LIC-16-0109, Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E (Attachments 1, 2, 4, 5 and 6)

Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E (Attachments 1, 2, 4, 5 and 6)

ML16356A578
Person / Time
Site:	Fort Calhoun
Issue date:	12/16/2016
From:	Marik S Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML17005A231	List: LIC-16-0109, Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E (Attachments 1, 2, 4, 5 and 6)
References
LIC-16-0109
Download: ML16356A578 (387)

Text

{{#Wiki_filter:Withhold from Public Disclosure under 10 CFR 2.390 10 CFR 50.12 10 CFR 50.47 10 CFR 50, Appendix E LIC-16-0109 December 16, 2016 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

Subject:

Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E

References:

1. Letter from OPPD (T. Burke) to USNRC (Document Control Desk), Certification of Permanent Cessation of Power Operations, dated June 24, 2016 (LIC-16-0043)

(ML16176A213)

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

(ML16242A127)

3. Letter from OPPD (T. Burke) to USNRC (Document Control Desk), Certification of Permanent Removal of Fuel from the Reactor Vessel, dated November 13, 2016 (LIC-16-0074) (ML16319A254)

Pursuant to 10 CFR 50.12, Omaha Public Power District (OPPD) requests exemptions from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR Part 50, Appendix E for the Fort Calhoun Station, Unit 1 (FCS). The requested exemptions would allow FCS to reduce emergency planning requirements consistent with the permanently defueled condition of the station. On June 24, 2016, OPPD certified that FCS would permanently cease power operations no later than December 31, 2016, in accordance with 10 CFR 50.82(a)(1)(i) (Reference 1). On August 25, 2016, OPPD certified that FCS would permanently cease power operations in accordance with 10 CFR 50.82(a)(1)(i) on October 24, 2016 (Reference 2). On November 13, 2016 (Reference 3), pursuant to 10 CFR 50.82(a)(1)(ii), OPPD certified that the fuel had been permanently removed from the reactor vessel and placed in the spent fuel pool. Therefore, pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for FCS no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel. ATTACHMENT 3 CONTAINS INFORMATION BEING WITHHELD FROM PUBLIC DISCLOSURE PER 10 CFR 2.390. UPON SEPARATION, THIS LETTER IS DECONTROLLED 444 South 16th Street Mall Omaha, NE 68102-2247

U.S. Nuclear Regulatory Commission LIC-16-0109 Page 2 Withhold from Public Disclosure under 10 CFR 2.390 The requested exemptions are permissible under 10 CFR 50.12 because they are authorized by law, will not present an undue risk to the public health and safety, are consistent with the common defense and security, and present special circumstances. More specifically, application of the portions of the regulations from which exemptions are sought is not necessary to ensure adequate emergency response capability for FCS and to achieve the underlying purpose of the rules. Furthermore, continued application of these portions of the regulations from which exemptions are sought would result in an undue hardship or other costs to OPPD and the FCS Decommissioning Trust Fund by requiring continued implementation of unnecessary emergency response capabilities. Finally, granting the requested exemptions would result in benefit to the public health and safety and would not result in a decrease in safety, because they would enhance the ability of the emergency response organization to respond to credible scenarios. The exemption requests are contained in Attachment 1 to this letter. OPPD has performed an analyses which show that 530 days (1 year, 165 days) after permanent cessation of power operations, the spent fuel stored in the spent fuel pool will have decayed to the extent that the requested exemptions may be implemented at FCS without any additional compensatory actions. Following the FCS shutdown, which occurred on October 24, 2016, 530 days after permanent cessation of power operations will occur on April 7, 2018. This analysis is included in . , Calculation FC08513, EAB Radiation Shine Dose 18 Months Post Shutdown with the SFP Drained, contains information that is proprietary to AREVA Inc. Accordingly, pursuant to 10 CFR 2.390, OPPD requests that Attachment 3 be withheld from public disclosure. is a nonproprietary version of Attachment 3 suitable for public disclosure. FCS also plans to submit a Permanently Defueled Emergency Plan and a Permanently Defueled Emergency Action Level scheme, for NRC review and approval pursuant to 10 CFR 50.54(q)(4) and 10 CFR Part 50, Appendix E, Section IV.B.2. In support of this exemption and the associated license amendment for the Permanently Defueled Emergency Plan, numerous discussions, both electronic and in person, have been held with the cognizant state (Nebraska and Iowa) and local response organizations. On October 13, 2016, FCS facilitated a presentation and discussion that included a line by line review of NSIR/DPR-ISG-02, Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants, which is the basis for the exemptions. Participants at this meeting include the States of Iowa and Nebraska, Washington and Douglas counties from Nebraska, and Regional leadership from the Federal Emergency Management Agency. Follow up conversations, via phone and email, have been made to address questions from that meeting. On December 8, 2016, FCS held the scheduled quarterly meeting with the States of Iowa and Nebraska, Douglas and Washington counties of Nebraska, Pottawattamie and Harrison Counties from Iowa, and Regional FEMA representatives. This meeting facilitated discussions on the process and plan changes that would be implemented at FCS as part of the Permanently Defueled Emergency Plan. ATTACHMENT 3 CONTAINS INFORMATION BEING WITHHELD FROM PUBLIC DISCLOSURE PER 10 CFR 2.390. UPON SEPARATION, THIS LETTER IS DECONTROLLED

U.S. Nuclear Regulatory Commission LIC-16-01 09 Page 3 Withhold from Public Disclosure under 10 CFR 2.390 OPPD requests review and approval of this request for exemptions by December 18, 2017, with an effective date of April?, 2018. Approval of these exemptions by December 18, 2017, will allow FCS adequate time to implement changes to the emergency plan and emergency response organization by the requested effective date. of this letter contains new regulatory commitments. If you should have any questions regarding this submittal, please contact Mr. Bradley H. Blome at (402) 533-7270. I declare under penalty of perjury that the foregoing is true and correct. Executed on December 16, 2016. Respectfully,

  ~1p~

Fo~ Shane M. Marik Site Vice President and CNO Ft. Calhoun Station SMM/epm Attachments: 1. Request for Exemptions from Portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2) and 10 CFR Part 50, Appendix E

2. Calculation FC081 04, Maximum Cladding Temperature Analysis for Adiabatic Heat-up of Spent Fuel Assembly

3. Calculation FC08513, EAB Radiation Shine Dose 18 Months Post Shutdown with the SFP Drained, without attachment E2 (Proprietary)

4. Calculation FC08513, EAB Radiation Shine Dose 18 Months Post Shutdown with the SFP Drained, without attachment E2 (Non-Proprietary)

5. Affidavit for Withholding Information Pursuant to 10 CFR 2.390

6. List of Regulatory Commitments c:

K.M. Kennedy, NRC Regional Administrator, Region IV C. F. Lyon, NRC Senior Project Manager S.M. Schneider, NRC Senior Resident Inspector Director of Consumer Health Services, Department of Regulation and Licensure, Nebraska Health and Human Services, State of Nebraska ATTACHMENT 3 CONTAINS INFORMATION BEING WITHHELD FROM PUBLIC DISCLOSURE PER 10 CFR 2.390. UPON SEPARATION, THIS LETTER IS DECONTROLLED

OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION DOCKET NUMBER 50-285 / LICENSE NUMBER CPR-40 ATTACHMENT 1 REQUEST FOR EXEMPTIONS FROM PORTIONS OF 10 CFR 50.47(b), 10 CFR 50.47(c)(2) AND 10 CFR PART 50, APPENDIX E

LIC-16-0109 Page 1 FORT CALHOUN STATION REQUESTS FOR EXEMPTIONS FROM PORTIONS OF 10 CFR 50.47(b); 50.47(c)(2); AND 10 CFR PART 50, APPENDIX E

Subject:

Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E 1.0

SUMMARY

DESCRIPTION

2.0 BACKGROUND

3.0 DETAILED DESCRIPTION

4.0 TECHNICAL EVALUATION

4.1 Accident Analysis Overview 4.2 Consequences of a Design Basis Event 4.3 Consequences of a Beyond Design Basis Event 4.4 Consequences of Other Analyzed Events 4.5 Comparison to NUREG-1738 Industry Decommissioning Commitments and Staff Decommissioning Assumptions 4.6 Consequences of a Beyond Design Basis Earthquake 4.7 Conclusion 5.0 Justification for Exemptions and Special Circumstances 5.1 Special Circumstances 5.2 Precedent

7.0 REFERENCES

LIC-16-0109 Page 2 1.0

SUMMARY

DESCRIPTION Pursuant to 10 CFR 50.12 Specific exemptions, Omaha Public Power District (OPPD) requests the following regulatory exemptions for the Fort Calhoun Station (FCS): Certain standards in 10 CFR 50.47(b) regarding onsite and offsite emergency response plans for nuclear power reactors; Certain requirements of 10 CFR 50.47(c)(2) to establish Plume Exposure and Ingestion Pathway Emergency Planning Zones (EPZs) for nuclear power plants; and Certain requirements of 10 CFR Part 50, Appendix E, which establishes the elements that make up the content of emergency plans. The requested exemptions would allow OPPD to revise the scope of the FCS Emergency Plan to reflect the permanently shutdown and defueled condition of the station. The current 10 CFR Part 50 regulatory requirements for emergency planning (developed for operating reactors) ensured protection of the health and safety of the public while FCS was licensed to operate. However, with the station permanently shut down, defueled, and in a state of decommissioning, some of these requirements exceed what is necessary to protect the health and safety of the public. The requested exemptions and justification for each are based on, and consistent with, Interim Staff Guidance (ISG) NSIR/DPR-ISG-02, Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants, issued May 11, 2015 (Reference 1).

2.0 BACKGROUND

FCS is located midway between Fort Calhoun and Blair, Nebraska, on the west bank of the Missouri River. The site consists of approximately 660.46 acres with an additional exclusion area of 582.18 acres on the northeast bank of the river directly opposite the plant buildings. FCS includes the Independent Spent Fuel Storage Installation (ISFSI), located within the protected area, approximately 200 meters north-northwest of the Containment Building. The distance from the reactor containment to the nearest site boundary is approximately 910 meters; and the distance to the nearest residence is beyond the site boundary. Except for the city of Blair and the villages of Fort Calhoun and Kennard, the area within a ten-mile radius is predominantly rural. The land use within the ten-mile radius 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 (USAR) describes the accident scenarios that are applicable to FCS. Many of the accident scenarios postulated in the USAR for operating power reactors involve failures or malfunctions of systems, which could affect the fuel in the reactor vessel, which in the most severe postulated accidents, would involve the release of large quantities of fission products. With the termination of reactor operations and the permanent removal of fuel from the reactor vessel, such accidents are no longer possible. Therefore, the postulated accidents involving failure or malfunction of the reactor, reactor cooling system, steam system, or turbine generator are no longer applicable.

LIC-16-0109 Page 3 On June 24, 2016 (Reference 2), OPPD certified that FCS would permanently cease power operations. On August 25, 2016 (Reference 3), pursuant to 10 CFR 50.82(a)(1)(i) and 10 CFR 50.4(b)(8), OPPD certified that FCS would permanently cease power operations on October 24, 2016. On November 13, 2016 (Reference 4), pursuant to 10 CFR 50.82(a)(1)(ii), OPPD certified that the fuel has been permanently removed from the reactor vessel and placed in the spent fuel pool (SFP). Upon docketing of the certifications for permanent cessation of power operations (10 CFR 50.82(a)(1)(i)) and permanent removal of fuel from the reactor vessel (10 CFR 50.82(a)(1)(ii)), pursuant to 10 CFR 50.82(a)(2), 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. Within two years following permanent cessation of power operations, FCS will submit a Post-Shutdown Decommissioning Activities Report (PSDAR), identifying the method FCS has selected for decommissioning. With the reactor permanently defueled, the reactor vessel assembly and supporting structures and systems will no longer be in operation and will have no function related to the safe storage and management of irradiated fuel in the SFP. A SFP cooling and clean-up system is provided to remove decay heat from spent fuel stored in the SFP and to maintain a specified water temperature, purity, and clarity. The irradiated fuel will be stored in the SFP and/or the ISFSI until it is removed by the Department of Energy (DOE). 3.0 DETAILED DESCRIPTION In order to allow a reduction in emergency planning requirements commensurate with the hazards associated with FCSs permanently shut down and defueled condition, exemptions from portions of 10 CFR 50.47(b); 50.47(c)(2); and 10 CFR Part 50, Appendix E, are needed. OPPD has performed an analysis indicating that 530 days (1 year, 165 days) after permanent cessation of power operations, the spent fuel stored in SFP will have decayed to the extent that the requested exemptions can be implemented at FCS without any compensatory actions. This analysis is included in Attachment 2. Considering the shutdown date of October 24, 2016, 530 days after permanent cessation of power operations will occur on April 7, 2018. OPPD will submit a Permanently Defueled Emergency Plan, including a Permanently Defueled Emergency Action Level scheme, for NRC review and approval pursuant to 10 CFR 50.54(q)(4) and 10 CFR 50, Appendix E, Section IV.B.2. The proposed emergency plan will be based on the exemptions requested herein. Based on the analyses detailed in Section 4.0, below, OPPD has concluded that the portions of 10 CFR 50.47(b); 10 CFR 50.47(c)(2); and 10 CFR Part 50, Appendix E identified in Tables 1 and 2 below will not be necessary to protect the health and safety of the public when FCS is in the permanently defueled condition and would be unduly burdensome. Approval of the exemptions requested in Tables 1 and 2 would not present an undue risk to the public or prevent an appropriate response in the event of an emergency at FCS.

LIC-16-0109 Page 4 EXEMPTIONS TO EMERGENCY PLAN REQUIREMENTS DEFINED BY 10 CFR 50.47 AND PART 50, APPENDIX E OPPD requests exemptions from portions of 10 CFR 50.47(b) and (c)(2) and 10 CFR Part 50, Appendix E to the extent that these regulations apply to specific provisions of onsite and offsite emergency planning that will no longer be applicable to FCS when the certifications required by 10 CFR 50.82(a)(1)(i) and (ii) have been submitted and sufficient decay of the spent fuel has occurred. The specific portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E from which exemptions are being requested are identified using strikethrough text in Table 1 (Exemptions Requested from 10 CFR 50.47(b) and (c)(2)) and Table 2 (Exemptions Requested from 10 CFR Part 50, Appendix E), below. The portions of regulation that are not identified using strikethrough text (i.e., those portions for which exemption is not being requested), will remain applicable to FCS. Details related to specific exemption requests are provided in the Basis for Exemption column in each table. Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2) Item # Regulation in 10 CFR 50.47 Basis for Exemption 1 10 CFR 50.47(b): The onsite and, except as provided in In the Statement of Considerations for the Final Rule for Emergency paragraph (d) of this section, offsite emergency response Planning requirements for Independent Spent Fuel Storage plans for nuclear power reactors must meet the following Installations (ISFSIs) and for monitored retrievable storage (MRS) standards: facilities (60 FR 32430; June 22, 1995) (Reference 5), the Commission responded to comments concerning offsite emergency planning for ISFSIs or an MRS and concluded that, the offsite consequences of potential accidents at an ISFSI or a MRS [monitored retrievable storage installation] would not warrant establishing Emergency Planning Zones. In a nuclear power reactors permanently defueled state, the accident risks are more similar to an ISFSI or MRS than an operating nuclear power plant. The draft proposed rulemaking in SECY-00-0145 (Reference 6) suggested that after at least one year of spent fuel decay time, the decommissioning licensee would be able to reduce its emergency planning program to one similar to that required for an MRS under 10 CFR 72.32(b) and additional emergency planning reductions would occur when: (1) approximately five years of spent fuel decay

LIC-16-0109 Page 5 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2) Item # Regulation in 10 CFR 50.47 Basis for Exemption time has elapsed; or (2) a licensee has demonstrated that the decay heat level of spent fuel in the pool is low enough that the fuel would not be susceptible to a zirconium fire for all spent fuel configurations. Because of the slow rate of the event scenarios in the postulated accident and postulated beyond design basis events analyses and because the duties of the on-shift personnel at a decommissioning reactor facility are not as complicated and diverse as those for an operating reactor, significant time is available to complete actions necessary to mitigate an emergency without impeding timely performance of emergency plan functions. Exemptions from offsite emergency planning requirements have been approved when the specific site analyses show that at least 10 hours is available from a partial drain down event where cooling of the spent fuel is not effective until the hottest fuel assembly reaches 900 degrees Celsius (°C). Because 10 hours allows sufficient time to initiate mitigative actions to prevent a zirconium fire in the SFP or to initiate offsite protective actions in accordance with a comprehensive approach to emergency planning, offsite emergency plans are not necessary for these permanently defueled nuclear power plant licensees. The OPPD analysis has demonstrated that within 10 days after permanent cessation of power operations, the radiological consequences of the postulated accident will not exceed the limits of the U.S. Environmental Protection Agency's (EPA) Protective Action Guides (PAGs) at the Exclusion Area Boundary (EAB). FCS has performed an analysis indicating that after the spent fuel has decayed for 530 days (1 year, 165 days), for beyond design

LIC-16-0109 Page 6 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2) Item # Regulation in 10 CFR 50.47 Basis for Exemption basis events where the SFP is partially drained, and air cooling is not possible, 10 hours is available to take mitigative actions or, if needed, implement offsite protective actions using a comprehensive approach to emergency planning from the time spent fuel cooling is lost until the hottest fuel assembly reaches a temperature of 900°C. FCS maintains procedures and strategies for the movement of any necessary portable equipment that will be relied upon for mitigating the loss of SFP water. These mitigative strategies are maintained in accordance with License Condition 3.G of the FCS Renewed Facility Operating License. These diverse strategies provide defense-in-depth and ample time to provide makeup water or spray to the SFP prior to the onset of zirconium cladding ignition when considering very low probability beyond design basis events affecting the SFP. 2 10 CFR 50.47(b)(1): Primary responsibilities for emergency See the basis for 10 CFR 50.47(b). response by the nuclear facility licensee and by State and local organizations within the Emergency Planning Zones have been assigned, the emergency responsibilities of the various supporting organizations have been specifically established, and each principal response organization has staff to respond and to augment its initial response on a continuous basis. 3 10 CFR 50.47(b)(2): On-shift facility licensee responsibilities No exemption is requested. 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 and the

LIC-16-0109 Page 7 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2) Item # Regulation in 10 CFR 50.47 Basis for Exemption interfaces among various onsite response activities and offsite support and response activities are specified. 4 10 CFR 50.47(b)(3): Arrangements for requesting and Discontinuing offsite emergency planning activities and reducing the effectively using assistance resources have been made, scope of onsite emergency planning is acceptable given the arrangements to accommodate State and local staff at the significantly reduced offsite consequences when FCS is in the licensee's Emergency Operations Facility have been made, permanently defueled condition. The FCS emergency plan will and other organizations capable of augmenting the planned continue to maintain arrangements for requesting and using response have been identified. assistance resources from offsite support organizations. Decommissioning power reactors present a low likelihood of any credible accident resulting in radiological releases requiring offsite protective measures because of the permanently shut down and defueled status of the reactor. An emergency operations facility (EOF) is not required. The control room or another location can provide for the communication and coordination with offsite organizations for the level of support required. Also see the basis for 10 CFR 50.47(b). 5 10 CFR 50.47(b)(4): A standard emergency classification and FCS will adopt the Permanently Defueled Emergency Action Levels action level scheme, the bases of which include facility system (EALs) consistent with those detailed in Appendix C of Nuclear and effluent parameters, is in use by the nuclear facility Energy Institute (NEI) 99-01, Development of EALs for Non-Passive licensee, and State and local response plans call for reliance Reactors, Revision 6 (Reference 7), endorsed by the NRC in a on information provided by facility licensees for determinations letter dated March 28, 2013 (Reference 8). FCS analysis shows that of minimum initial offsite response measures. after the spent fuel has decayed for 530 days (1 year, 165 days), for beyond design basis events where the SFP is partially drained, and air cooling is not possible, 10 hours is available to take mitigative or, if needed, offsite protective actions using a comprehensive

LIC-16-0109 Page 8 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2) Item # Regulation in 10 CFR 50.47 Basis for Exemption approach to emergency planning from the time spent fuel cooling is lost until the hottest fuel assembly reaches a temperature of 900°C. No offsite protective actions are anticipated to be necessary. Therefore, classification above the Alert level will no longer be required. Also see the basis for 10 CFR 50.47(b). 6 10 CFR 50.47(b)(5): Procedures have been established for Per SECY-00-0145 (Reference 6), after approximately 1 year of notification, by the licensee, of State and local response spent fuel decay time (and as supported by the SFP analysis), the organizations and for notification of emergency personnel by NRC staff believes an exception to the offsite EPA PAG standard is all organizations; the content of initial and followup messages justified for a zirconium fire scenario considering the low likelihood of to response organizations and the public has been this event together with time available to take mitigative or protective established; and means to provide early notification and clear actions between the initiating event and before the onset of a instruction to the populace within the plume exposure pathway postulated fire. SECY-13-0112, Consequence Study of a Beyond-Emergency Planning Zone have been established. Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor, (Reference 9) provides that depending on the size of the pool liner leak, releases could start anywhere from eight hours to several days after the leak starts, assuming that mitigation measures are unsuccessful. If 10 CFR 50.54(hh)(2)-type mitigation measures are successful, releases could only occur during the first several days after the fuel was removed from the reactor. As previously indicated, an FCS analysis shows that after the spent fuel has decayed for 530 days (1 year, 165 days), for beyond design basis events where the SFP is partially drained, and air cooling is not possible, 10 hours is available to take mitigative or, if needed, offsite protective actions using a comprehensive approach to emergency planning from the time spent fuel cooling is lost until the hottest fuel assembly reaches a

LIC-16-0109 Page 9 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2) Item # Regulation in 10 CFR 50.47 Basis for Exemption temperature of 900°C.Therefore, offsite emergency plans are not necessary for permanently defueled nuclear power plants. Also see the basis for 10 CFR 50.47(b). 7 10 CFR 50.47(b)(6): Provisions exist for prompt See the basis for 10 CFR 50.47(b). communications among principal response organizations to emergency personnel and to the public. 8 10 CFR 50.47(b)(7): Information is made available to the See the basis for 10 CFR 50.47(b). public on a periodic basis on how they will be notified and what their initial actions should be in an emergency (e.g., listening to a local broadcast station and remaining indoors), [T]he principal points of contact with the news media for dissemination of information during an emergency (including the physical location or locations) are established in advance, and procedures for coordinated dissemination of information to the public are established. 9 10 CFR 50.47(b)(8): Adequate emergency facilities and No exemption is requested. equipment to support the emergency response are provided and maintained. 10 10 CFR 50.47(b)(9): Adequate methods, systems, and See the basis for 10 CFR 50.47(b). equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use.

LIC-16-0109 Page 10 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2) Item # Regulation in 10 CFR 50.47 Basis for Exemption 11 10 CFR 50.47(b)(10): A range of protective actions has been In the unlikely event of a SFP accident, the iodine isotopes which developed for the plume exposure pathway EPZ for contribute to an offsite dose from an operating reactor accident are emergency workers and the public. In developing this range of not present, so potassium iodide (KI) distribution offsite would no actions, consideration has been given to evacuation, longer serve as an effective or necessary supplemental protective sheltering, and, as a supplement to these, the prophylactic action. Protective actions will be maintained for emergency workers use of potassium iodide (KI), as appropriate. Evacuation time and any offsite emergency responders who would respond to the estimates have been developed by applicants and licensees. site. Licensees shall update the evacuation time estimates on a The Commission responded to comments in its Statement of periodic basis. Guidelines for the choice of protective actions Considerations for the Final Rule for Emergency Planning during an emergency, consistent with Federal guidance, are requirements for ISFSIs and MRS facilities (60 FR 32435), and developed and in place, and protective actions for the concluded that, the offsite consequences of potential accidents at ingestion exposure pathway EPZ appropriate to the locale an ISFSI or a MRS would not warrant establishing Emergency have been developed. Planning Zones. Additionally, in the Statement of Considerations for the Final Rule for Emergency Planning requirements for ISFSIs and for MRS facilities (60 FR 32430) (Reference 5), the Commission responded to comments concerning site-specific emergency planning that includes evacuation of surrounding population for an ISFSI not at a reactor site, and concluded that, The Commission does not agree that as a general matter emergency plans for an ISFSI must include evacuation planning. Because the NRC concludes that evacuation planning is not needed for a decommissioning reactor site that meets the criteria for an exemption from offsite EP requirements as discussed in the exemption from 10 CFR 50.47(b), evacuation time estimates are also not needed.

LIC-16-0109 Page 11 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2) Item # Regulation in 10 CFR 50.47 Basis for Exemption Also see the basis for 10 CFR 50.47(b) detailing the low likelihood of any credible accident resulting in radiological releases requiring offsite protective measures and the basis for Section IV.1 exemptions for a discussion on the similarity between a permanently defueled reactor and a non-power reactor. 12 10 CFR 50.47(b)(11): Means for controlling radiological No exemption is requested. exposures, in an emergency, are established for emergency workers. The means for controlling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving Activity Protective Action Guides. 13 10 CFR 50.47(b)(12): Arrangements are made for medical No exemption is requested. services for contaminated injured individuals. 14 10 CFR 50.47(b)(13): General plans for recovery and reentry No exemption is requested. are developed. 15 10 CFR 50.47(b)(14): Periodic exercises are (will be) No exemption is requested. conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected. 16 10 CFR 50.47(b)(15): Radiological emergency response No exemption is requested. training is provided to those who may be called on to assist in an emergency.

LIC-16-0109 Page 12 Table 1 Exemptions Requested from 10 CFR 50.47(b) and 50.47(c)(2) Item # Regulation in 10 CFR 50.47 Basis for Exemption 17 10 CFR 50.47(b)(16): Responsibilities for plan development No exemption is requested. and review and for distribution of emergency plans are established, and planners are properly trained. 18 10 CFR 50.47(c)(2): Generally, the plume exposure pathway FCS has developed an analysis indicating that 530 days (1 year, EPZ for nuclear power plants shall consist of an area about 10 165 days) after permanent cessation of power operations, no miles (16 km) in radius and the ingestion pathway EPZ shall credible accident at FCS will result in radiological releases requiring consist of an area about 50 miles (80 km) in radius. The exact offsite protective actions. The analysis of the potential radiological size and configuration of the EPZs surrounding a particular impact of the postulated accident for FCS in a permanently defueled nuclear power reactor shall be determined in relation to local condition indicates that any releases beyond the site boundary are emergency response needs and capabilities as they are limited to small fractions of the EPA PAG exposure levels. According affected by such conditions as demography, topography, land to the EPAs Protective Action Guides and Planning Guidance for characteristics, access routes, and jurisdictional boundaries. Radiological Incidents, Draft for Interim Use and Public Comment, The size of the EPZs also may be determined on a case-by- dated March 2013 (PAG Manual), EPZs are not necessary at those case basis for gas cooled nuclear reactors and for reactors facilities where it is not possible for PAGs to be exceeded off-site. with an authorized power level less than 250 MW thermal. The (Reference 10). plans for the ingestion pathway shall focus on such actions as Also see the basis for 10 CFR 50.47(b). are appropriate to protect the food ingestion pathway.

LIC-16-0109 Page 13 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 19 10 CFR 50 Appendix E No exemption is requested. III. The Final Safety Analysis Report; Site Safety Analysis Report The final safety analysis report or the site safety analysis report for an early site permit that includes complete and integrated emergency plans under § 52.17(b)(2)(ii) of this chapter shall contain the plans for coping with emergencies. The plans shall be an expression of the overall concept of operation; they shall describe the essential elements of advance planning that have been considered and the provisions that have been made to cope with emergency situations. The plans shall incorporate information about the emergency response roles of supporting organizations and offsite agencies. That information shall be sufficient to provide assurance of coordination among the supporting groups and with the licensee. The site safety analysis report for an early site permit which proposes major features must address the relevant provisions of 10 CFR 50.47 and 10 CFR part 50, appendix E, within the scope of emergency preparedness matters addressed in the major features. The plans submitted must include a description of the elements set out in Section IV for the emergency planning zones (EPZs) to an extent sufficient to demonstrate that the plans provide reasonable assurance that adequate protective measures can and will be taken in the event of an emergency.

LIC-16-0109 Page 14 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 20 10 CFR 50 Appendix E Following docketing of the Certification of Permanent Removal of Fuel from the Reactor Vessel, in accordance with 10 CFR IV. Content of Emergency Plans 50.82(a)(1)(i) and (ii), FCS became a permanently shut down facility

1. The applicant's emergency plans shall contain, but not with spent fuel stored in the SFP. In the EP Final Rule (76 FR necessarily be limited to, information needed to 72596, Nov. 23, 2011) (Reference 11), the NRC defined hostile demonstrate compliance with the elements set forth action as, in part, an act directed toward a nuclear power plant or its below, i.e., organization for coping with radiological personnel. This definition is based on the definition of "hostile action" emergencies, assessment actions, activation of provided in NRC Bulletin 2005-02. NRC Bulletin 2005-02 was not emergency organization, notification procedures, applicable to nuclear power reactors that have permanently ceased emergency facilities and equipment, training, maintaining operations and have certified that fuel has been removed from the emergency preparedness, and recovery, and onsite reactor vessel. The NRC excluded non-power reactors (NPRs) from protective actions during hostile action. In addition, the the definition of hostile action at that time because an NPR is not a emergency response plans submitted by an applicant for nuclear power plant and a regulatory basis had not been developed a nuclear power reactor operating license under this part, to support the inclusion of NPR in that definition. Similarly, a or for an early site permit (as applicable) or combined decommissioning power reactor or ISFSI is not a nuclear reactor license under 10 CFR part 52, shall contain information as defined in the NRCs regulations.

needed to demonstrate compliance with the standards The following similarities between FCS and NPRs show that the described in § 50.47(b), and they will be evaluated against FCS facility should be treated in a similar fashion as an NPR. Similar those standards. to NPRs, FCS will pose lower radiological risks to the public from accidents than do power reactors because: (1) FCS will be a permanently shut down facility (with fuel stored in the SFP and ISFSI) and will no longer generate fission products; 2) Fuel stored in the FCS SFP will have lower decay heat resulting in lower risk of fission product release in the event of a beyond design basis boil off or drain down event; and 3) no credible accident at FCS will result in radiological releases requiring offsite protective actions.

LIC-16-0109 Page 15 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 21 IV.2 This nuclear power reactor license applicant shall also See the basis for 10 CFR 50.47(b)(10). provide an analysis of the time required to evacuate various sectors and distances within the plume exposure pathway EPZ for transient and permanent populations, using the most recent U.S. Census Bureau data as of the date the applicant submits its application to the NRC. 22 IV.3 Nuclear power reactor licensees shall use NRC approved See the basis for 10 CFR 50.47(b)(10). evacuation time estimates (ETEs) and updates to the ETEs in the formulation of protective action recommendations and shall provide the ETEs and ETE updates to State and local governmental authorities for use in developing offsite protective action strategies. 23 IV.4 Within 365 days of the later of the date of the availability See the basis for 10 CFR 50.47(b)(10). of the most recent decennial census data from the U.S. Census Bureau or December 23, 2011, nuclear power reactor licensees shall develop an ETE analysis using this decennial data and submit it under § 50.4 to the NRC. These licensees shall submit this ETE analysis to the NRC at least 180 days before using it to form protective action recommendations and providing it to State and local governmental authorities for use in developing offsite protective action strategies. 24 IV.5 During the years between decennial censuses, nuclear See the basis for 10 CFR 50.47(b)(10). power reactor licensees shall estimate EPZ permanent resident population changes once a year, but no later than 365 days from the date of the previous estimate, using the

LIC-16-0109 Page 16 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. These licensees shall maintain these estimates so that they are available for NRC inspection during the period between decennial censuses and shall submit these estimates to the NRC with any updated ETE analysis. 25 IV.6 If at any time during the decennial period, the EPZ See the basis for 10 CFR 50.47(b)(10). permanent resident population increases such that it causes the longest ETE value for the 2-mile zone or 5-mile zone, including all affected Emergency Response Planning Areas, or for the entire 10-mile EPZ to increase by 25 percent or 30 minutes, whichever is less, from the nuclear power reactor licensee's currently NRC approved or updated ETE, the licensee shall update the ETE analysis to reflect the impact of that population increase. The licensee shall submit the updated ETE analysis to the NRC under § 50.4 no later than 365 days after the licensee's determination that the criteria for updating the ETE have been met and at least 180 days before using it to form protective action recommendations and providing it to State and local governmental authorities for use in developing offsite protective action strategies. 26 IV.7 After an applicant for a combined license under part 52 of No exemption is requested. FCS is not an applicant for a combined this chapter receives its license, the licensee shall conduct at license. Therefore, this regulation is not applicable to FCS and an least one review of any changes in the population of its EPZ at exemption is not necessary. least 365 days prior to its scheduled fuel load. The licensee shall estimate EPZ permanent resident population changes

LIC-16-0109 Page 17 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. If the EPZ permanent resident population increases such that it causes the longest ETE value for the 2-mile zone or 5-mile zone, including all affected Emergency Response Planning Areas, or for the entire 10-mile EPZ, to increase by 25 percent or 30 minutes, whichever is less, from the licensee's currently approved ETE, the licensee shall update the ETE analysis to reflect the impact of that population increase. The licensee shall submit the updated ETE analysis to the NRC for review under § 50.4 of this chapter no later than 365 days before the licensee's scheduled fuel load. 27 A. Organization No exemption is requested. 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 and the means for notification of such individuals in the event of an emergency. Specifically, the following shall be included: 28 A.1. A description of the normal plant operating organization. Following docketing of the certifications required by 10 CFR 50.82(a)(1)(i) and (ii), FCS will not be a facility that can be operated to generate electrical power. Therefore, FCS will not have a plant operating organization. Rather, the station will be maintained by a defueled on-shift staff.

