases beyond the ISFSI controlled area are expected to be less than the EPA PAGs. The controlled area is

Ll C-19-000 1 Page 16 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 with all spent 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. Exposure levels, which warrant pre-planned response measures, are limited to onsite areas. For this reason, radiological emergency planning is focused onsite.

4.4 ISFSI Emergency Action Levels Enclosure 1, Attachment 4 of this submittal provides the FCS ISFSI EAL Technical Bases Document, which contains the proposed FCS ISFSI EAL scheme for NRC review and approval. . The proposed ISFSI EAL scheme is to be implemented by the FCS ISFSI Emergency Plan (provided in Enclosure 1).

Deletions from the proposed Permanently Defueled EAL scheme are identified in Table 1, "Emergency Plan Initiating Conditions Being Deleted," in Section 3.1, "Elimination of SFPs Initiating Conditions and EALs," above.

Related Documents Supporting evaluations/calculations for establishing appropriate radioactive material administrative control limits are provided in Attachment 1 to this submittal.

Operating Modes and Applicability The proposed ISFSI EALs are only applicable after the final spent nuclear fuel assembly has been transferred out of the SFP and placed in dry 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. Prior to implementation of the EAL scheme proposed in this License Amendment Request (LAR), FCS will provide an overview of the new classification scheme to State and local emergency management officials in accordance with 10 CFR 50, Appendix E, Section IV.B.1.

5.0 REGULATORY EVALUATION

The proposed emergency plan does not meet all standards of 10 CFR 50.47(b) and requirements of 10 CFR Part 50, Appendix E. However, FCS previously received 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 December 11 , 2017 (Reference 7.3). The basis for these exemptions has not changed and

LIC-19-0001 Page 17 remains in effect for the emergency plan changes requested in this document. Considering the previously approved exemptions, the emergency plan, as revised, will continue to meet the remaining applicable requirements in 10 CFR Part 50, Appendix E and the remaining applicable planning standards of 10 CFR 50.47(b ).

5.1 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," OPPD requests NRC approval of a reduction in effectiveness of the site Emergency Plan by the removal of several EALs and corresponding changes to the emergency plan, to be implemented after all spent fuel has been removed from the SFP and placed in dry storage within the ISFSI. The proposed IOEP and ISFSI EAL Technical Bases Document are commensurate with the reduction in radiological source term at FCS.

The PSDAR documents the time period that FCS expects to have all spent fuel transferred to the ISFSI. To comport to the reduced scope of potential radiological accidents with all spent fuel in dry cask storage within the ISFSI, FCS proposes a new emergency plan and corresponding EAL scheme.

Pursuant to 10 CFR 50.92, OPPD has reviewed the proposed changes and concludes that the changes do not involve a significant hazards consideration because the proposed changes satisfy the criteria in 10 CFR 50.92( c). These criteria require that operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The proposed changes would revise the FCS emergency plan and EAL scheme commensurate with the hazards associated with a permanently shut down and defueled facility that has transferred all spent fuel from the SFP to dry cask storage within the ISFSI.

The discussion below addresses each of these criteria and demonstrates that the proposed amendment does not constitute a significant hazard.

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

LIC-19-0001 Page 18 Because FCS has permanently ceased power operations, the generation of fission products has ceased and the remaining source term continues to decay. This continues to significantly reduce the consequences of previously evaluated postulated accidents.

Therefore, the proposed change does not invoive a significant increase in the probability or consequences of an accident previously evaluated.

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 FCS.

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 SSC 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 change 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 FCS 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 spent nuclear fuel transferred out of wet storage from the SFP and placed in dry storage within the ISFSI, a fuel handling accident is no longer credible. There are no credible events that would result in radiological releases beyond the site boundary exceeding the EPA PAG exposure levels, as detailed in the EPA's PAG Manual "Protective Action Guides and Planning Guidance for Radiological Incidents" dated January 2017 (EPA 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 SSCs 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 these margins.

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

LIC-19-0001 Page 19 Based on the above, OPPD 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, considering the previously requested exemptions are discussed below.

Title 10 of the Code of Federal Regulations (10 CFR), Section 50.47, "Emergency Plans,"

set 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 ... "

• 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 change reduces 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)(1)(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 ... "

LIC-19-0001 Page 20 Section IV.C.1 of Appendix E requires that each emergency plan define the emergency classification levels that determine the extent of participation of the emergency response organization.

Section IV.E of Appendix Estates, in part: "Adequate provisions shall be made and described for emergency facilities and equipment ... ". As identified in 10 CFR 72.13, "Applicability," the applicable emergency plan requirements for an ISFSI associated with a general license are specified in 10 CFR 72.32(c) and (d).

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

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.

5.3 Precedent Similar changes to emergency plans and associated EAL schemes approved by the NRC for plants that have transitioned to ISFSI-only status include: 1) the La Crosse Boiling Water Reactor (LACBWR) facility on September 8, 2014 (Reference 7.15); 2) the Zion Facility on May 14, 2015 (Reference 7.16); and 3) Duke Energy Florida, Inc. for the Crystal River Unit 3 Nuclear Generating Station on August 12, 2016 (Reference 7.17).

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.

LIC-19-0001 Page 21

6.0 ENVIRONMENTAL CONSIDERATION

S This amendment request meets the eligibility criteria for categorical exclusion from environmental review set forth in 10 CFR 51.22(c)(9) as follows:

(i) The amendment involves no significant hazards consideration.

As described in Section 5.1 of this evaluation, the proposed changes involve no significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

The proposed changes do not involve any physical alterations to the plant configuration or any changes to the operation of the facility that could lead to a change in the type or amount of effluent release offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed changes do not involve any physical alterations to the plant configuration or any changes to the operation of the facility that could lead to a significant increase in individual or cumulative occupational radiation exposure.

Based on the above, OPPD concludes that the proposed change meets the eligibility criteria for categorical exclusion as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

7.0 REFERENCES

7.1 OPPD Letter (T. Burke) to USNRC (Document Control Desk)- "Certification of Permanent Cessation of Power Operations," dated August 25, 2016 (LIC-16-0067)

(ADAMS Accession No. ML16242A127) 7.2 OPPD Letter (T. Burke) to USNRC (Document Control Desk)- "Certifications of Permanent Cessation of Power Operations and Permanent Removal of Fuel from the Reactor Vessel," dated November 13,2016 (LIC-16-0074) (ADAMS Accession No.

ML16319A254) 7.3 Letter US NRC (J. Kim) to OPPD (M. Fisher)- "Fort Calhoun Station, Unit No. 1, Exemptions From Certain Emergency Planning Requirements and Related Safety Evaluation", dated December 11,2017 (LIC-16-0109) (CAC No. MF9067)

(ML17263B198; ML17263B191; ML17278A178) 7.4 OPPD Letter (T. Burke) to USNRC (Document Control Desk)- "License Amendment Request 16-05 to Revise the Fort Calhoun Station Emergency Plan to the Permanently Defueled Emergency Plan and Permanently Defueled Emergency Action Level Scheme," dated December 16, 2016 (LIC-16-0108) (ADAMS Accession No. ML16351A464)

LIC-19-0001 Page 22 7.5 Letter USNRC (J. Kim) to OPPD (M. Fisher) - "Fort Calhoun Station, Unit No. 1, Post-Shutdown Decommissioning Activities Report", dated March 23, 2017 (LIC 0033) (CAC No. 9536) (ML18011A687) 7.6 Nuciear Reguiatory Commission to AREVA TN Americas' CoC i004, Amendment 14, CoC, dated March 31,2017, effective April25, 2017. (ADAMS Accession No. ML17191A236) 7.7 U.S. Environmental Protection Agency, "Protective Action Guide and Planning Guidance for Radiological Incidents," dated January 2017 (PAG Manual) 7.8 Nuclear Energy Institute (NEI) 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," dated November 2012. (ADAMS Accession No.

ML12326A805) 7.9 NRC Interim Compensatory Measures (ICM) Order EA-02-026, Section B.5.b mitigation strategies (dated February 25, 2002) (ADAMS Accession No.

ML020510635) 7.10 NRC Bulletin (BL) 2005-02, "Emergency Preparedness and Response Actions for Security Based Events," dated July 18, 2005 (ADAMS Accession No. ML051740058) 7.11 NUREG-0586, "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," Supplement 1, Volume 1, dated November 2002

7. 12 NUREG/CR-347 4, "Long-Lived Activation Products in Reactor Materials," dated August 2000 7.13 NUREG-0654, FEMA-REP-1, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants,"

Revision 1, published November 1980 7.14 Letter, Mark Thaggard (USNRC) 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) 7.15 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," dated September 8, 2014 (ADAMS Accession No. ML14155A112) 7.16 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," dated May 14, 2015 (ADAMS Accession No. ML15092A423) 7.17 Memo, Office of Nuclear Security and Incident Response, Reactor Licensing Branch, Division of Preparedness and Response to Office of Nuclear Materials Safety and Safeguards, Division of Decommissioning, Uranium Recovery and Waste

LIC-19-0001 Page 23 Programs, Reactor Decommissioning Branch, "Safety Evaluation Input for the Crystal River Unit 3 Independent Spent Fuel Storage Installation Only Emergency Plan (CAC No L53129}," dated August 12, 2016 (ADAMS Accession No. ML16201A135) 7.18 Letter from U.S. Nudear Regulatory Commission for Holders of Licenses for Operating Power Reactors "Rescission or Partial Rescission of Certain Power Reactor Security Orders Applicable to Nuclear Power Plants," dated November 28, 2011 (ADAMS Accession No. ML111220447)

OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION DOCKET NUMBER 50-285 I LICENSE NUMBER DPR-40 ATTACHMENT 1 SUPPORTING EVALUATIONS AND CALCULATIONS

CC-FC-309-1 001 Revision 1 Page 1 of 1 ATTACHMENT 1 Design Analysis Cover Sheet page 1 0 f 1 Design Analysis

- I Last Page No. e 32 Analysis No.:' FC08566 Revision:

• 0 Major 12:1 MinorO Title: 3 Dose Consequences of a High Integrity Container (HIC) Drop Event EC No.: 4 70115 Revision: e 0 Statlon(s):

• FCS Component('* 14 Unit No.: e 1 Discipline:

• Nuc Descrlp. Code/Keyword: 10 RW Safety/QA Class: " Safety Related System Code: 12 NA Structure: ,. NA l CONTROLLED DOCUMENT REFERENCES 15

'

Document No.: From/To Document No.: From/To RP Calc FC-17-001 From CH-ODCM-0001 From FC08790 From Is this Desig,n Analysis Safeguards Information? 1" YesO No [gl If yes, see SY-FC-101-106

• Does this Design Ana!yals contain Unverified Assumptions? 17 YesO No~ If yes, ATI/AR#:

• This Design Ans!ysls SUPERCEDES: ,. NA In Its entirety.

~-----~-...

Description of Revision (list changed pages when all pages of original analysis were not changed): 10 Preparer: 20 Carol Waszak Prlnt Namo

~~. W;:;N~ 10/ Z."\ I)!('

Date Method of Review: 21 Detailed Review ~ ftJtemate ~lculaflni!tachecl) 0 TestingO ~

Reviewer:~ Jim Carlson Pr.ntName

~ 'hfl.~ --

._.. Sign Nlfme t_c> Lz.,_ be;*

loste f Review Notes: 2s Independent review lXI Peer/eview 0 (For Exlemal Ar.ai:~Ses Only)

External Approver: "" _NA _ _ _-..:==----

PrintName Sign Name Oa!G FCS Reviewer: 26 NA Print Name Si nName Date Independent 3rn Party Review Reqd?,. Yes 0 FCS Approver: 27 ~ry ShiA ck. /tJ-CJJ-1~

__ . _.Prlnl Namo Date

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev.O Page 2 of32 TABLE OF CONTENTS

• 1.0 PURPOSE ........................................................................................................................... 3 2.0 INPUTS ............................................................................................................................... 3 3.0 ASSUMPTIONS .................................................................................................................. 3 4.0 IDENTIFICATION OF COMPUTER PROGRAMS .............................................................. 4

5.0 REFERENCES

.................................................................................................................... 4 6.0 METHOD OF ANALYSIS .................................................................................................... 5 7.0 NUMERIC ANALYSIS ......................................................................................................... 7 8.0 RESULTS ............................................................................................................................ 9

9.0 CONCLUSION

..................................................................................................................... 9 10.0 ATTACHMENTS ............................................................................................................... 10

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. 0 Pa e 3 of 32 1.0 PURPOSE The purpose of this calculation is to determine the radiation dose to the public due to a postulated High Integrity Container (HIC) drop event. In addition, this calculation will assess if the dose is below the acceptance criteria listed below.

Acceptance Criteria:

1) Less than 1 rem TEDE over 4 days at the Control Area Boundary based on the EPA PAG (ref 5.10) for the Early Phase release .

2) Less than 10 mrem TEDE at the Control Area Boundary based on NEI 99-01 (Ref 5.11 ),

Appendix C, Table PD-1.

2.0 INPUTS

• Respirable Airborne Release Fraction is 1E-3 (See att. 1 and ref 5.8)

• BR = 3.50E-04 m3/sec - Breathing rate of reference man is in accordance with Reg Guide 1.183 (ref 5.5)

• Nuclides in a Resin Mix (Attachment 3) 10 CFR Part 61 analyses for previous resin shipments were used to create a bounding resin-HIC which represents the maximum values for observed ratios of hard to detect nuclides.

• Nuclides in Plant Mix- The non-resin HIC uses ratios previously determined in FC-1 7-001 (ref 5.3).

3.0 ASSUMPTIONS

• Damage Fraction is 1.0 (100% damage to container)(ref 5.8)

• Leakage Fraction is 1.0 (all contents of container potentially at risk)(ref 5.8)

• Maximum gross weight of HIC was used as weight of contents inside the HIC, ignoring density.

• A bounding value of 1000 curies is used as the content of the HI C. This is conservative based on the weight of the HICs used at Fort Calhoun

• No ventilation or filtration is credited for reducing release

• Release of airborne material during the event occurred as a 'puff release in which all of the material is released at once.

• 100% of the activity is available for release. Although the bulk of the activity in inert metals that may be loaded into the container are internal in the metal itself, it is tremendously conservative to assume 100% of the activity is available to be released. In reality, only the loose surface contamination would be released on a dropped container. There is no way to predict the size, shape, surface area, thickness or portion of the activity that is due to surface contamination vs fixed or internal contamination , thus the assumption is that the contents are of a granular or powder like substance where all of the activity of the material could potentially become airborne. This is likewise for resins, the radioactive particles are ionically bonded to the resin bead media.

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. 0 Pa e 4 of32

• The shortest distance from the drop location to the control area boundary is 400 meters. This is reasonable due to the Aux Building Stack distance to the control area boundary is 464.82 meters (Ref 5.12).

• A 20% factor of conservatism is applied to the summary values . For dose values, the reported mrem values are 20% greater than calculated values .

• Both HICs and other plant hardware will be analyzed as ratios can vary dramatically in resins .

4.0 IDENTIFICATION OF COMPUTER PROGRAMS XOQDOQ, Version 2 was used to calculate the X/Q. This computer program is used by the NRC in its independent meteorological evaluation of continuous and anticipated intermittent release from commercial nuclear power reactors. The program implements the assumptions outlined in Section C of NRC Regulatory Guide 1. 111. (Ref. 5. 13).

XOQDOQ, Version 2 is maintained approved in OPPD SWIMS per SCRC0000132018 .

5.0 REFERENCES

5. 1 CH-ODCM-0001 , Rev 28, Offsite Dose Calculation Manual (ODCM) 5.2 10 CFR Part 61 , Licensing Requirements for Land Disposal of Radioactive Waste.

5.3 FC-17-001 Evaluation of Instrument Response to Measured Plant Radionuclide Mix" Fort Calhoun Station Calculation, January 2017.

5.4 RG 1. 195 Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors.

5.5 RG 1.183 Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.

5.6 EPA Federal Guidance Report 11 Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation Submersion and Ingestion. (September 1988) 5.7 EPA Federal Guidance Report 12 External Exposure to Radionuclides in Air, Water, and Soil.

(September 1993) 5.8 ANL EAD 1M-53 Supplemental Analysis of Accident Sequences and Source Terms for Waste Treatment and Storage Operations and Related Facilities for the U.S. Department of Energy Waste Management Programmatic Environmental Impact Statement.

5. 9 40CFR190 EPA Environmental Radiation Protection Standards for Nuclear Power Operations.

5.10 EPA-400/R-17/001 , PAG Manual: Protective Action Guides and Planning Guidance for Radiological Incidents 5.11 NEI 99-01 , Rev 6, Development of Emergency Action Levels for Non-Passive Reactors

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. 0 Page 5 of 32 5.12 FC08790, Rev 0, Atmospheric Dispersion Factors (X/Qs) at the Decommissioning Exclusion Area Boundary (EAB) for Radiological Releases from the Fort Calhoun Station NUREG/CR-2919, XOQDOQ: Computer Program fer the Meteorological Evaluation of Routine Effluent Releases at Nuclear Power Stations (September 1982) 5.14 RG 1.145 Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants.

6.0 METHOD OF ANALYSIS 6.1 X/Q Analysis The X/Q is calculated using the same methodology as the ODCM (Ref 5.1). The long term atmospheric dispersion factor, X/Q, for normal effluent releases was used because the Unusual Event initiating condition from Ref. 5.11 was defined in terms of ODCM limits and calculations used to assess compliance with those limits use a non-accident dispersion factor.

The NRC approved computer code XOQDOQ, Version 2 (Ref. 5.13) was used to calculate the X/Q, as discussed in Section 6.4 of the ODCM (Ref 5.1 ).

6.2 Source Term All Calculations performed follow guidance set forth by RG 1.183 (ref 5.5) and RG 1.1 95 (ref 5.4).

The Integrated Activity of Release (IAR) is equivalent to the 'source term' from the DOE guide on the Supplemental Analysis of Accident Sequences and Source Terms (ref 5.8). The equation for Source term is as follows:

=

LPF Leak Path Factor, which is a term where a reduction in the source during the event is credited. This value is not credited in the evaluation and thus the value is 1.0 (1 00% of the material at risk is available to be released).