LIC-16-0109 Page 18 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 29 A.2. A description of the onsite emergency response No exemption is requested. organization (ERO) with a detailed discussion of:

a. Authorities, responsibilities, and duties of the individual(s) who will take charge during an emergency;

b. Plant staff emergency assignments;

c. Authorities, responsibilities, and duties of an onsite emergency coordinator who shall be in charge of the exchange of information with offsite authorities responsible for coordinating and implementing offsite emergency measures.

30 A.3. A description, by position and function to be performed, of The number of staff at FCS during the decommissioning process will the licensee's headquarters personnel who will be sent to the be small but commensurate with the need to safely store spent fuel plant site to augment the onsite emergency organization. at the facility in a manner that is protective of public health and safety. FCS will maintain a level of emergency response that does not require response by headquarters personnel. The on-shift and emergency response positions will be defined in the Permanently Defueled Emergency Plan. 31 A.4. Identification, by position and function to be performed, of FCS has developed an analysis indicating that 530 days (1 year, persons within the licensee organization who will be 165 days) after permanent cessation of power operations, no responsible for making offsite dose projections and a credible accident at FCS will result in radiological releases requiring description of how these projections will be made and the offsite protective actions. results transmitted to State and local authorities, NRC, and FCS will maintain the capability to determine if a radiological release other appropriate governmental entities. is occurring and perform dose projections. If a release is occurring, FCS will communicate release and dose projection information to offsite authorities for their consideration. The offsite organizations

LIC-16-0109 Page 19 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption are responsible for deciding what, if any, protective actions should be taken. 32 A.5. Identification, by position and function to be performed, of As indicated by the FCS adiabatic heatup analysis, the time other employees of the licensee with special qualifications for available to initiate compensatory actions in the event of a loss of coping with emergency conditions that may arise. Other SFP cooling or inventory precludes the need to identify and describe persons with special qualifications, such as consultants, who the special qualifications of these individuals in the emergency plan. are not employees of the licensee and who may be called The number of staff at FCS when it is in the permanently defueled upon for assistance for emergencies shall also be identified. state will be small but will be commensurate with the need to The special qualifications of these persons shall be described. operate the facility in a manner that is protective of public health and safety. 33 A.6. A description of the local offsite services to be provided in No exemption is requested. support of the licensee's emergency organization. 34 A.7. By June 23, 2014, identification of, and a description of A decommissioning power reactor has a low likelihood of a credible the assistance expected from, appropriate State, local, and accident resulting in radiological releases requiring offsite protective Federal agencies with responsibilities for coping with measures. For this reason and those described in the basis for emergencies, including hostile action at the site. For purposes Section IV.1 of 10 CFR Part 50, Appendix E, a decommissioning of this appendix, "hostile action" is defined as an act directed power reactor is not a facility that falls within the definition of "hostile toward a nuclear power plant or its personnel that includes the action." use of violent force to destroy equipment, take hostages, Similarly, for security, risk insights can be used to determine which and/or intimidate the licensee to achieve an end. This includes targets are important to protect against sabotage. A level of security attack by air, land, or water using guns, explosives, commensurate with the consequences of a sabotage event is projectiles, vehicles, or other devices used to deliver required and is evaluated on a site-specific basis. The severity of the destructive force. consequences declines as fuel ages, and over time, the underlying

LIC-16-0109 Page 20 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption concern that a sabotage attack could cause offsite radiological consequences is removed. Although, the analysis described above and in the basis for 10 CFR Part 50, Appendix E, Section IV.1 provides a justification for exempting FCS from "hostile action" related requirements, some EP requirements for security-based events will be maintained. The classification of security-based events, notification of offsite authorities, and coordination with offsite agencies under a comprehensive emergency management plan concept will still be required. FCS will maintain appropriate actions for the protection of onsite personnel in a security-based event. The scope of protective actions will be appropriate for the defueled plant status, but will not be the same as actions necessary for an operating power plant. 35 A.8. Identification of the State and/or local officials responsible Offsite emergency measures are limited to support provided by local for planning for, ordering, and controlling appropriate police, fire departments, and ambulance and hospital services as protective actions, including evacuations when necessary. appropriate. An FCS analysis has been developed indicating that 530 days (1 year, 165 days) after permanent cessation of power operations, no credible accident at FCS will result in radiological releases requiring offsite protective actions. Therefore, protective actions such as evacuation should not be required. Also see the basis for 10 CFR 50.47(b)(10). 36 A.9. By December 24, 2012, for nuclear power reactor Responsibilities of the on-shift and emergency response personnel licensees, a detailed analysis demonstrating that on shift will be detailed in the Permanently Defueled Emergency Plan and personnel assigned emergency plan implementation functions implementing procedures and will be regularly tested through drills

LIC-16-0109 Page 21 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption are not assigned responsibilities that would prevent the timely and exercises, audited, and inspected by FCS and the NRC. The performance of their assigned functions as specified in the duties of the on-shift personnel at a decommissioning reactor facility emergency plan. are not as complicated and diverse as those for an operating power reactor. In the EP Final Rule (Reference 11), the NRC acknowledged that the staffing analysis requirement was not necessary for non-power reactor licensees because staffing at non-power reactors is generally small, which is commensurate with operating the facility in a manner that is protective of the public health and safety. The minimal systems and equipment needed to maintain the spent nuclear fuel in the SFP or in a dry cask storage system in a safe condition requires minimal personnel and is governed by Technical Specifications. Because of the slow rate of the event scenarios in the postulated accident and postulated beyond design basis events analyses and because the duties of the on-shift personnel at a decommissioning reactor facility are not as complicated and diverse as those for an operating reactor, significant time is available to complete actions necessary to mitigate an emergency without impeding timely performance of emergency plan functions. For these reasons, it can be concluded that a decommissioning NPP is exempt from the requirement of 10 CFR Part 50, Appendix E, Section IV.A.9. 37 B. Assessment Actions FCS will develop EALs consistent with the Permanently Defueled EALs detailed in Appendix C of NEI 99-01, Revision 6 (Reference B.1. The means to be used for determining the magnitude of, 7). FCS proposes to continue to review EALs with the State of and for continually assessing the impact of, the release of Nebraska and Washington County on an annual basis. However, radioactive materials shall be described, including emergency based upon the reduced scope of EALs for the permanently

LIC-16-0109 Page 22 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption action levels that are to be used as criteria for determining the defueled facility, the scope of the annual review of EALs is expected need for notification and participation of local and State to be limited (i.e., informal mailings, etc.). agencies, the Commission, and other Federal agencies, and Also see the basis for Section IV.1 for the justification from the the emergency action levels that are to be used for requirements in Appendix E related to hostile action. determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The emergency action levels shall be based on in-plant conditions and instrumentation in addition to onsite and offsite monitoring. By June 20, 2012, for nuclear power reactor licensees, these action levels must include hostile action that may adversely affect the nuclear power plant. The initial emergency action levels shall be discussed and agreed on by the applicant or licensee and state and local governmental authorities, and approved by the NRC. Thereafter, emergency action levels shall be reviewed with the State and local governmental authorities on an annual basis. 38 B.2. A licensee desiring to change its entire emergency action No exemption is requested. level scheme shall submit an application for an amendment to its license and receive NRC approval before implementing the change. Licensees shall follow the change process in § 50.54(q) for all other emergency action level changes. 39 C. Activation of Emergency Organization The Permanently Defueled EALs, developed consistent with Appendix C of NEI 99-01, Revision 6 (Reference 7), will be adopted, C.1. The entire spectrum of emergency conditions that involve as previously described. This scheme eliminates the Site Area the alerting or activating of progressively larger segments of Emergency and General Emergency event classifications. the total emergency organization shall be described. The

LIC-16-0109 Page 23 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption communication steps to be taken to alert or activate Additionally, the need to base EALs on containment parameters is emergency personnel under each class of emergency shall be no longer appropriate. The EAL scheme presented in NEI 99-01, described. Emergency action levels (based not only on onsite Revision 6 was endorsed by the NRC in a letter dated March 28, and offsite radiation monitoring information but also on 2013 (ML12346A463). No offsite protective actions are anticipated readings from a number of sensors that indicate a potential to be necessary, so classification above the Alert level is no longer emergency, such as the pressure in containment and the required. In the event of an accident at a defueled facility that meets response of the Emergency Core Cooling System) for the conditions for relaxation of emergency planning requirements, notification of offsite agencies shall be described. The there will be available time for event mitigation, and if necessary, existence, but not the details, of a message authentication implementation of offsite protective actions using a comprehensive scheme shall be noted for such agencies. The emergency approach to emergency planning. See the basis for 10 CFR 50.47(b) classes defined shall include: (1) Notification of unusual detailing the low likelihood of any credible accident resulting in events, (2) alert, (3) site area emergency, and (4) general radiological releases requiring offsite protective measures. emergency. These classes are further discussed in NUREG-Containment parameters will not provide an indication of the 0654/FEMA-REP-1. conditions at FCS and emergency core cooling systems will no longer be required. Other indications, such as SFP level or temperature, will be used while there is spent fuel in the SFP. In the Statement of Considerations for the Final Rule for Emergency Planning requirements for ISFSIs and for MRS facilities (60 FR 32430) (Reference 5), the Commission responded to comments concerning a General Emergency at an ISFSI and MRS, and concluded that, an essential element of a General Emergency is that a release can be reasonably expected to exceed EPA Protective Action Guidelines exposure levels off site for more than the immediate site area. The probability of a condition reaching the level above an emergency classification of Alert is very low. In the event of an accident at a defueled facility that meets the conditions for relaxation of EP requirements, there will be time to take

LIC-16-0109 Page 24 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption measures to protect the public in accordance with a comprehensive approach to emergency planning. As stated in NUREG-1738, Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants (February 2001) (Reference 12) for instances of small SFP leaks or loss of cooling scenarios, these events evolve very slowly and generally leave many days for recovery efforts. Offsite radiation monitoring will be performed as the need arises. Due to the decreased risks associated with defueled plants, offsite radiation monitoring systems are not required. 40 C.2. By June 20, 2012, nuclear power reactor licensees shall In the Proposed Rule (74 FR 23254) (Reference 13) to amend establish and maintain the capability to assess, classify, and certain emergency planning requirements for 10 CFR Part 50, the declare an emergency condition within 15 minutes after the NRC asked for public comment on whether the NRC should add availability of indications to plant operators that an emergency requirements for non-power reactor licensees to assess, classify, action level has been exceeded and shall promptly declare the and declare an emergency condition within 15 minutes and promptly emergency condition as soon as possible following declare an emergency condition. The NRC received several identification of the appropriate emergency classification level. comments on these issues. The NRC believed there may be a need Licensees shall not construe these criteria as a grace period for the NRC to be aware of security-related events early on so that to attempt to restore plant conditions to avoid declaring an an assessment of the likelihood that the event is part of a larger emergency action due to an emergency action level that has coordinated attack can be made. However, the NRC determined that been exceeded. Licensees shall not construe these criteria as further analysis and stakeholder interactions are needed prior to preventing implementation of response actions deemed by the changing the requirements for non-power reactor licensees. licensee to be necessary to protect public health and safety Therefore, the NRC did not include requirements in the 2011 EP provided that any delay in declaration does not deny the State Final Rule (Reference 11) for non-power reactor licensees to and local authorities the opportunity to implement measures assess, classify, and declare an emergency condition within 15 necessary to protect the public health and safety. minutes and promptly declare an emergency condition.

LIC-16-0109 Page 25 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption FCS will maintain the capability to assess, classify, and declare an emergency condition within 30 minutes after the availability of indications to operators that an EAL threshold has been reached. Emergency declaration is required to be made as soon as conditions warranting classification are present and recognizable, but within 30 minutes in all cases of conditions being present. In the permanently defueled condition, the rapidly developing scenarios associated with events initiated during reactor power operation are no longer credible. The consequences resulting from the only remaining events (e.g., fuel handling accident) develop over a significantly longer period. As such, the 15 minute requirement to classify and declare an emergency is unnecessarily restrictive. Because of the geographic location of FCS, emergency planning and responsibilities have historically involved coordination with the States of Nebraska and Iowa. Decommissioning-related emergency plan submittals for FCS have been discussed with offsite response organizations since OPPD provided notification that it would permanently cease power operations. These discussions have addressed changes to onsite and offsite emergency preparedness throughout the decommissioning process, including the proposed 30-minute emergency declaration time. Emergency management officials with both states have agreed that this declaration time is appropriate. See the basis for 10 CFR 50.47(b) detailing the low likelihood of any credible accident resulting in radiological releases requiring offsite protective measures and 10 CFR Part 50, Appendix E, Section IV.1

LIC-16-0109 Page 26 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption for discussion on the similarity between a permanently defueled reactor and a non-power reactor. 41 D. Notification Procedures See the basis for 10 CFR 50.47(b) and 10 CFR 50.47(b)(10). D.1. Administrative and physical means for notifying local, State, and Federal officials and agencies and agreements reached with these officials and agencies for the prompt notification of the public and for public evacuation or other protective measures, should they become necessary, shall be described. This description shall include identification of the appropriate officials, by title and agency, of the State and local government agencies within the EPZs. 42 D.2. Provisions shall be described for yearly dissemination to See the basis for Section IV.D.1. the public within the plume exposure pathway EPZ of basic emergency planning information, such as the methods and times required for public notification and the protective actions planned if an accident occurs, general information as to the nature and effects of radiation, and a listing of local broadcast stations that will be used for dissemination of information during an emergency. Signs or other measures shall also be used to disseminate to any transient population within the plume exposure pathway EPZ appropriate information that would be helpful if an accident occurs. 43 D.3. A licensee shall have the capability to notify responsible While the capability needs to exist for the notification of offsite State and local governmental agencies within 15 minutes after government agencies within a specified time period, previous declaring an emergency. The licensee shall demonstrate that exemptions have allowed for extending the State and local

LIC-16-0109 Page 27 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption the appropriate governmental authorities have the capability to government agencies notification time up to 60 minutes based on make a public alerting and notification decision promptly on the site-specific justification provided. being informed by the licensee of an emergency condition. FCS proposes to complete emergency notification to the State of Prior to initial operation greater than 5 percent of rated thermal Nebraska within 60 minutes after an emergency declaration or a power of the first reactor at a site, each nuclear power reactor change in classification. This timeframe is consistent with the 10 licensee shall demonstrate that administrative and physical CFR 50.72(a)(3) notification to the NRC and is appropriate because means have been established for alerting and providing in the permanently defueled condition, the rapidly developing prompt instructions to the public within the plume exposure scenarios associated with events initiated during reactor power pathway EPZ. The design objective of the prompt public alert operation are no longer credible and there is no need for State or and notification system shall be to have the capability to local response organizations to implement any protective actions. essentially complete the initial alerting and initiate notification of the public within the plume exposure pathway EPZ within Because of the geographic location of FCS, emergency planning about 15 minutes. The use of this alerting and notification and responsibilities have historically involved coordination with the capability will range from immediate alerting and notification of States of Nebraska and Iowa. Decommissioning-related emergency the public (within 15 minutes of the time that State and local plan submittals for FCS have been discussed with offsite response officials are notified that a situation exists requiring urgent organizations since OPPD provided notification that it would action) to the more likely events where there is substantial permanently cease power operations. These discussions have time available for the appropriate governmental authorities to addressed changes to onsite and offsite emergency preparedness make a judgment whether or not to activate the public alert throughout the decommissioning process, including the proposed and notification system. The alerting and notification capability 60-minute notification to the State of Nebraska. Emergency shall additionally include administrative and physical means management officials with both states have agreed that the for a backup method of public alerting and notification capable proposed notification to Nebraska within 60 minutes is appropriate. of being used in the event the primary method of alerting and FCS analyses demonstrate that 530 days (1 year, 165 days) after notification is unavailable during an emergency to alert or permanent cessation of power operations, no remaining postulated notify all or portions of the plume exposure pathway EPZ accidents at FCS will result in radiological releases requiring offsite population. The backup method shall have the capability to protective actions, or in the event of beyond design basis accidents, alert and notify the public within the plume exposure pathway 10 hours is available to take mitigative actions, and if needed, EPZ, but does not need to meet the 15-minute design implement offsite protective actions using a comprehensive

LIC-16-0109 Page 28 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption objective for the primary prompt public alert and notification emergency management plan. Therefore, there is no need to system. When there is a decision to activate the alert and maintain an Alert and Notification System. notification system, the appropriate governmental authorities Also see the basis for 10 CFR 50.47(b) and 10 CFR 50.47(b)(10). will determine whether to activate the entire alert and notification system simultaneously or in a graduated or staged manner. The responsibility for activating such a public alert and notification system shall remain with the appropriate governmental authorities. 44 D.4. If FEMA has approved a nuclear power reactor site's alert See the basis for Section IV.D.3 regarding the alert and notification and notification design report, including the backup alert and system requirements. notification capability, as of December 23, 2011, then the backup alert and notification capability requirements in Section IV.D.3 must be implemented by December 24, 2012. If the alert and notification design report does not include a backup alert and notification capability or needs revision to ensure adequate backup alert and notification capability, then a revision of the alert and notification design report must be submitted to FEMA for review by June 24, 2013, and the FEMA-approved backup alert and notification means must be implemented within 365 days after FEMA approval. However, the total time period to implement a FEMA-approved backup alert and notification means must not exceed June 22, 2015. 45 E. Emergency Facilities and Equipment No exemption is requested. Adequate provisions shall be made and described for emergency facilities and equipment, including:

LIC-16-0109 Page 29 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption E.1. Equipment at the site for personnel monitoring; 46 E.2. Equipment for determining the magnitude of and for No exemption is requested. continuously assessing the impact of the release of radioactive materials to the environment; 47 E.3. Facilities and supplies at the site for decontamination of No exemption is requested. onsite individuals; 48 E.4. Facilities and medical supplies at the site for appropriate No exemption is requested. emergency first aid treatment; 49 E.5. Arrangements for medical service providers qualified to No exemption is requested. handle radiological emergencies onsite; 50 E.6. Arrangements for transportation of contaminated injured No exemption is requested. individuals from the site to specifically identified treatment facilities outside the site boundary; 51 E.7. Arrangements for treatment of individuals injured in No exemption is requested. support of licensed activities on the site at treatment facilities outside the site boundary; 52 E.8.a(i) A licensee onsite technical support center and an FCS analyses demonstrate that 530 days (1 year, 165 days) after emergency operations facility from which effective direction permanent cessation of power operations, no remaining postulated can be given and effective control can be exercised during an accidents at FCS will result in radiological releases requiring offsite emergency; protective actions, or in the event of beyond design basis accidents, 10 hours is available to take mitigative actions, and if needed, implement offsite protective actions using a comprehensive

LIC-16-0109 Page 30 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption emergency management plan. Therefore, there is no need to maintain a TSC or an EOF. Offsite agency response will not be required at an EOF and onsite actions may be directed from the Control Room or another location, without the requirements imposed on a Technical Support Center (TSC). An onsite facility will continue to be maintained, from which effective direction can be given and effective control may be exercised during an emergency. The FCS emergency plan will continue to maintain arrangements for requesting assistance and using resources from appropriate offsite support organizations. 53 E.8.a(ii) For nuclear power reactor licensees, a licensee onsite NUREG-0696, Functional Criteria for Emergency Response operational support center; Facilities, (Reference 14) provides that the operational support center (OSC) is an onsite area separate from the Control Room and the TSC where licensee operations support personnel will assemble in an emergency. For a permanently shut down and defueled power plant, an OSC is no longer required to meet its original purpose of an assembly area for plant logistical support during an emergency. A single onsite facility will continue to be maintained at FCS, from which Control Room support, emergency mitigation, radiation monitoring, and effective control may be exercised during an emergency.

LIC-16-0109 Page 31 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 54 E.8.b. For a nuclear power reactor licensee's emergency In accordance with paragraph 8.e. the requirements of paragraph operations facility required by paragraph 8.a of this section, 8.b.(1) - (5) do not apply to the FCS EOF because it was an either a facility located between 10 miles and 25 miles of the approved facility prior to December 23, 2011. However, the nuclear power reactor site(s), or a primary facility located less exemption is requested to clearly reflect that the requirement no than 10 miles from the nuclear power reactor site(s) and a longer applies to FCS in a permanently shut down and defueled backup facility located between 10 miles and 25 miles of the condition. nuclear power reactor site(s). An emergency operations See also basis for 10 CFR 50.47(b)(3). facility may serve more than one nuclear power reactor site. A licensee desiring to locate an emergency operations facility more than 25 miles from a nuclear power reactor site shall request prior Commission approval by submitting an application for an amendment to its license. For an emergency operations facility located more than 25 miles from a nuclear power reactor site, provisions must be made for locating NRC and offsite responders closer to the nuclear power reactor site so that NRC and offsite responders can interact face-to-face with emergency response personnel entering and leaving the nuclear power reactor site. Provisions for locating NRC and offsite responders closer to a nuclear power reactor site that is more than 25 miles from the emergency operations facility must include the following: 55 E.8.b.(1) Space for members of an NRC site team and Federal, State, and local responders 56 E.8.b.(2) Additional space for conducting briefings with emergency response personnel;

LIC-16-0109 Page 32 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 57 E.8.b.(3) Communication with other licensee and offsite emergency response facilities; 58 E.8.b.(4) Access to plant data and radiological information; and 59 E.8.b.(5) Access to copying equipment and office supplies; 60 E.8.c. By June 20, 2012, for a nuclear power reactor See the basis for 10 CFR 50.47(b)(3). licensee's emergency operations facility required by paragraph 8.a of this section, a facility having the following capabilities: (1) The capability for obtaining and displaying plant data and radiological information for each reactor at a nuclear power reactor site and for each nuclear power reactor site that the facility serves; 61 E.8.c.(2) The capability to analyze plant technical information and provide technical briefings on event conditions and prognosis to licensee and offsite response organizations for each reactor at a nuclear power reactor site and for each nuclear power reactor site that the facility serves; and 62 E.8.c.(3) The capability to support response to events occurring simultaneously at more than one nuclear power reactor site if the emergency operations facility serves more than one site; and

LIC-16-0109 Page 33 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 63 E.8.d. For nuclear power reactor licensees, an alternative See the basis for Section IV.1 regarding hostile action. facility (or facilities) that would be accessible even if the site is under threat of or experiencing hostile action, to function as a staging area for augmentation of emergency response staff and collectively having the following characteristics: the capability for communication with the emergency operations facility, control room, and plant security; the capability to perform offsite notifications; and the capability for engineering assessment activities, including damage control team planning and preparation, for use when onsite emergency facilities cannot be safely accessed during hostile action. The requirements in this paragraph 8.d must be implemented no later than December 23, 2014, with the exception of the capability for staging emergency response organization personnel at the alternative facility (or facilities) and the capability for communications with the emergency operations facility, control room, and plant security, which must be implemented no later than June 20, 2012. 64 E.8.e. A licensee shall not be subject to the requirements of See the basis for 10 CFR 50.47(b)(3) and Appendix E, Section paragraph 8.b of this section for an existing emergency IV.E.8.b. operations facility approved as of December 23, 2011; 65 E.9. At least one onsite and one offsite communications See the basis for 10 CFR 50.47(b) and (b)(10). system; each system shall have a backup power source. All FCS will maintain communications with the State of Nebraska, communication plans shall have arrangements for Washington County, and the NRC. The onsite response facilities will emergencies, including titles and alternates for those in be combined into a single facility, as described in IV.E.8.a(ii). charge at both ends of the communication links and the

LIC-16-0109 Page 34 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption primary and backup means of communication. Where consistent with the function of the governmental agency, these arrangements will include: E.9.a. Provision for communications with contiguous State/local governments within the plume exposure pathway EPZ. Such communications shall be tested monthly. 66 E.9.b. Provision for communications with Federal emergency No exemption is requested. response organizations. Such communications systems shall be tested annually. 67 E.9.c. Provision for communications among the nuclear power FCS analyses demonstrate that 530 days (1 year, 165 days) after reactor control room, the onsite technical support center, and permanent cessation of power operations, no remaining postulated the emergency operations facility; and among the nuclear accidents at FCS will result in radiological releases requiring offsite facility, the principal State and local emergency operations protective actions, or in the event of beyond design basis accidents, centers, and the field assessment teams. Such 10 hours is available to take mitigative actions, and if needed, communications systems shall be tested annually. implement offsite protective actions using a comprehensive emergency management plan. Therefore, there is no need for the TSC, EOF, or field assessment teams. Also see justification for 10 CFR 50.47(b)(3). The provisions remaining in 10 CFR Part 50, Appendix E, Section IV.E.9.a, b, and d include the necessary requirements. Communication with State and local EOCs will be maintained to coordinate assistance on site, if required.

LIC-16-0109 Page 35 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 68 E.9.d. Provisions for communications by the licensee with The functions of the Control Room, EOF, TSC, and OSC may be NRC Headquarters and the appropriate NRC Regional Office combined into one or more locations due to the smaller facility staff Operations Center from the nuclear power reactor control and the greatly reduced interaction required with State and local room, the onsite technical support center, and the emergency emergency response facilities. An onsite facility will continue to be operations facility. Such communications shall be tested maintained, from which effective command and control will be monthly. maintained and direction can be given during an emergency. FCS will maintain communications capability with the NRC. Also see the basis for 10 CFR 50.47(b). 69 F. Training No exemption is requested. F.1. The program to provide for: (a) The training of employees and exercising, by periodic drills, of emergency plans to ensure that employees of the licensee are familiar with their specific emergency response duties, and (b) The participation in the training and drills by other persons whose assistance may be needed in the event of a radiological emergency shall be described. This shall include a description of specialized initial training and periodic retraining programs to be provided to each of the following categories of emergency personnel: 70 F.1.i. Directors and/or coordinators of the plant emergency organization; 71 F.1.ii. Personnel responsible for accident assessment, including control room shift personnel; 72 F.1.iii. Radiological monitoring teams;

LIC-16-0109 Page 36 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 73 F.1.iv. Fire control teams (fire brigades); 74 F.1.v. Repair and damage control teams; 75 F.1.vi. First aid and rescue teams; 76 F.1.vii. Medical support personnel; 77 F.1.viii. Licensee's headquarters support personnel; The number of staff at FCS during the decommissioning process will be small but commensurate with the need to safely store spent fuel at the facility in a manner that is protective of public health and safety. FCS will maintain a level of emergency response that does not require additional response by headquarters personnel. The on-shift and emergency response positions are defined in the Permanently Defueled Emergency Plan and will be regularly tested through drills and exercises, audited, and inspected by FCS and the NRC. Also see the basis for 10 CFR 50.47(b). Therefore, exempting licensee's headquarters personnel from training requirements is considered to be reasonable. 78 F.1.ix. Security personnel. No exemption is requested. 79 F.1. In addition, a radiological orientation training program Because there will no longer be any expected actions that must be shall be made available to local services personnel; e.g., local taken by the public during an emergency, it is no longer necessary emergency services/Civil Defense, local law enforcement to pre-plan the dissemination of this information to the public or to personnel, local news media persons. provide radiological orientation training to local news media persons.