=

DF Damage Factor, which is a term for the percentage of damage to the cask. This value is not credited in the evaluation thus the value is 1.0 (1 00% damage i.e., the container is completely demolished releasing all of its contents instead of the more likely scenario of a crack forming and releasing only a fraction of its contents).

=

RARF Respirable Airborne Release Fraction is a combination of the respirable fraction (RF) which involves estimating the Aerodynamic Mean Aerosol Diameter(AMAD) compared with the particle size that can be inhaled and remain in the human body. The RF is then multiplied by the Airborne Release Fraction {ARF) which is the fraction of the material that is released into the air.

The ARF is a function of both the material composition inside the container and the type of damage that has occurred to the container. Reference 5.8 conveniently combines these values into 1 term, the RARF. The RARF for the types of materials that may be found in a HIC at FCS each have an RARF of 1E-03. Attachment 1 lists the RARFs for various materials.

MAR = Material At Risk which is the curie contents inside the container which are at risk of being released . It is conservatively assumes that all of the curie content inside the container is at risk of being released. In reality, only the loose surface contamination would be released on a dropped container. There is no way to predict the size, shape, surface area, thickness or

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. 0 Page 6 of 32 portion of the activity that is due to surface contamination vs fixed or internal contamination, thus the assumption is that the contents are of a granular or powder like substance where all of the activity of the material could potentially become airborne . This is likewise for resins, the radioactive particles are ionically bonded to the iesin bead media.

Note on selection of MAR: If dose calculations were performed individually on each radionuclide, the MAR of each individual nuclide would have to be calculated . As this document uses the DCFett value to calculate dose, only one MAR is calculated . The MAR calculated is in units of j.JCi of Total Activity as these units are required by our first set of DCFett-TA with units of mrem/

j.JCi of Total Activity.

Source Term (IAR) =The amount of radioactive material that becomes airborne and is of a respirable AMAD that can be inhaled by reference man. The units of Source Term (IAR) is identical to the units of MAR as the other terms are unitless ratios. Source Term as defined by Ref 5.8 is identical to the Integrated Activity of Release (IAR) of Ref. 5.4.

6.3 Dose equations for EAB Inhalation doses at unrestricted area boundary (CEDE and Organ):

Inbalati'Dn o1:e (nu~i::"l) =

• f't)in \ . )( ( sec)

CF!lf ( --;:::-) "lA ( C:i)" -Q - 3 "BR -

(m:$ )

• ...1

• n1

• sec Submersion doses from standing in semi-infinite cloud at unrestricted area boundary (DDE and Skin):

em* m j) x sec

• me -sion ose (mre

,

= CF'!! . (r.:u11'-J*sec

,..

1

• "fAR (!!Ci) - (3 )

• 6.4 Description of DCFett:

The effective dose conversion factor (DCFett) described is the sum of the isotope ratio-weighted dose conversion factors. By using this value, the dose equations can be simplified by performing the calculation one time instead of performing the dose calculation -21 times. Using these effective values also permit great simplification in reverse calculating curie content based on mrem value, as this also would have to be performed -21 times and then the results summed.

Using DCFett is an automatic sum of dose consequence of all nuclides including hard to detect nuclides. Additionally Am-241 in-growth is included in the decay correction of the values. The mathematical representation for DCFettiS as follows:

DCFett Based on Total Activity (DCFett-TA)

This equation also applies to submersion dose, simply by adding m 3/sec to the original DCF.

6.5 DCF values for each nuclide come directly from EPA FGR 11 (Ref 5.6) for inhalation doses and EPA FGR 12 for submersion dose. The units for inhalation DCFs in EPA FGR 11 are in Sv/Bq which can be converted to mrem/llCi by multiplying by 3.7E+09. Similarly submersion DCF values in EPA FGR 12 (Ref 5.7) are in Sv-m 3/Bq-sec which can be converted to mrem-m 3/llCi-sec by multiplying by the same factor. The converted units are listed inAtt. 4

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. 0 Pa e 7 of 32 6.6 Ratios that are an input to DCFett The method of determining the ratios of nuclide ito Total Activity is the same. Multiple lab reports are referenced. For resins, the actual lab reports are included in Att. 3. For plant mix ratios, the values can be found in FCS RP document FC-17-001 (ref 5.3). To determine the worst case ratio, each set of 10 CFR part 61 data ratios were first decay corrected (including ingrowth of Am-241 from Pu-241) to the same date. Ratios were then performed on each set of data then compared. Whichever set of data had the higher ratio was the ratio used in the

'bounding' ratio. Thus the ratio of the nuclides in the resins is not an average, but a conservative maximum. These ratios of each nuclide to the Total Activity was performed such that the values could be used in practical manner as well as theoretical.

7.0 NUMERIC ANALYSIS 7.1 X/Q:

XOQDOQ, Version 2 was run using a distance of 400 meters from the release point. The meteorological data from 2009 was used as this is the highest value that corresponds to the value in the ODCM (Ref 5.1 ).

The computer run results are contained in Attachment 5. The worst case X/Q from the computer run is 8.1 OE-05 sec/m 3 . The computer output file is contained in Attachment 6.

7.2 MAR:

The MAR is the total activity in the HIC. For the resin container the total activity (or MAR) was calculated on the attached excel sheet for the analysis. The value for the MAR for the resin container was determined to be 2.62E+0811Ci.

The value for the MAR for the plant mix container was assumed to be 1 11Ci. This was used to find the contributions from each nuclide.

7.3 IAR: The IAR input to the dose equations were calculated in this manner.

• W A='HU:: j o Al Ae

• W .ii.ru ?' .w?P'11 F Table 7.0 Ratio Source Term Contents method MAR(11Ci) RARF LPF OF (IAR} (11Ci) Notes:

rrhese values are from the 1.00E actual resin Resin TA 2.62E+08 1 1 2.62E+05 03 HICs recently shipped. See Attachment 3.

As no HIC with Plant Mix materials has been shipped, 1.00E 111Ci was Plant Mix TA 1.00E+O 1 1 1.00E-03 03 used to determine IAR per 111Ci of container contents.

Dose Consequences of a High Integrity Container (HIC} Drop Event FC08566 Rev. 0 Page 8 of 32 7.4 DCFeff equation:

DCFeff is dependent on the surrogate nuclide to which ratios are created. Additionally, since ratios differ between resin and plant mix, there needs be 2 sets of DCFeff. Thus the following equation was performed 16 times. The calculations are performed in Excel. The original DCFs along with their conversion to mrem/jJCi can be found in Att. 4.

DCFfrf = I CFi( .. ioi to s rroga e 7.5 Equation used to calculate ratios: ratios can be found in the attached excel sheet.

7.6 Decay correction equation and Am-241 ingrowth equation. Half-lives and decay corrected ratios can be found in Att. 4.

ACt" =A(:o) ~~i':t.

ln(2)

Jt(days* ) = h.alr-lire(days)

Since Pu-241 decays to Am-241, Am-241 activity will slowly build up to a maximum and then decay. No equilibrium is achieved as Am-241 has a longer half-life than Pu-241. The Activity for combined ingrowth and decay is as follows:

7.7 Dose equations: Table 7.6.1 was calculated using these equations.

CEDE (mrem) = (mrem)

CFd'f.CEDE --;::::-

l.h .. l x ( sec)

• JAR ( .C:i)* n_ -y BR -

'.!: m (ms) sec Bone (mr~;,l:)=: Cf~*b!IM (~~~) "LitR (J!::i)* ij (:;) BR ( : :)

Table 7 .6.1 Dose per Curies of Total Activity Resin Plant Mix Curies of Organ Organ Total Dose Skin Dose Skin Activity -Bone Dose TEDE -Bone Dose TEDE (Ci) (mrem) (mrem) (mrem) (mrem) (mrem) (mrem) 1000 1.43E+01 9.16E-03 2.62E+OO 8.62E+01 1.38E-02 7.10E+OO

Dose Consequences of a High Integrity Container (HI C) Drop Event FC08566 Rev. 0 Pa e 9 of 32 8.0 RESULTS The total effective dose equivalent (TEDE) at the control area boundary after a drop of a High Integrity Container (HI C) containing resin with a 1000 curies of total activity with 20%

conservatism applied is 2.62 mrem.

The total effective dose equivalent (TEDE) at the control area boundary after a drop of a High Integrity Container (HI C) containing plant components with a 1000 curies of total activity with 20% conservatism applied is 7.10 mrem.

9.0 CONCLUSION

The conclusion and interpretation of the results show that the expected resultant dose from a radioactive waste handling event (dropped HIC) of 1,000 Curies of total activity for both the isotopic mix contained in resins or the isotopic mix contained in other plant components are less than the 1 Rem Criterion for the EPA PAGs and less than the 10 mRem criterion for the NEI 99-01 guidance at the control area boundary.