LIC-16-0109 Page 37 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption The phrase "Civil Defense" is no longer a commonly used term and is no longer applicable as an example in the regulation. 80 F.2. The plan shall describe provisions for the conduct of FCS analyses demonstrate that 530 days (1 year, 165 days) after emergency preparedness exercises as follows: permanent cessation of power operations, no remaining postulated accidents at FCS will result in radiological releases requiring offsite Exercises shall test the adequacy of timing and content of protective actions, or in the event of beyond design basis accidents, implementing procedures and methods, test emergency 10 hours is available to take mitigative actions, and if needed, equipment and communications networks, test the public alert implement offsite protective actions using a comprehensive and notification system, and ensure that emergency emergency management plan. Therefore, the public alert and organization personnel are familiar with their duties.3 notification system will not be used and no testing would be required. Also see the basis for 10 CFR 50.47(b). 81 F.2.a. A full participation exercise4 which tests as much of the FCS will continue to invite the State of Nebraska and Washington licensee, State, and local emergency plans as is reasonably County to participate in the periodic drills and exercises conducted achievable without mandatory public participation shall be to assess their ability to perform responsibilities related to an conducted for each site at which a power reactor is located. emergency at FCS, to the extent defined by the FCS emergency Nuclear power reactor licensees shall submit exercise plan. Because the need for offsite emergency planning is relaxed scenarios under § 50.4 at least 60 days before use in a full due to the low probability of the postulated accident or other credible participation exercise required by this paragraph 2.a. events that would be expected to result in an offsite radioactive release that would exceed the EPA PAGs and the available time for 82 F.2.a.(i) For an operating license issued under this part, this event mitigation, no offsite emergency plans will be in place to test. exercise must be conducted within two years before the The intent of submitting exercise scenarios at power reactors is to issuance of the first operating license for full power (one verify that licensees utilize different scenarios in order to prevent the authorizing operation above 5 percent of rated power) of the preconditioning of responders at power reactors. For defueled sites, first reactor and shall include participation by each State and there are limited events that could occur and the previously routine local government within the plume exposure pathway EPZ

LIC-16-0109 Page 38 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption and each state within the ingestion exposure pathway EPZ. If progression to General Emergency in power reactor site scenarios is the full participation exercise is conducted more than 1 year not applicable to a decommissioning site. prior to issuance of an operating licensee for full power, an OPPD considers FCS to be exempt from F.2.a.(i) - (iii) because FCS exercise which tests the licensee's onsite emergency plans will be exempt from the umbrella provision of Section IV.F.2.a. must be conducted within one year before issuance of an operating license for full power. This exercise need not have State or local government participation. 83 F.2.a.(ii) For a combined license issued under part 52 of this chapter, this exercise must be conducted within two years of the scheduled date for initial loading of fuel. If the first full participation exercise is conducted more than one year before the scheduled date for initial loading of fuel, an exercise which tests the licensee's onsite emergency plans must be conducted within one year before the scheduled date for initial loading of fuel. This exercise need not have State or local government participation. If FEMA identifies one or more deficiencies in the state of offsite emergency preparedness as the result of the first full participation exercise, or if the Commission finds that the state of emergency preparedness does not provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, the provisions of § 50.54(gg) apply. 84 F.2.a.(iii) For a combined license issued under part 52 of this chapter, if the applicant currently has an operating reactor at the site, an exercise, either full or partial participation,5 shall be conducted for each subsequent reactor constructed on the

LIC-16-0109 Page 39 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption site. This exercise may be incorporated in the exercise requirements of Sections IV.F.2.b. and c. in this appendix. If FEMA identifies one or more deficiencies in the state of offsite emergency preparedness as the result of this exercise for the new reactor, or if the Commission finds that the state of emergency preparedness does not provide reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, the provisions of § 50.54(gg) apply. 85 F.2.b. Each licensee at each site shall conduct a subsequent See the basis for Section IV.F.2.a. exercise of its onsite emergency plan every 2 years. Nuclear The low probability of the postulated accident or other credible power reactor licensees shall submit exercise scenarios under events that would result in an offsite radioactive release that would

        § 50.4 at least 60 days before use in an exercise required by exceed the EPA PAGs and the available time for event mitigation at this paragraph 2.b. The exercise may be included in the full FCS during decommissioning render the TSC, OSC, and EOF participation biennial exercise required by paragraph 2.c. of unnecessary. The principal functions required by regulation can be this section. In addition, the licensee shall take actions performed at a single onsite location that does not meet the necessary to ensure that adequate emergency response requirements of the TSC, OSC, or EOF. The onsite response capabilities are maintained during the interval between facilities at FCS will be combined into a single facility.

biennial exercises by conducting drills, including at least one drill involving a combination of some of the principal functional FCS will continue to conduct biennial exercises and will invite the areas of the licensee's onsite emergency response State of Nebraska and local support organizations (firefighting, law capabilities. The principal functional areas of emergency enforcement, and ambulance/medical services) to participate in response include activities such as management and periodic drills and exercises to assess their ability to perform coordination of emergency response, accident assessment, responsibilities related to an emergency at FCS, to the extent event classification, notification of offsite authorities, defined by the FCS emergency plan. assessment of the onsite and offsite impact of radiological The intent of submitting exercise scenarios for use by power reactor releases, protective action recommendation development, licensees is to check that licensees utilize different scenarios in

LIC-16-0109 Page 40 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption protective action decision making, plant system repair and order to prevent the preconditioning of responders at power mitigative action implementation. During these drills, activation reactors. In FCSs permanently shut down and defueled condition, of all of the licensee's emergency response facilities there are limited events that could occur and the previously routine (Technical Support Center (TSC), Operations Support Center progression to General Emergency in scenarios is not applicable to (OSC), and the Emergency Operations Facility (EOF)) would a decommissioning site. not be necessary, licensees would have the opportunity to consider accident management strategies, supervised instruction would be permitted, operating staff in all participating facilities would have the opportunity to resolve problems (success paths) rather than have controllers intervene, and the drills may focus on the onsite exercise training objectives. 86 F.2.c. Offsite plans for each site shall be exercised biennially See the basis for Section IV.F.2.a. with full participation by each offsite authority having a role under the radiological response plan. Where the offsite authority has a role under a radiological response plan for more than one site, it shall fully participate in one exercise every two years and shall, at least, partially participate in other offsite plan exercises in this period. If two different licensees each have licensed facilities located either on the same site or on adjacent, contiguous sites, and share most of the elements defining co-located licensees,6 then each licensee shall: 87 F.2.c.(1) Conduct an exercise biennially of its onsite emergency plan;

LIC-16-0109 Page 41 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 88 F.2.c.(2) Participate quadrennially in an offsite biennial full or partial participation exercise; 89 F.2.c.(3) Conduct emergency preparedness activities and interactions in the years between its participation in the offsite full or partial participation exercise with offsite authorities, to test and maintain interface among the affected State and local authorities and the licensee. Co-located licensees shall also participate in emergency preparedness activities and interaction with offsite authorities for the period between exercises; 90 F.2.c.(4) Conduct a hostile action exercise of its onsite emergency plan in each exercise cycle; and 91 F.2.c.(5) Participate in an offsite biennial full or partial participation hostile action exercise in alternating exercise cycles. 92 F.2.d. Each State with responsibility for nuclear power reactor See the basis for Section IV.2. emergency preparedness should fully participate in the ingestion pathway portion of exercises at least once every exercise cycle. In States with more than one nuclear power reactor plume exposure pathway EPZ, the State should rotate this participation from site to site. Each State with responsibility for nuclear power reactor emergency preparedness should fully participate in a hostile action exercise at least once every cycle and should fully participate

LIC-16-0109 Page 42 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption in one hostile action exercise by December 31, 2015. States with more than one nuclear power reactor plume exposure pathway EPZ should rotate this participation from site to site. 93 F.2.e. Licensees shall enable any State or local government See the basis for Section IV.2. located within the plume exposure pathway EPZ to participate in the licensee's drills when requested by such State or local government. 94 F.2.f. Remedial exercises will be required if the emergency FEMA is responsible for evaluating the adequacy of an offsite plan is not satisfactorily tested during the biennial exercise, response exercise. No action is expected from State or local such that NRC, in consultation with FEMA, cannot (1) find government organizations in response to an event at a reasonable assurance that adequate protective measures can decommissioning site other than receiving notification of the and will be taken in the event of a radiological emergency or emergency and firefighting, law enforcement, and (2) determine that the Emergency Response Organization ambulance/medical response services. Letters of Agreement will (ERO) has maintained key skills specific to emergency continue to be in place for those services. Offsite response response. The extent of State and local participation in organizations will continue to implement actions to protect the health remedial exercises must be sufficient to show that appropriate and safety of the public as they would at any other industrial site. corrective measures have been taken regarding the elements of the plan not properly tested in the previous exercises. 95 F.2.g. All exercises, drills, and training that provide No exemption is requested. performance opportunities to develop, maintain, or demonstrate key skills must provide for formal critiques in order to identify weak or deficient areas that need correction. Any weaknesses or deficiencies that are identified in a critique of exercises, drills, or training must be corrected.

LIC-16-0109 Page 43 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 96 F.2.h. The participation of State and local governments in an No exemption is requested. emergency exercise is not required to the extent that the applicant has identified those governments as refusing to participate further in emergency planning activities, pursuant to § 50.47(c)(1). In such cases, an exercise shall be held with the applicant or licensee and such governmental entities as elect to participate in the emergency planning process. 97 F.2.i. Licensees shall use drill and exercise scenarios that At FCS, there will be limited events that could occur that could result provide reasonable assurance that anticipatory responses will in radiological releases that exceed the EPA PAGs and the not result from preconditioning of participants. Such scenarios previously routine progression to General Emergency in power for nuclear power reactor licensees must include a wide reactor site scenarios will not be applicable. Therefore, FCS does spectrum of radiological releases and events, including hostile not expect to demonstrate response to a wide spectrum of events. action. Exercise and drill scenarios as appropriate must Also see the basis for 10 CFR 50.47(b) detailing the low likelihood of emphasize coordination among onsite and offsite response any credible accident resulting in radiological releases requiring organizations. offsite protective measures and basis for Section IV.1 regarding hostile action. 98 F.2.j. The exercises conducted under paragraph 2 of this See the basis for Section IV.F.2. section by nuclear power reactor licensees must provide the Periodic drills and exercises will be completed to demonstrate ERO opportunity for the ERO to demonstrate proficiency in the key proficiency in key skills necessary to implement the principal skills necessary to implement the principal functional areas of functional areas of emergency response as applicable for the emergency response identified in paragraph 2.b of this permanently defueled plant status. Critiques will follow each drill or section. Each exercise must provide the opportunity for the exercise activity. FCS will continue to invite the State of Nebraska ERO to demonstrate key skills specific to emergency and local support organizations to participate in the periodic drills response duties in the control room, TSC, OSC, EOF, and and exercises to assess their ability to perform responsibilities joint information center. Additionally, in each eight calendar

LIC-16-0109 Page 44 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption year exercise cycle, nuclear power reactor licensees shall related to an emergency at FCS to the extent defined by the FCS vary the content of scenarios during exercises conducted emergency plan. under paragraph 2 of this section to provide the opportunity for the ERO to demonstrate proficiency in the key skills necessary to respond to the following scenario elements: hostile action directed at the plant site, no radiological release or an unplanned minimal radiological release that does not require public protective actions, an initial classification of or rapid escalation to a Site Area Emergency or General Emergency, implementation of strategies, procedures, and guidance developed under § 50.54(hh)(2), and integration of offsite resources with onsite response. The licensee shall maintain a record of exercises conducted during each eight year exercise cycle that documents the content of scenarios used to comply with the requirements of this paragraph. Each licensee shall conduct a hostile action exercise for each of its sites no later than December 31, 2015. The first eight-year exercise cycle for a site will begin in the calendar year in which the first hostile action exercise is conducted. For a site licensed under Part 52, the first eight-year exercise cycle begins in the calendar year of the initial exercise required by Section IV.F.2.a. 99 G. Maintaining Emergency Preparedness No exemption is requested. Provisions to be employed to ensure that the emergency plan, its implementing procedures, and emergency equipment and supplies are maintained up to date shall be described.

LIC-16-0109 Page 45 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 100 H. Recovery No exemption is requested. Criteria to be used to determine when, following an accident, reentry of the facility would be appropriate or when operation could be resumed shall be described. 101 I. Onsite Protective Actions During Hostile Action See the basis for Section IV.1. By June 20, 2012, for nuclear power reactor licensees, a range of protective actions to protect onsite personnel during hostile action must be developed to ensure the continued ability of the licensee to safely shut down the reactor and perform the functions of the licensees emergency plan. 102 10 CFR 50 Appendix E No exemption is requested. V. Implementing Procedures No less than 180 days before the scheduled issuance of an operating license for a nuclear power reactor or a license to possess nuclear material, or the scheduled date for initial loading of fuel for a combined license under part 52 of this chapter, the applicants or licensee's detailed implementing procedures for its emergency plan shall be submitted to the Commission as specified in § 50.4.

LIC-16-0109 Page 46 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 103 10 CFR 50 Appendix E The regulation that identifies the requirement to maintain the Emergency Response Data System (ERDS) is not applicable to VI. Emergency Response Data System nuclear power facilities that are permanently shut down.

1. The Emergency Response Data System (ERDS) is a direct When FCS is permanently defueled, this system will no longer be near real-time electronic data link between the licensee's necessary to transmit safety system parameter data. No exemption onsite computer system and the NRC Operations Center that is requested because this change in the ERDS data requirement is provides for the automated transmission of a limited data set identified in 10 CFR Part 50, Appendix E, Section VI.2.

of selected parameters. The ERDS supplements the existing voice transmission over the Emergency Notification System (ENS) by providing the NRC Operations Center with timely and accurate updates of a limited set of parameters from the licensee's installed onsite computer system in the event of an emergency. When selected plant data are not available on the licensee's onsite computer system, retrofitting of data points is not required. The licensee shall test the ERDS periodically to verify system availability and operability. The frequency of ERDS testing will be quarterly unless otherwise set by NRC based on demonstrated system performance.

2. Except for Big Rock Point and all nuclear power facilities that are shut down permanently or indefinitely, onsite hardware shall be provided at each unit by the licensee to interface with the NRC receiving system. Software, which will be made available by the NRC, will assemble the data to be transmitted and transmit data from each unit via an output port on the appropriate data system.

LIC-16-0109 Page 47 Table 2 Exemptions Requested from 10 CFR 50, Appendix E Item # Regulation in Part 50, Appendix E Basis for Exemption 104 10 CFR 50 Appendix E OPPD considers FCS to be exempt from Footnotes 3, 4, 5, and 6 because FCS will be exempt from the umbrella provisions of Section Footnotes 3, 4, 5, and 6 are proposed for exemption. F.2.

LIC-16-0109 Page 48

4.0 TECHNICAL EVALUATION

4.1 Accident Analysis Overview 10 CFR 50.82(a)(2) specifies that the 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel after docketing the certifications for permanent cessation of power operations and permanent removal of fuel from the reactor vessel in accordance with 10 CFR 50.82(a)(1)(i) and (ii). Following the termination of power operations at FCS and the permanent removal of the fuel from the reactor vessel, the postulated accidents involving failure or malfunction of the reactor and supporting structures, systems, and components are no longer applicable. A summary of the postulated radiological accidents analyzed for the permanently shut down and defueled condition is presented below. According to the EPA, Protective Action Guides and Planning Guidance for Radiological Incidents, Draft for Interim Use and Public Comment, dated March 2013 (Reference 10), Section 2.3.5, PAGs and Nuclear Facilities Emergency Planning Zones (EPZ), EPZs are not necessary at those facilities where it is not possible for PAGs to be exceeded offsite. Section 5.0 of ISG-02 (Reference 1) indicates that site-specific analyses should demonstrate that: (1) the radiological consequences of the remaining applicable postulated accident would not exceed the limits of the EPA PAGs at the EAB; (2) in the event of a beyond design basis event resulting in the partial drain down of the SFP to the point that cooling is not effective, there is at least 10 hours (assuming an adiabatic heat up) from the time that the fuel is no longer being cooled until the hottest fuel assembly reaches 900C; (3) adequate physical security is in place to assure implementation of security strategies that protect against spent fuel sabotage; and (4) in the unlikely event of a beyond design basis events resulting in a loss of all SFP cooling, there is sufficient time to implement pre-planned mitigation measures to provide makeup or spray to the SFP before the onset of a zirconium cladding ignition. Table 3 contains a listing of seven analyses that are expected to be evaluated by a decommissioning power reactor licensee requesting exemption of emergency planning requirements. The table also contains a description of how FCS addresses each of these analyses.

LIC-16-0109 Page 49 TABLE 3 Interim Staff Guidance-02 Comparison Analysis ISG-02 Description Response 1 Applicable design DBAs (i.e., fuel The postulated accident that will remain applicable to FCS and handling accident in the spent could contribute to dose upon implementation of the requested fuel storage facility, waste gas exemptions is the fuel handling accident (FHA) in the Auxiliary system release, and cask Building, where the SFP is located. The results of the analysis handling accident if the cask indicate that the dose at the EAB would not exceed the EPA handling system is not licensed PAGs within 10 days after permanent cessation of power as single-failure-proof) (Indicates operations. that any radiological release This analysis is described in Section 4.2. would not exceed the limits of EPA PAGs at EAB); 2 Complete loss of SFP water FCS performed an analysis that conservatively evaluates the inventory with no heat loss length of time (number of hours) it takes for uncovered spent (adiabatic heatup) demonstrating fuel assemblies in the SFP to reach the temperature at which a minimum of 10 hours is the zirconium cladding would fail. Based on the limiting fuel available before any fuel cladding assembly for decay heat and adiabatic heat up analysis, at 530 temperature reaches 900 days (1 year, 165 days) after permanent cessation of power degrees Celsius from the time all operations, the time for the hottest fuel assembly to reach cooling is lost (Demonstrates 900°C is 10 hours after spent fuel cooling is lost. sufficient time to mitigate events This analysis is described in Section 4.3 and is included in that could lead to a zirconium Attachment 2. cladding fire); 3 Loss of SFP water inventory FCS performed an analysis to determine the offsite radiological resulting in radiation exposure at impact of a complete loss of SFP water. It was determined that the EAB and control room; the gamma radiation dose rate at the EAB would be limited to (Indicates that any release is less small fractions of the EPA PAG exposure levels. than EPA PAGs at EAB); and This analysis is described in Section 4.4 and is included in Attachment 3. 4 Considering the site-specific FCS conducted a seismic evaluation in response to a NRC seismic hazard, either an request for information pursuant to 10 CFR 50.54(f) regarding evaluation demonstrating a high Recommendation 2.1 of the Near-Term Task Force (NTTF) confidence of a low-probability Review of Insights from the Fukushima Dai-ichi Accident. The (less than 1 x 10-5 per year) of seismic evaluation included all structures including the SFP, seismic failure of the spent fuel and was prepared and submitted for NRC review. The OPPD storage pool structure or an submittal (LIC-14-0047) (Reference 15) documents the seismic analysis demonstrating the fuel evaluation in conformance with NTTF Recommendation 2.1 has decayed sufficiently that including the high confidence of a low-probability of seismic natural air flow in a completely failure (HCLPF) values and the 1 x 10-5 per year hazard level. drained pool would maintain The Staff review of the NTTF submittal, specifically for the SFP peak cladding temperature below Evaluation associated with the reevaluated seismic hazard 565 degrees Celsius (the point of implementing NTTF Recommendation 2.1 (CAC No. MF3735) incipient cladding damage) is documented in NRC-16-0068 (ML16182A361) (Reference (Indicates that any release is less 16). The NRC staff concluded that the assessment was than EPA PAGs at EAB). performed consistent with the NRC-endorsed (ML15350A158) (Reference 17) SFP Evaluation Guidance Report (Reference

18) and provided sufficient information, including the SFP

LIC-16-0109 Page 50 Analysis ISG-02 Description Response integrity evaluation, to meet the SFP Evaluation Guidance (Item 9 in Enclosure 1 of the NRCs 50.54(f) letter), thus supporting SDA No. 6 of NUREG-1738. Therefore, the air-cooled analysis is not being performed. The analysis is described in SDA-6 of Table 5. 5 The analyses and conclusions IDCs and SDAs are addressed in Section 4.5 and Tables 4 and described in NUREG-1738 are 5. predicated on the risk reduction measures identified in the study as Industry Decommissioning Commitments (IDC) and Staff Decommissioning Assumptions (SDA), listed in Tables 4.1-1 and 4.1-2 of that document. The staff should ensure that the licensee has addressed these IDCs and SDAs for the decommissioning site if they are storing fuel in an SFP. 6 Verify that the licensee presents The onsite restoration plans for repair of the SFP cooling a determination that there is system and to provide makeup water to the SFP are sufficient resources and incorporated into FCS procedures. adequately trained personnel There are multiple ways to initiate mitigative actions and add available on-shift to initiate makeup water to the SFP within the 10-hour minimum time mitigative actions within the 10-period with or without entry to the SFP floor. hour minimum time period that will prevent an offsite radiological Refer to SDA 2 in Table 5. release that exceeds the EPA PAGs at the EAB. 7 Verify that mitigation strategies FCS maintains procedures and strategies for the movement of are consistent with that required any necessary portable equipment that will be relied upon for by the Permanently Defueled mitigating the loss of SFP water. These mitigative strategies Technical Specifications or by were developed in response to 10 CFR 50.54(hh)(2) and are retained license conditions. maintained in accordance with License Condition 3.G of the FCS Renewed Facility Operating License. These diverse strategies provide defense-in-depth and ample time to provide makeup water or spray to the SFP prior to the onset of zirconium cladding ignition when considering very low probability beyond design basis events affecting the SFP. Refer to SDA 4 in Table 5.

LIC-16-0109 Page 51 4.2 Consequences of a Postulated Accident While spent fuel remains in the SFP, the only postulated accident that will remain applicable to FCS that could contribute to dose upon implementation of the requested exemptions is the FHA in the Auxiliary Building, where the SFP is located. FCS maintains an analysis (Calculation FC08557, Fuel Handling Accident in the Spent Fuel Pool Site Boundary and Control Room Dose (Reference 29)) that has determined the EAB dose due to a FHA occurring in the Auxiliary Building. The FHA analysis is performed using selective application of the Alternative Source Term (AST), the guidance in Regulatory Guide 1.183, Appendix B (Reference 19), and Total Effective Dose Equivalent (TEDE) dose criteria. The results of the analysis indicate that the EAB dose is within regulatory allowable limits for a FHA occurring in the Auxiliary Building within 10 days after shutdown. The results of this analysis may be applied after November 13, 2016, the date that OPPD certified that all fuel has been permanently removed from the reactor vessel and placed in the SFP (Reference 4). 4.3 Consequences of a Beyond Design Basis Event With respect to beyond design basis events, FCS analyzed a partial drain down of the SFP water that would effectively impede any decay heat removal (adiabatic heatup). The analysis (Calculation FC08104, Maximum Cladding Temperature Analysis for Adiabatic Heat-up of Spent Fuel Assembly (Reference 30)) compares the conditions for the hottest fuel assembly stored in the FCS SFP to a criterion proposed in SECY-99-168 (Reference 20) applicable to offsite emergency response for a unit in the decommissioning process. This criterion considers the time for the hottest assembly to heat up from 30°C to 900°C adiabatically. If the heat up time is greater than 10 hours from the time spent fuel cooling is lost, then offsite emergency preplanning involving the plant is not necessary. Based on the limiting fuel assembly decay heat and adiabatic heat up analysis, 530 days (1 year, 165 days) after permanent cessation of power operations, the time for the hottest fuel assembly to reach 900°C is 10 hours after the assemblies have been uncovered. As stated in NUREG-1738, Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants (February 2001) (Reference 12), 900°C is an acceptable temperature to use for assessing the onset of fission product release under transient conditions (to establish the critical decay time for determining availability of 10 hours to evacuate) if fuel and cladding oxidation occurs in air. Because of the length of time it would take for the adiabatic heat up to occur, there is ample time to respond to any partial drain down event that might cause such an occurrence by restoring SFP cooling or makeup, or providing spray. As a result, the likelihood that such a scenario would progress to a zirconium fire is not deemed credible. The analysis is included in Attachment 2.

LIC-16-0109 Page 52 4.4 Consequences of Other Analyzed Events FCS analyzed a drain down event of the SFP to determine a dose rate curve at the EAB and Control Room. NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, (Reference 21) Supplement 1, Section 4.3.9, identifies that a SFP drain down event is a beyond design basis event. Although Calculation FC08104, Maximum Cladding Temperature Analysis for Adiabatic Heat-up of Spent Fuel Assembly, demonstrated a significant release of radioactive material from the spent fuel is not possible in the absence of water cooling after 530 days (1 year, 165 days) following permanent cessation of power operations, the potential exists for radiation exposure to an offsite individual in the event that shielding of the fuel is lost. The SFP water and the concrete pool structure serve as radiation shielding. A loss of water shielding above the fuel could increase the offsite radiation levels because of the gamma rays streaming up out of the SFP being scattered back to a receptor at the site boundary. The offsite and Control Room radiological impacts of a postulated complete loss of SFP water were assessed in Calculation FC08513, EAB Radiation Shine Dose 18 Months Post Shutdown with the SFP Drained (Reference 31). It was determined that the gamma radiation dose rate at the EAB would be limited to small fractions of the EPA PAGs. The EPA PAGs were developed to respond to a mobile airborne plume that could transport and deposit radioactive material over a large area. In contrast, the radiation field formed by scatter from a drained SFP would be stationary rather than moving and would not cause transport or deposition of radioactive materials. The extended period required to exceed the EPA PAG limit of 1 Rem TEDE would allow sufficient time to develop and implement onsite mitigative actions and provide confidence that additional offsite measures could be taken without planning if efforts to reestablish shielding over the fuel are delayed. Based on the data presented in Calculation FC08513, 530 days (1 year, 165 days) following permanent cessation of operations, the dose rate in the Control Room during an event involving a complete loss of SFP water will be below 2.32 x 10-3 mRem/hr, which is less than 15 mRem/hr. Calculation FC08513 is included in Attachment 3 and 4. 4.5 Comparison to NUREG-1738 Industry Decommissioning Commitments and Staff Decommissioning Assumptions Although the limited scope of the postulated accident and beyond design basis events that remain applicable to FCS justify a reduction in the necessary scope of emergency response capabilities, OPPD also evaluated the IDCs and SDAs contained in NUREG-1738 (Reference 12).

LIC-16-0109 Page 53 NUREG-1738 contains the results of the NRC staffs evaluation of the potential accident risk in SFPs at decommissioning plants in the United States. The study was undertaken to support development of a risk-informed technical basis for reviewing regulatory exemption requests and a regulatory framework for integrated rulemaking. The NRC staff performed analyses and sensitivity studies on evacuation timing to assess the risk significance of relaxed offsite emergency preparedness requirements during decommissioning. The staff based its sensitivity assessment on the guidance in Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis (Reference 22). The staff's analyses and conclusions apply to decommissioning facilities with SFPs that meet the design and operational characteristics assumed in the risk analysis. The study found that the risk of a potential SFP accident at decommissioning plants is low and well within the Commission's Safety Goals. The risk is low because of the very low likelihood of a zirconium fire (resulting from a postulated irrecoverable loss of SFP cooling water inventory) even though the consequences from a zirconium fire could be serious. The study provided the following assessment: The staff found that the event sequences important to risk at decommissioning plants are limited to large earthquakes and cask drop events. For emergency planning (EP) assessments, this is an important difference relative to operating plants where typically a large number of different sequences make significant contributions to risk. Relaxation of offsite EP a few months after shutdown resulted in only a "small change" in risk, consistent with the guidance of RG 1.174. Figures ES-1 and ES-2 [in NUREG-1738] illustrate this finding. The change in risk due to relaxation of offsite EP is small because the overall risk is low, and because even under current EP requirements, EP was judged to have marginal impact on evacuation effectiveness in the severe earthquakes that dominate SFP risk. All other sequences including cask drops (for which emergency planning is expected to be more effective) are too low in likelihood to have a significant impact on risk. For comparison, at operating reactors, additional risk-significant accidents for which EP is expected to provide dose savings are on the order of 1x10-5 per year, while for decommissioning facilities, the largest contributor for which EP would provide dose savings is about two orders of magnitude lower (cask drop sequence at 2x10-7 per year). The Executive Summary in NUREG-1738 states, in part, "the staff's analyses and conclusions apply to decommissioning facilities with SFPs that meet the design and operational characteristics assumed in the risk analysis. These characteristics are identified in the study as IDCs and SDAs. Provisions for confirmation of these characteristics would need to be an integral part of rulemaking." The IDCs and SDAs are listed in Tables 4.1-1 and 4.1-2, respectively, of NUREG-1738. Tables 4 and 5 describe how the FCS SFP meets or compares with each of these IDCs (Table 4) and SDAs (Table 5). Attachment 4 includes a new regulatory commitment to update the FCS USAR with this information.

LIC-16-0109 Page 54 4.6 Consequences of a Beyond Design Basis Earthquake NUREG-1738 (Reference 12) identifies beyond design basis seismic events as the dominant contributor to events that could result in a loss of SFP coolant that uncovers fuel for plants in the Central and Eastern United States. Additionally, NUREG-1738 identifies a zirconium fire resulting from a substantial loss-of-water inventory from the SFP, as the only postulated scenario at a decommissioning plant that could result in a significant offsite radiological release. The scenarios that lead to this condition have very low frequencies of occurrence (i.e., on the order of one to tens of times in a million years) and are considered beyond design basis events because the SFP and attached systems are designed to prevent a substantial loss of coolant inventory under accident conditions. However, the consequences of such accidents could potentially lead to an offsite radiological dose in excess of the EPA PAGs (Reference 10) at the EAB. However, the risk associated with zirconium cladding fire events decreases as the spent fuel ages, decay time increases, decay heat decreases, and short-lived radionuclides decay away. As decay time increases, the overall risk of a zirconium cladding fire continues to decrease due to two factors: (1) the amount of time available for preventative actions increases, which reduces the probability that the actions would not be successful; and (2) the increased likelihood that the fuel is able to be cooled by air, which decreases the reliance on actions to prevent a zirconium fire. The results of research conducted for NUREG-1738 and NUREG-2161, Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor, dated September 2014 (Reference 23), suggest that, while other radiological consequences can be extensive, a postulated accident scenario leading to a SFP zirconium fire, where the fuel has had significant decay time, will have little potential to cause offsite early fatalities, regardless of the type of offsite response (i.e., formal offsite radiological emergency preparedness plan or Comprehensive Emergency Management Plan). The purpose of NUREG-2161 (Reference 23) was to determine if accelerated transfer of older, colder spent fuel from the SFP at a reference plant to dry cask storage significantly reduces the risks to public health and safety. The study states that "this study's results are consistent with earlier research studies' conclusions that spent fuel pools are robust structures that are likely to withstand severe earthquakes without leaking cooling water and potentially uncovering the spent fuel. The study shows the likelihood of a radiological release from the spent fuel after the analyzed severe earthquake at the reference plant to be about one time in 10 million years or lower. If a leak and radiological release were to occur, this study shows that the individual cancer fatality risk for a member of the public is several orders of magnitude lower than the Commission's Quantitative Health Objective of two in one million (2 x 10-6/year). For such a radiological release, this study shows public and environmental effects are generally the same or smaller than earlier studies." The reference plant for the study (a General Electric Type 4 BWR with a Mark I containment) generated approximately 3500 MWt and the SFP contained 2844 fuel assemblies. FCS was licensed to generate 1500 MWt, and the SFP has the capacity to hold 1083 fuel assemblies. The SFP holds 944 fuel assemblies following permanent cessation of power operations and transfer of all fuel from the reactor vessel to the SFP. Based on these differences, the risk and the consequences of an event involving the SFP at FCS are lower than those in the NUREG-2161 study.

LIC-16-0109 Page 55 FCS conducted a seismic evaluation in response to a NRC request for information pursuant to 10 CFR 50.54(f) regarding Recommendation 2.1 of the NTTF Review of Insights from the Fukushima Dai-ichi Accident. The seismic evaluation included all structures including the SFP, and was prepared and submitted for NRC review. The OPPD submittal (LIC-14-0047) (Reference 15) documents the seismic evaluation in conformance with NTTF Recommendation 2.1 including the HCLPF values and the 1 x 10-5 per year hazard level. The Staff review of the NTTF submittal, specifically for the SFP Evaluation associated with the reevaluated seismic hazard implementing NTTF Recommendation 2.1 (CAC No. MF3735) is documented in NRC-16-0068 (ML16182A361) (Reference 16). The NRC staff concluded that the assessment was performed consistent with the NRC-endorsed (ML15350A158) (Reference 17) SFP Evaluation Guidance Report (Reference 18) and provided sufficient information, including the SFP integrity evaluation, to meet the SFP Evaluation Guidance (Item 9 in Enclosure 1 of the NRCs 50.54(f) letter), thus supporting SDA No. 6 of NUREG-1738. 4.6 Conclusion Based on the above, FCS has demonstrated that no credible accident will result in radiological releases requiring offsite protective actions. Additionally, there is sufficient time, resources and personnel available to initiate mitigative actions that will prevent an offsite release that exceeds EPA PAGs.

LIC-16-0109 Page 56 TABLE 4 Industry Decommissioning Commitments (IDCs) Comparison IDC Industry Commitments Response 1 Cask drop analyses will be performed The FCS Auxiliary Building crane with its main hook is used to move the spent fuel casks into the or single failure-proof cranes will be in SFP. The capacity of the crane lifting system using the main hook is 106 tons to accommodate use for handling of heavy loads (i.e., the ISFSI spent fuel casks. The design of this lifting system is single failure-proof. Therefore, the phase II of NUREG-0612 will be likelihood of dropping the spent fuel casks into the SFP or in the Auxiliary Building is extremely implemented). low. The crane, main hook and lifting system meet the requirements of NUREG-0612, NUREG-0554, ANSI-30.2-1976, Regulatory Guide 1.104, CMAA-70, and ASME NOG-1-2004 as a single failure-proof system. The licensing basis for the control of heavy loads associated with the Refueling Area Crane consists of a commitment to NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," - Phase 1 requirements. The Auxiliary Building Crane large hook conforms to the design requirements of NUREG-0554 "Single-Failure-Proof Cranes for Nuclear Power Plants", as demonstrated through conformance with the generic topical report for nuclear safety-related X SAM crane EDR-1. The basis for FCSs defense-in-depth approach for the control of heavy loads in the SFP area consists of preventing load drops due to crane failure, rigging failure, or human error through a combination of the use of the single failure-proof crane and compliance with the following guidelines from NUREG-0612.

1. Definition of safe load paths

2. Development of load handling procedures

3. Periodic inspection and testing of cranes

4. Qualifications, training and specified conduct of operators

5. Special lifting devices should satisfy the guidelines of ANSI N14.6

6. Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9

7. Design of cranes to ANSI B30.2 or CMAA-70 Rigging equipment meeting Guideline Nos. 5 and 7 must be used to maintain single failure-proof status. The licensing basis with regard to these guidelines requires that off the shelf and specially designed rigging equipment meet one of the following requirements:

LIC-16-0109 Page 57 IDC Industry Commitments Response

1. Provide redundancy or duality such that a single lift point failure will not result in uncontrolled lowering of the load; lift points must have a design factor of safety with respect to ultimate strength of five times the maximum load. The following is an example for meeting this requirement. Lifting a 10,000 lb. load requires two sets of rigging, each rated for at least 10,000 lbs. (with a 5:1 factor of safety). The total combined factor of safety would then be 10:1.

or

2. A non-redundant or non-dual lift point system must have a design safety factor of ten times the maximum load. This can be met by using one set of rigging rated at 20,000 lbs. (with a 5:1 factor of safety) to lift a 10,000 lb. load. The total combined factor of safety would then be 10:1.

Because the auxiliary building crane is single failure-proof, an accidental load drop is considered not to be a credible event such that condition 5.1.2(1) of NUREG-0612 is satisfied and analysis of cask drop accidents in accordance with condition 5.1.2(4) of NUREG-0612 is not required. 2 Procedures and training of personnel FCS procedures are in place to ensure onsite and offsite resources can be brought to bear will be in place to ensure that onsite during an event, including: AOP-01, Acts of Nature, AOP-06, Fire Emergency, AOP-37, Security and offsite resources can be brought to Event, AOP-31, 161kV Grid Malfunctions, AOP-38, Blair Water Main Trouble, and OCAG-1/2/3/4, bear during an event. Operational Contingency Action Guidelines. EP-FC-112-100-F-01, Command and Control Checklist - Control Room, directs the Control Room to implement EP-FC-114-100, Off-Site notifications. The Shift Manager or other designated Control Room personnel activates the ERO Notification System in accordance with EP-FC-112-100-F-06, ERO Notification or Augmentation. This process ensures FCS emergency personnel report to their proper locations. Periodic Emergency Plan drills are conducted with opportunities for off-site response organization participation in accordance with EP-FC-122, Drills and Exercises. Training requirements are outlined in EP-FC-10, Emergency Preparedness Program Description, and executed per TQ-FC-113, ERO Training and Qualification. The post-shutdown on-shift operations staff, including Certified Fuel Handlers (CFH) and Non-Certified Operators (NCO) will be appropriately trained on these procedures. The FCS CFH training program was submitted for NRC review and approval by letter dated January 15, 2016 (Reference 24). Finally, the systematic approach to training is implemented to ensure Operations and other appropriate personnel receive initial and continuing training on B.5.b event related procedures

LIC-16-0109 Page 58 IDC Industry Commitments Response and strategies credited in the Mitigation Strategy License Condition under 10 CFR 50.54 (hh)(1), 10CFR 50.54 (hh)(2), and Attachment 2 to NRC order EA-06-137. Additionally, ERO decision-makers receive initial and continuing training in accordance with TQ-AA-113, ERO Training and Qualification. Training is provided on strategies and command and control aspects of a 10 CFR 50.54 (hh)(1) and 10 CFR 50.54(hh)(2) B.5.b event that includes the following: Initial operational response, Initial damage assessments, appropriate offsite notifications, and notification of the ERO. 3 Procedures will be in place to establish FCS maintains the following procedures to provide guidance for establishing and maintaining communication between onsite and communications between offsite agencies and the onsite ERO during severe weather and offsite organizations during severe seismic events: weather and seismic events. 1. EP-FC-10, Emergency Preparedness Program Description

2. EP-FC-11, Operating Stations Emergency Preparedness Process Description

3. EP-FC-111, Emergency Classification and Protective Action Recommendations

4. EP-FC-114-100, Off-Site Notifications

5. EP-FC-114-100-F-1, Fort Calhoun Station - State/Local Event Notification Form

6. EP-FC-114-100-F-2, Notification/Update of States and Counties

7. AOP-01, Acts of Nature

8. OCAG-1, Operational Contingency Action Guideline ***10 CFR 2.390*****

9. OCAG-2, Operational Contingency Action Guideline for Dam Failure Predicted Flood Greater than 1014 Feet MSL ***10 CFR 2.390***

10. OCAG-3, Mitigating Beyond design Basis Flooding Greater than EL. 1036' with RCS Intact ***10 CFR 2.390***

11. OCAG-4, Mitigating Beyond design Basis Flooding Greater than EL. 1036' with RCS Open ***10 CFR 2.390*****

4 An offsite resource plan will be The following procedures provide guidance for access to offsite resources including a listing of developed which will include access to providers, contact information and resources, as well as mitigation strategies for SFP damage portable pumps and emergency power and water supply: to supplement onsite resources. The 1. OP-AA-201-010-1001, B.5.b Mitigating Strategies Equipment Expectations plan would principally identify 2. OCAG-1, Operational Contingency Action Guideline ***10 CFR 2.390***** organizations or suppliers where offsite

3. OCAG-2, Operational Contingency Action Guideline for Dam Failure Predicted Flood resources could be obtained in a timely Greater than 1014 Feet MSL ***10 CFR 2.390***

manner.

LIC-16-0109 Page 59 IDC Industry Commitments Response

4. OCAG-3, Mitigating Beyond design Basis Flooding Greater than EL. 1036' with RCS Intact ***10 CFR 2.390***

5. OCAG-4, Mitigating Beyond design Basis Flooding Greater than EL. 1036' with RCS Open ***10 CFR 2.390*****

External support resources are validated on an annual basis per OP-AA-201-010-1001. 5 SFP instrumentation will include The FCS design is consistent with this IDC, which includes multiple readouts and alarms for SFP readouts and alarms in the control temperature, water level, and area radiation monitors. Additionally, FCS maintains procedure room (or where personnel are AOP-36, Loss of Spent Fuel Pool Cooling, to guide the response in the event that cooling or stationed) for SFP temperature, water water level is lost in the SFP. The Control Room and local operating panel (AI-100) are also level, and area radiation levels. alerted by alarm when a SFP cooling pump is not running. This allows for prompt operation action prior to having an issue with cooling. SFP temperature indication in the Control Room and remote locations is monitored from a point in the SFP cooling path on the outlet of the heat exchanger. Additionally, the Component Cooling water temperature through the heat exchanger is also monitored. Both of these indications will cause alarms in the Control Room on high temperature. Additionally, temperature indications from the 1013-, 1022-, and 1035-foot elevations of the SFP are displayed and recorded on the Emergency Response Facility (ERF) computer. These indications have high, high-high, and rate of change alarms. SFP water level is monitored each shift as part of required Operator rounds using an installed level indication and visual verification. This level indication also drives a Control Room alarm on both high and low level. There is also a ruler mounted indication in the SFP for monitoring level locally. In the event of a beyond design basis flooding event, additional level indication is installed in the SFP as driven through OCAG-2/3/4. SFP area radiation levels are monitored via area radiation monitors that provide indication and alarm annunciation in the Control Room. A local alarm to notify personnel of high area radiation levels is also in place. 6 SFP seals that could cause leakage The design of the SFP and its cooling system and connections to the pool are such that the SFP leading to fuel uncovery in the event of cannot be drained below the level of the top of the stored fuel when in its storage rack. seal failure shall be self-limiting to Additionally, drainage grooves are provided behind the stainless steel liner to permit detection of leakage or otherwise engineered so any liner leakage. On a quarterly basis, the liner leakage is evaluated per preventative that drainage cannot occur. maintenance work instructions. The hydraulic design of the Spent Fuel Pool Cooling System is such that the single failure of any line or other single component will not drain the SFP.

LIC-16-0109 Page 60 IDC Industry Commitments Response The top of the fuel assemblies in the rack is at elevation 1008-6", which is the same as the elevation of the bottom of the gate connecting the pool with the fuel transfer canal. A plate has been installed across the bottom of the gate to raise the minimum possible water level in the pool to 1009-8.5". The gate is a taper design that is 74 wide at the top and tapers to 46 wide at the bottom. The gate is constructed of carbon steel with stainless steel plate. The lower one-third is filled with concrete to withstand hydrostatic forces. The gate is sealed using a rubber gasket at the gate-to-SFP liner interface. This design provides a passive seal (e.g., in comparison to the active seal of designs using inflatable bladders). The SFP cooling system has two suction points. The upper suction line enters at elevation 1034-0 and is the normal suction point. The lower suction line enters at elevation 1011-4" which is above the top of the stored fuel. The lower suction normally has a closed valve, but if the valve is inadvertently left open, the upper cooling water suction line and strainer would serve as a vacuum breaker and prevent draining of the pool. A rupture of the suction line upstream of the valve would only drain the pool to a level still above the spent fuel assemblies. The SFP cooling water discharges to the pool at 1034-0" and ends at 1031-7" (a 90 degree elbow to direct flow downward). The line is provided with a 1/2 inch hole to act as a vacuum breaker in the event that a discharge piping rupture occurs causing the line to act as a siphon. The transfer tube is isolated by a flange on the reactor side and by gate valve in the fuel transfer canal. The isolation valve is a 36-inch, double sliding stem gate valve, manually operated from the 1038-foot level of the spent fuel deck area. 7 Procedures or administrative controls to Administrative controls are in place that drive procedure use and adherence and risk reduce the likelihood of rapid management. Specifically, HU-AA-104-101, Procedure Use and Adherence, establishes the draindown events will include (1) expectations and requirements for procedure adherence and usage for all personnel performing prohibitions on the use of pumps that activities. Additionally, all work activities are subject to the work process controls and Integrated lack adequate siphon protection or (2) Risk Management per WC-AA-104 where the activities are analyzed and managed for risk (e.g., controls for pump suction and address SFP activities). discharge points. The functionality of The SFP is designed to reduce the likelihood of rapid drain down events. Specifically, the SFP anti-siphon devices will be periodically cooling lower suction, which is still well above the top of stored fuel, has a manual valve that is verified. controlled/locked closed. If the valve were to be inadvertently left open, the upper suction line and strainer would serve as a vacuum breaker and prevent draining of the SFP. A rupture of the suction line upstream of the valve would only drain the SFP to a level still above the spent fuel assemblies. The SFP cooling discharge line is provided with a 1/2 inch hole to act as a vacuum breaker in the event that a discharge piping rupture occurs causing the line to act as a siphon.

LIC-16-0109 Page 61 IDC Industry Commitments Response FCS maintains an administrative requirement that the SFP be maintained at an elevation of 1033' to 1037'-9" at all times. This corresponds to 37.5' and 42.3' on local level indication. This administrative requirement is placed in procedures as a precaution/limitation. Additionally, the ISFSI equipment design is such that there are no SFP operations that have the potential to cause a rapid drain down. The ISFSI controlling procedures carefully control water volume move in and out of the SFP during operations. 8 An onsite restoration plan will be in FCS practices align with this IDC. The onsite restoration plan for repair of the SFP cooling place to provide repair of the SFP system and for makeup water to the SFP are incorporated into procedures AOP-36, Loss of cooling systems or to provide access Spent Fuel Pool Cooling, and OCAG-1, Operational Contingency Action Guideline ***10 CFR for makeup water to the SFP. The plan 2.390**** (also OCAGs-2/3/4) will provide for remote alignment of the AOP-36 establishes multiple makeup sources from onsite that include: makeup source to the SFP without requiring entry to the refuel floor. 1. Demineralized Water

2. Blair Water (city water)

3. Fire Water Storage Tank

4. Diesel Fire Pump (river water)

OCAG-1 establishes multiple onsite and offsite makeup sources. OCAG-2/3/4 establish additional makeup sources from onsite in the event of major flooding. There are multiple ways to add makeup water to the SFP with or without entry to the pool floor as depicted above. 9 Procedures will be in place to control Administrative controls are in place that drive procedure use and adherence and risk SFP operations that have the potential management. Specifically, HU-AA-104-101, Procedure Use and Adherence, establishes the to rapidly decrease SFP inventory. expectations and requirements for procedure adherence and usage for all personnel performing These administrative controls may activities. Additionally, all work activities are subject to the work process controls and Integrated require additional operations or Risk Management per WC-AA-104 where the activities are analyzed and managed for risk (e.g., management review, management address SFP activities). physical presence for designated FCS maintains an administrative requirement that the SFP be maintained at an elevation of 1033' operations or administrative limitations to 1037' 9" at all times. This corresponds to 37.5' and 42.3' on local level indication. This such as restrictions on heavy load administrative requirement is placed in procedures as a precaution/limitation. Additionally, the movements. ISFSI equipment design is such that there are no SFP operations that have the potential to

LIC-16-0109 Page 62 IDC Industry Commitments Response cause a rapid drain down. The ISFSI controlling procedures carefully control water volume move in and out of the SFP during operations. The fuel handling procedures at FCS require additional personnel to verify fuel assemblies are correctly grappled prior to movement. Additionally, the fuel handling tools are designed with a J-Hook that prevents a fuel assembly from separating from the tool. The heavy loads and crane usage around the SFP are controlled through GM-OI-HE-002, Auxiliary Building Crane HE-2 Normal Operation. Attachment 13 of this procedure summarizes the licensing basis for the auxiliary building crane being single failure-proof. FCS maintains a defense-in-depth approach to control heaving loads in the SFP area. The basis for FCSs defense-in-depth approach for the control of heavy loads in the SFP area consists of preventing load drops due to crane failure, rigging failure, or human error through a combination of the use of the single failure-proof crane and through compliance with the following guidelines from NUREG-0612.

1. Definition of safe load paths

2. Development of load handling procedures

3. Periodic inspection and testing of cranes

4. Qualifications, training and specified conduct of operators

5. Special lifting devices should satisfy the guidelines of ANSI N14.6

6. Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9

7. Design of cranes to ANSI B30.2 or CMAA-70 Rigging equipment meeting guideline Nos. 5 and 7 must be used to maintain single failure-proof status. The licensing basis with regard to these guidelines requires that off the shelf and specially designed rigging equipment meet one of the following requirements:

1. Provide redundancy or duality such that a single lift point failure will not result in uncontrolled lowering of the load; lift points must have a design factor of safety with respect to ultimate strength of five times the maximum load. The following is an example for meeting this requirement. Lifting a 10,000 lb. load requires two sets of rigging, each rated for at least 10,000 lbs. (with a 5:1 factor of safety). The total combined factor of safety would then be 10:1.

or

LIC-16-0109 Page 63 IDC Industry Commitments Response

2. A non-redundant or non-dual lift point system must have a design safety factor of ten times the maximum load. This can be met by using one set of rigging rated at 20,000 lbs.

(with a 5:1 factor of safety) to lift a 10,000 lb. load. The total combined factor of safety would then be 10:1. 10 Routine testing of the alternative fuel FCS practices align with this IDC. SFP makeup strategies are outlined in AOP-36 and OCAG pool makeup system components will documents as previously outlined. One of the primary makeup strategies is from the Fire be performed and administrative Protection System. The Fire Protection System has redundant pumping capability and power controls for equipment out of service supplies to ensure alternate SFP makeup functions. The system is supplied by redundant will be implemented to provide added pumps, one diesel driven and one electric motor driven, each design rated for 2000 gpm at 125 assurance that the components would psig discharge pressure. Both pumps take suction from the plant intake cooling water structure be available, if needed. from the Missouri River. The fire protection header is normally maintained at greater than 130 psig by a jockey pump. If pressure decreases in the system, the fire pumps are automatically started by their initiation logic to maintain the fire protection system header pressure. The electric motor driven pump starts when system pressure decreases to 110 psig and the diesel driven pump is started in 10 seconds if the electric motor driven pump does not start or when system pressure drops to 100 psig. The onsite Fire Truck can also take suction from the Missouri River or a hydrant to provide an alternate source of makeup water to the SFP. The Fire Protection System can also be cross-connected with the Blair Water System (city water). The Fire Protection System provides defense-in-depth, and is routinely tested to ensure capability is maintained. OCAG-1 establishes multiple makeup sources from onsite and offsite as discussed previously. OP-AA-201-010-1001 contains the maintenance, verification and testing requirements of B.5.b Equipment. Administrative controls are in place that drive procedure use and adherence and risk management. Specifically, HU-AA-104-101, Procedure Use and Adherence, establishes the expectations and requirements for procedure adherence and usage for all personnel performing activities. Additionally, all work activities are subject to the work process controls and Integrated Risk Management per WC-AA-104 where the activities are analyzed and managed for risk (e.g., address SFP activities). Once all fuel is removed from containment, a large inventory of water above the fuel still exists to extend the time to boil or until fuel is uncovered. The SFP cooling system has far fewer interfaces than the RCS that could impact the ability of the system to perform its function; however, there are also fewer redundant makeup sources available which constitutes a higher risk. Therefore, SO-O-21, Shutdown Operations Protection Plan, has special requirements to adhere to or

LIC-16-0109 Page 64 IDC Industry Commitments Response additional clarifying information when crediting or considering available systems and equipment to fulfill SFP only shutdown conditions.

LIC-16-0109 Page 65 TABLE 5 Staff Decommissioning Assumptions (SDAs) Comparison SDA Staff Assumptions Response 1 Licensee's SFP cooling design will be The FCS design is consistent with this SDA. The SFP cooling system, SFP racks, and Auxiliary at least as capable as that assumed in Building which houses the SFP were designed and constructed to Seismic Class I standards. the risk assessment, including The SFP cooling system heat exchangers are ultimately cooled by the Missouri River (ultimate instrumentation. Licensees will have at heat sink) using redundant pumps that are supplied by redundant power sources. The pumps least one motor-driven and one diesel-are normally powered from offsite power, but can be supplied from an alternate reliable power driven fire pump capable of delivering source. The makeup system is adequate to provide water at the required capacity. inventory to the SFP. The SFP makeup system can provide 500 gpm, and additional water is available from both the demineralized water system and the fire protection system using hoses. The station design includes a motor-driven fire pump and a diesel-driven fire pump, both of which will be maintained until all fuel is removed from the SFP. Each fire pump has the capability to deliver 200 to 500 gallons per minute (gpm) of makeup water to the SFP. FCS also has a pumper type fire truck, which is able to provide 200 to 500 gpm of makeup water to the SFP. 2 Walk-downs of SFP systems will be FCS operators perform a walk down of the SFP systems once per shift as driven by operator performed at least once per shift by the rounds and by surveillance testing procedure. operators. Procedures will be The onsite restoration plan for repair of the SFP cooling system and for makeup water to the developed for and employed by the SFP are incorporated into procedures AOP-36, Loss of Spent Fuel Pool Cooling, and OCAG-1, operators to provide guidance on the Operational Contingency Action Guideline ***10CFR2.390****. (also OCAG-2/3/4) capability and availability of onsite and offsite inventory makeup sources and AOP-36 establishes multiple makeup sources from onsite that include: time available to initiate these sources

1. Demineralized Water for various loss of cooling or inventory events. 2. Blair Water (city water)

3. Fire Water Storage Tank

4. Diesel Fire Pump (river water)

OCAG-1 establishes multiple onsite and offsite makeup sources. OCAG-2/3/4 establish additional makeup sources from onsite in the event of major flooding.

LIC-16-0109 Page 66 SDA Staff Assumptions Response There are multiple ways to add makeup water to the SFP with or without entry to the pool floor. 3 Control room instrumentation that The FCS design is consistent with this SDC, which include multiple readouts and alarms for SFP monitors SFP temperature and water temperature, water level, and area radiation monitors. Additionally, FCS maintains procedure level will directly measure the AOP-36, Loss of Spent Fuel Pool Cooling, to guide the response in the event that cooling or parameters involved. Level level is lost in the SFP. The Control Room is also alerted by alarm when a SFP cooling pump is instrumentation will provide alarms at not running. This allows for prompt operation action prior to having an issue with cooling. levels associated with calling in offsite SFP temperature indication in the Control Room is monitored from a point in the SFP cooling resources and with declaring a general path on the outlet of the heat exchanger. Additionally, the component cooling water temperature emergency. through the heat exchanger is also monitored. Both of these indications will cause alarms in the control room on high temperature. Additionally, temperature indications from the 1013-, 1022-, and 1035-foot elevations of the SFP are displayed and recorded on the plant computer. These indications have high, high-high, and rate of change alarms. SFP water level is monitored each shift as part of surveillance testing using an installed level indication. The level indication includes an alarm in the control room for both high and low SFP level. This alarm will be changed to reflect the requirement of Emergency Plan limits. There is a field installed ruler type level indication mounted to the SFP liner and can be used to manually monitor the SFP level. In the event of a beyond design basis flooding event, additional level indication is available for installation in the SFP as driven through OCAG-2/3/4. FCS will continue to use the approved NRC EAL scheme, based on NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6. FCS will adopt to the Permanently Defueled EALs based on Appendix C of NEI 99-01. Based on the current EAL scheme a declaration of General Emergency would be based on radiation levels from a fuel handling accident. 4 Licensee determines that there are no The normal pool operating level is 1035-6. The SFP cooling system is normally supplied drain paths in the SFP that could lower through an 8 pipe at 1034-0. The SFP cooling system discharges through an 8 pipe. The the pool level (by draining, suction, or discharge pipe enters the SFP at 1034-0 and ends at 1031-7 (a 90 degree elbow to direct flow pumping) more than 15 feet below the downward). This discharge line is provided with a 1/2 inch hole to act as a vacuum breaker. The normal pool operating level and that discharge line is also the normal borated makeup path for the SFP. licensee must initiate recovery using The SFP cooling system has an additional lower suction point at an elevation of 1011-4, which offsite sources. is more than 15 feet below the normal pool operating level, but above the top of the fuel assemblies, which are at an elevation of 1008-6. The lower suction valve is locked closed and controlled closed. Additionally, FCS maintains an administrative requirement that the SFP be

LIC-16-0109 Page 67 SDA Staff Assumptions Response maintained at an elevation of 1033 to 1037-9 at all times. This ensures compliance with Technical Specifications, maintaining greater than 23 feet of water above the top of the fuel. The gate connecting the SFP to the fuel transfer canal is designed such that the minimum possible water level in the SFP is 1009-8.5. The gate is a taper design that is 74 wide at the top and tapers to 46 wide at the bottom. The gate is constructed of carbon steel with stainless steel plate. The lower one-third is filled with concrete to withstand hydrostatic forces. The gate is sealed using a rubber gasket at the gate-to-SFP liner interface. This design provides a passive seal (e.g., in comparison to the active seal of designs using inflatable bladders). The fuel transfer canal has two openings that are more than 15 feet below the normal pool operating level. These lower openings are only a credible drain path when the SFP gate is removed. The lowest entry point to the fuel transfer canal when the gate is removed is at the 1009-8.5 elevation (i.e., cannot drain below this elevation). The first opening is the fuel transfer tube, where the opening centerline is at 1001-6 and the bottom of the transfer tube is at 1000. The transfer tube is isolated, except during refueling, by a flange on the reactor side and by a gate valve in the fuel transfer canal. The isolation valve is a 36 inch, double sliding stem gate valve, manually operated from the 1038 elevation of the spent fuel deck area. The second opening is the drain at the bottom of the fuel transfer canal, which is at 995-6. This opening has a locked closed isolation valve that is controlled closed except during specific operations (e.g., transfer to waste, draining transfer canal). Both of these openings cannot drain below an elevation of 1009-8.5 since that is the bottom of the SFP gate opening. FCS maintains procedures and guidelines in place to obtain offsite assistance if necessary for mitigation of events that result in significant loss of SFP inventory. These mitigating strategies were implemented as part of AOP-36 and are also included in B.5.b requirements. 5 Load Drop consequence analyses will The FCS design is in alignment with this description. The heavy loads and crane usage around be performed for facilities with the SFP are controlled through GM-OI-HE-002, Auxiliary Building Crane HE-2 Normal Operation. nonsingle failure-proof systems. The Attachment 13 of this procedure summarizes the licensing basis for the Auxiliary Building crane analyses and any mitigative actions being single failure-proof. FCS maintains a defense-in-depth approach to control heaving loads necessary to preclude catastrophic in the SFP area. The basis for FCSs defense-in-depth approach for the control of heavy loads in damage to the SFP that would lead to a the SFP area consists of preventing load drops due to crane failure, rigging failure, or human rapid pool draining would be sufficient error through a combination of the use of the single failure-proof crane and through compliance to demonstrate that there is high with the following guidelines from NUREG-0612. confidence in the facilities ability to

1. Definition of safe load paths withstand a heavy load drop.

2. Development of load handling procedures

LIC-16-0109 Page 68 SDA Staff Assumptions Response

3. Periodic inspection and testing of cranes

4. Qualifications, training and specified conduct of operators

5. Special lifting devices should satisfy the guidelines of ANSI N14.6

6. Lifting devices that are not specially designed should be installed and used in accordance with the guidelines of ANSI B30.9

7. Design of cranes to ANSI B30.2 or CMAA-70 Rigging equipment meeting guideline Nos. 5 and 7 must be used to maintain single failure-proof status. The licensing basis with regard to these guidelines requires that off the shelf and specially designed rigging equipment meet one of the following requirements:

1. Provide redundancy or duality such that a single lift point failure will not result in uncontrolled lowering of the load; lift points must have a design factor of safety with respect to ultimate strength of five times the maximum load. The following is an example for meeting this requirement. Lifting a 10,000 lb. load requires two sets of rigging, each rated for at least 10,000 lbs. (with a 5:1 factor of safety). The total combined factor of safety would then be 10:1.

or

2. A non-redundant or non-dual lift point system must have a design safety factor of ten times the maximum load. This can be met by using one set of rigging rated at 20,000 lbs. (with a 5:1 factor of safety) to lift a 10,000 lb. load. The total combined factor of safety would then be 10:1.

Because the Auxiliary Building crane is single failure-proof, an accidental load drop is considered not to be a credible event such that condition 5.1.2(1) of NUREG-0612 is satisfied and analysis of cask drop accidents in accordance with condition 5.1.2(4) of NUREG-0612 is not required. 6 Each decommissioning plant will FCS conducted a seismic evaluation in response to a NRC request for information pursuant to successfully complete the seismic 10 CFR 50.54(f) regarding Recommendation 2.1 of the NTTF Review of Insights from the checklist provided in Appendix 2B to Fukushima Dai-ichi Accident. The seismic evaluation included all structures including the SFP, this study [NUREG-1738]. If the and was prepared and submitted for NRC review. The OPPD submittal (LIC-14-0047) checklist cannot be successfully (Reference 15) documents the seismic evaluation in conformance with NTTF Recommendation completed, the decommissioning plant 2.1 including the HCLPF values and the 1 x 10-5 per year hazard level. The Staff review of the will perform a plant specific seismic risk NTTF submittal, specifically for the SFP Evaluation associated with the reevaluated seismic assessment of the SFP and hazard implementing NTTF Recommendation 2.1 (CAC No. MF3735) is documented in NRC demonstrate that SFP seismically 0068 (ML16182A361) (Reference 16). The NRC staff concluded that the assessment was

LIC-16-0109 Page 69 SDA Staff Assumptions Response induced structural failure and rapid loss performed consistent with the NRC-endorsed (ML15350A158) (Reference 17) SFP Evaluation of inventory is less than the generic Guidance Report (Reference 18) and provided sufficient information, including the SFP integrity bounding estimates provided in this evaluation, to meet the SFP Evaluation Guidance (Item 9 in Enclosure 1 of the NRCs 50.54(f) study (<1 x10-5 per year including non- letter), thus supporting SDA No. 6 of NUREG-1738. seismic events). 7 Licensees will maintain a program to The FCS spent fuel racks utilize Boral rather than Boraflex as a neutron absorbing material. provide surveillance and monitoring of There are two regions or types of racks in the SFP that both contain Boral as described in USAR Boraflex in high-density spent fuel 9.5. The Boral is attached as panels between each storage cell. The panels are protected with a racks until such time as spent fuel is no stainless steel sheath. longer stored in these high-density The technical specifications describe the surveillance test requirements, which requires poison racks. sample coupons to be tested for dimensional change, weight, neutron attenuation change, and specific gravity change. This is required 1, 2, 4, 7, and 10 years after installation, and every 5 years thereafter. The purpose of the coupon sample program is to ensure that the required amount of boron remains in the neutron absorber material throughout the life of the racks. These coupon samples are sent out for analysis to an independent firm whose results are reviewed and accepted by FCS.

LIC-16-0109 Page 70 5.0 JUSTIFICATION FOR EXEMPTIONS AND SPECIAL CURCUMSTANCES 10 CFR 50.12 states that the Commission may, upon application by any interested person or upon its own initiative, grant exemptions from the requirements of the regulations of 10 CFR Part 50 which are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. 10 CFR 50.12 also states that the Commission will not consider granting an exemption unless special circumstances are present. As discussed below, this request for exemptions satisfies the provisions of Section 50.12. A. The exemptions are authorized by law 10 CFR 50.12 allows the NRC to grant exemptions from the requirements of 10 CFR Part 50. The proposed exemptions would not result in a violation of the Atomic Energy Act of 1954, as amended, or the Commission's regulations. Therefore, the exemptions are authorized by law. B. The exemptions will not present an undue risk to public health and safety The underlying purpose of 10 CFR 50.47(b); 10 CFR 50.47(c)(2); and 10 CFR Part 50, Appendix E, is to ensure that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, to establish Plume Exposure and Ingestion Pathway EPZs for nuclear power plants, and to ensure that licensees maintain effective offsite and onsite emergency plans. As discussed in this request, FCS performed an analysis indicating that within 10 days after shutdown, the radiological consequences of the postulated accident will not exceed the limits of the EPA PAGs at the EAB. In addition, FCS has developed an analysis for beyond design basis events related to the SFP which show that 530 days (1 year, 165 days) after permanent cessation of power operations, the analyzed event is either not credible, is capable of being mitigated, or the radiological consequences of the event will not exceed the limits of the EPA PAGs at the EAB. Additionally, the offsite and Control Room radiological impacts of a postulated complete loss of SFP water were assessed. It was determined that the gamma radiation dose rate at the EAB would be limited to small fractions of the EPA PAG exposure levels and the dose rate in the Control Room will be below 2.32 x 10-3 mRem/hr, which is less than 15 mRem/hr. For these reasons, offsite emergency response plans will no longer be needed for protection of the public beyond the EAB 530 days after permanent cessation of power operations. Based on the reduced consequences of radiological events possible at FCS in the permanently defueled condition, the scope of the onsite emergency preparedness organization and corresponding offsite requirements in the emergency plan may be accordingly reduced without an undue risk to the public health and safety. Therefore, the underlying purpose of the regulations will continue to be met. Because the underlying purpose of the rules will continue to be met, the exemptions will not present an undue risk to the public health and safety.

LIC-16-0109 Page 71 C. The exemptions are consistent with the common defense and security The reduced consequences of radiological events that will remain possible at FCS when it is in the permanently defueled condition allows for a corresponding reduction in the scope of the onsite emergency preparedness organization and associated reduction of requirements in the emergency plan. These reductions will not adversely affect FCSs ability to physically secure the site or protect special nuclear material. Physical security measures at FCS are not affected by the requested exemptions. Therefore, the proposed exemptions are consistent with the common defense and security. 5.1 Special Circumstances Pursuant to 10 CFR 50.12(a)(2), the NRC will not consider granting an exemption to its regulations unless special circumstances are present. OPPD has determined that special circumstances are present as discussed below. Special circumstances will exist at FCS because the plant will be permanently shut down and defueled and the radiological source term at the site will be reduced from that associated with reactor power operation. With the reactor permanently shut down and defueled, the accidents postulated to occur during reactor operation will no longer be possible. In particular, the potential for a release of a large radiological source term to the environment from the high pressures and temperatures associated with reactor operation will no longer exist. A. Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule. (10 CFR 50.12(a)(2)(ii)) The underlying purpose of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR Part 50, Appendix E is to ensure that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency, to establish Plume Exposure and Ingestion Pathway EPZs for nuclear power plants, and to ensure that licensees maintain effective offsite and onsite emergency plans. The standards and requirements in 10 CFR 50.47(b); 10 CFR 50.47(c)(2); and 10 CFR Part 50, Appendix E, were developed taking into consideration the risks associated with operation of a nuclear power reactor at its licensed full power level. These risks include the potential for a reactor accident with offsite radiological dose consequences. The radiological consequences of the postulated accident that will remain possible at FCS upon permanent shutdown of power operations are substantially lower than those at an operating plant. The upper bounds of the analyzed dose consequences limits the highest attainable emergency class to the Alert level. In addition, because of the reduced consequences of radiological events that will still be possible at the site, the scope of the onsite emergency preparedness organization may be reduced accordingly. Thus, the underlying purpose of the regulations will not be adversely affected by eliminating offsite emergency planning activities or reducing the scope of onsite emergency planning as described in this request.