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. 0 Page 10 of 32 10.0 ATTACHMENTS 10.1 Attachment 1 - RARF Factors (1 page) 10.2 Attachment 2 - HIC Specifications (13 pages) 10.3 Attachment 3- 10 CFR Part 61 Analyses (4 pages)

Two separate copies of results from resin analyses are included :

• LIMS #: 7148167 reflects the results from Resin shipment number 17-20. The analysis was performed on 10/17/2011 .

• Ll MS #: L48167 reflects the results from Resin shipment number 17-16. The analysis was performed on 01/20/2017.

10.4 Attachment 4- DCF from FGR and mrem/~Ci conversion of them (2 pages) 10.5 Attachment 5- XOQDOQ run Results (1 page) 10.6 Attachment 6- XOQDOQ output file (imbedded txt file)

TABLE D.l WM PElS Waste BARFs for LLW, LLMW, and TRUl"'l Sror.ge and Handling Sll'elllol*

MecbanicaUy Driven Releases Plre &plolivcly Driveu Releases 0

0 Free-Fall Crush Over- High (f)

(!)

CatcgoriCIISubcategoriesb Spill Impact pn:ssurization Small Large Bl.aBt Shock Presure ()

0

I (f)

l. Org. combullible liq. (!)

IE-4 IE-4 IE-4 IE-2 l.B-1 lB-1 Mass1NTEq. 6B-4 .0

a. Solutions c 4B-S 4B-S IB-4 fiB.S 6E-S 6B-S MassTNTEq. 6B-4 (!)

b. Slurries  ::I

()

2. Aqueous liquids (!)

(f)

a. Solutioos IB-4 IE-4 JH-4 2B-3 :ZS.3 IE-4 MassTNTHq. 2E-3 4B-S tB-4 IE-4 Mass TNT P.q. 6B-4 Q.

b. Slunics 48-S 28-3 28-3 Ill

3. Powder*, noncombust. 6B-4" (£4" 28-3" 6E-S" 6E-SC 1B-2" 0.2 [mass TNT Bq.f 7E-ZC I

<5.

4. Combustible solids  ::::r

a. DAW IE-3 IE-3 lB-3 SE-4 25-211 SE-4 Ma.sr TNT Bq. 18-3 t:J~

b. Plastics (incl. elast.) JB-3 IB-3 IB-3 lE-2 2B-2d lB-2 MassTNTEq. 18*3 t!,.,g c:l\-,

c. Cellulosics IE-3 IB-3 IE-3 JB-2 2B-2d . IB-2 MurTNTEq. IE*3

~

d. Polystyrene IE-3 lB-3 IB-3 lE-2 2E-:t IB-2 Mas1TNTEq. lB-3 ()

s. Metals 0

I

a. Inert lE-3 lE-3 28-3 68-5 ISB*S 7E-2 MassTNTBq. 7B-2 ors*

b. Reactive 18-3 lB-3 28-3 6B-S' J.B-28 7B-2 M1111 TNT Bq.c 1B*2 (!)

..,

6. NODCOIIlb. aggn:plce 28-llpgh 21!--11 pgh 18-3 D.2.1.1.9 D.2.1.1.9 7B-2 MassTNTBq. 1B-*2

'I

1. HSPA filters SE-4 SB-4 IB-2 lE-4 lB-4 lB-2 25-6 JB..2 _§ 0

• Stresses aa defined in Mueller et al. (1996). a "0

b Physical forms 111 defined by Mue11cr(1994). m

<

(!)

c Of material; all other& must be multiplied by !he CQC!centration of the matcrial-of-conc:em in lhe matrix. ~

d See Section D.2.2.1.

e p

• Wllllte density, g =gravitational constant (981 em/s 2), h "'height.

VJ N

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. 0 Page 12 of 32 Attachment 2- HIC Specs Usable volume Total volume Max internal (CF) Gross Empty Liner type (CF) volume (CF) dewater/solid weight (lbs) Weight (lbs)

PL8-120 FR 120.3 107.6 101 10000 650 PL 10-160 FR 145.7 129.8 121.8 9500 750 Waste Volume Waste Weight Shipment (cf) (cc) (lbs) (g) 17-16 100 2.83E+06 3350 1.52E+06 17-20 100 2.83E+06 3100 1.41E+06

FC08566 Rev. 0

~ Page13of32 Dose Consequences of a High Integrity Container (HI C) Drop Event ~

EnergySolutions HICs and Liners ENERGYSOLUTJONS Liner Codes Suffax Suffax c Conical EXM Expanded Metal Bottom BT Barrel Top F Flat CMT Cement IF Internal Foam E fu>oxy MT Empty EDX Ecodex p Powdex WM Wide Mouth ....

0 Bead Resin Polyethylene Liners Volumes/Weights

Burial Max Internal Usable Vol (CF) Gross Empty Weight Polyethylene Liners Vol (CF) Vol (CF) Dewater/Solid Wei2ht (lbs) _(lb~

PL6-80 MT 83.4 73.3 NA 7,500 500 PL6-80MTIF 83.4 64.1 NA 7,500 525 PL6-80 FR 83.4 73.3 64/NA 7,500 550 PL 6-80 FP/FEDX 83.4 73.3 62/NA 7,500 625 PL 8-120 MT 120.3 107.6 NA 10,000 600 PL 8-120 MTIF 120.3 95.8 NA 10,000 625 PL 8-120 FR 1 120.3 107.6 101 /NA 10,000 650 PL 8-120 FP/FEDX 120.3 107.6 99/NA 10,000 725 PL8-120CMT 120.3 107.6 101/NA 14,000 720 PL 10-160 MT 145.7 129.8 NA 9,500 700 PL 10-160 MTIF 145.7 115.6 NA 9,500 735 PL 10-160 FR 145.7 129.8 (121.8 9,500 750 PL 10-160 FP/FEDX 145.7 129.8 119.4 9,500 825 PL 10-160 CMT 145.7 129.8 121.8 16,000 820 PL 14-170 MT 170.8 150.3 NA 10,800 800 PL 14-170 MTIF 170.8 134.9 NA 10,800 850 PL 14-170 FR 170.8 150.3 141/NA 10,800 850 Pagel

FC08566 Rev. 0 Page 14 of32 Dose Consequences of a High Integrity Container (HIC) Drop Event ~

EnergySolutions HICs and Liners ENERGYSOLUTIONS Burial Max Internal Usable Vol (CF) Gross Empty Weight Polyethylene Liners Vol (CF) Vol (CF) Dewater/Solid Weight (lbs) (lbs)

PL 14-170 FP/FEDX 170.8 150.3 138/NA 10,800 1000 PL 14-170 CMT 170.8 150.3 141/NA 18,000 1000 PL 14-195 MT 194.1 171.4 NA 12,200 850 PL 14-195 MTIF 194.1 154.6 NA 12,200 900 PL 14-195 FR 194.1 171.4 162/NA 12,200 900 PL 14-195 FP/FEDX 194.1 171.4 159/NA 12,200 1050 PL 14-195 CMT 194.0 171.4 N/A 19,000 1,000 PL 14-215 MT 205.8 189.2 NA 13,000 1200 PL 14-215 MTIF 205.8 171.7 NA 13,000 1250 PL 14-215 FR 205.8 189.2 177/NA 13,000 1250 PL 14-215 FP/FEDX 205 .8 189.2 174/NA 13,000 1400 PL21-300 MT 314.2 285 .1 NA 18,750 1100 PL 21-300 MTIF 314.2 262.1 NA 18,750 1175 PL 21-300 FR 314.2 285 .1 269/NA 18,750 1150 PL 21-300 FP/FEDX 314.2 285 .1 264/NA 18,750 1350 60 GAL OVERPACK 10.2 8.4 NA 1,200 125 SMALL OVERPACK 28.0 23 .8 NA 2,500 250 246 GAL O.P. 36.5 32.7 NA 2,500 285 MEDIUMO.P. 38.3 33.5 NA 2,500 305 Page 2

FC08566 Rev. 0 Page 15 of 32 Dose Consequences of a High Integrity Container (HI C) Drop Event  ;:::::=-

Energy Solutions HICs and Liners ENERGYSOLUTIONS Suffix HUH HUHIC I EL Envirolene I NOTE: All liners below are empty- without internals.