LIC-16-0109 Page 72 Radiological analysis indicates that within 10 days after shutdown, the radiological consequences of the postulated accident that will remain possible at FCS upon permanent removal of fuel from the reactor will not exceed the limits of the EPA PAGs at the EAB. In addition, an analysis has been developed for beyond design basis events related to the SFP which show that 530 days (1 year, 165 days) after permanent cessation of power operations, the analyzed event is either not credible, is capable of being mitigated, or the radiological consequences of the event will not exceed the limits of the EPA PAGs at the EAB. Therefore, application of all of the standards and requirements in 10 CFR 50.47(b); 10 CFR 50.47(c)(2); and 10 CFR Part 50, Appendix E are not necessary to achieve the underlying purpose of those rules. Because the underlying purposes of the rules would continue to be achieved even with FCS being permitted to reduce the scope of emergency preparedness requirements consistent with placing the facility in the permanently defueled condition, application of the rules is not necessary to achieve the underlying purpose, and the special circumstances are present as defined in 10 CFR 50.12(a)(2)(ii). B. Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated. (10 CFR 50.12(a)(2)(iii)) Application of all of the standards and requirements in 10 CFR 50.47(b); 10 CFR 50.47(c)(2); and 10 CFR Part 50, Appendix E is not necessary for adequate emergency response capability and is excessive for a permanently shut down and defueled facility. Application of all of these standards and requirements would result in undue costs being incurred for the maintenance of an ERO in excess of that actually needed to respond to the diminished scope of credible events. Other licensed sites similarly situated, such as Entergy Nuclear Operation, Inc.s (ENO) Vermont Yankee Nuclear Power Station (VY), Southern California Edison Companys San Onofre Nuclear Generating Station (SONGS), Duke Energy Florida, Inc.s Crystal River Unit 3 Nuclear Generating Station (CR3), and Dominion Energy Kewaunee, Inc.s Kewaunee Power Station (KPS), have recently been granted similar exemptions. Therefore, compliance with the rule would result in an undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated. The special circumstances required by 10 CFR 50.12(a)(2)(iii) exist. C. The exemptions would result in benefit to the public health and safety that compensates for any decrease in safety that may result from the grant of the exemptions. (10 CFR 50.12(a)(2)(iv)) The plant will be permanently shut down and defueled and the radiological source term at the site will be reduced from that associated with reactor power operation. With the reactor permanently shut down and defueled, the postulated accidents that could occur during reactor operation will no longer be possible. In particular, the potential for a release of a large radiological source term to the environment from the high pressures and temperatures associated with reactor operation will no longer exist.

LIC-16-0109 Page 73 The proposed exemptions would allow FCS to revise the onsite emergency plan to correspond to the reduced scope of remaining accidents and events. As such, the emergency plan would no longer need to address response actions for events that would no longer be possible. The revised emergency plan would thereby enhance the ability of the ERO to respond to those scenarios that remain credible because emergency preparedness training and drills would focus only on applicable activities. Elimination of requirements for classification of EALs for events that were no longer possible would enhance the ability of the ERO to correctly classify those events that remain credible. As the proposed exemptions will enhance the ability of the organization to respond to credible events, a resultant benefit to the public health and safety is realized. Therefore, because granting the exemptions would result in benefit to the public health and safety and would not result in a decrease in safety, the special circumstances required by 10 CFR 50.12(a)(2)(iv) exist. 5.2 Precedent The exemption requests for 10 CFR 50.47(b); 10 CFR 50.47(c)(2); and 10 CFR Part 50, Appendix E requirements are consistent with exemptions on the same emergency planning requirements that recently have been issued by the NRC for other nuclear power reactor facilities beginning decommissioning. Specifically, the NRC granted similar exemptions to ENO for VY (Reference 25); to Southern California Edison Company for SONGS, Units 1, 2, and 3 (Reference 26); to Duke Energy Florida, Inc. for CR3 (Reference 27); and to Dominion Energy Kewaunee, Inc. for KPS (Reference 28). Similar to the current request, these precedents each resulted in exemptions from certain emergency planning requirements in 10 CFR 50.47(b); 10 CFR 50.47(c)(2); and 10 CFR Part 50, Appendix E, related to the elimination of offsite radiological emergency plans and reduction in the scope of the onsite emergency planning activities. For the same reasons that the NRC recently issued these exemptions, OPPD seeks approval of the enclosed proposed exemption requests.

6.0 ENVIRONMENTAL CONSIDERATION

The proposed exemptions meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(25), because the proposed exemptions involve: (i) no significant hazards consideration; (ii) no significant change in the types or significant increase in the amounts of any effluents that may be released offsite; (iii) no significant increase in individual or cumulative public or occupational radiation exposure; (iv) no significant construction impact; (v) no significant increase in the potential for or consequences from radiological accidents; and (vi) requirements of an administrative, managerial, or organizational nature. Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed exemptions. (i) No Significant Hazards Consideration Determination Omaha Public Power District (OPPD) has evaluated the proposed exemptions to determine whether or not a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92 as discussed below:

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

The proposed exemptions have no effect on structures, systems, and components (SSCs) and no effect on the capability of any plant SSC to perform its design function. The proposed exemptions would not increase the likelihood of the malfunction of any plant SSC. When the exemptions become effective, there will be no credible events that would result in doses to the public beyond the Exclusion Area Boundary (EAB) that would exceed the Environmental Protection Agency (EPA) Protective Action Guides (PAGs). The probability of occurrence of previously evaluated accidents is not increased, because most previously analyzed accidents will no longer be able to occur and the probability and consequences of the remaining postulated accident, a fuel handling accident (FHA), is unaffected by the proposed exemptions. Therefore, the proposed exemptions do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Do the proposed exemptions create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed exemptions do not involve a physical alteration of the plant. No new or different type of equipment will be installed and there are no physical modifications to existing equipment associated with the proposed exemptions. Similarly, the proposed exemptions will not physically change any SSCs involved in the mitigation of any accidents. Thus, no new initiators or precursors of a new or different kind of accident are created. Furthermore, the proposed exemptions does not create the possibility of a new accident as a result of new failure modes associated with any equipment or personnel failures. No changes are being made to parameters within which the plant is normally operated, or in the set points which initiate protective or mitigative actions, and no new failure modes are being introduced. Therefore, the proposed exemptions do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Do the proposed exemptions involve a significant reduction in a margin of safety?

The proposed exemptions do not alter the design basis or any safety limits for the plant. The proposed exemptions do not impact station operation or any plant SSC that is relied upon for accident mitigation. Therefore, the proposed exemptions do not involve a significant reduction in a margin of safety. Based on the above, OPPD concludes that the proposed exemptions present no significant hazards consideration, and, accordingly, a finding of "no significant hazards consideration" is justified.

LIC-16-0109 Page 75 (ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite. There are no expected changes in the types, characteristics, or quantities of effluents discharged to the environment associated with the proposed exemptions. There are no materials or chemicals introduced into the plant that could affect the characteristics or types of effluents released offsite. In addition, the method of operation of waste processing systems will not be affected by the exemptions. The proposed exemptions will not result in changes to the design basis requirements of SSCs that function to limit or monitor the release of effluents. The SSCs associated with limiting the release of effluents will continue to be able to perform their functions. Therefore, the proposed exemptions will not result in changes to the types or significant increases in the amount of any effluents that may be released offsite. (iii) There is no significant increase in individual or cumulative public or occupational radiation exposure. The exemptions will result in no expected increases in individual or cumulative occupational radiation exposure on either the workforce or the public. There are no expected changes in normal occupational doses. Likewise, the dose of the postulated accident is not impacted by the proposed exemptions. (iv) There is no significant construction impact. No construction activities are associated with the proposed exemptions. (v) There is no significant increase in the potential for or consequences from radiological accidents. See the no significant hazards considerations discussion in Item (i)(1) above. (vi) Requirements of an administrative, managerial, or organizational nature. The proposed exemptions will form the basis for a reduction in the size of the FCS ERO commensurate with the reduction in consequences of radiological events that will be possible at FCS when the facility is in the permanently defueled condition. They also will modify the requirements for emergency planning. Therefore, the exemptions address requirements of an administrative, managerial, or organizational nature.

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7.0 REFERENCES

1. NSIR/DPR-ISG-02, Interim Staff Guidance, Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants, dated May 11, 2015 (ML14302A490)

2. OPPD Letter (T. Burke) to USNRC (Document Control Desk) - Certification of Permanent Cessation of Power Operations, dated June 24, 2016 (LIC-16-0043) (ML16176A213)

3. OPPD Letter (T. Burke) to USNRC (Document Control Desk) - Certification of Permanent Cessation of Power Operations, dated August 25, 2016 (LIC-16-0067) (ML16242A127)

4. OPPD Letter (T. Burke) to USNRC (Document Control Desk), Certification of Permanent Removal of Fuel from the Reactor Vessel, dated November 13, 2016 (LIC-16-0074)

(ML16319A254)

5. Federal Register Notice, Vol. 60, No. 120 (60 FR 32430), Emergency Planning Licensing Requirements for Independent Spent Fuel Storage Facilities (ISFSI) and Monitored Retrievable Storage Facilities (MRS), dated June 22, 1995

6. USNRC, Integrated Rulemaking Plan for Nuclear Power Plant Decommissioning, Commission Paper SECY-00-0145, dated June 28, 2000 (ML003721626)

7. NEI 99-01, Development of Emergency Action Levels for Non-Passive Reactors, Revision 6, dated November 2012 (ML12326A809)

8. 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 (ML12346A463)

9. Commission Paper SECY-13-0112, Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor, dated October, 2013 (ML13256A334)

10. Environmental Protection Agency Protective Action Guides and Planning Guidance for Radiological Incidents, Draft for Interim Use and Public Comment, dated March 2013

11. Federal Register Notice, Vol. 76, No. 226 (76 FR 72596), Enhancements to Emergency Preparedness Regulations, dated November 23, 2011 (ML13091A112)

12. NUREG-1738, Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants, dated February 2001 (ML010430066)

13. Federal Register Notice, Vol. 74, No. 94 (74 FR 23254), Enhancements to Emergency Preparedness Regulations, dated May 18, 2009

14. NUREG-0696, Functional Criteria for Emergency Response Facilities, dated February 1981 (ML051390358)

15. OPPD Letter (L. Cortopassi) to USNRC (Document Control Desk) - Omaha Public Power District (OPPD) Seismic Hazard and Screening Report (CEUS Sites), Response NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident, dated March 31, 2014 (LIC-14-0047)(ML14097A087)

16. USNRC Letter to OPPD (S. Marik) - Fort Calhoun Station, Unit 1 - Staff Review of Spent Fuel Pool Evaluation Associated with Reevaluated Seismic Hazard Implementing Near-Term Task Force Recommendation 2.1 (CAC NO. MF3735), dated August 4, 2016 (ML16182A361)

LIC-16-0109 Page 77

17. Letter, Jack R. Davis (USNRC) to Joseph E. Pollock (NEI), Endorsement of Electric Power Research Institute Report 3002007148, Seismic Evaluation Guidance: Spent Fuel Pool Integrity Evaluation, dated March 17, 2016 (ML15350A158)

18. EPRI, Seismic Evaluation Guidance: Spent Fuel Pool Integrity Evaluation, Electric Power Research Institute Technical Update 3002007148, February, 2016 (ML16055A021)

19. USNRC Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000 (ML003716792)

20. Commission Paper SECY-99-168, Improving Decommissioning Regulations for Nuclear Power Plants, dated June 30, 1999

21. NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, dated October 2002

22. Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated May 2011 (ML100910006)

23. NUREG-2161, Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor, dated September 2014 (ML010430066)

24. OPPD Letter to USNRC, Request for Approval of Certified Fuel Handler Training Program, dated July 7, 2016 (LIC-16-0049) (ML16190A208)

25. Federal Register Notice, Vol. 80, No. 242 (80 FR 78776), Entergy Nuclear Operations, Inc.;

Vermont Yankee Nuclear Power Station, Exemption; issuance, dated December 17, 2015

26. Federal Register Notice, Vol. 80, No. 113 (80 FR 33558), Southern California Edison Company; San Onofre Nuclear Generating Station, Units 1, 2, and 3, and Independent Spent Fuel Storage Installation, Exemption; issuance, dated June 12, 2015

27. Federal Register Notice, Vol. 80, No. 69 (80 FR 19358), Duke Energy Florida, Inc.; Crystal River Unit 3 Nuclear Generating Station, Exemption; issuance, dated April 10, 2015

28. Federal Register Notice, Vol. 79, No. 214 (79 FR 65715), Dominion Energy Kewaunee, Inc.; Kewaunee Power Station, Exemption; issuance, dated November 5, 2014

29. FCS Calculation FC08557, Fuel Handling Accident in the Spent Fuel Pool Site Boundary and Control Room Dose, (Proprietary)

30. FCS Calculation FC08104, Maximum Cladding Temperature Analysis for Adiabatic Heat-up of Spent Fuel Assembly

31. FCS Calculation FC08513, EAB Radiation Shine Dose 18 Months Post Shutdown with the SFP Drained, without attachment E2 (Proprietary)

OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION DOCKET NUMBER 50-285 / LICENSE NUMBER CPR-40 ATTACHMENT 2 CALCULATION FC08104, MAXIMUM CLADDING TEMPERATURE ANALYSIS FOR ADIABATIC HEAT-UP OF SPENT FUEL ASSEMBLY

CC-AA-309-1 001 Revision 8 Design Analysis Cover Sheet Page 1 44 TOTAL PAGES Design Analysis I Last Page No.* p. 2 of Attachment 3 Analysis No.: ' FC08104 Revision: 2 0 Major~ Minor 0

Title:

' Maximum Cladding Temperature Analysis for Adiabatic Heat-Up of Spent Fuel Assembly Revision : 5 Station(s): 7 Fort Calhoun Component(s): " Unit No.:

• 1 Discipline:
• ME Descrip. Code/Keyword: ' 0 CALC Safety/QA Class: " Non-CQE System Code: ' 2 AC-SFP Structure: " Auxiliary Building CONTROLLED DOCUMENT REFERENCES ' 5 Document No.: From/To Document No.: From/To TOOl 16-048 NFA, Rev. 0 From USAR §3.1, Rev. 10 From TODI16-059 NFA, Rev. 0 From USAR Fig. 3. 1-2 From EA14-006, Rev. 0 From EA14-012, Rev. 0 From Is this Design Analysis Safeguards Information ? '" Yes D No ~ If yes, see SY-AA-1 01-1 06 Does this Design Analysis contain Unverified Assumptions? '

7 Yes D No ~ If yes, ATIIAR#:

                                                                                                                                     ------l This Design Analysis SUPERCEDES: '"                                      N/A                                                     in its entirety.

Description of Revision (list changed pages when all pages of original analysis were not changed): '" Initial issue. 44 pages. Preparer: 20 See S&L Cover Sheet Pnnt Name Sign Name Date Method of Review: 2

                                 '    Detailed Review ~                       Alternate Calculations (attached)          D    Testin g  D Reviewer:         22                  See S&L Cover Sheet Pnnt Name                                          Sign Name                               Date Review Notes:            2
                          '          Independent review [gj                        Peer review D (For External Analyses Only)

External Approver: 2' _se_e_S _&_L_C_o_v-=e_r. . ,.,sh Pnnt Name t_ _ _

                                                                ,.....e_e_

Sign Name

                                                                                                       """       ~

1 Date _lf)-2-()-1 /, Exelon Reviewer: 25 Tristan McDonald t Pnnt Name Sign Name Date Independent 3'd Party Review Reqd? 20 Yes D No ~ 4'./ I ' /r/-.)tJ .*:f~ Exelon Approver: 27

                                   .5tE.vf. ~ /w:ff/,..,. ft,,~'l.-. ~ Lt/Jk/~                                                          JO

• 2n* J~

Date Print Name I Sign Nan'le _U_ FC08104 REV 0 EC 68969 PAGE 1

CC-AA-1 03-1 003 Revision 12 ATTACHMENT 2 Owner's Acceptance Review Checklist for External Design Analyses Page 1 of3 Design Analysis No.: _F_C-...:0=-:8;..:1~0..:..4_ _ _ _ _ _ __ Rev: _o;..___ Page: -=2:.....-_ Contract#: ....::1;..:.1~32=2::..:1:....-_ _ _ _ __ Release #: 128

                                                                            ~~-------------------------

No Question Instructions and Guidance Ye~No/N/A 1 Do assumptions have All Assumptions should be stated in clear terms with enough ~ D D sufficient documented justification to confirm that the assumption is conservative. rationale? For example, 1) the exact value of a particular parameter may not be known or that parameter may be known to vary over the range of conditions covered by the Calculation. It is appropriate to represent or bound the parameter with an assumed value. 2) The predicted performance of a specific piece of equipment in lieu of actual test data. It is appropriate to use the documented opinion/position of a recognized expert on that equipment to represent predicted equipment performance. Consideration should also be given as to any qualification testing that may be needed to validate the Assumptions. Ask yourself, would you provide more justification if you were performing this analysis? If yes, the rationale is likely incomplete. / Are assumptions Ensure the documentation for source and rationale for the IZI D D 2 compatible with the assumption supports the way the plant is currently or will be way the plant is operated post change and they are not in conflict with any operated and with the design parameters. If the Analysis purpose is to establish a licensing basis? new licensing basis, this question can be answered yes, if the assumption supports that new basis. / 3 Do all unverified If there are unverified assumptions without a tracking D D B assumptions have a mechanism indicated, then create the tracking item either tracking and closure through an ATI or a work order attached to the implementing mechanism in place? WO. Due dates for these actions need to support verification prior to the analysis becoming operational or the resultant plant change being op authorized. / 4 Do the design inputs The origin of the input, or the source should be identified and ~ D D have sufficient be readily retrievable within Exelon's documentation system. rationale? If not, then the source should be attached to the analysis. Ask yourself, would you provide more justification if you were performing this analysis? If yes, the rationale is likely incomplete. / 5 Are design inputs The expectation is that an Exelon Engineer should be able to IZI D D correct and reasonable clearly understand which input parameters are critical to the with critical parameters outcome of the analysis. That is, what is the impact of a identified, if change in the parameter to the results of the analysis? If the appropriate? impact is large, then that parameter is critical. / 6 Are design inputs Ensure the documentation for source and rationale for the 0 D D compatible with the inputs supports the way the plant is currently or will be way the plant is operated post change and they are not in conflict with any operated and with the design parameters. licensing basis? FC08104 REV 0 EC 68969 PAGE 2

CC-AA-1 03-1003 Revision 12 ATTACHMENT 2 Owner's Acceptance Review Checklist for External Design Analyses Page 2 of3 Design Analysis No.: .....F. .C;. .;,; ;0;,; :;8. . ,10 - . . 4 - - - - - - - - - - - - Rev: ..o;....___ Page: .....3__ No Question Instructions and Guidance Yes I No/ N/~ 7 Are Engineering See Section 2.13 in CC-AA-309 for the attributes that are D D ~ Judgments clearly sufficient to justify Engineering Judgment. Ask yourself, documented and would you provide more justification if you were performing justified? this analysis? If yes, the rationale is likely incomplete. / 8 Are Engineering Ensure the justification for the engineering judgment D D 0 Judgments compatible supports the way the plant is currently or will be operated with the way the plant is post change and is not in conflict with any design operated and with the parameters. If the Analysis purpose is to establish a new licensing basis? licensing basis, then this question can be answered yes, if the judgment supports that new basis. / 9 Do the results and Why was the analysis being performed? Does the stated ~ D D conclusions satisfy the purpose match the expectation from Exelon on the proposed purpose and objective of application of the results? If yes, then the analysis meets the Design Analysis? the needs of the contract. / 10 Are the results and Make sure that the results support the UFSAR defined 0 D D conclusions compatible system design and operating conditions, or they support a with the way the plant is proposed change to those conditions. If the analysis operated and with the supports a change, are all of the other changing documents licensing basis? included on the cover sheet as impacted documents? / 11 Have any limitations on Does the analysis support a temporary condition or D D B the use of the results procedure change? Make sure that any other documents been identified and needing to be updated are included and clearly delineated in transmitted to the the design analysis. Make sure that the cover sheet appropriate includes the other documents where the results of this organizations? analysis provide the ing_ut. / 12 Have margin impacts Make sure that the impacts to margin are clearly shown D D B been identified and within the body of the analysis. If the analysis results in documented reduced margins ensure that this has been appropriately appropriately for any dispositioned in the EC being used to issue the analysis. negative impacts (Reference ER-AA-2007)? / 13 Does the Design Are there sufficient documents included to support the (;ZI D D Analysis include the sources of input, and other reference material that is not applicable design basis readily retrievable in Exelon controlled Documents? documentation? / 14 Have all affected design Determine if sufficient searches have been performed to D D 0 analyses been identify any related analyses that need to be revised along documented on the with the base analysis. It may be necessary to perform Affected Documents List some basic searches to validate this. (ADL) for the associated Configuration Change? / 15 Do the sources of inputs Compare any referenced codes and standards to the current ~ D D and analysis design basis and ensure that any differences are reconciled. methodology used meet If the input sources or analysis methodology are based on committed technical and an out-of-date methodology or code, additional reconciliation regulatory may be required if the site has since committed to a more reQuirements? recent code FC08104 REV 0 EC 68969 PAGE 3

CC-AA-1 03-1 003 Revision 12 ATTACHMENT 2 Owner's Acceptance Review Checklist for External Design Analyse~ Page 3 of 3 Design Analysis No.: _F_c_oo.; :;s.:. 10;::;..4~--------- Rev:...;:o;....___ Page: ..:4:.....--_ No Question Instructions and Guidance Yes I No I Nl~ 16 Have vendor supporting Based on the risk assessment performed during the pre-job D D ~ technical documents brief for the analysis (per HU-AA-1212), ensure that and references sufficient reviews of any supporting documents not provided (including GE DRFs) with the final analysis are performed. been reviewed when necessary? / 17 Do operational limits Ensure the Tech Specs, Operating Procedures, etc. contain 0 D D support assumptions operational limits that support the analysis assumptions and and inputs? inputs. Create an SFMS entry as required by CC-AA-4008. SFMS Number: - - - - - - - - FC08104 REV 0 EC 68969 PAGE 4

Sargent & Lundy ' " ISSUE

SUMMARY

Form SOP-0402-07. Revision 11 DESIGN CONTROL

SUMMARY

CLIENT: Omaha Public Power District UNIT NO.: 1 PAGE NO.: 5 PROJECT NAME: Fort Calhoun Station S&L NUCLEAR QA PROGRAM PROJECT NO.: 07751-388 APPLICABLE 181 YES 0 NO CALC. NO..: 2016-10694 (FC08104) SAFETY RELATED 0 YES 181 NO

                                                                                                                                          ~.

TITLE: Maximum Cladding Temperature Analysis lor Adiabatic Heat-Up of Spent Fuel Assembly y EQUIPMENT NO.: IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDNOIDED & REVIEW METHOD Initial Issue. 44 pages. INPUTS/ ASSUMPTIONS 181 VERIFIED 0 UNVERIFIED REVIEW METHOD: Detailed REV.: 0 STATUS: 181 APPROVED 0 SUPERSEDED BY CALCULATION NO. O VOID DATE FOR REV.: 10/19/2016

                                            ~-~4                                                                             DATE: )CL/?/2.0/(p PREPARER:           Helmut R. Kopke REVIEWER:           DanielS. Elegant                                                                                         DATE:  /olf:i /2.cJ I 6 APPROVER:           Robert J. Peterson       Jl)>j§(L~~}                                                                     DATE:   [cJ-JCI-/6 IDENTIFICATION OF PAGES ADDEDIREVISED/SUPERSEDEDNOIDED & REVIEW METHOD INPUTS/ ASSUMPTIONS 0     VERIFIED 0     UNVERIFIED REVIEW METHOD:                                                                                                               REV.:

STATUS: 0 APPROVED 0 SUPERSEDED BY CALCULATION NO. OVOID DATE FOR REV.: PREPARER: DATE: REVIEWER: DATE: APPROVER: DATE: IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDNOIDED & REVIEW METHOD INPUTS/ ASSUMPTIONS 0 VERIFIED 0 UNVERIFIED REVIEW METHOD: REV.: STATUS: 0APPROVED 0 SUPERSEDED BY CALCULATION NO. OVOID DATE FOR REV.: PREPARER: DATE: REVIEWER: DATE: APPROVER: DATE: NOTE: PRINT AND SIGN IN THE SIGNATURE AREAS SOP040207.DOC Rev. Date: 02-02-2015 FC08104 REV 0 EC 68969 PAGE 5

Omaha Public Power District S&L Calculation 2016-10694 {FC08104) Fort Calhoun Station Revision 0 Page6 Table of Contents Exelon Cover Sheet .......................................................................................................... 1 Exelon Owner's Acceptance Review Sheets .................................................................... 2 Sargent & Lundy LLC Cover Sheet ................................................................................... 5 Table of Contents .............................................................................................................. 6 1.0 Purpose ............................................................................................................................. 7 2.0 Inputs ................................................................................................................................ 8 3.0 Assumptions ...........................................................................................................,_,.,u ........ 12 4.0 References ...................................................................................................................... 13 5.0 Identification of Computer Programs ............................................................................... 14 6.0 Method of Analysis .......................................................................................................... 15 7.0 Numeric Analysis ............................................................................................................ 18 8.0 Results and Conclusions ................................................................................................. 19 Attachments Attachment 1: Adiabatic Spent Fuel Heat-Up Computation ............................ (14 pages) Attachment 2: OPPD Letter NED-16-048 NFA I TODI16-048 NFA ................. (8 pages) Attachment 3: OPPD TODI16-059 NFA .......................................................... (2 pages) Total Pages= 44 (20 pages main body+ 24 pages attachments) FC08104 REV 0 EC 68969 PAGE 6

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page7 1.0 Purpose The purpose of this calculation is to perform an adiabatic heat-up analysis of the hottest spent fuel assembly to conservatively determine the decay time after reactor shutdown when it takes at least 10 hours for uncovered spent fuel assemblies to heat-up to 900°C (1652°F). This calculation satisfies the NRC expectation set forth in Item 2 of Section 5.0 of the Interim Staff Guidance for emergency planning exemption requests (Ref. 4.1) which states that an adiabatic heat-up analysis be used to determine when a minimum of 10 hours is available from the time cooling is lost before the cladding temperature reaches 900°C. This analysis conservatively assumes that there is no radiative or air cooling of the fuel assemblies: the flow paths that would provide natural circulation cooling are assumed to be blocked and no credit is taken for radiation or conduction to adjacent assemblies or spent fuel pool racks. FC08104 REV 0 EC 68969 PAGE 7

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 PageS 2.0 Inputs 2.1 Maximum Fuel Cladding Temperature Item 2 of Section 5.0 of the Interim Staff Guidance for emergency planning exemption requests (Ref. 4.1) states that an adiabatic heat-up analysis be used to determine when a minimum of 10 hours is available from the time cooling is lost before the cladding temperature reaches 900°C. Therefore, 900°C is the maximum fuel cladding temperature investigated in this calculation. This is the temperature where "runaway oxidation" is expected to occur per NUREG-1738 (Ref. 4.6, p. 3-7). 2.2 Fuel Cladding Properties The fuel cladding material is M5' Zirconium alloy (Ref. 4.3). The density to be used for the cladding is 6.55 g/cm3 (Ref. 4.3) or 408.9 lbm/ft3 [=6.55 * (30.48 cm/ft) 3 I 453.59 gllbm]. The specific heat of M5' Zirconium alloy increases as temperature increases over most of the temperature range of interest (Table 3 of Ref. 4.3). The mass based specific heat below is computed by dividing the provided volumetric heat capacity by the density. Table 2-1: Fuel Cladding Heat Capacity Temperature Volumetric Heat Capacity Mass Based Heat Capacity OF oc Btu/ft3-°F Btu/lbm-°F 31.7 -0.2 27.23 0.0666 1520.3 826.8 39.98 0.0978 1592.3 866.8 88.69 0.2169 1790.3 976.8 32.57 0.0797 2.3 Spent Fuel Pool Temperature I Spent Fuel Initial Temperature The spent fuel pool maximum normal temperature is 140°F (Ref. 4.3). This temperature is appropriate to use as the initial cladding temperature for this analysis. In the generic analyses in both SECY-99-168 (Ref. 4.4) and NUREG-1738 (Ref. 4.6, pg. 2-2), the starting water/fuel assembly temperature was set at 30°C (86°F). Both documents state that the analysis starts at the time of fuel uncovery. Using 140°F as the starting temperature of the analysis is conservative compared to the initial temperature used in NUREG-1738 and SECY-99-168 because more heat energy is required to heat the assembly from 86°F to 900°C than is required to heat the assembly from 140°F to 900°C. Therefore, the decay heat required for a given heat up time (e.g. 10 hrs) will be reached earlier after shutdown for lower initial temperatures. Since the result of this analysis is a time after shutdown and a greater time is conservative, use of a higher initial temperature is conservative. Furthermore, using the spent fuel pool maximum normal temperature for the initial temperature of the spent fuel is a source of conservatism: as the decay time increases FC08104 REV 0 EC 68969 PAGE 8

Omaha Public Power District S&L Calculation 2016-10694 {FC08104) Fort Calhoun Station Revision 0 Page9 {i.e. as the plant has been shutdown longer) the fuel heat generation rate will be lower and the maximum spent fuel pool temperature would likely be lower. 2.4 Fuel Assembly Geometry (Ref. 4.3) Ta bl e 2 -2 : F ue I Assem blIY G eometry t Fuel Type 14x14 Combustion Engineering Fuel Pellet Diameter 0.3805 inches Inner Diameter of Cladding 0.387 inches Outer Diameter of Cladding 0.440 inches Number of Full Length Rods per Assembly 176 Number of Partial Length Rods per Assembly 0 Active Fuel Length of Full Length Rods 129.3 inches Number of Other Tubes in Assembly 4 guide tubes and 1 instrument tube Inner Diameter of Guide & Instrument Tubes 1.035 inches Outer Diameter of Guide & Instrument Tubes 1.115 inches In addition, Reference 4.7(b) shows that the fuel rod pitch {center to center distance) is 0.580 inches, resulting in an orthogonal spacing of 0.140 inches [=0.580-0.440] between fuel rods (distance between cladding). Based on Figure 3.1 of Reference 4.9, Cycle 28 (the final fuel cycle) consists of fuel from batches DO, EE, FF, and GG. The fuel assemblies corresponding to these batches are htp_24, htp_25, htp_26, and htp_27, respectively, based on p. 119 in Attachment 4 of Reference 4.1 0. The guide tube material for each of these fuel assembly types is provided as parameter assy_gt_matl on pp. 67, 79, 91, and 102, respectively, of Attachment 4 of Reference 4.10. Similarly, the instrument tube material for each of these fuel assembly types is provided as parameter assy_it_matl on pp. 68, 80, 91, and 103, respectively, of Attachment 4 of Reference 4.1 0. The above information is used to determine that the guide and instrument tube material is M5' Zirconium alloy. 2.5 Fuel Properties The fuel is uranium dioxide, U02 {Ref. 4.3). The fuel density is equal to 96o/o of the theoretical density of 10.96 g/cm 3 {Ref. 4.3). Thus, the fuel density is 10.5216 g/cm3 [=0.96

• 10.96] or 656.8 lbm/ft3 [=10.5216 *

{30.48 cm/ft) 3 I 453.59 g/lbm]. The specific heat of uranium dioxide fuel increases as temperature increases over the temperature range of interest {Table 2 of Ref. 4.3). The mass based specific heat below is computed by dividing the provided volumetric heat capacity by the fuel density. FC08104 REV 0 EC 68969 PAGE 9