Burial Max Internal Usable Vol (CF) Gross EmptyWt Empty Wt (lbs) Empty Wt (lbs)

Voi(CF) Vol (CF) Dewater/Solid Wt (lbs) (lbs) (w/Std Lift Assembly) (.,./Grapple Lift Assembly)

Radlok Series Radlok 500 135.8 111.0 110.7 9,500 680 Rad1ok 179 178.9 156.8 156.4 18,500 1,090 Rad1ok 195 195.2 172.8 172.4 18,500 1,150 Envirolene Series EL-50 51.2 41.0 4,200 600 909 EL-142 132.4 113.6 8,250 882 1,456 EL-190 174.3 150.6 14,800 1,025 1,843 EL-210 202.1 176.7 17,300 1, 138 2,072 Page3

FC08566 Rev.O Page 16 of32 Attachment 2 Dose Consequences of a High Integrity Container (HIC) Drop Event ~

EnergySolutions HICs and Liners ENERGYSOLUTIOM' Polyethylene Liner Dimensions

.  ;=i~ Max Diameter Standard Lift Max Dia meter Standard Lift Max Diameter Grapple Lift Max Diameter Grapple Lift Max Diameter Grapple Stack Lift Max Diameter Grapple Stack Lift

]J Max Bail with Bail without Bail with Bail without Bail with Bail without Polyethylene Height Centering Tabs Centering Tabs Centering Tabs Centering Tabs Centering Tabs Centering Tabs Liners (inches) (inches) (inches) (inches) (inches) (inches) (inches)

PL6-80MT 57.5 58.75 58 59.5 58 60 58.5 PL6-80MTIF 57.5 58.75 58 59.5 58 60 58.5 PL6-80 FR 57.5 58.75 58 59.5 58 60 58.5 PL 6-80 FP/FEDX 57.5 58.75 58 59.5 58 60 58.5 PL8-120MT 74.5 61.75 61 N/A 61 N/A 61.5 PL 8-120 MTIF 74.5 61.75 61 N/A 61 N/A 61.5 JPL'8-120 FR 74.5 61.75 61 NIA 61 NIA 61.5 PL 8-120 FP/FEDX 74.5 61.75 61 N/A 61 N/A 61.5 PL8-120CMT 74.5 61.75 61 N/A 61 N/A 61.5 PL 10-160MT 76.25 67.25 65.5 67.5 65.5 N/A N/A PL 10-160 MTIF 76.25 67.25 65.5 67.5 65.5 N/A N/A (PL- 10-160 FR 76.25 67.25 65 .5 67.5 65 .5 N/A N/A PL 10-160 FP/FEDX 76.25 67.25 65.5 67.5 65.5 N/A N/A PL 10-160 CMT 76.25 67.25 65.5 67.5 65.5 N/A N/A PL l4-170MT 72.75 75.25 73.5 75.5 73.5 75.5 74 PL14-170MTIF 72.75 75.25 73.5 75.5 73.5 75.5 74 PL14-170FR 72.75 75.25 73.5 75.5 73.5 75.5 74 PL 14-170FP/FEDX 72.75 75.25 73.5 75.5 73.5 75.5 74 PL 14-170CMT 72.75 75.25 73.5 75.5 73.5 75.5 74 PL 14-195 MT 79.5 76.75 75 77 75 77 75 PL 14-195 MTIF 79.5 76.75 75 77 75 77 75 PL 14-195 FR 79.5 76.75 75 77 75 77 75 PL 14-195 FP/FEDX 79.5 76.75 75 77 75 77 75 PL 14-195 CMT 79.5 76.75 75 77 75 77 75 Page4

FC08566 Rev. 0 Page 17 of32 Attachment 2 Dose Consequences of a High Integrity Container (HI C) Drop Event ~

EnergySolutions HICs and Liners ENERGYSOLUTIONS 11£::111~

'---,0-,J 1:'11 ~ Max Diameter M ax Diameter Max Diameter Max Diameter Max Diameter Max Diameter Standard Lift Standard Lift Grapple Lift Grapple Lift Grapple Stack Lift Grapple Stack Lift Max Bail with Bail without Bail with Bail without Bail with Bail without

"" Polyethylene Liners Height (inches)

Centering Tabs (inches)

Centering Tabs (inches)

Centering Tabs (inches)

Centering Tabs (inches)

Centering Tabs (inches)

Centering Tabs (inches)

PL 14-215 MT 79.5 N/A 76.125 N/A 76.375 N/A N/A PL 14-215 MTIF 79.5 N/A 76.125 N/A 76.375 N/A N/A PL 14-215 FR 79.5 N/A 76.125 N/A 76.3 75 N/A N/A PL 14-215 FP/FEDX 79.5 N/A 76.125 N/A 76.375 N/A N/A PL21-300 MT 108.5 82.75 81 82.75 8i.5 82.5 82 PL 21-300 MTIF 108.5 82.75 81 82.75 81.5 82.5 82 PL21-300 FR 108.5 82.75 81 82.75 81.5 82 .5 82 PL 21-300 FP/FEDX 108.5 82.75 81 82.75 81.5 82.5 82 60 GAL OVERPACK 35 N/A 25.5 N/A N/A N/A N/A SMALL OVERPACK 57 N/A 34 N/A N/A N/A N/A 246GALO.P. 74.25 N/A 34 N/A N/A N/A N/A MEDIUMO.P. 78 N/A 34 N/A N/A N/A N/A PageS

FC08566 Rev. 0 Page 18 of 32 Dose Consequences of a High Integrity Container (HIC) Drop Event  ;:::::=-

Energy Solutions HICs and Liners ENERGYSOLUTIONS

'

-gw.,~~

.. , Max Height Include slings (inches)

Outside Diameter (inches)

~~

Outside Height (inches)

Manway Opening Dia meter (inches)

Fill Port Opening Diameter (inches)

Lift Assembly Outside Diamtter w /Std Lift Assembl)*

(inches)

Lift Assembly Outside Diameter w/Grapplc Lift Assembly

-(inches)

Radlok Series Radlok 500 115.875 64.5 71.875 16.0 8.375 Radlok 179 119.50 73.50 72.875 19.250 10.125 Radlok 195 126.625 73.50 79.50 19.250 10.125 1'-t.. r r - ~ - ~ :c Envirolene Series EL-50 74.0 46.5 51.0 19.8 64.5 65 .0 EL-142 101.0 64.0 70.0 19.8 73.5 74.5 EL-190 107.0 73.0 71.0 19.8 75.5 76.5 EL-210 114.0 75.0 78.0 19.8 NIA N/A Note: The exact liner height and diameter can be verified by contacting EnergySolutions Liner Operations.

To verify the Dimensions the liner serial number will be required.

Page6

FC08566 Rev. 0 Page 19 of 32 Dose Consequences of a High Integrity Container (HI C) Drop Event ~

EnergySolutions HICs and Liners ENERGYSOLUTJONS Liner Codes Prefix Suffix L Carbon Steel c Conical R Resin (Bead) BT Barrel Top p Powdex CMT Cement EDX Ecodex E Epoxy F Flat EXM Expanded Metal Bottom MT Empty Steel Liners Height Diameter Burial Max Internal Usable Gross Empty Steel Liners (inches) (mches) Voi(CF) Voi(CF) Voi(CF) Wt (lbs) Wt (lbs)

L 6-80 MT 57.0 58.0 87.2 82.9 NA 9,900 1,000 L6-80CMT 57.0 58.0 87.2 82.9 NA/80 9,900 1,150 L 6-80 IN-SITU 57.0 58.0 87.2 49.8 64/NA 9,900 3,500 L6-80 FP 57.0 58.0 87.2 82.9 62/NA 9,900 1,050 L 6-80 FP/FEDX 57.0 58.0 87.2 82.9 NA 9,900 1,225 L 8-120 MT 74.0 61.0 125.2 120.2 NA 14,500 1,200 L 8-120 CMT 74.0 61.0 125 .2 120.2 NA/117 14,500 1,350 L 8-120 IN-SITU 74.5 61.0 126.0 80.3 NA 14,500 4,200 L 8-120 FR 74.0 61.0 125.2 120.2 114/NA 14,500 1,250 L 8-120 FP/FEDX 74.0 61.0 125.2 120.2 112/NA 14,500 1,325 L 14-170 TVA 73 .25 69.0 158.5 151.3 NA/147 20,750 1,450 L 14-170 MT 71.375 74.5 180.1 172.7 NA 20,750 1,550 L 14-170 CMT 71.375 74.5 180.1 172.7 NA/168 20,750 1,750 L 14-170 IN-SITU 74.0 74.5 186.7 66.1 NA 20,750 TBD L 14-170 FR 71.375 74.5 180.1 172.7 163/NA 20,750 1,600 L 14-170 FP/FEDX 71.375 74.5 180.1 172.7 160/NA 20750 1750 Page 7

FC08566 Rev. 0 Page 20 of32 Dose Consequences of a High Integrity Container (HIC) Drop Event  ;:::::=-

Energy Solutions HICs and Liners ENERGYSOLUTIONS Height Diameter Burial Max Internal Usable Gross Empty Steel Liners (inches) (inches) Vol (CF) Voi(CF) Voi(CF) Wt (lbs) Wt{lbs)

L 14-195 MT 79.0 76.0 207.4 199.6 NA 23700 1650 L 14-195 CMT 79.0 76.0 207.4 199.6 NA/195 23700 1850 L 14-195 IN-SITU 78.5 76.0 206.1 138.5 NA 23700 6300 L 14-195 FR 79.0 76.0 207.4 199.6 190/NA 23700 1700 L 14-195 FP/FEDX 79.0 76.0 207.4 199.6 187/NA 23700 1850 L 21-300 MT 108.0 82.0 330.1 320.6 NA 27250 2200 L 21-300 FP/FEDX 108.0 82.0 .330.1 320.6 303/NA 27250 2450 1-13G INSERT 43 .75 19.25 7.4 6.2 6.2/NA 5000 300 1-13 LINER 51.375 25 .0 14.6 14.6 12/NA 5000 500 3-55 LINER 109.25 34.0 57.4 57.4 52/NA 7800 945 PV-24-79 78.875 24.0 20.7 18.3 16/NA 1900 570 PV-24-72 72.0 24.0 18.8 16.6 14/NA 1750 530 PV-24-51 51.0 24.0 13.4 11.3 9/NA ll50 415 I~~efix I ISuffix O.T.