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 10 Table 2-3: Uranium Dioxide Fuel Heat Capacity Temperature Volumetric Heat Capacity Mass Based Heat Capacity OF oc Btu/ft3-°F Btu/Ibm-oF 100 37.8 37.75 0.0575 200 93.3 40.48 0.0616 300 148.9 42.38 0.0645 400 204.4 43.78 0.0667 500 260.0 44.86 0.0683 600 315.6 45.71 0.0696 700 371.1 46.41 0.0707 BOO 426.7 47.01 0.0716 900 482.2 47.53 0.0724 1000 537.8 47.98 0.0731 1100 593.3 48.4 0.0737 1200 648.9 48.78 0.0743 1300 704.4 49.13 0.0748 1400 760.0 49.46 0.0753 1500 815.6 49.78 0.0758 1600 871.1 50.08 0.0762 1700 926.7 50.38 0.0767 The fuel stack density is 10.3743 g/cm3 (Ref. 4.3) or 647.6 lbm/ft3 [=10.3743 * (30.48 cm/ft) 3 /453.59 g/lbm]. The fuel stack density is slightly less dense than the fuel since it accounts for the fact that fuel pellets are not perfectly cylindrical since they have dished ends, chamfered edges, and an outward land taper (OLT). The fuel stack density is 98.6% of the fuel density since the volume reduction (relative to a right circular cylinder) due to the aforementioned geometric features is 1.4% (Table 3.3 and Section 5.1.2 of Ref. 4.9). When computing the mass of fuel, the stack density should be used with the pellet stack volume assuming a right circular cylinder. It is recognized that Table 3.3 of Reference 4.9 lists dish, OLT, and chamfer volumes of both 1.2% and 1.4o/o. These volumes are the reductions in pellet stack volume relative to a right circular cylinder. The 1.2°Al volume is only applicable to blanket pellets, while the 1.4o/o volume is applicable to enriched pellets. Blanket pellets are those at the top and bottom of the fuel rods. Since the overwhelming majority of pellets are enriched pellets, 1.4% is the appropriate volume to use. Furthermore, use of 1.4% is conservative as it results in less fuel mass and hence a faster heat-up time. FC08104 REV 0 EC 68969 PAGE 10

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 11 2.6 Heat Load Reference 4.5 states that the maximum decay heat load from a single assembly occurs for Fuel Assembly DD07. The table below shows the maximum fuel assembly heat generation rate for several different times after shutdown. Table 2-4: Heat Generated by Highest Heat Load Fuel Assembly (0007) Cooling Time (years) Cooling Time (months) Decay Heat (watts) 1.0 12 5043 1.0833 13 4764 1.167 14 4515 "-Y 1.25 15 4291 1.33 16 4086 1.4167 17 3898 1.5 18 3726 2.0 24 2921 2.5 30 2380 3.0 36 2002 The reactor contains 133 fuel assemblies (Ref. 4.7(a)) and has a 100o/o reactor power level of 1500 Mwt (Ref. 4.8). Therefore, the average heat generation for an assembly at full power in the reactor is 11.3 megawatts [=1500/133]. The heat loads in the table above are on the order of 0.04o/o to 0.02o/o of the peak reactor power. NUREG-1738 (Ref. 4.6) mentions that additional heat load due to oxidation of the cladding may exist at temperatures less than 900°C, although it is not a significant heat source below 600°C (p. A1A-5 of NUREG-1738). However, the oxidation heat load is not included in the adiabatic analysis in NUREG-1738. For the purposes of comparing the adiabatic heat-up results for the Fort Calhoun fuel to the generic results in NUREG-1738, no additional oxidation heat load is applied herein. Exclusion of oxidation heat loads is consistent with the Staff guidance for adiabatic spent fuel heat-up (Section 4.0 of Ref. 4.1 ). The Staff recognized the exclusion of oxidation as non-conservative, but stated that the overall adiabatic heat-up analysis reasonably represents conditions that may occur in the event of an SFP accident involving the loss of water inventory. The Staff conclusion was based on weighing the conservatisms of the adiabatic heat-up analysis methodology with the non-conservatisms. FC08104 REV 0 EC 68969 PAGE 11

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 12 3.0 Assumptions 3.1 The heat-up time does not credit the drain down time of the pool. This is conservative because no credit is taken for the thermal capacitance of the water. 3.2 It is assumed that the properties for uranium dioxide fuel adequately represent all metal oxides in spent fuel, which contains uranium as well as measureable amounts of other elements such as neptunium, plutonium, americium, and curium (Tables 5.6 and 5. 7 of Ref. 4.11). Based on Tables 5.4 through 5.7 of Reference 4.11, the mass of uranium dioxide at discharge and 10 years after discharge is 93-95% of the total initial mass of fuel. Given that some mass is converted to energy during the fuel cycle, it can be concluded that the mass of uranium dioxide in spent fuel is greater than 93% of the total spent fuel mass. Tables 5.6 and 5.7 present the results of ORIGEN runs for spent fuel with 3.2o/o and 4.2%, uranium 235 enrichment, respectively. These enrichments are similar to Fort Calhoun which utilizes fuel with enrichments ranging from 3.3 to 4.5°/o uranium 235 (pp. 48 and 108 of Ref. 4.9). FC08104 REV 0 EC 68969 PAGE 12

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 13 4.0 References 4.1 NSIRIDPR-ISG-02 Interim Staff Guidance, "Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants, dated May 11, 2015. ADAMS Accession No. ML14106A057. 4.2 lncropera, Frank P., and David P. DeWitt, Fundamentals of Heat and Mass Transfer, Third Edition, John Wiley & Sons, 1990. ISBN 0-471-61246-4. 4.3 OPPD Memorandum NED-16-048 NFA,

Subject:

Transmittal of Fuels-Related Data for Adiabatic Spent Fuel Heat-Up Analysis, dated July 8, 2016. Includes Transmittal of Design Information TODI16-048 NFA, Revision 0. Included as Attachment 2. 4.4 SECY-99-168, "Improving Decommissioning Regulations for Nuclear Power Plants," June 30, 1999. 4.5 Transmittal of Design Information TODI 16-059 NFA, Revision 0,

Subject:

Hottest Assembly Decay Heat Data for Adiabatic Heatup Analysis, dated October 10, 2016. Includes OPPD Letter NED-16-059 NFA. Included as Attachment 3. 4.6 NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," published February 2001. ADAMS Accession No. ML010430066.

4. 7 Fort Calhoun Updated Safety Analysis Report, Revision 10

a. Section 3, "Reactor," Sub-Section 3.1, "Summary of Description."

b. Figure 3.1-2, "Reactor Core Cross Section."

4.8 Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, Unit 1, Renewed Facility Operating License No. DPR-40. ADAMS Accession No. ML053110488. 4.9 Analysis EA14-006, Revision 0, "Cycle 28 Design Depletions." 4.10 Analysis EA14-012, Revision 0, "Cycle 28 Plant Database." 4.11 Oak Ridge National Laboratory Report ORNLITM-11 018, "Standard- and Extended-Bumup PWR and BWR Reactor Models for the ORIGEN2 Computer Code," published December 1989. FC08104 REV 0 EC 68969 PAGE 13

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 14 5.0 Identification of Computer Programs The analyses performed herein utilize Microsoft Excel Version 14.0.7015.1000 (32-bit) which is included with Microsoft Office Professional Plus 2010. Microsoft Excel is commercially available. The validation of Excel is implicit in the detailed review of all spreadsheets used in this analysis. All Excel computer runs were performed using S&L PC No. ZL 10734 under the Windows 7 Enterprise (Build 7601) operating system with Service Pack 1. FC08104 REV 0 EC 68969 PAGE 14

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 15 6.0 Method of Analysis 6.1 Adiabatic Heat-Up Methodology This analysis determines the heat-up time of the fuel assembly using the thermal capacity of materials (Based on Section 2.3 of Ref. 4.2).

          .      V *C      ~T q=p*            *-                                                                  Equation 6-1 P     t Where:

q heat generation rate (Btu/hr) p material density (lbm/ft3) V material volume (fe) cp material specific heat (Btu/lbm-°F)

        ~T        temperature increase (°F) t         heat-up time (hr)

For this analysis, there are two materials being heated: uranium dioxide fuel pellets and zirconium alloy. Zirconium alloy is in the cladding and the guide/instrument tubes. The zirconium and the uranium dioxide are modeled as heating up at the same rate since:

1) the fuel pellets and zirconium alloy are immediately adjacent to each other, 2) the heat is generated throughout the assembly, 3) the transient is relatively slow, and 4) the materials have high thermal conductivities. Therefore, the ~Tit will be the same for both materials.

Equation 6-2 Where: Xu signifies the property is for yranium dioxide Xz signifies the property is for ~irconium alloy Given that the volume of uranium dioxide computed below is based on a right circular cylinder, the appropriate uranium dioxide density to use is the fuel stack density. This calculation seeks the heat-up time, so Equation 6-2 is solved for t. t= fl.; *4>u,stack

• Vu,c:yl
• Cp,u + Pz
• Vz
• Cp,z) Equation 6-3 This approach is taken instead of calculating the heat generation rate required for a 10 hour heat-up time since the heat capacity is temperature dependent. The heat-up time is calculated for discrete temperature increments of 100°F or less in order to more accurately model the heat capacity. For each temperature increment, the minimum specific heats of zirconium alloy and uranium dioxide are used. Minimizing the specific FC08104 REV 0 EC 68969 PAGE 15

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 16 heat is conservative as it results in a faster heat-up. The total heat-up time is then the sum of the heat-up times for the increments up to the temperature of interest (900°C). The volume of uranium dioxide is given below. V cyl = (1t ._DP_2 ) . Nh . L . _1_ft_2_ Equation 6-4 u, 4 r 144 in 2 Where: Dp diameter of the uranium dioxide fuel pellet (in) Nhr number of heated rods per assembly L heated length of the rods (ft) The volume of zirconium alloy in the heated rods and in the guide/instrument tubes is given below. The length of the cladding and tubes that is heated is conservatively modeled as being the same as the heated length of uranium dioxide. In reality, the cladding and guide/instrument tubes are longer than the length of the uranium dioxide pellets (i.e. the heated length of the fuel is less than the total length of the assembly). Thus, the actual mass of zirconium in a fuel assembly is greater than that which is modeled. Modeling less zirconium results in faster heat-up times, which are conservative. V z,c

               =  (1t .Dc,o 24- Dc/ ) .N        hr
                                                   . L . 1 ft2 144 in2                                  Equation 6-5 V =

z,t (1t . D t,o 2 - Du2 4 J.N . L . 1441 ft2in2 t Equation 6-6 Equation 6-7 Where: Vz,c volume of zirconium alloy in the fuel rod cladding (ft3 ) Dc,o outer diameter of the cladding (in) Dc,i inner diameter of the cladding (in) Vz,t volume of zirconium alloy in the guide/instrument tubes (ft3) Dt,o outer diameter of the guide/instrument tubes (in) Dt.i inner diameter of the guide/instrument tubes (in) Nt number of tubes (guide plus instrument) per assembly The temperature increase {8T) for this analysis is from the initial temperature of the pool, 140°F (Input 2.3), to the cladding failure temperature of interest, 1652°F (Section 1 and Input 2.1). FC08104 REV 0 EC 68969 PAGE 16

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 17 Temperatures in the fuel assembly are modeled as uniform before and during the event since: 1) the heat generation rate is relatively uniform and low (see Input 2.6), 2) the diameter of the fuel rods is small, and 3) the spacing between rods is small (0.14 inches per Input 2.4). The heat-up time is calculated as a function of the decay time after reactor shutdown. 6.2 Acceptance Criteria There are no specific acceptance criteria for this analysis. However, the goal is to determine the decay time after shutdown at which it takes 10 hours for the fuel to heat up to 900°C. This supports the analysis requirement set forth in Item 2 of Section 5.0 of the Interim Staff Guidance for emergency planning exemption requests (Ref. 4.1 ), repeated below. Complete loss of SFP water inventory with no heat loss (adiabatic heatup) demonstrating a minimum of 10 hours is available before any fuel cladding temperature reaches 900 degrees Celsius from the time all cooling is lost (Demonstrates sufficient time to mitigate events that could lead to a zirconium cladding fire); ... Reference 4.1 is consistent with SECY-99-168 (Ref. 4.4) which suggests that "10 hours [is] sufficient time to take mitigative actions." SECY-99-168 performed a generic analysis that found that for PWRs, 2.5 years is expected to be the decay time needed to reach a 10 hour heat-up time from 30°C to 900°C while NUREG-1738 shows that a 10 hour adiabatic heat up time from 30°C to 900°C for a PWR would occur at less than 2 years (Ref. 4.6, Fig. 2.2). The results of this analysis are compared to the generic analyses. FC08104 REV 0 EC 68969 PAGE 17

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 18 7.0 Numeric Analysis The numeric analysis is performed using Microsoft Excel and is presented in Attachment 1. This attachment implements the inputs and methodology described above. The equations behind each cell are provided. FC08104 REV 0 EC 68969 PAGE 18

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 19 8.0 Results and Conclusions The heat-up times to 900°C for the hottest fuel assembly (calculated in Attachment 1) at various times after shutdown are shown below. Table 8-1 : Results Cooling Time Cooling Time Heat-Up Time to 900°C (1652°F) (Months) (Years) (hours) 12 1.0 7.6 13 1.0833 8.0 14 1.167 8.5 15 1.25 8.9 16 1.33 9.4 17 1.4167 9.8 18 1.5 10.3 24 2.0 13.1 30 2.5 16.1 36 3.0 19.1 By interpolating, the heat-up time to 900°C is at least 10 hours at a decay time of 1.453 years (1 year, 165 days) after shutdown. This duration is based on 365 days per year. As stated in Section 6.2, SECY-99-168 performed a generic analysis that found that for PWRs, 2.5 years is expected to be the decay time needed to reach a 10 hour heat-up time from 30°C to 900°C. NUREG-1738 shows that a 10 hour heat up time to 900°C for a PWR would occur at less than 2 years (Ref. 4.6, Fig. 2.2). The results calculated here are consistent with the NUREG-1738 analysis. Figure 8-1 shows the heat-up time to 900°C as a function of decay time. FC08104 REV 0 EC 68969 PAGE 19

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Revision 0 Page 20 Final Figure 8-1: Heat-Up Time vs. Decay Time 20

                                                                                                     ~l
                 -~    900°C (1652°F)                                                              ~
                                                                                                 ~

18 1---

                                                                                              ~
                  &    10 hr to soooc                                                  , ,""'

16

                                                                          , , ~
   ~0 14
                                                                       ~,
                                                             'f',

5. _, ,,l p0 12 , ,'

0 en , ,

   .s 10 CD                   ~ ,tr
                               ,o-

_4 , 1.453, 10.0 E , .,A-' I i= 8 Q. r--

   ~
    =6 4

2 0 1 1.25 1.5 1.75 2 2.25 2.5 2.75 3 Years Since Shutdown Several noteworthy conservatisms in this analysis are listed below.

• Convective heat transfer (air cooling) is not credited.
• No heat transfer (convection, radiation, or conduction) from the hottest assembly to adjacent colder assemblies is credited.
• Metal heat capacitance for the fuel rods and guide/instrument tubes beyond the heated length, spacer grids, the upper and lower end fittings, and channel walls in the fuel assembly is ignored.

Water thermal capacitance during drain down is ignored. FC08104 REV 0 EC 68969 PAGE 20

Omaha Public Power District S&L Calculation 2016-10694 (FCOB104) Attachment 1: Adiabatic Spent Fuel Heat-Up Computation Fort Calhoun Station Revision 0 Page 1 of 14 A 8 c D E F G H I J K L M N 1 I I ZrAIIoy uo, 2 Heated Rods per Assembl 176 Rods Input 2.4 Temperature (* F) c;,,, (Btunbm-* F) c;,,u (Btunbm-*F) 3 Number of Guide/Instrument Tubes 5 Rods Input 2.4 31.7 0.0666 4 Heated Length of Fuel 129.3 in Input 2.4 100 0.0680 0.0575 5 Heated Length of Fuel 10.8 feet -in+ 12 in/ft 140 0.0689 0.0591 6 200 0.0701 0.0616 0.25 7 Diameter of U02 Fuel Pellets! 0.3805 jinches j lnput 2.4 0.0645 B Volume of U02 Fuel 1.497 ft Equation 6-4 300 400 0.0722 0.0743 0.0667 0.2

                                                                                                                                                                                    ~ -- ZrAIIoy l t                -
                                                                                                                                                                         ~              ...... U0 2                                              - -

9 10 11 Outer Diameter of Claddin Inner Diameter of Claddin 0.440 0.387 inches inches Input 2.4 Input 2.4 500 600 700 0.0764 0.0785 0.0806 0.0683 0.0696 0.0707

                                                                                                                                                                         ..,e
                                                                                                                                                                         ~ 0 . 15                                               \                -

e

                                                                                                                                                                                                                                   \
                                                                                                                                                                                                                                                 ~

12 Volume of Zirconium Alloy in Heated Rods 13 Outer Diameter of Guide/Instrument Tubes 0.453 1.115 n' inches Equation 6-5 Input 2.4 BOO 900 0.0827 0.0848 0.0716 0.0724 ..

I: 0.1 c--

14 Inner Diameter of Guidellnstrument Tubes 1.035 inches Input 2.4 1000 0.0869 0.0731 u

                                                                                                                                                                                                                                  .., ~

15 Volume of Zirconium Alloy in Gil Tubes 0.051 n' Equation 6-6 1100 0.0890 0.0737 *;:; 16 Total Volume of Zirconium Alloy 0.504 n' Equati on 6-7 1200 0.0911 0.0743 ~ 0.05 17 1300 0.0932 0.0748 lbm/rt* 18 U02 Fuel Stack Densi ty (* Fuel Density) 647.6 Input 2.5 1400 0.0953 0.0753 3 0 - 19 Density of Zirconium Alloy! 408.9 l 1bm/ft !Input 2.2 1500 0.0974 0.0758 20 1520.3 0.0978 0.0759 0 500 1000 1500 2000 --- 21 Initial Temperature 140 *F Input 2.3 1592.3 0.2169 0.0762 Temperature (*F) --- 22 Final Temperature at 9oo* c 1652 *F Input 2.1 1600 0.2116 0.0762 23 Total temperature Increase to 9oo*c 1512 *F = Initial

• Final 1652 0.1755 0.0765 l l 24 1700 0.1423 0.0767 25 1790.3 0.0797 I I 26 I I Italicized, bold, blue values are interpolated. Other values are taken from Input 2.2 (Zr Alloy) or 2.5 (UO:z). I

~ 27

                      -                                               ~                           l                                                                                                                                     J

~ Specific Heats Over Each Tem erature Increment 30 T1 (*F) 140 200 300 400 500 600 700 800 900 31 T2 (*F 200 300 400 500 600 700 800 900 1000 6T(. F) ~ 60 100 100 100 100 100 100 100 100 33 c,,, at T1 (Btu/lbm-*F) 0.0689 0.0701 0 .0722 0.0743 0.0764 0.0785 0.0806 0.0827 0.0848 fJ4 - c,_, at T2 (Btu/lbm-*F) 0.0701 0.0722 0 .0743 0.0764 0.0785 0.0806 0.0827 0.0848 0.0869 "35 - -- c**'*"'" (Btunbm-*F) 0.0689 0.0701 0.0722 0.0743 0.0764 0.0785 0.0806 0.0827 0.0848 36 Cp,u at T1 (Btu/lbm-*F) 0.0591 0.0616 0 .0645 0.0667 0.0683 0.0696 0.0707 0 .0716 0.0724 37 cp.u at T2 (Btu/Ibm-*F) 0.0616 0.0645 0 .0667 0 .0683 0.0696 0.0707 0.0716 0 .0724 0.0731 38 C,.u.mn (Btunbm-*F) 0.0591 0.0616 0.0645 0.0667 0.0683 0.0696 0.0707 0.0716 0.0724 39 I I I I Heat-Up FC08104 REV 0 EC 68969 PAGE 21

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation Revision o Page2 of14 A B c I D E F G H I J K L M N 40 1 W ., 1 J/sec

• 3600 seclhr + 1055.06 J/Btu I CooUng Decay Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time Cooling Time (months) Time Heat Decay Heat to900*C fromT1 to T2 from T1 toT2 from T1 toT2 fromT1 toT2 from T1 to T2 fromT1 toT2 fromT1 toT2 from T1 toT2 fromT1 toT2 41 (years) (Watts) (Btu/hr) (hours) (hrs) (hrs) (hrs) (hrs) (hrs) (hrs) (hrs) (hrs) (hrs) 42 12 1.0 5,043 17,207 7.6 0.2 0.4 0.4 0.5 0.5 0.5 0.5 0.5 0.5 43 13 1.0833 4,764 16,255 8.0 0.3 0.5 0.5 0.5 0.5 0.5 0.5 0.5 0.5 44 14 1.167 4,515 15,406 8.5 0.3 0.5 0.5 0.5 0.5 0.5 0.6 0.6 0.6 45 15 1.25 4,291 14,641 8.9 0.3 0.5 0.5 0.5 0.6 0.6 0.6 0.6 0.6 46 16 1.33 4,086 13,942 9.4 0.3 0.5 0.6 0.6 0.6 0.6 0.6 0.6 0.6 47 17 1.4167 3,898 13,300 9.8 0.3 0.6 0.6 0.6 0.6 0.6 0.6 0.7 0.7 48 18 1.5 3,726 12,714 10.3 0.3 0.6 0.6 0.6 0.6 0.7 0.7 0.7 0.7 49 24 2.0 2,921 9,967 13.1 0.4 0.7 0.8 0.8 0.8 0.8 0.9 0.9 0.9 50 30 2.5 2,380 8,121 16.1 0.5 0.9 1.0 1.0 1.0 1.0 1.0 1.1 1.1 51 36 3.0 2,002 6,831 19.1 0.6 1.1 1.1 1.2 1.2 1.2 1.2 1.3 1.3 52 53 Determine the Cooling Time at which Heat-Up to Temperature of Interest= 10 hrs

                                                                                                                                                                           --rI Heat-Up Time Interpolated Cooling Time (years)     to900*c         Years        Days 54                                                                              (hours)                                                                                                        I               I 55                                                        I       1.453           10.0           1           165 (I

Heat-Up FC08104 REV 0 EC 68969 PAGE 22

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation Revision 0 Page 3 of 14 0 p a R s T u 2 3 ----------- ------------- *- * - - - - - - - - - - - - - - - - - - 1 4 5 6--- --***----- -----~t-------f-------------1-----t----~t-------1 7 8 ~9~~-----t------------~---------1-----~-----~---~ ~ -------t-----~t~~~------t-----1---- 11

• ---------- -=-=----=-. -=-- .]::=__ .

12 13 I 1f ---- -------- --- - I 18 I 19 I W I

~

22 -~-------* ----~-- ---------- I

                                                                                      ---~------ ---- -

• -------------+--------------- -- -j- -----t-------

25 1 I I 26 .g 28 29

    -------t------jt-
                                              --------------+---

30 1000 1100 1200 1300 1400 1500 1600 31 1100 1200 1300 1400 1500 1600 1652 32 100 100 100 100 100 100 52 33 0.0889 0.0890 0.0911 0.0932 0.0953 0.0974 0.2118 34 0.0890 0.0911 0.0932 0.0953 0.0974 0.2118 0.1755 35 0.0869 0.0890 0.0911 0.0932 0.0953 0.0974 0.1755 36 0.0731 0.0737 0.0743 0.0748 0.0753 0.0758 0.0762 37 0.0737 0.0743 0.0748 0.0753 0.0758 0.0762 0.0765 38 0.0731 0.0737 0.0743 0.0748 0.0753 0.0758 0.0762 39 Heat-Up FC08104 REV 0 EC 68969 PAGE 23

Omaha Public Power District S&l Calculation 2016-10694 (FC08104) Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation Revision 0 Page4 of 14 0 p a R s T u 40 Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time Heat-Up Time fromT1 toT2 fromT1 toT2 from T1 toT2 fromT1 toT2 from T1 toT2 fromT1 toT2 from T1 to T2 41 (hrs) (hrs) (hrs) (hrs) (hrs) (hrs) (hrs) 42 0.5 0.5 0.5 0.5 0.5 0.5 0.3 43 0.5 0.6 0.6 0.6 0.6 0.6 0.4 44 0.6 0.6 0.6 0.6 0.6 0.6 0.4 45 0.6 0.6 0.6 0.6 0.6 0.6 0.4 46 0.6 0.6 0.7 0.7 0.7 0.7 0.4 47 0.7 0.7 0.7 0.7 0.7 0.7 0.4 48 0.7 0.7 0.7 0.7 0.7 0.7 0.5 49 0.9 0.9 0.9 0.9 0.9 0.9 0.6 50 1.1 1.1 1.1 1.1 1.1 1.2 0.7 51 1.3 1.3 1.3 1.3 1.4 1.4 0.8 52 I 1 53 I I 54 55 I Heat-Up FC08104 REV 0 EC 68969 PAGE 24

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation Revision 0 PageS of 14 A B c D E F 1 2 Heated Rods per Assembly 176 Rods lnput2.4 3 Number of Guidennstrument Tubes 5 Rods lnput2.4 4 Heated Length of Fuel 129.3 in lnput2.4 5 Heated Length of Fuel =C4/12 feet =In+ 121nm 6 7 Diameter of U02 Fuel Pellets 0.3805 inches Input 2.4 8 Volume of U02 Fuel =PI()*C7 11 214*C2*C5/144 ft.. Equation 6-4 9 10 Outer Diameter of Cladding 0.44 inches lnput2.4 ~~~-~- 11 Inner Diameter of Cladding 0.387 Inches Input 2.4 12 Volume of Zirconium Alloy In Heated Rods =PI ()*(C 1OA2-C 11 112)/4*C2*C5/144 ft3 Equation 6-5 13 Outer Diameter of Guide/Instrument Tubes 1.115 inches Input 2.4 14 Inner Diameter of Guide/Instrument Tubes 1.035 Inches Input 2.4 15 Volume of Zirconium Alloy in Gil Tubes =PI ()*(C 13112-C 14112)/4*C3*C5/144 ft3 Equation 6-6 ~~-----*--- -- 16 Total Volume of Zirconium Alloy =C12+C15 ft3 Equation 6-7 17 18 U02 Fuel Stack Density (* Fuel Density) 647.6 Ibm/ft.. lnput2.5 19 Density of Zirconium Alloyi.W8.9 lbmlft3 lnput2.2 20 21 Initial Temperature 140 *F lnput2.3 22 Final Temperature at eoo*c c9Q0/5*9+32 *F Input 2.1 23 Total temperature Increase to eoo*c =C22-C21 *F  ::: Initial - Final 24 25 I 26 I 27 I I 28 I 29 I Specific Heats Over Each Temperature I 30 I T1 (*F) 140 31 I T2 (*F) 200 32 6T (°F)  :::F31-F30 33 Cp.z at T1 (Btu/Ibm-*F)  :::VLOOKUP(F30,$G$3:$1$25,2) 34 Cp.z at T2 (Btullbm-*F)  :::VLOOKUP(F31,$G$3:$1$25,2) 35 Cp,z.mn (BtuJlbm-*F) =MIN(F33:F34) 36 Cp,u at T1 (Btullbm-*F)  :::VLOOKUP(F30,$G$3:$1$25,3) 37 Cp,u at T2 (Btu/ibm-*F)  :::VLOOKUP(F31,$G$3:$1$25,3) 36 Cp,u.mn (Btullbm-*F) =MIN(F36:F37) 39 Heat-Up (Eqs) FC08104 REV 0 EC 68969 PAGE 25

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Attachment 1: Adiabatic Spent Fuel Heat-Up Computation RevlsionO Fort Calhoun Station Page 6of14 A B c D E F 40 1 W = 1 J/sec

• 3600 seclhr + 1055.06 Cooling Time Cooling Time (years)

(months) Heat-Up Time to 41 Decay Heat (Watts) Decay Heat (Btu/hr) 900"C (hours) Heat-Up Time from T1 to T2 (hrs) 42 12 =A42/12 5043 =C42*3600/1 055.06 =SUM F42:U42 =F$32* $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D42 43 =A42+1 =A43112 4764 =C43*3600/1 055.06 =SUM F43:U43 =F$32* $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D43 44 =A43+1 =A44/12 4515 =C44*3600/1055.06 =SUM F44:U44 =F$32* $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D44 45 =A44+1 =A45112 4291 =C45*3600/1 055.06 =SUM F45:U45 =F$32* $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D45 46 =A45+1 =A46/12 4086 =C46*3600/1055.0S =SUM F46:U46 =F$32* $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D46 47 =A46+1 =A47/12 3898 =C47*3600/1055.06 =SUM F47:U47 =F$32* $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$047 48  ::::A47+1 =A48/12 3726 =C48*3600/1 055.08 =SUM F46:U48 =F$32* $C$18*$C$8*F$38+$C$19*$C$16*F$35 1$048 49 24 =A49/12 2921 =C49*3600/1 055.06 =SUM F49:U49 =F$32* $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D49 50 30  ::::A50f12 2380 =C50*3600/1 055.08 =SUM F50:U50 =F$32* $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D50 51 36 =A51/12 2002 =C51*3600/1 055.06  :::SUM F51:U51 "'F$32* $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D51 52 ---------- 53 Determlnet Interpolated Cooling Time (years) Years Heat-Up Time to 54 900"C (hours) 55 =B47+(B48-B47)/(E48-E47)*(E55-E47) 10 =FLOOR(D55,1) Heat-Up (Eqs) FC08104 REV 0 EC 68969 PAGE 26

Omaha Public Pow er District S&L Calcula tion 2016-10694 (FC08104) Attachment 1: Adiabatic Spent Fue l Heat-Up Computation Revision 0 Fort Calhoun Station Page 7 of 14 G I H I J 1 Zr AIIo y uo, 2 Temperature (' F) C,.z (Btunbm-' F) _C,u (Btunbm-' F) 3 31 .7 4 100 0.0666

                                                                   =HS3+(HS20-HS3)/(G$20-G$3) '(G4-G $3)              0.0575 5 140                                                             =H$3+(H$20-H$3)/(G$20.G$3

• G5-G$3 -14+(16-14 I(G6-G4)'(G5- G4 6 200 =HS3+(H$20-H$3)/(G$20-G$3) ' (G6-GS3) 0 .0616 0.25 7 =G6+100 =H$3+(H$20-H$3)1(G$20.G$3)'(G7-G$3) 0.0645 8 =G7+100 =H$3+(HS20-H$3)/(GS20.GS3) '(G8-G$3) 0.0667 ~ 0.2 9

10

    =G8+100
    =G9+100
                                                                   -H$3+(H$20-H$3)/(G$20-G$3) ' (G9-G$3)
                                                                   =H$3+(H$20-H$3 /(GS20-G$3) ' (G10-G$3) 0 .0683 0.0696 e
                                                                                                                                                                                                 .D 11  =G10+100                                                       -HS3+{H$20-H$3)1(GS20-G$3) ' (G11*GS3)             0 .0707                                                                    ~      0. 15 e

12 =G11 +100 13 =G12+100

                                                                   =H$3+(H$20.H$3)/(G$20.G$3) '(G12-G$3)
                                                                   =H$3+(H$20.H$3)/(G$20-G$3)'(G13-G$3) 0 .0716 0 .0724

t: 0 .1 14 =G13+100 -H$3+(H$20-H$3)/(G$20.G$3) ' (G14--G$3) 0.0731 u 15 =G14+100 =H$3+(H$20-H$3)/(G$20.G$3) '(G 15-G$3) 0.0737 "'*c.. lit-16 =G15+100 =H$3+(H$20-H$3)/(G$20-G$3) ' (G16-G$3) 0 .0743 .ll- 0.05 17 =G16+100 =H$3+(H$20-H$3)/(G$20-G$3)' (G17-G$3) 0 .0748 18 =G17+100 =HS3+(H$20-H$3)/(GS20.G$3) ' (G 18-G $3) 0 .0753 0