IOpen Top Height Diameter Burial Max Internal Gross Empt}

Carbon & Stainless Steel (inthes) (inches) Voi(CF) Voi{CF) Wt (lbs) Wt (lbs)

ES-50 51.00 47.25 52.00 49.30 4,200 250 ES-142 69.75 63 .50 128.3 122.20 10,000 1,100 ES-190 71.00 72.50 170.2 162.40 16,800 1,285 ES-210 78 .25 74.75 199.4 191.00 20,000 1,475 ES-210 O.T. 79.00 75.00 202.00 189.50 20,000 1,710 PageS

FC08566 Rev. 0 Page 21 of 32 Dose Consequences of a High Integrity Container (HIC) Drop Event  ;::::=-

Energy Solutions HICs and Liners ENERGYSOLUTIONS Steel Wide-Mouth Liners Non-Stackable, Wide Mouth Steel Liner, Standard and In-Situ, Slings or Grapple Large Diameter Round Lid, Bolt On Opening Gross Max Height Diameter Diameter Empty Weight Internal Steel Wide Mouth L6-80 (inches) 57 (inches) 58 (inches) 47 7/8 ..

Weight (lbs)

.

(lbs)

..

(CF) 82.9 L8-120 Ll4-170 74 721 /4 61 74 1/2 50 7/8 64 3/8 i

.. .

120.2 172.7 Ll4-195 78 1/2 76 65 7/8 199.6 Non-Stackable, Steel, Square Lid (low profile) Liner, Slings Only, Bolt on lid Opening Empty Gross Max Height Diameter Square Weight Weight Internal Steel Wide Mouth L6-80 (inches) 54 (inches) 58 (in2) 32 X 32 ..

(lbs)

..

(lbs) (CF) 82.9 L8-120 Ll4-170 71 1/2 67 7/8 61 74 1/2 36 X 36 48 x48 .. .. 120.2 172.7 Ll4-195 L21-300 75 1/2 104 1/2 76 82 48 X 48 56 X 56 . . 199.6 320.6

• Gross and empty weights are very close to the same Size MT contamer configuratiOn.

Non-Stackable, Steel Wide Mouth Liner, Round Raised Lid, Bolt On, Slings Only Opening Empty Gross Max Height Diameter Diameter Weight Weight Internal Steel Wide Mouth (inches) (inches) (inches) (lbs) (lbs) (CF)

L6-80 52 58 43 1102 10,052 82.5 L8-120 69 1/2 61 46 1275 14,675 118 Ll4-170 67 3/8 74112 59 112 1830 21 ,030 175 L14-195 73 112 76 61 1945 22,050 198 L21-300 102 1/2 82 67 2550 27,600 317 Page9

FC08566 Rev. O Page 22 of 32 Dose Consequences of a High Integrity Container (HI C) Drop Event ~

EnergySolutions HICs and Liners ENERGYSOLUTIONS Liner & Cask Compatibility Table Max Max . Burial Height Diameter Volume Cask Shield Equiv Liners (i.n~hes} {IJM:bes) (CF) Cask Compatibility & Max Rllp-PL 6-80 57.5 60 83.4 6-80 5.0"/1.860R/hr EL-50 51.0 47.0 51.2 6-80 5.0"/1.860R/hr PL 8-120 74.5 61.75 120.3 8-120AorB 4.5"/880R/hr.

EL-142 70.0 64.5 132.4 EL-142- will not fit in 8-120 cask because ofht. 8-120 3.13"/78 R/hr.

liner will not fit in OH-142 cask because of diameter. (I0-160B)

Difference bet. I0-160B interior hgt and liner is 7".

PLI4-170 72.75 75.5 170.8 14-170 2.13"/15R/hr.

NUHIC-136 71 65 136.3 14-170 w/shoring 2.13"/15R/hr.

EL-190 71 73 .5 174.3 14-170 series II or III 2.13"/15R/hr.

Radlok P-500 71.0 71.1 163.3 14-170 2.13"/15R/hr.14-195 2.75"/21R/hr.14-215 w/shoring 2.73"/20R/hr.

PL 14-195 79.5 76.75 194.1 14-195 2.75"/21R/hr.

EL-210 78.0 75.5 176.7 14-195 2.75"/21R/hr.14-215 2.73"/20R/hr.

Radlok P-195 79.5 73.5 195.2 14-195 2.75"/21R/hr.14-215 2.73"/20R/hr.

PL 14-215 79.5 76.75 205.8 14-215 2.73"/20R/hr.

EL-210 78.0 75 .5 202.1 14-195 2.75"/21R/hr.14-215 2.73"/20R/hr.

Radlok P-179 72.9 73 .5 178.9 14-195 2.75"/21R/hr.14-215 2.73"/20R/hr.

Radlok P-195 79.5 73 .5 195.2 14-215 2.73"/20R/hr.

PL 21-300 108.5 82.75 314.2 21-300 1.5"/1.9R/hr.

• Note: If using grapple bails, or steel inserts, listed liner may not fit. Consult Transportation and Liner Operations prior to ordering a shipping cask.

Page 10

FC08566 Rev. 0 Page 23 of32 Dose Consequences of a High Integrity Container (HIC) Drop Event ~

EnergySolutions HICs and Liners ENERGYSOLUTIONS Max Opening Max Internal Burial Empty Max Height Diameter Diameter Volume Volume Weight Liner (OP-Overpack) (incheS) (inches) (inches) (CF) (CF) Obs) 60 gal (OP) 35 25.5 NA 8.4 10.2 125 Small OP 57 34 NA 23.8 28 250 MediumOP 78 34 NA 33.5 38.3 305 246 g_al OP 74.25 34 NA 32.7 36.5 285 PL 6-80 57.5 60 22.5 73.3 83 .4 500 1-13G 43 .75 19.25 NA 6.2 7.4 300 1-13 51.375 25 NA 12.5 14.6 500 3-55 72 24 NA 16.6 18.8 945 NUHIC-55 (OP) 41.62 31 27 14.8 18.8 65 EL-50 51.0 47.0 25 41 51.2 500 PL 8-120 74.5 61.75 22.5 107.6 120.3 600 EL-142 70.0 64.5 25 113.6 132.4 650 P[l0-160) 76.25 67.5 22.5 129.8 11 45.8 825 PL 14-170 72.75 75.25 22.5 150.3 170.8 800 NUHIC-136 71 65 22 127 136.3 600 EL-190 71 73.5 25 150.6 174.3 800 Radlok P-500 71.0 71.1 163.3 695 PL 14-195 79.5 76.75 22.5 171.4 194.1 850 PL 14-215 79.5 76.75 22.5 189.2 205.8 1200 EL-210 78.0 75.5 25 202.1 900 Radlok P-179 72.9 73.5 179.4 1090 Radlok P-195 79.5 73.5 195.7 1150 PL 21 -300 108.5 82.75 22.5 285.1 314.2 llOO Papll

FC08566 Rev. 0 Page 24 of32 Dose Consequences of a High Integrity Container (HIC) Drop Event  ;:::::=-

Energy Solutions HICs and Liners ENERGYSOLUTION.S' Cask and Liner Compatibility Table

-  ;-~~ Shielding Equiv/Max Rad Level Limit I

Internal Dimension (Hf'ight x Diametf'r) Payload

[!ll* *m""i n Cask (Ribr) Type *(inches) Obs) ~ Compatible Liners ~'

6-80-2 5.0/1,860 DOT-7A 58 X 59 7,500 6-80 6/100L 6.00/ DOT-7A 62 X 61 12,000 6-80 6/100H 6.00/ DOT-7A 62 X 61 12,000 6-80 8-120 4.5/880 DOT-7A 75 X 62 20,000 8-120, 6-80*

8-120 4.5/880 B 75 X 62 14,680 8-120, 6-80*

101140 3.6/ DOT7A 73 X 66 15,000 6-80*

10-140MB 3.25/ B 73 x66 15,000 6-80*

10-142 4.25/350 CoC Expired 72x66 10,000 6-80*

10-160 3.13/78 B 75 x67 14,500 8-120, 6-80*

14-170-II 2.13115 DOT-7A 73.25 X 75.5 14,000 14-170, 14-150*, 7-100*, 6-80*

14-170 III 2.13115 DOT-7A 73.25 X 75.5 17,500 141190L 2.0017 DOT-7A 73.38 X 75.5 20,000 14-170, 14-150*, 7-100*, 6-80*

14/190M 2.25110 DOT-7A 73 .38 X 75.5 20,000 14-170 14-150* 7-100* 6-80*

14/190H 3.50/150 DOT-7A 73.38 X 75.5 20,000 14-170, 14-150*, 7-100*, 6-80*

14-190H 3.50/60 DOT-7A 73 .38 X 75.5 20,000 14-170 14-150* 7-100* 6-80*

14-195H 2.75/21 DOT-7A 80 X 77 17,700 14-215, 14-195, 14-170*, 14-150*,

8-120*, 7-100* 6-80*

14/210L 2.00/7 DOT-7A 80.25 X 77.25 20,000 14-215, 14-195, 14-170*, 14-150*,

8-120*, 7-100* 6-80*

14/210H 2.73/20 DOT-7A 80.25 X 77.25 20,000 14-215, 14-195, 14-170*, 14-150*,

8-120* 7-100* 6-80*

14-215H 2.73/20 DOT-7A 80.25 X 77.25 20,000 14-215, 14-195, 14-170*, 14-150*,

8-120*, 7-100* 6-80*

21-300 1.5/1.9 DOT-7A 109 X 83 27,250 21-300, 14-215*14-195*, 14-170*,

14-150*, 8-120*, 7-100*, 6-80*

• Proper shoring must be installed in cask.