19 =G18+100 =H$3+(H$20.H$3)/(G$20.G$3) ' (G19-G$3) 0 .0758 0 20 1520.3 0 .0978 =1$19+ 1$22-1$19/(G$22-G$19 '(G20.G$19 21 1592.3 0.2169 -I$19+(1$22-1$19)/(G$22-G$19)'(G21*G$19) 22 =G19+100 H$21+(H$25-H$21)/(G$25-G$21 '(G22-G$21) 0 .0762 23 1652 =HS21+(H$25-H$21 I(G$25-G$21 '(G23-G$21 =122+(124--122 ~ G24--G22 ' G23-G22 24 =G22+100 -HS21+(H$25-H$21)1(G$25-G$21) ' {G24--GS21) 0 .0767 25 1790.3 0 .0797 26 Italicized, bold, blue values are interpolated. Other values a r4 28

                                                                      --                                          -  1-                                 -      --          --            --                    -

29 I 30 =F31 =G31 =H31 =131 31 =G30+100 =H30+100 =130+100 =J30+100 32 =G31-G30 =H31-H30 =131 -130 =J31-J30 33 =VLOOKUP(G30.SG$3:SIS25.2) =VLOOKU P(H30.$G$3:$1$25,2) =VLOOKUP(I30,SG$3:$1$25,2) =VLOOKUP(J30,SG$3 :$1$25,2) 34 =VLOOKUP(G31,$G$3:$1$25.2) =VLOOKUP(H31,$G$3:$1$25,2) =VLOOKUP(I31.$GS3:SI$25,2) =VLOOKUP(J31,$G$3:$1$25,2) 35 =MIN(G33:G34) =MIN(H33:H34) =MIN(/33:134) =MIN(J33:J34) 36 =VLOOKUP(G30.SG$3:SIS25,3) =VLOOKUP(H30,$GS3:SIS25,3) =VLOOKUP(I30,SGS3:SIS25,3) =VLOOKUP(J30.SG$3:SIS25,3) 37 =VLOOKUP(G31,$GS3:SIS25,3) =VLOOKUP(H31.SGS3:SI$25.3) =VLOOKUP(I31 ,SGS3:$1$25,3) =VLOOKUP(J31,$G$3:SI$25,3) 38 =MIN(G36:G37) =MIN(H36:H37) =MIN(I36:137) =MIN(J36:J37) 39 I I Heat-Up (Eqs) FC08104 REV 0 EC 68969 PAGE 27

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Attachment 1: Adiabatic Spent Fuel Heat-Up Computation Fort Calhoun Station Revision 0 PageS of 14 G H I J 40 41 Heat-Up Time from T1 to T2 (hrs) Heat-Up Time from T1 to T2 (hrs) Heat-Up Time from T1 to T2 (hrs) Heat-Up Time from T1 to T2 (hrs) 42 ~:G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 1$042 ~:H$32* $C$18*$C$8*H$38+$C$19*$C$16*H$35 /$042 "'1$32* $C$18*$C$8*1$38+$C$19*$C$16*1$35 /$D42  ::J$32* $C$18*$C$8*J$38+$C$19*$C$16*J$35 /$042 43 ~:G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 /$043 =H$32* $C$18*$C$8*H$38+$C$19*$C$16*H$35 /$043 =1$32* $C$18*$C$8*1$38+$C$19*$C$16*1$35 /$D43  ;::J$32* $C$18*$C$8*J$38+$C$19*$C$16*J$35 /$043 44 =G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 /$044 =H$32* $C$18*$C$8*H$38+$C$19*$C$16*H$35 /$044 =1$32* $C$18*$C$8*1$38+$C$19*$C$16*1$35 /$D44  ;::J$32* $C$18*$C$8*J$38+$C$19*$C$16*J$35 /$044 45 =G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 /$045 =H$32* $C$18*$C$8*H$38+$C$19*$C$16*H$35 /$045 =1$32* $C$18*$C$8*1$38+$C$19*$C$16*1$35 /$D45 =J$32* $C$18*$C$8*J$38+$C$19*$C$16*J$35 /$045 48 =G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 /$046 ~:H$32* $C$18*$C$8*H$38+$C$19*$C$16*H$35 /$046 =1$32* $C$18*$C$8*1$38+$C$19*$C$16*1$35 /$046 =J$32* $C$18*$C$B*J$38+$C$19*$C$16*J$35 /$046 47 =G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 /$047 =H$32* $C$18*$C$B*H$38+$C$19*$C$18*H$35 /$047 =1$32* $C$1 B*$C$8*1$38+$C$19*$C$16*1$35 /$047 =J$32* $C$18*$C$B*J$38+$C$19*$C$16*J$35 /$047 48 =G$32* $C$18*$C$B*G$38+$C$19*$C$16*G$35 /$048  ::H$32* $C$18*$C$B*H$38+$C$19*$C$16*H$35 1$048 =1$32* $C$18*$C$8*1$38+$C$19*$C$16*1$35 /$048 =J$32* $C$18*$C$8*J$38+$C$19*$C$16*J$35 /$048 49 =G$32* $C$18*$C$B*G$3B+$C$19*$C$16*G$35 /$049 ~:H$32* $C$18*$C$8*H$38+$C$19*$C$16*H$35 /$049 =1$32* $C$18*$C$8*1$38+$C$19*$C$16*1$35 /$049 =J$32* $C$18*$C$8*J$38+$C$19*$C$16*J$35 /$049 50 =G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 /$050 =H$32* $C$18*$C$B*H$38+$C$19*$C$16*H$35 /$050 =1$32* $C$1 B*$C$8*1$38+$C$19*$C$16*1$35 /$D50 =J$32* $C$18*$C$8*J$38+$C$19*$C$16*J$35 /$050 51 =G$32* $C$18*$C$B*G$3B+$C$19*$C$16*G$35 /$051 =H$32* $C$18*$C$B*H$38+$C$19*$C$16*H$35)/$051 ~:1$32* $C$18*$C$8*1$38+$C$19*$C$16*1$35 /$D51 =J$32* $C$18*$C$8*J$38+$C$19*$C$16*J$35 /$051 52 53 Days 54 55 =(D55-F55)*365 Heat-Up (Eqs) FC08104 REV 0 EC 68969 PAGE 28

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Attachment 1: Adiabatic Spent Fuel Heat-Up Computation Fort Calhoun Station Revision 0 Page 9 of 14 K L M N 1 2 3 4 --- 1- - I --- --- ~ --- r-4--- r-- 7 8 - -ZrAIIoy - - - U02 I ,.._ r-g* - f-10 111 r& ~ ~ 13 15 r-!Z-18 I I ~ 200 400 600 800 1000 1200 1400 1600 1800 2000 ~ Temperature (' F) ~ 22 r#24 -- - - - --- 25 26 ~ 28

                       -                        -j                                                                         -  -   -

29 30 =J31 =K31 =L31 =M31 31 =K30+100 =L30+100 =M30+100 =N30+100 32 =K31 -K30 =L31-L30 =M31-M30 =N3 1-N 30 33 - V LOOKUP(K30,$G$3:SI$25,2) =VLOOKUP(l30,SG$3:$1$25,2) =VLOOKUP(M30,$G$3:$1$25.2) =VLOOKUP(N30,$G$3:$1$25.2) 34 =VLOOKUP(K31 .SG$3:$1$25.2) =VL OOKU P(l31 ,SG$3:$1$25,2) =VLOOKUP(M31.SG$3:$ 1$25,2) =VLOOKUP(N31.SG$3:$1$25,2) 35 -MIN(K33:K34) =M/N(L33:L34) -MIN(M33:M34) =M/N(N33:N34) 36 =VLOOKUP(K30,SGS3:SIS25,3) =VLOOKUP(l30,SGS3 :SIS25,3) =VLOOKUP(M30,SGS3:SIS25,3) =VLOOKUP(N30.SGS3:SIS25,3) 37 - VLOOKUP(K31 .SG$3:$1$25,3) =VLOOKUP(l 31 ,SGS3:SIS25,3) =VLOOKUP(M31,SGS3:SI$25,3) =VLOOKUP(N31 ,SG$3:$1$25,3) 38 =M/N(K36:K37) =MIN(L36:L37) =MIN(M36:M37) =MIN(N36:N37) 39 (i Heat-Up (Eqs) FC08104 REV 0 EC 68969 PAGE 29

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation Revision 0 Page 10 of 14 K T M N 40

e Time from T1 to T2 e. Time from T1 to T2 le Time from T1 to T2 S) rime from toT2 1:;:

1:::: 1:::: lc

• 1:;:

1:;:

                                                                                     !l:"l.l;i\
                                                                                     ~':l'i\j
      ,.,..,                                                                         ~':l'i\j
      , ........                                      1::::                          S35\J
      ,........                            , ....,... lc                             !l:"l.J>\1
                                                                                                --------         --*           *-----                    --------     -~--

54 55 Heat-Up (Eqs) FC08104 REV 0 EC 68969 PAGE 30

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation Revision 0 Page 11 of 14 0 p a R 1 I 2 3 ----~--~----- 4 6 7 8

                                                                                                                            ----------                                  --*-*~--------

9 10 - - - - - - - -- "'i1 12 13 14 ---------------------- -~-

                                                                                                          -- ~                   .. ---- -                                                  - --

~-17

         - ~----

18 19 20 21 - --~----~---~ *-- *- -- 22 ~ 24 25 I 26 27 *---------- --~---- --- -----~--- ------ 28 I 29 I 30 =N31 =031 =P31 =031 31 1100 =P30+100 =030+100 =R30+100 32 =031-030 "'P31-P30 =031-030 =R31-R30 33 =VLOOKUP(030,$G$3:$1$25,2) =VLOOKUP(P30,$G$3:$1$25,2) =VLOOKUP(Q30,$G$3:$1$25,2) =VlOOKUP(R30,$G$3:$1$25,2) 34 =VLOOKUP(031,$G$3:$1$25,2) =VLOOKUP(P31,$G$3:$1$25,2) =VLOOKUP(Q31,$G$3:$1$25,2) =VlOOKUP(R31,$G$3:$1$25,2) 35 =MIN(033:034} ""MIN(P33:P34} =MIN(Q33:Q34) =MIN(R33:R34} 36 =VLOOKUP(030,$G$3:$1$25,3) =VLOOKUP(P30,$G$3:$1$25,3) =VLOOKUP(Q30,$G$3:$1$25,3) =VlOOKUP(R30,$G$3:$1$25,3) 37 =VLOOKUP(031,$G$3:$1$25,3) =VLOOKUP(P31,$G$3:$1$25,3) =VLOOKUP(Q31,$G$3:$1$25,3) "'VlOOKUP(R31,$G$3:$1$25,3) 38 =MIN(036:037} =MIN(P36:P37} =MIN(Q36:Q37} =MIN(R36:R37} 39 Heat-Up (Eqs) FC08104 REV 0 EC 68969 PAGE 31

Omaha PubUc Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation Revision 0 Page 12 of 14 0 p a R 40 41 Heat-Up Time from T1 to T2 (hrs) Heat-Up Time from T1 to T2 (hrs) Heat-Up Time from T1 to T2 (hrs) Heat-Up Time from T1 to T2 (hrs) 42 =0$32* $C$18*$C$8*0$38+$C$19~C$16*0$35 /$042 =P$32* $C$18*$C$8~$38+$C$19~C$16*P$35 /$042 =Q$32* $C$18*$C$8*Q$38+$C$19~C$16*Q$35 /$042 =R$32* $C$18~C$8*R$38+$C$19*$C$16*R$35 /$042 43 =0$32* $C$18*$C$8*0$38+$C$19~C$16*0$35 /$043 =P$32* $C$18*$C$8*P$38+$C$19~C$16*P$35 /$043 =Q$32* $C$18*$C$8*Q$38+$C$19~C$16*Q$35 /$043 =R$32* $C$18~C$8*R$38+$C$19~C$16*R$35 /$043 44 =0$32* $C$18*$C$8*0$38+$C$19*$C$16*0$35 /$044 =P$32* $C$18*$C$8~$38+$C$19~C$16*P$35 /$044 =Q$32* $C$18*$C$8*Q$38+$C$19*$C$16*Q$35 11$044 cR$32* $C$18*$C$8*R$38+$C$19*$C$16*R$35 /$044 45 =0$32* $C$18~C$8*0$38+$C$19*$C$16*0$35 /$045 =P$32* $C$18*$C$8~$38+$C$19*$C$16*P$35 /$045 cQ$32* $C$18*$C$8*0$38+$C$19~C$16*Q$35 /$045 =R$32* $C$18~C$8*R$38+$C$19~C$16*R$35 /$045 46 =0$32* $C$18~C$8*0$38+$C$19~C$16*0$35 /$046 =P$32* $C$18*$C$8~$38+$C$19~C$16*P$35 /$046 =Q$32* $C$18*$C$8*Q$38+$C$19*$C$16*Q$35 /$046 =R$32* $C$18*$C$8*R$38+$C$19~C$16*R$35 /$048 47 =0$32* $C$18~C$8*0$38+$C$19~C$16*0$35 1$047 =P$32* $C$18*$C$8~$38+$C$19~C$16*P$35 /$047 =Q$32* $C$18*$C$8*Q$38+$C$19*$C$16*Q$35 /$047 =R$32* $C$18~C$8*R$38+$C$19~C$16*R$35 /$047 48 =0$32* $C$18*$C$8*0$38+$C$19*$C$16*0$35 /$048 =P$32* $C$18*$C$8~$38+$C$19*$C$16*P$35 /$048 =Q$32* $C$18*$C$8*Q$38+$C$19~C$16*Q$35 /$048 =R$32* $C$18*$C$8*R$38+$C$19*$C$16*R$35 /$048 49 =0$32* $C$18*$C$8*0$38+$C$19*$C$16*0$35 /$049 "'P$32* $C$18*$C$8~$38+$C$19*$C$16*P$35 /$049 =Q$32* $C$18*$C$8*Q$38+$C$19~C$16*Q$35 /$049 =R$32* $C$18*$C$8*R$38+$C$19~C$16*R$35 /$049 50  :;:Q$32* $C$18*$C$8*0$38+$C$19*$C$16*0$35 /$050 =P$32* $C$18*$C$8*P$38+$C$19~C$16*P$35 /$050 =Q$32* $C$18*$C$8*Q$38+$C$19~C$16*Q$35 /$050 "'R$32* $C$18*$C$8*R$38+$C$19*$C$16*R$35 /$050 51 =0$32* $C$18*$C$8*0$38+$C$19*$C$16*0$35 /$051 "'P$32* $C$18*$C$8~$38+$CS19*$C$16*P$35 /$051 =Q$32* $C$18*$C$8*Q$38+$C$19*$C$16*Q$35 /$051 =R$32* $C$18*$C$8*R$38+$C$19*$C$16*R$35 /$051 52 *---*---- .. ~-~-

                                                                                                                                              ...    ----~--  -~                                            ----

53 54 55 Heat-Up (Eqs) FC08104 REV 0 EC 68969 PAGE 32

Omaha PubUc Power Dlstrld S&L Calculation 2016-10894 (FC08104) Attachment 1: Adiabatic Spent Fuel Heat-Up Computation RevlslonO Fort Calhoun Station Page 13 of 14 s T u 1 2 3 -- 4 5 --~~---~- -- 6 7 8

                                                     --~------ ----~--*------*--------

9 ~ ----~-- ---~

                                    -                                                                      ------------**--* --    ----~~~

11 12 13 I 14 I ---- - ---------- 15 ---~-- I -------- --------- 16 -- - - - - ------- I ------------ 17 I I 18 19 20 21 - -- *- --. --- 22 23 -----* ----- 24 25 26 27 -- 28 29 30  ::R31  ::531  :::f31 31  :::530+100  :::f3Q+100 1652 32  :::531-830 =T31-T30  :::U31-U30 33  ::VLOOKUP(S30,$G$3:$1$25,2) =VLOOKUP(T30,$G$3:$1$25,2) =VLOOKUP(U30,$G$3:$1$25,2) 34  :::VLQOKUP(S31 ,$G$3:$1$25,2)  :::VLOOKUP(T31 ,$G$3:$1$25,2)  ::VLOOKUP(U31 ,$G$3:$1$25,2) 35  ::M/N(S33:S34) =MIN(T33:T34)  ::M/N(U33:U34) 36 =VLOOKUP(S30,$G$3:$1$25,3) =VLOOKUP(T30,$G$3:$1$25,3) =VLOOKUP(U30,$G$3:$1$25,3) 37 =VLOOKUP(S31 ,$G$3:$1$25,3)  :::VLOOKUP(T31 ,$G$3:$1$25,3) =VLOOKUP(U31 ,$G$3:$1$25,3) 38  ::M/N(S36:S37) =MIN(T36:T37) =MIN(U36:U37) 39 Heat-Up (Eqs) FC08104 REV 0 EC 68969 PAGE 33

Omaha Public Power District S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation RevlsionO Page 14of14 s T u 40 Time from T1 to T2 Time from T1 to T2 h s) Time from T1 to T2 I"' I'" I= 4 I* 1=1 1= 1"'1 I"' 1::1 I"' 1::1

     -~60                                  I"'        j1"                          1=1 u~~-
     -~ .....                              I"'                                     1::1 u~~-

M 55 I {} Heat-Up (Eqs) FC08104 REV 0 EC 68969 PAGE 34

Omaha Public Power District Attachment 2: S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station OPPD Letter NED-16-048 NFA/ Revision 0 TOOl 16-048 NFA Page 1 of 8 Omaha Public Power Oisiiict MEMORANDUM Date: July 8, 2016 NED-16-048 NFA From: Carol Waszak To: Russell J. Pleskunas

Subject:

Transm ittal of Fuels-Related Data for Adiabatic Spent Fuel Heat-Up Analysis

Reference:

1. EA15-006, Revision 0, PWR Fuel Design Criteria Review for Fort Calhoun Unit 1 Reload FCA1-28 (FTC-14) and Cycle 28 Assemblies.

2. Calculation FC6800, "Bounding Composite Equilibrium Core Inventory W ith Initial U-235 Enrichments of 3.5 w/o to 5.0 w/o", Revision 1, October, 2006.

3. Fort Calhoun USAR, Section 9.6.1.

4. 02-1023224-02, "Material Specification M5 Sheer , October, 2003.

5. 08-9062240-000, "Uranium Dioxide Pellets", November, 2008.

6. EA14-037, "Cycle 28 S-RELAP5 Update", August, 2015.

7. EA14-006, "Cycle 28 Design Depletions", August, 2015.

Dear Mr. Pleskunas,

The purpose of th is letter is transm it the data required for the spent fuel heatup calculation for the Fort Calhoun Nuclear Station Decommissioning. The data corresponds only to the adiabatic case, and the balance of the data for the air-cooled cases will provided in a later transmittal. Please note that Fort Calhoun Station does not have maximum assembly heat loads as a function of time available, as previously discussed in our phone calls. This data will need to be generated separately. If you have any questions or comments concerning this, please contact myself at (402) 533-6482 or Thomas Luedeke at (402) 533-6179. Sincerely, (1~ lJru-; At-Carol Waszak Supervisor - Nuclear Fuels and Analysis CLW/TPL cc: NED Correspondence FC08104 REV 0 EC 68969 PAGE 35

Omaha Public Power District Attachment 2: S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station OPPD Letter NED-1 6-048 NFA I Revision 0 TODI 16-048 NFA Page 2 of 8 Transmittal of Design Information CC-FC-31 0-F-0 1 Rev. 0 Transmittal of Design Information (TOOl) Refer to CC-AA-31 0 for requirements TODI No.: TODI* 16-048 NFA Revision:-o- - Date: 8-July-2016 Transmittal Letter Number (if applicable): NED-16..()48 NFA From: Carol L. Waszak {OPPD) To: Russell J . Pleskunas (SarQent & Lundy)

Subject:

Transmittal of fuels-related data for Adiabatic spent fuel heatup analysis Information Provided and Intended Use: (List any reference documents or attachments)

4) EA15-006, Table 3.1 1) 08-9062240-000

5) FC06800, Section 4 2) EA14-037

6) Fort Calhoun USAR, Section 9.6.1 3) EA14-006

7) 02-1 023224-{)2 Classification: @ Non-QA Source and Basis of transmittal Status of Information:

[RJ Approved for use 0 Preliminary, Scheduled for Confirmation on _ _ _ __ Limitations on Use: TOOl Prepare (Thomas P. luedeke) Date 8-July-2016

                                                                               --~--------------

Distribution: Per RM-AA-1 01 Project File Use TOOl Number in any subsequent correspondence concerning input FC08104 REV 0 EC 68969 PAGE 36

Omaha Public Power District Attachment 2: S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station OPPD Letter NED-16-048 NFA I Revision 0 TODI16-048 NFA Page 3 of8 JulyS, 2016 NED-16-048 NFA Table 1: Input Data for Fort Calhoun Decommissioning Spent Fuel Pool Heatup Analysis Parameter Value Reference Comments 14 x 14 assembly with five large Fuel assembly 14 x 14 Combustion water holes for guide/instrument Engineering 1, Table 3-1 tubes. USAR Figure 3-1.2 has a configuration depiction. Number of active full-length fuel rods 176 1, Table 3-1 - In assembly Number of active partial-length fuel 0 Fort Calhoun does not utilize part-NIA rods in assembly length rods Number of water 4 guide rods I guide tubes/ tubes/assembly Fort Calhoun does not utilize non-other non-powered 1, Table 3-1 powered rods (other than the 1 instrument water holes) rods tube/assembly This is a fresh fuel pellet outer diameter, and does not account Fuel pellet for dimensional changes due to 0.3805 inches 1, Table 3-1 exposure. diameter Applicable to U02, blanket, and gadolinia-bearing pellets. Cladding inner This is a fresh fuel dimension, and 0.387 inches 1, Table 3-1 does not account for changes with diameter exposure such as creep. Cladding outer This is a fresh fuel dimension, and 0.440 inches 1, Table 3-1 does not account for changes with diameter exposure such as creep. FC08104 REV 0 EC 68969 PAGE 37

Omaha Public Power District Attachment 2: S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station OPPD Letter NE0-16-048 NFA I Revision 0 TOOl 16-048 NFA Page4 of8 July 8, 2016 NED-16-048 NFA Parameter Value Reference Comments

                                                              ~

Length of active fuel for full-length Applicable to rods with and without rods in assembly, axial blankets. 129.3 inches 1, Table 3-1 length of active fuel for part-length rods Fort Calhoun does not use part-in assembly length rods. Value corresponds to both guide Inner diameter of and instrument tubes. 1.035 Inches 1, Table 3-1 non-powered rods Fort Calhoun does not use non-powered rods. Value corresponds to both guide Outer diameter of and Instrument tubes. 1.115 inches 1, Table 3-1 non-powered rods Fort calhoun does not use non-powered rods. Cladding wall M5TM Zirconium material Alloy 1, Table 3-1 - Non-powered rod Fort Calhoun does not use non-wall material N/A - powered rods. Corresponds to the stack density 10.3743 glee 7, Table 5.1.1, for central zone fuel with no p.30 gadollnla (U02 only). Fuel density Pellet density specification. 96% of theoretical 1, Table 3-1 = Theoretical density 10.96 glee, per p. 7 of Reference 5. This U02 volumetric heat capacity Fuel specific heat See Table 2 can be converted to a specific 6,p. 139 heat using the density of the material. This corresponds to Zr density. 6.55 g/cm3 Per Attachment A of Reference 4, Cladding density 2, Section 4 zirconium is an extremely high percentage of the alloy. FC08104 REV 0 EC 68969 PAGE 38

Omaha Public Power District Attachment 2: S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station OPPD Letter NED-16-048 NFA/ Revision 0 TODI16-048 NFA Page 5 of8 JulyS, 2016 NED-16-048 NFA Parameter Value Reference Comments This MS volumetric heat capacity Cladding specific can be converted to a specific heat See Table 3 e. p. 142-143 heat using the density of the material. Maximum single assembly heat load Parameter Is not currently as a function of Not available N/A available and will need to be time following generated. shutdown Spent fuel pool maximum normal operating 140°F 3, Section 9.6.1 - temperature Table 2: Heat Capacity for U02 Fuel Volumetric Heat Capacity Temperature (°F) (Btuift3-°F) 0 33.64 100 37.75 200 40.48 300 42.38 400 43.78 500 44.86 600 45.71 700 46.41 800 47.01 900 47.53 1000 47.98 FC08104 REV 0 EC 68969 PAGE 39

Omaha Public Power District Attachment 2: S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station OPPD Letter NED-16-048 NFA I Revision 0 TODI16-048 NFA Page 6 of8 a. July 2016 NED-16-048 NFA Volumetric Heat Capacity Temperature (°F) (Btuift3-°F) 1100 48.4 1200 48.78 1300 49.13 1400 49.46 1500 49.78 1600 50.08 1700 50.38 1800 50.68 1900 50.98 2000 51.30 2100 51.63 2200 51.99 2300 52.39 2400 52.84 2500 53.35 2600 53.93 2700 54.59 2800 55.35 2900 56.23 3000 57.22 3100 58.35 3200 59.62 3300 61.04 3400 62.62 FC08104 REV 0 EC 68969 PAGE 40

Omaha Public Power District Attachment 2: S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station OPPD Letter NED-16-048 NFA I Revision 0 TOOl 16-048 NFA Page 7 of8 JulyS, 2016 NED-16-048 NFA Volumetric Heat Capacity Temperature (°F) (Btulft3-°F) 3500 64.38 3600 66.31 3700 68.42 3800 70.72 3900 73.20 4000 75.87 4100 78.73 4200 81.77 4300 85.00 4400 88.41 4500 92.00 4600 95.75 4700 99.68 4800 103.77 4900 108.00 5000 112.39 5100 116.91 FC08104 REV 0 EC 68969 PAGE 41

Omaha Public Power District Attachment 2: S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station OPPD Letter NED-16-048 NFA I Revision 0 TODI16-048 NFA Page 8 of8 July 8, 2016 NED-16-048 NFA Table 3: Volumetric Heat Capacity for M5' Cladding Volumetric Heat Capacity Temperature COF) (Btufft3-°F) 31.7 27.23 1520.3 39.98 1592.3 88.69 1790.3 32.57 2420.3 35.87 FC08104 REV 0 EC 68969 PAGE 42

Omaha Public Power District Attachment 3: S&L Calculation 2016-10694 (FC08104) Fort Calhoun Station OPPD TODI16-059 NFA Revision 0 Page 1 of2 Transmittal of Design Information CC-FC-31 0-F-01 Rev.O Transmittal of Design lnfonnation (TOOl) Refer to CC-AA-31 0 for requirements TOOl No.: TODI-..:.;16-0:.....::.;5=9~N:..:.;F:..:.A-=------ Revision:.=--0_ __ Date:10:....;f1-=0/~1=-6_ __ Transmittal Letter Number (if applicable): NED-16-059 NFA From: Tristan McDonald To: Robert Peterson (Sargent & Lundy>

Subject:

Hottest Assembly Decay Heat Data for Adiabatic Heatun Analvsis Information Provided and Intended Use: (Ust any reference documents or attachments) Decay heat data for the assembly with the highest heat load in the spent fuel pool for use in adiabatic spent fuel heatup analysis. Classification: QA ~n-Source and Basis of transmittal FC08514 Revision 0 Status of Information: 181 Approved for use D Preliminary, Scheduled for Confirmation on ---- Limitations on Use: None TOOl PrepareJ:-'- z;_ H Date \O I\" / "Z.. '" TOOl Approver OJM...f. rJlAkfJ_. Date \1) I Jo { 2.Dt\e Distribution: Per RM-AA-101 Project File Use TOOl Number in any subsequent correspondence concerning input FC08104 REV 0 EC 68969 PAGE 43

Omaha Public Power District Attachment 3: S&L Calculation 2016-1 0694 (FC08104) Fort Calhoun Station OPPD TODI 16-059 NFA Revision 0 Page 2 of 2 October 10, 2016 NED-16-059 NFA Mr. Robert Peterson Senior Manager* Nuclear Plant Analysis Sargent & Lundy 55 East Monroe Street Chicago, IL 60603

SUBJECT:

Hottest Assembly Decay Heat Data for Adiabatic Heatup Analysis

References:

1. FC08514 Revision 0, "Eighteen Month Shutdown: Hottest Assembly-Heat Load, Core Offload and Spent Fuel Pool Gamma and Neutron Source Term"