Page 12

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. O Page 25 of32 Attachment 3 - 1 0 CFR part 61 Analyses Page 1 of4 Rc:port ~~r Analysis l~OliiiU* !

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Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev.O Page 26 of 32 Attachment 3-10 CFR part 61 Analyses Page 2 of4 Rerort of Analy~is

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Dose Consequences of a High Integrity Container (HI C) Drop Event FC08566 Rev. 0 Page 29 of 32 Attachment 4 - DCF from FGR and mrem/JJ.Ci Conversion Page 1 of2 EPA FGR 11, Table 2.1- Inhalation Dose Factors

!conversion! 3.70E+09lmrem-Bq/Sv-uCi I CEDE Bone Lung CEDE Bone Lung Nuclide Sv/Bq Sv/Bq Sv/Bq mrem/uCi mrem/uCi mrem/uCi C-14 5.64E-10 5.64E-10 5.64E-10 2.09E+00 2.09E+OO 2.09E+OO Fe-55 7.26E-10 5.14E-10 1.06E-09 2.69E+00 1.90E+OO 3.92E+OO Ni-59 3.58E-10 3.51E-10 1.20E-09 1.32E+OO 1.30E+OO 4.44E+OO Ni-63 8.39E-10 8.22E-10 3.07E-09 3.10E+OO 3.04E+OO 1.14E+01 Sr-89 1.12E-08 8.37E-09 8.3SE-08 4.14E+01 3.10E+01 3.09E+02 Sr-90 3.51E-07 7.27E-07 2.86E-06 1.30E+03 2.69E+03 1.06E+04

.Tc-99 2.25E-09 4.52E-11 1.67E-08 8.33E+00 1.67E-01 6.18E+01 Cr-51 9.03E-11 2.74E-11 5.34E-10 3.34E-01 1.01E-01 1.98E+OO Mn-54 1.81E-09 2.56E-09 6.66E-09 6.70E+OO 9.47E+OO 2.46E+01 Co-58 2.94E-09 6.93E-10 1.60E-08 1.09E+Ol 2.56E+OO 5.92E+01 Fe-59 4.00E-09 2.91E-09 1.38E-08 1.48E+01 1.08E+01 5.11E+01 Co-60 5.91E-08 1.35E-08 3.4SE-07 2.19E+02 S.OOE+01 1.28E+03 Zn-65 5.51E-09 3.36E-09 2.10E-08 2.04E+01 1.24E+01 7.77E+01 Nb-95 1.57E-09 2.42E-09 8.32E-09 5.81E+OO 8.9SE+OO 3.08E+01 Zr-95 6.39E-09 1.03E-07 4.07E-08 2.36E+01 3.81E+02 1.51E+02 Ag-llOm 2.17E-08 5.19E-09 1.20E-07 8.03E+01 1.92E+01 4.44E+02 Sb-124 6.80E-09 3.41E-09 4.14E-08 2.52E+01 1.26E+01 1.53E+02 Sb-125 3.30E-09 2.73E-09 2.17E-08 1.22E+01 1.01E+01 8.03E+01 Cs-134 1.2SE-08 1.10E-08 1.18E-08 4.63E+01 4.07E+01 4.37E+01 Cs-137 8.63E-09 7.94E-09 8.82E-09 3.19E+01 2.94E+01 3.26E+01 Am-241 1.20E-04 2.17E-03 1.84E-OS 4.44E+OS 8.03E+06 6.81E+04 Cm-242 4.67E-06 4.87E-OS l.SSE-05 1.73E+04 1.80E+OS 5.74E+04 Cm-243/4 8.30E-OS 1.47E-03 1.94E-OS 3.07E+OS 5.44E+06 7.18E+04 Pu-238 1.06E-04 1.90E-03 3.20E-04 3.92E+OS 7.03E+06 1.18E+06 Pu-239/4C 1.16E-04 2.11E-03 3.23E-04 4.29E+OS 7.81E+06 1.20E+06 Pu-241 2.23E-06 4.20E-OS 3.18E-06 8.25E+03 l.SSE+OS 1.18E+04 Nb-94 1.12E-07 1.97E-08 7.48E-07 4.14E+02 7.29E+01 2.77E+03 1-129 4.69E-08 1.38E-10 3.14E-10 1.74E+02 S.llE-01 1.16E+OO Ra-226 2.32E-06 7.59E-06 1.61E-05 8.58E+03 2.81E+04 5.96E+04 Co-57 2.4SE-09 4.52E-10 1.69E-08 9.07E+OO 1.67E+OO 6.2SE+01 Ce-144 1.01E-07 4.54E-08 7.91E-07 3.74E+02 1.68E+02 2.93E+03 H-3 1.73E-11 1.73E-11 1.73E-11 6.40E-02 6.40E-02 6.40E-02

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. 0 Page 30 of 32 Attachment 4 - DCF from FGR and mrem/J.LCi Conversion Page 2 of 2 EPA FGR 12, Table 111.1- Submersion Dose Factors Conversion 3. 70E+09 mrem-m3-Bq/Sv-uCi-sec DDE Skin DDE Skin Sv-m3/Bq- Sv-m3 mrem-m3 mrem-m3 Nuclide sec /Bq-sec /uCi-sec /uCi-sec C-14 2.24E-19 2.43E-16 8.29E-10 8.99E-07 Fe-55 O.OOE+OO 0 O.OOE+OO O.OOE+OO Ni-59 0 0 O.OOE+OO O.OOE+OO Ni-63 0 0 O.OOE+OO O.OOE+OO Sr-89 7.73E-17 3.69E-14 2.86E-07 1.37E-04 Sr-90 7.53E-18 9.2E-15 2.79E-08 3.40E-05 Tc-99 1.62E-18 2.74E-15 5.99E-09 1.01E-05 Cr-51 1.51E-15 1.75E-15 5.59E-06 6.48E-06 Mn-54 4.09E-14 4.67E-14 1.51E-04 1.73E-04 Co-58 4.76E-14 5.58E-14 1.76E-04 2.06E-04 Fe-59 5.97E-14 7.13E-14 2.21E-04 2.64E-04 Co-60 1.26E-13 1.45E-13 4.66E-04 5.37E-04 Zn-65 2.9E-14 3.29E-14 1.07E-04 1.22E-04 Nb-95 3.74E-14 4.3E-14 1.38E-04 1.59E-04 Zr-95 3.60E-14 4.5E-14 1.33E-04 1.67E-04 Ag-llOm 1.36E-13 1.57E-13 5.03E-04 5.81E-04 Sb-124 9.15E-14 1.26E-13 3.39E-04 4.66E-04 Sb-125 2.02E-14 2.65E-14 7.47E-05 9.81E-05 Cs-134 7.57E-14 9.45E-14 2.80E-04 3.50E-04 Cs-137 7.74E-18 8.63E-15 2.86E-08 3.19E-05 Am-241 8.18E-16 1.28E-15 3.03E-06 4.74E-06 Cm-242 5.69E-18 4.29E-17 2.11E-08 1.59E-07 Cm-243/4 5.88E-15 9.79E-15 2.18E-05 3.62E-05 Pu-238 4.88E-18 4.09E-17 1.81E-08 1.51E-07 Pu-239/4C 4.75E-18 3.92E-17 1.76E-08 1.45E-07 Pu-241 7.25E-20 1.17E-19 2.68E-10 4.33E-10 Nb-94 7.70E-14 9.52E-14 2.85E-04 3.52E-04 1-129 3.80E-16 1.10E-15 1.41E-06 4.07E-06 Ra-226 3.15E-16 4.79E-16 1.17E-06 1.77E-06 Co-57 5.61E-15 6.63E-15 2.08E-05 2.45E-05 Ce-144 8.53E-16 2.93E-15 3.16E-06 1.08E-05 H-3 3.31E-19 0 1.22E-09 O.OOE+OO

Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. 0 Page 31 of 32 Attachment 5 - XOQDOQ Run Results

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Dose Consequences of a High Integrity Container (HIC) Drop Event FC08566 Rev. 0 Page 32 of 32 Attachment 6- XOQDOQ Output File (imbedded txt file)

FORT CALHOUN STATION DOCKET NUMBER 50-285 I LICENSE NUMBER DPR-40 ATTACHMENT 2 COMPARISON MATRIX FOR ISFSI EALS BASED ON THE PROPOSED REGULATORY GUIDE DG-1346 "EMERGENCY PLANNING FOR DECOMMISSIONING NUCLEAR REACTORS" TO THE PROPOSED FCS EMERGENCY CLASSIFICATION SYSTEM AND ISFSI EALS

LIC-19-0001 Attachment 2 Page 1 DG-1346, Appendix A Proposed EAL Matrix for FCS Comparison ICs/EALs EU2 EU2

• ECL: Unusual Event ECL: Unusual Event Initiating Condition: Damage to a loaded cask Initiating Condition: Damage to a loaded CONFINEMENT BOUNDARY. cask CONFINEMENT BOUNDARY.

Operating Mode Applicability:

Not Applicable Example Emergency Action Emergency Action Levels: 1

• Provided FCS-specific radiation Levels: levels that conform to

1. Damage to a loaded cask confinement 10 CFR 20.1301 allowable levels (1) Damage to a loaded cask BOUNDARY as indicated by an based on calculations.

CONFINEMENT abnormal radiation reading of >2 BOUNDARY as indicated mRem/hr. (gamma) within the ISFSI Protected Area or on a Horizontal by a radiation monitor Storage ivioduie (HSM) concrete reading greater than surface.

NORMAL background at or near the cask.

Basis: Basis:

• Added FCS specific basis This IC addresses an event that This IC addresses an event that results in information.

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

factors, and configuration changes which could cause The existence of "damage" is determined challenges in removing the cask by radiological survey. The radiation limits or fuel from storage. listed in the EAL reflect calculations based on 10 CFR 20.1301(a)(2) radiation dose The existence of "damage" is limits to the public. In addition to aligning determined by radiological with 10 CFR 20.1301 limits, the radiation survey. The emphasis for this levels chosen are a reasonable indication classification is the degradation that actual cask confinement boundary has in the level of safety of the spent occurred due to the level being greater fuel cask and not the magnitude than calculated levels.

of the associated dose or dose rate. It is recognized that in the Security-related events for ISFSis are case of extreme damage to a covered under ICs EU1 and EA1.

loaded cask, the IC may be determined based on measurement of a dose rate at some distance from the cask.

LIC-19-0001 Page 2 DG-1346, Appendix A Proposed EAL Matrix for FCS .* Comparison ICs/EALs Security-related events for ISFSis are covered under ICs EU1 and EA1.

EU1 EU1 ECL: Unusual Event ECL: Unusual event Initiating Condition: Confirmed SECURITY CONDITION, or Initiating Condition: Confirmed threat, at the rndependent spent SECURITY CONDITION, or threat, at the storage installation {ISFSI). independent spent storage installation (ISFSI).

Applicability: IOEP Example Emergency Action Emergency Action Levels: 1 or 2

• Removed the term "HOSTILE Levels : (1 or 2 or 3) ACTION" as it does not apply to

• 1. A SECURITY CONDITION as an ISFSI Only Facility

( 1) A SECURITY CONDITION reported by the security force and that does not involve a impacting the ISFSI.

• Deleted EAL 3 related to aircraft HOSTILE ACTION as threat reported by the (site- 2. Notification of a credible security specific security shift threat directed at the ISFSI.

supervision) and impacting the ISFSI.

(2) Notification of a credible security threat directed at the ISFSI.

(3) A validated notification from the NRC providing information of an aircraft threat.

Basis: Basis:

• Deleted reference to communicating with the Control This initiating condition (IC) This IC addresses events that pose a Room and referenced addresses events that pose a threat to facility personnel or spent fuel, communicating with the ISFSI threat to plant personnel and , and thus represent a potential degradation Shift Supervisor/Emergency thus, represents a potential in the level of facility safety. Security Director degradation in the level of plant events which do not meet one of these safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR 73.71 or

• Deleted wording associated with aircraft threats emergency action levels (EALs) 10 CFR 50.72. Security events assessed are adequately addressed by the as ADVERSARIAL ACTION are

• Deleted wording regarding requirements of Section 73.71, classifiable under IC EA 1. security-sensitive information "Reporting of safeguards Timely and accurate communication events," of Title 10 of the Code between the security force and the ISFSI of Federal Regulations (10 CFR)

Shift Supervisor/Emergency Director is Part 73, "Physical Protection of Plants and Materials," or Section essential for proper classification of a 50.72, "Immediate notification security-related event. Classification of requirements for operating these events will initiate appropriate threat-nuclear power reactors," of

LIC-19-0001 Attachment 2 Page 3 DG-1346, Appendix A Proposed EAL Matrix for FCS Comparison ICs/EALs 10 CFR Part 50, "Domestic related notifications to site personnel and Licensing of Production and Offsite Response Organizations (OROs).

Utilization Facilities."

Security plans and terminology are based Timely and accurate on the guidance provided by NEI 03-12, communications between Template for the Security Plan, Training Security Shift Supervision and and Qualification Plan, Safeguards the Control Room is essential for Contingency Plan [and Independent Spent proper classification of a Fuel Storage Installation Security security-related event. Program].

Classification of these events will initiate appropriate threat-related EAL #1 references the security force notifications to plant personnel because these are the individuals trained and offsite response to confirm that a security event is occurring organizations (OROs). or has occurred. Training on security event confirmation and classification is controlled Security plans and terminology due to the nature of Safeguards and are based on the guidance 10 CFR 2.390 information.

provided by NEI 03-12 "Template for the Security Plan, EAL #2 addresses the receipt of a credible Training and Qualification Plan, security threat. The credibility of the threat Safeguards Contingency Plan is assessed in accordance with Security

[and Independent Spent Fuel procedures.

Storage Installation Security Escalation of the emergency classification Program]" . level would be via IC EA 1.

EAL #1 references (site-specific security shift 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 Section 2.390, "Public inspections, exemptions, and requests for withholding," of 10 CFR Part 2, "Agency Rules of Practice and Procedure,"

information.

EAL #2 addresses the receipt of a credible security threat directed at the ISFSI. The credibility of the threat is assessed in accordance with (site-specific procedure).

EAL #3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by North American Aerospace

LIC-19-0001 Attachment 2 Page 4 DG-1346, Appendix A Proposed EAL Matrix for FCS Comparison ICs/EALs Defense Command (NORAD) through the NRC. Validation of the threat is performed in accordance with (site-specific procedure).

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.

This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.

Security-sensitive information should be contained in non-public documents such as the Security Plan.

Escalation of the emergency classification level would be via IC EA1 .

EA1 EA1

• Changed Initiating Condition wording ECL: Alert ECL: Alert Initiating Condition: HOSTILE Initiating Condition: ADVERSARIAL

• Deleted reference to airborne threat ACTION within the OWNER ACTION is occurring or has occurred.

CONTROLLED AREA or airborne attack threat within 30 minutes.

Applicability: IOEP Example Emergency Action Emergency Action Levels: 1

• Reworded to make EAL specific Levels: to FCS ISFSI facility

( 1) A HOSTILE ACTION is 1. An ADVERSARIAL ACTION is occurring or has occurred as reported

• Deleted Example EAL 2 related to occurring or has occurred aircraft threat within the ISFSI as reported by the security force.

by the (site-specific security shift supervision).

(2) A validated notification from NRC of an aircraft attack threat within 30 minutes of the site.

Basis: Basis:

• Changed wording to reflect FCS ISFSI EAL wording This IC addresses the This IC addresses the occurrence of an occurrence of a HOSTILE ADVERSARIAL ACTION.

• Deleted wording associated with ACTION within the ISFSI or aircraft threats

LIC-19-0001 Attachment 2 Page 5 DG-1346, Appendix A Proposed EAL Matrix for FCS Comparison ICs/EALs notification of an aircraft attack Timely and accurate communication ~ Deleted reference to threat. This event will require between the security force and the ISFSI communicating with the Control rapid response and assistance Shift Supervisor/Emergency Director is Room and referenced due to the possibility of the essential for proper classification of a communicating with the ISFSI attack compromising stored security-related event. Shift Supervisor/Emergency spent fuel or damaging the Director storage casks, or the need to As time and conditions allow, these events prepare the plant and staff for a require a heightened state of readiness by

• Deleted wording regarding potential aircraft impact. the facility staff and implementation of security-sensitive information onsite protective measures (e.g.,

Timely and accurate evacuation, dispersal or sheltering). The communications between Alert declaration will also heighten the Security Shift Supervision and awareness of Offsite Response the Control Room is essential for Organizations (OROs), allowing them to be proper classification of a better prepared should it be necessary to security-related event. consider further actions.

Security plans and terminology are based on the guidance provided by NEI 03-12.

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of possible onsite protective measures (e.g.,

evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of OROs, 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.

EAL #1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the ISFSI.

EAL #2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes.

LIC-19-0001 Attachment 2 Page 6 DG-1346, Appendix A ICs/EALs . Proposed EAL Matrix for FCS Comparison The intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with (site-specific procedure).

The NRC HOO will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact within the ISFSI was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration should not be unduly delayed while awaiting notification by a Federal agency.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information.

This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location.

Security-sensitive information should be contained in non-public documents such as the Security Plan.