Dear Mr. Peterson,

Calculation FC08514 was completed to generate decay heat data as a function of time for the assembly with the highest decay heat load In the Fort Calhoun Station spent fuel pool after permanent cessation of operations. The calculation has determined that the hottest assem.bly in the spent fuel pool is Assembly DD07. The decay heat In Watts from 1 year after shutdown to 3 years after shutdown is shown below from Reference 1, p.44. The decay heat data presented above is suitable for use by Sargent & Lundy to perform calculations for the adiabatic heatup of the highest decay heat spent fuel assembly. Please contact me with any questions or if you require any further information.

  ~~~ Tristan McDonald FCS Decommissioning Transition Team C:       Russel Plesskunas (S&L)

Helmut Kopke {S&L) 444 SOUTii 16TH STREET MALL

• OMAHA, NE 68102-2247 FC08104 REV 0 EC 68969 PAGE 44

OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION DOCKET NUMBER 50-285 / LICENSE NUMBER CPR-40 ATTACHMENT 4 CALCULATION FC08513, EAB RADIATION SHINE DOSE 18 MONTHS POST SHUTDOWN WITH THE SFP DRAINED, without attachment E2 NON-PROPRIETARY

I I

 \ ATTACHMENT 1 Design Analysis Cover Sheet Page 1 of 5 Design Analysis                                                                          I Last Page No. 6 269 Analysis No.:          1 FC08513                                    Revision:       2 0 Major.           MinarD

Title:

' EAB RADIATION SHINE DOSE 18 MONTHS POST SHUTDOWN WITH THE SFP DRAINED EC/ECR No.:

• EC68969 Revision: 5 Station(s): ' FCS Component(s): ,.

Unit No.:

• 1 Discipline: ' RAD Descrip. Code/Keyword: 10 SFP,EAB Safety/QA Class: 11 Non-Safety Related System Code: 12 Structure: " Spent Fuel Pool CONTROLLED DOCUMENT REFERENCES 15 Document No.: From/To Document No.: From/To FC07586 From Technical Specification 4.1 & 5.2.2 From FC07688 From TDB-111.35 From FC08514 From TDB-1.8.1 From EA96-001 From From From From

, Is this Design Analysis Safeguards Information? Does this Design Analysis contain Unverified Assumptions? r-rom 16 17 YesO No. YesO No. If yes, ATI/AR#: From If yes, see SY-AA-101-106 This Design Analysis SUPERCEDES: " N/A in its entirety. Description of Revision (list changed pages when all pages of original analysis were not changed): " This is a new calculation

                                                                                                   ~~

Preparer: 20 Jan Bostelman, P.E. Print Name 0 / cill &JsS_IdJnaA.- Sign Namll'=' Ia ~/_rfJ.CJ IL Oat<# Method of Review: 21 Detailed Review [gl Altt~un;ttached) D Testing D Reviewer: 22 Steve Gebers, CHP Print Name Sign Name'

                                                                                                                                  / 1/zbNL Date Review Notes:                         Independent review~                Peer review       13 (For External Analyses Only)  External Approver: "

Print Name Sign Name Date Exelon Reviewer: " Print Name Sign Name Date No~~~ Wo.HtL-- Independent J*d Party Review Reqd? ,. YesD Exelon Approver: " Ca. (b I Wa_514L Print Name Sian Name L 1d of~,t-Date R0 11-10-16 FC08513, R0 EC 68969 1

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 2 of 269 CONTROLLED DOCUMENT REFERENCES (continued) " Document No.: I= rom/To Document No.: From/To 11405-A-5 From 11405-S-251 , Sheet 1 r:rom 11405-A-6 I= rom 11405-S-251 , Sheet 2 "'rom 11405-A-7 From 11405-S-251, Sheet 3 r:rom 11405-A-8 I= rom 11405-S-251, Sheet 4 "'rom 11405-A-9 From 11405-S-251, Sheet 5 From 11405-A-1 0 I= rom 11405-S-251, Sheet 6 I= rom 11405-A-11 From 11405-S-251, Sheet 7 From 11405-A-12 I= rom 11405-S-251, Sheet 8 I= rom 11405-A-13 From 11405-S-251, Sheet 9 From 11405-A-15 From 11405-S-251, Sheet 10 I= rom 11405-A-20 From 11405-S-251 , Sheet 11 From 11405-S-2 From 11405-S-251, Sheet 12 From 11405-S-6 I= rom 1000 From 11405-S-8 From 1001 From 11405-S-11 I= rom 1002 From 11405-S-12 From 1003 From 11405-S-13 I= rom 1004 From 11405-S-47 I= rom 1005 From 11405-S-48 r:rom 1007 I= rom 11405-S-49 "'rom 13007.01-EA-2A From 11405-S-50 .-rom 13007.01-EA-1 B I= rom 11405-S-51 "'rom 232552 I= rom 11405-S-52 .-rom G-576, Sheet 1 From 11405-S-53 r:rom D-4693, Sheet 1 .-rom 11405-S-54 "'rom D-4693, Sheet 5 .-rom 11405-S-55 "'rom G-576, Sheet 2 .-rom 11405-S-56 .-rom G-576, Sheet 20 .-rom 11405-S-57 From USAR Figure 1.2-1 .-rom 11405-S-58 From USAR Figure 1.2-2 .-rom 11405-S-59 I= rom USAR Figure 2.3-1 .-rom 11405-S-60 From C-4493 r:rom 11405-S-61 From ,...rom 11405-S-63 From From 11405-S-64 From I= rom 11405-S-65 From From 11405-S-66 From From 11405-S-90 From From FC08513, R0 EC 68969 2

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 3 of269 Table of Contents 1.0 Purpose .......................................................................................................................................7 1.1 Objective ..................................................................................................................................7 2.0 Inputs ............... ... .............................. ........................................ ........................................ .......... 8 3.0 References ........ .......................................................................................................................... 8 4.0 Assumptions .............................................................................................................................. 13 5.0 Method of Analysis and Acceptance Criteria ............................................................................. 16 5.1 Method of Analysis ................................................................................................................. 16 5.2 Acceptance Criteria ................................................................................................................ 17 5.3 Identification of Computer Code ............................................................................................ 18 5.4 MCNP Benchmark ................................................................................................................. 18 6.0 Model Geometry Overview ........................................................................................................ 19 6.1.1 Photon Flux-to-Dose Rate Conversion Factors ............................................................... 23 6.1.2 Material Specifications used in MCNP ............................................................................ 25 6.1.3 M3 Material Composition .................................................................................................42 6.1.4 Source Term ........................ ............................................................................................ 54 6.1.5 Source Term Output Dose Important Nuclides as a Function of Cooling Time ....... ........ 61 6.2 Plant Geometry ....................... ............................................................................................... 64 6.2.1 Model Geometry .............................................................................................................. 66 7.0 Conclusions ............................................................................................................................... 84 Attachment A: MCNP Input Files ....................................................................................................... 88 : MCNP Tallie Output. ................................................................................................. 117 Attachment C: Review of NRC RAis as they pertain to SFP shine dose .......................... ............... 120 Attachment D: Excel Spreadsheets ................................................................................................. 129 Attachment E: Additional References .. ............................................................................................219 Attachment F: File Listings ..............................................................................................................259 Attachment G: Reviewer Alternate Calculation (Micro Skyshine) ..................................................... 266 FC08513, R0 EC 68969 3

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 4 of269 List of Figures Figure6.1.1-1 ...................................................................................................................................... 19 Figure 6.1.1-2 ......................................................................................................................................20 Figure 6.1.1-3 ......................................................................................................................................21 Figure 6.1.1-4 ...................................................................................................................................... 22 Figure 6.1.1-1 ...................................................................................................................................... 24 Figure 6.1.1-2 .......................................... ............................................................................................ 25 Figure 6.1.3-1 ...................................................................................................................................... 53 Figure 6.1.5-1 ......................................................................................................................................61 Figure 6.1.5-2 ...................................................................................................................................... 63 Figure 6.1.5-1 ...................................................................................................................................... 64 Figure 6.1.5-2 ......................................................................................................................................64 Figure 6.1.5-3 ...................................................................................................................................... 65 Figure 6.1.5-4 ......................................................................................................................................66 Figure 6.2.1-1 .............. ............ ........ .................................................................................................... 67 Figure 6.2.1-2 ......................................................................................................................................66 Figure 6.2.1-3 .............. ........................................................................................................................ 68 Figure 6.2.1-4 ......................................................................................................................................71 Figure 6.2.1-5 ...................................................................................................................................... 72 Figure 6.2.1-6 ...................................................................................................................................... 73 Figure 6.2.1-7 ...................................................................................................................................... 74 Figure 6.2.1-8 ...................................................................................................................................... 75 Figure 6.2.1-9 ...................................................................................................................................... 77 Figure 6.2.1-10 .................................................................................................................................... 78 Figure 6.2.1-11 ....................................................................................................................................80 Figure 6.2.1-12 ............... ..................................................................................................................... 80 Figure 6.2.1-13 ....................................................................................................................................82 Figure 6.2.1-14 .................................................................................................................................... 83 FC08513, R0 EC 68969 4

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 5 of269 List of Tables Table 6.1.1-1 .......................................................................................................................................23 Table 6.1.1-2 .......................................................................................................................................24 Table 6.1.2-1 .......................................................................................................................................26 Table 6.1.2-2 ....................................................................................................................................... 27 Table 6.1.2-3 .......................................................................................................................................27 Table 6.1.2-4 ....................................................................................................................................... 28 Table 6.1.2-5 .......................................................................................................................................29 Table 6.1.2-6 .......................................................................................................................................31 Table 6.1.2-7 .......................................................................................................................................33 Table 6.1.2-8 ....................................................................................................................................... 34 Table 6.1.2-9 .......................................................................................................................................37 Table 6.1.2-10 ..................................................................................................................................... 37 Table 6.1.2-11 ..................................................................................................................................... 38 Table 6.1.2-12 .....................................................................................................................................38 Table 6.1.2-13 ..................................................................................................................................... 39 Table 6.1.2-14 ... ..................................................................................................................................41 Table 6.1.2-15 ..................................................................................................................................... 41 Table 6.1.3-1 .......................................................................................................................................42 Table 6.1.3-2 ....................................................................................................................................... 43 Table 6.1.3-3 ....................................................................................................................................... 43 Table 6.1.3-4 .......................................................................................................................................44 Table 6.1.3-5 .......................................................................................................................................45 Table 6.1.3-6 .......................................................................................................................................45 Table 6.1.3-7 .......................................................................................................................................45 Table 6.1.3-8 .......................................................................................................................................46 Table 6.1.3-9 .......................................................................................................................................47 Table 6.1.3-10 .....................................................................................................................................48 Table 6.1.3-11 .............................................................................. ....................................................... 49 Table 6.1.3-12 ..................................................................................................................................... 50 Table 6.1.3-13 ..................................... ................................................................................................ 51 Table 6.1.3-14 ..................................................................................................................................... 52 Table 6.1.4-1 .......................................................................................................................................55 Table 6.1.4-2 .......................................................................................................................................56 Table 6.1.4-3 .......................................................................................................................................57 Table 6.1.4-4 .......................................................................................................................................58 Table 6.1.4-5 .......................................................................................................................................59 Table 6.1.4-6 .......................................................................................................................................60 Table 6.2.1.2-1 ....................................................................................................................................69 Table 6.2.1.2-2 ....................................................................................................................................69 Table 6.2.1.3-3 ....................................................................................................................................76 Table 6.2.1.3-4 .................................................................................................................................... 79 Table 6.2.1.4-5 .................................................................................................................................... 81 Table 6.2.1.4-6 ....................................................................................................................................82 Table 7-1 ............................................................................................................................................. 85 Table 7-2 .............................................................................................................................................85 FC08513, R0 EC 68969 5

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 6 of269 Table 7-3 .............................................................................................................................................87 FC08513, R0 EC 68969 6

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 7 of269 1.0 Purpose The purpose for this calculation will be to show that the maximum dose at the EAB will be well below the acceptance criterion, which is taken to be based upon EPA Protective Action Guidelines (PAGs, Reference 21). The MCNP calculation determines the dose rate as a function of time following a full drain down of the spent fuel pool. This calculation does not determine a dose rate during the drain-down and utilizes a single time point which is used in calculating the time varying dose. The time of the event is estimated to be 18 months after December 1, 2016, i.e. June 1, 2018. The dose rate will be calculated using the code MCNP6.1 (References 1-4). The dose rates to be calculated were the direct and scattered gamma and neutron radiation as a result of a complete drain down of water from the Fort Calhoun Station (FCS) spent fuel pool (SFP). The complete drain down of water in the SFP is considered a beyond design basis event as it is not part of the FCS Updated Safety Analysis Report (USAR). 1.1 Objective Determine the gamma and neutron an average dose rate to the plant staff in the Control Room (CR) and to the public at the Exclusion Area Boundary (EAB). FC08513, R0 EC 68969 7

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 8 of 269 2.0 Inputs This calculation utilizes various drawings, references and assumptions to develop an MCNP model to determine dose rates at the EAB and CR. These inputs are show in the calculation where the input is referenced and utilized. 3.0 References

1. LA-UR-03-1987, MCNP- A General Monte Carlo N-Particle Transport Code, Version 5, Volume 1: Overview and Theory, Los Alamos National Laboratory, 2008.

2. LA-CP-03-0245, MCNP- A General Monte Carlo N-Particle Transport Code, Version 5, Volume II: User's Guide, Los Alamos National Laboratory, 2008.

3. LA-UR-13-22934, Initial MCNP6 Release Overview- MCNP6 version 1.0, Los Alamos National Laboratory, 2013.

4. LA-UR-10-06235, MCNP5-1 .60 Release Notes, Los Alamos National Laboratory, 2010.

5. PNNL-15870, Revision 1, Compendium of Material Composition Data for Radiation Transport Modeling, Pacific Northwest National Laboratory, 2011.

6. Contract 759, Concrete Structures, Containment, Structural Steel and Miscellaneous Facilities, Fort Calhoun Station Unit 1, GIBBS, HILL, DURHAM and RICHARDSON, INC.

7. FC07586, Maximum Gamma Energy Deposition rates in Spent Fuel Storage Pool Walls and Floor with EPU Spent Fuel Assemblies, Rev. 0

8. FC07688, Rev. draft, Ft. Calhoun Station Unit 1 EPU Spent Fuel Pool Decay Heat Calculation, AREVA Document No. 32-9128125-000, AREVA Letter AREVA-11-01908, dated July 28, 2011 Transmittal of Project Deliverable.

9. FC08514, Eighteen Month Shutdown: Hottest Assembly-Heat Load , Core Offload and Spent Fuel Pool Gamma and Neutron Source Term, Rev. 0

10. AREVA Engineering Information Record, 51-9124707-000, Fuel Assembly Design Parameters for Fort Calhoun Extended Power Uprate (EPU), 11/03/2009

11. TODI-BE-16-0PPD-001, Revision 1, August 15, 2016

12. Southern California Edison , Docket No. 50-206, 50-361, and 72-041 Response to Request for Additional Information Regarding Emergency, Planning Exemption request San Onofre Nuclear Generation Station, Units 1, 2, 3 and ISFSI, page 25

13. NED-DEN-99-0153, Radiological Parameter Table for Precursor Analyses. July, 1999.

14. Drawings 14.1. 11405-A-5, GENERAL ARRANGEMENT PRIMARY PLANT BASEMENT FLOOR PLAN P AND ID (12162), Sheet 1, Revision 45 14.2. 11405-A-6, PRIMARY PLANT GROUND FLOOR PLAN EL 1013 FT 0 IN P AND ID (12163), Sheet 1, Revision 88 FC08513, R0 EC 68969 8

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 9 of 269 14.3. 11405-A-7, PRIMARY PLANT INTERMEDIATE & OPERATING FLOOR PLANS EL 1025FT OIN (12164), Sheet 1, Revision 31 14.4. 11405-A-8, GENERAL ARRANGEMENT PRIMARY PLANT OPERATING FLOOR PLAN EL 1045 FT 0 IN (12165), Sheet 1, Revision 53 14.5. 11405-A-9, PRIMARY PLANT ROOF AND UPPER LEVELS PLAN (12166), Sheet 1, Revision 19 14.6. 11405-A-10, PRIMARY PLANT SOUTH ELEVATION,AUXILARY AND CONTAINMENT (12167), Sheet 1, Revision 6 14.7. 11405-A-11, PRIMARY PLANT WEST ELEVATION,AUXILARY AND CONTAINMENT (12168), Sheet 1, Revision 10 14.8. 11405-A-12, PRIMARY PLANT NORTH ELEVATION,AUXILARY AND CONTAINMENT (12169), Sheet 1, Revision 7 14.9. 11405-A-13, PRIMARY PLANT SECTION A-A,AUXILARY AND CONTAINMENT ( 12170), Sheet 1, Revision 13 14.10. 11405-A-15, PRIMARY PLANT WALL SECTIONS AND FINISH SCHEDULE (12172), Sheet 1, Revision 5 14.11. 11405-A-20, PRIMARY PLANT CONTROL ROOM PLAN ELEVATIONS & DETAILS (12177), Sheet 1, Revision 32. 14.12. 11405-S-2, CONTAINMENT STRUCTURE STEEL LINER SHEET 1,DOME PLAN AND CONTAINMENT REINFORCING SECTION (16381), Sheet 1, Revision 15 14.13. 11405-S-6, CONTAINMENT STRUCTURE CONCRETE OUTLINE SHEET 1, DOME PLAN, DEVELOPED EXTERIOR ELEVATION, WALL (16385), Sheet 1, Revision 9 14.14. 11405-S-8, CONTAINMENT STRUCTURE CONVENTIONAL REINFORCEMENT, SHEET 1, DOME REINFORCEMENT PLAN (16387), Sheet 1, Revision 7 14.15. 11405-S-11, CONTAINMENT STRUCTURE PRESTRESSING SYSTEM SHEET 1, CONDUIT AND TENDON ANCHORAGES (16390), Sheet 1, Revision 5 14.16. 11405-S-12, CONTAINMENT STRUCTURE PRESTRESSING SYSTEM SHEET 2, DEVELOPED INTERIOR ELEVATION, TENDON GEOMETRY (16391), Sheet 2, Revision 4 14.17. 11405-S-13, CONTAINMENT STRUCTURE INSTRUMENTATION, EQUIPMENT ACCESS HATCH, PERSONNEL AIR LOCK, DOME PLAN (16392), Sheet 1, Revision 9 14.18. 11405-S-47, AUXILIARY BUILDING FOUNDATION PLAN ELEV 989FT OIN, OUTLINE, SHEET 1 (16432), Sheet 1, Revision 15 14.19. 11405-S-48, AUXILIARY BUILDING FOUNDATION PLAN ELEV 989FT OIN, OUTLINE, SHEET 2 (16433), Sheet 2, Revision 2 14.20. 11405-S-49, AUXILIARY BUILDING FOUNDATION PLAN ELEV 971FT OIN OUTLINE, SHEET 3 (16434), Sheet 1, Revision 11 FC08513, R0 EC 68969 9

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 10 of 269 14.21. 11405-S-50, AUXILIARY BUILDING GROUND FLOOR PLAN ELEV 1007FT OIN , 1011FT 0 IN, 1013 FT 0 IN, OUTLINE, SHEET 1 (16435), Sheet 1, Revision 17 14.22. 11405-S-51, AUXILIARY BUILDING GROUND FLOOR PLAN ELEV 1007 FT 0 IN, OUTLINE, SHEET 2 (16436), Sheet 2, Revision 17 14.23. 11405-S-52, AUXILIARY BUILDING GROUND FLOOR SECTION AND DETAIL SHEET 3 (16437), Sheet 3, Revision 7 14.24. 11405-S-53, AUXILIARY BUILDING INTERMEDIATE FLOOR ELEV 1025FT OIN, OUTLINE, SHEET 1 (16438), Sheet 1, Revision 6 14.25. 11405-S-54, AUXILIARY BUILDING INTERMEDIATE FLOOR ELEV 1025FT OIN, OUTLINE, SHEET 2 (16439), Sheet 5, Revision 11 14.26. 11405-S-55, AUXILIARY BUILDING OPERATING FLOOR ELEV 1036FT OIN OUTLINE, SHEET 3 (16440), Sheet 3, Revision 19 14.27. 11405-S-56, AUXILIARY BUILDING ROOF PLAN ELEV 1057 FT 0 IN , OUTLINE, SHEET 1 (16441 ), Sheet 1, Revision 16 14.28. 11405-S-57, AUXILIARY BUILDING ROOF PLAN ELEV 1044'-0" OUTLINE (16442), Sheet 2, Revision 10 14.29. 11405-S-58, AUXILIARY BUILDING ROOF PLAN ELEV 1083FT OIN , OUTLINE , SHEET 1 (16443), Sheet 1, Revision 3 14.30. 11405-S-59, AUXILIARY BUILDING EXTERIOR WALL OUTLINE SHEET 1 (16444), Sheet 1, Revision 3 14.31. 11405-S-60, AUXILIARY BUILDING EXTERIOR WALL OUTLINE, SHEET 2 (16445), Sheet 2, Revision 7 14.32. 11405-S-61, AUXILIARY BUILDING SPENT FUEL WELL OUTLINE (16446), Sheet 1, Revision 14 14.33. 11405-S-63, AUXILIARY BUILDING SECTION SHEET 1 (16448), Sheet 1, Revision 15 14.34. 11405-S-64, AUXILIARY BUILDING SECTION SHEET 2 (16449}, Sheet 2, Revision 11 14.35. 11405-S-65, AUXILIARY BUILDING EXTERIOR WALL REINFORCEMENT SHEET 1 (16450), Sheet 1, Revision 10 14.36. 11405-S-66, AUXILIARY BUILDING EXTERIOR WALL REINFORCEMENT SHEET 1 (16451), Sheet 2, Revision 7 14.37. 11405-S-90, AUXILIARY BUILDING ROOF PLAN EL 1083 FT 0 IN REINFORCEMENT (16474), Sheet 3, Revision 3 14.38. 11405-S-251, SITE PLAN TOPOGRAPHY (16487), Sheet 1, Revision 9 14.39. 11405-S-251, SITE PLAN TOPOGRAPHY (45711), Sheet 2, Revision 5 14.40. 11405-S-251, SITE PLAN TOPOGRAPHY (45712), Sheet 3, Revision 3 14.41. 11405-S-251, SITE PLAN TOPOGRAPHY (45713), Sheet 4, Revision 0 FC08513, R0 EC 68969 10

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 11 of 269 14.42. 11405-S-251, SITE PLAN TOPOGRAPHY (45714), Sheet 5, Revision 3 14.43. 11405-S-251, SITE PLANT TOPOGRAPHY (45951 ), Sheet 6, Revision 2 14.44. 11405-S-251, SITE PLAN TOPOGRAPHY (46501), Sheet 7, Revision 0 14.45. 11405-S-251, SITE PLAN TOPOGRAPHY (46502), Sheet 8, Revision 0 14.46. 11405-S-251, SITE PLAN TOPOGRAPHY (46503), Sheet 9, Revision 0 14.47. 11405-S-251, SITE PLAN TOPOGRAPHY (46504), Sheet 10, Revision 0 14.48. 11405-S-251, SITE PLAN TOPOGRAPHY (46505), Sheet 11, Revision 0 14.49. 11405-S-251, SITE PLAN TOPOGRAPHY (46506), Sheet 12, Revision 0 14.50. 11405-S-270, SITE PLAN EXCAVATION AND PILING (16502), Sheet 1, Revision 6 14.51. 1000, POOL LAYOUT SPENT FUEL STORAGE RACKS (42762), Sheet 1, Revision 1 14.52. 1001 , RACK CONSTRUCTION REGION 1 SPENT FUEL RACKS (42764) , Sheet 1, Revision 1 14.53. 1002, RACK CONSTRUCTION REGION 1 SPENT FUEL RACKS (42765), Sheet 1, Revision 1 14.54. 1003, RACK CONSTRUCTION- REGION II SPENT FUEL STORAGE RACKS (42766), Sheet 1, Revision 1 14.55. 1004, RACK CONSTRUCTION - REGION II SPENT FUEL STORAGE RACKS (42767), Sheet 1, Revision 1 14.56. 1005, SUPPORT LEVELING TOOL SPENT FUEL STORAGE RACKS (42768), Sheet 1, Revision 1 14.57. 1007, SUPPORT DETAILS SPENT FUEL STORAGE RACKS (42953), Sheet 1, Revision 1 14.58. 13007.01-EA-2A, MAIN PLANT, KEY PLANS AND DOOR SCHEDULE (11298), Sheet 1, Revision 13 14.59. 13007.01-EY-1B, FENCE AND GATE DETAILS SITE SECURITY (11343), Sheet 1, Revision 6 14.60. 232552, STEEL ROLLING DOOR, RAILROAD SIDING FOR AUXILIARY BUILDING (12103), Sheet 5, Revision 1 14.61. G-576, GENERAL ARRANGEMENT EXT. MT TYPE 1 VL 2PC FLOOR (41771), Sheet 1, Revision 0 14.62. D-4693, ATTACHMENT TO AUXILIARY BUILDING (41866), Sheet 1, Revision 0 14.63. D-4693, PANEL Al-69 FOR DOOR 1004-1A (41871), Sheet 5, Revision 1 14.64. G-576, DOOR PANEL FRAME V.L. TYPE 1 EXT. MOUNT 2 PIECE (42072), Sheet 2, Revision 0 FC08513, R0 EC 68969 11

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 12 of 269 14.65. G-576, DOOR PANEL SHEETING V.L. TYPE 1 EXT MOUNT 2 PIECE (42077), Sheet 20, Revision 0 14.66. USAR Figure 1.2-1, SITE PLAN (36030) USAR FIGURE, Sheet 1, Revision 9 14.67. USAR Figure 1.2-2, RESTRICTED AREA (36031), Sheet 1, Revision 7 14.68. FIG. 2.3-1, SITE TOPOGRAPHY (36046), Sheet 1, Revision 2 14.69. C-4493, SITE PLAN TAG NUMBERS FOR MAJOR PLANT STRUCTURES (64944), Sheet 1, Revision 0

15. "American National Standard Neutron and Gamma-Ray Flux-to-Dose Rate Factors," ANSI/IANS-6.1.1-1977, American Nuclear Society, La Grange Park, Illinois, March 1977.

16. OPPD Analysis, EA-FC-96-001," Criticality Safety Evaluation of the Ft. Calhoun Spent Fuel Storage Rack for Maximum Enrichment Capability," Rev. 1.

17. Technical Specification, Omaha Public Power District, Fort Calhoun Station Unit No. 1 Operating License No. DPR-40, T.S 5.2.2.

18. "Nuclear Chemical Engineering," M. Benedict, T. Pigford & H. Levi, Second Edition McGraw-Hill, 1981, Page 941. (Attachment E1)

19. OPPD Procedure, TDB-111.35, TECHNICAL DATA BOOK, "SPENT FUEL POOL RACK MEASUREMENTS," Revision 1b.

20. OPPD Procedure, TDB-I.B.1, TECHNICAL DATA BOOK, "SPENT FUEL POOL LAYOUT and FH-12 COORDINATES," Revision 72.

21. "MANUAL OF PROTECTIVE ACTION GUIDES *AND PROTECTIVE ACTIONS FOR NUCLEAR INCIDENTS," Office of Radiation Programs, United States Environmental Protection Agency, Second printing, May 1992.

22. "NUREG-0737, "Clarification of TMI Action Plan Requirements" Division of Licensing, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, November 1980.

FC08513, R0 EC 68969 12

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 13 of 269 4.0 Assumptions

1. All structures are constructed of Ordinary Concrete.

Basis: Ordinary Concrete (Reference 5, Material 95 Concrete, Ordinary (NBS 03)) is a generic concrete of granite/sand aggregate and the most commonly available cement. In most nuclear power plants, the grade of concrete is specified during construction. With varying concretes densities ranging from 2.18 to 5.56 grams/cm 3 depending on application (Reference 5). The FCS contract 759 (Reference 6) specified a minimum of 145 lbs/ft3 for type A and B concretes (equivalent to 2.322 grams/cm 3 ) and 224 lbs/ft3 for type C concrete (equivalent to 3.588 grams/cm 3). Since most of the construction utilized Type A and B concretes and the Compendium of Material Composition Data for Radiation Transport Modeling has ordinary concretes with varying densities, the 2.35 grams/cm 3 was chosen as it has a similar density to the Type A and B concrete specifications.

2. Any volume not specifically specified as concrete, steel or fuel mixture was defined as dry air.

Basis: Utilizing air without moisture simplifies the problem, additionally the scope of the evaluation is the SFP is completely drained. All water is removed.

3. The railroad siding rollup door (1 004-1A) a rolling door of overlapping carbon steel design was modeled as single sheet of carbon steel. (Carbon Steel is used as the drawing title, Reference 14.60 is "STEEL ROLLING DOORS- RAILROAD SIDING FOR AUXILIARY BUILDING")

Basis: Due to the design of the door is overlapping carbon steel design. To calculate the percent of overlap is excessive detail for this type of model, and carbon steel is used since stainless steel would have been too expensive. This approach allows for more transmission of particles though door, thus the calculation will calculate a higher dose rate .

4. The railroad siding vertical lift door (1 004-1 C) was modeled similarly to the design of the door without utilizing the internal structural steel within the door or overlapping the panels.

Basis: Due to the design of the door, The Polyisocyanurate insulation wrapped by the 14-gauge carbon steel simplifies the model. This approach also allows for more transmission of particles though door, thus the calculation will calculate a slightly higher dose estimate.

5. The reinforcing steel rebar in the concrete walls was not modeled due to the specific details that are difficult and excessive detail to model.

Basis: The steel of the rebar is a more effective photon reflector/scatterer given its relatively high density compared to concrete. The lack of rebar allows for more transmission of particles though walls of the auxiliary building and spent fuel area, thus the calculation will calculate a slightly higher dose estimate. FC08513, R0 EC 68969 13

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 14 of 269

6. The equipment in the affected rooms and hallways was disregarded in terms of shielding or scattering.

Basis: This is conservative due to the materials would scatter/absorb particles (photons/ neutrons) and reduce the particle's effective energy, thus reduce dose from that particle. The affected buildings and hallways contain significant amounts of equipment. For example, room 69 contains: the various tanks storing water, component cooling water pumps and piping , as well as ventilation equipment. Equipment like this occupies space and provides additional scattering of the source photons/neutrons.

7. Modeling the auxiliary building and control room floors and walls was simplified by excluding the 2.5 feet concrete beams.

Basis: By averaging these beams into the thickness of the floors and walls would increase the thickness and potentially decrease the dose. Incorporating the beams changes the scattering similarly to ignoring the equipment within the structures.

8. Only the areas adjacent to SFP at the 1025' elevation of the auxiliary building was modeled ,

eliminating room 68 (the cask decontamination room). Basis: The remaining concrete structures below this level have a considerable amount of reinforced concrete and will have little impact to particles transporting through the exterior walls of the auxiliary building and to the control room.

9. The control operators occupy the control room twenty-four (24) hours a day, seven (7) days a week. (720 hours).

Basis: Per current plant Technical Specification Table 5.2-1, "MINIMUM SHIFT CREW COMPOSITION," at least one of these individuals (or the second senior licensed operator, if both senior licensed operators are in the control room) must be present at the controls at all times. This consistent with NUREG-0737, Reference 22 .

10. No axial or radial peaking factors are applied in developing the source, since this is a homogenized source term.

Basis: The source term represents the spent fuel as a homogenized volume with each plane of the source representing a planar source with no self-shielding.

11. The Polyisocyanurate foam (PNNL-15870 material249 page 240 of 357) is an equivalent replacement for the Trymar 9501.

Basis: OPPD Drawing G-576 shows Trymar 9501 as the insulation component for door 1004-1A. The DOW Chemical MSDS shows Trymar 9501 as a "POLYMERIZED POLYURETHANE MODIFIED POLYISOCYANURATE RIGID CELLULAR PLASTIC." FC08513, R0 EC 68969 14

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 15 of 269 Since Trymar 9501 is not a commonly known material, research shows the following from the DOW MSDS sheet (Trymer): DOW CHEMICAL CO-- TRYMER 9501 -4 RIGID FOAM INSULATION, 02837 -- 9320-00N056603

12. This calculation assumes (a) no cladding failure during or after the drain down and (b) any pre-existing cladding breaches (fuel pins with holes identified prior to Cycle 21) have released the gap activity or decayed significantly to impact overall radiation dose.

Basis: (a) Spent fuel pool (SFP) accidents involving a sustained loss of coolant have the potential for leading to significant fuel heat up and resultant release of fission products to the environment. This heatup is precluded because the fuel in the SFP is assumed to be beyond the point of heating up enough the fuel clad temperature will not exceed 900C for 10-hours after loss of SFP inventory, and; (b) Any fuel that has cladding breaches (i.e. from grid to rod fretting, etc.) has released radionuclides from the gap during storage in the pool water over the years, and was subsequently filtered through the SFP demineralizer and filtration system, along with pool skimmers. Cycle 20 (September 2003) was the last operating cycle with known fuel failures. This assumption maximizes the direct exposure from shine and bound the site-boundary and control room dose rate from direct shine. FC08513, R0 EC 68969 15

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 16 of 269 5.0 Method of Analysis and Acceptance Criteria 5.1 Method of Analysis MCNP6, "Monte Carlo N-Particle Transport Code," a radiation transport computer code, was utilized to determine dose rates from direct and scattered radiation at Fort Calhoun Station from spent fuel assemblies within the spent fuel pool following a complete loss of SFP water. This calculation does not determine airborne immersion or inhalation doses due to releases of radioactive material from the spent fuel assemblies following drain down of the SFP. Gamma and neutron cases are utilized to determine "shine" dose rates. A complete MCNP model was developed to evaluate specifically the FCS pool, auxiliary building, control room and containment. The simplified model assumes the main structures of the facility are present ignoring plant components such as piping, valves, pumps, ventilation equipment and sub-structures like electrical panels and internal non-structural walls. The MCNP model developed used existing FCS fuel information that was derived for the Extended Power Uprate (EPU) calculations (Reference 8) in terms of pool geometry and fuel modeling to expedite and simplify the process. EPU operating parameters were not utilized in this analysis. Because the fuel cladding is assumed intact, at any point after 18-months post shutdown, the dominant noble gas in the source term within the fuel is Kr-85. Because Kr-85 is a noble gas, no shine from ground will be included, because the fuel cladding remains intact. The FCS model was developed to include various dose point tallies and detectors inside the control room, in and around the SFP area and outside of the buildings. Simplification of geometry was utilized when excessive details could not be modeled and did not reduce the expected attenuation before radiation would exit the auxiliary building. A homogenized fuel source was used which includes the 944 fuel assemblies to be stored in the FCS spent fuel pool after final operation of the Cycle 28. In addition to the fuel, the homogenized source incorporates the stainless-steel racks and air to approximate self-absorption of the radiation source. The radiation source term for the fuel region noted above was derived from the ORIGEN-ARP/ORIGEN-S modules within SCALE6.1 (Reference 9). The conversion factors used to convert both energy dependent gamma and neutron flux to dose rates will be based upon ANSI/ANS-6.1.1-1977. Point detectors were used in the model at the spent fuel walkway, Auxiliary building roof, at ground level 10 feet and 100 feet from the railroad siding door (1 004-1 ), the control room operating space and the closest approach to the FCS EAB (-900 meters south southwest measuring from containment centerline). Only the two detectors which were of importance to the purpose of the calculation were utilized in the final case runs. The others were of initial interest in comparisons to the San Onofre for benchmark comparisons. The dose to the plant staff and public is significantly different with a pool drain down event at 18 months after shutdown than immediately after reactor shutdown. Therefore, the last aspect of this calculation is to document the available source term of the fuel discharged from the reactor (assuming a full core offload prior to December FC08513, R0 EC 68969 16

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 17 of 269 1, 2016) and the subsequent decay of that fuel to June 1, 2018. Important for dose considerations during a loss of all SFP water inventory is also the fuel that is currently in the SFP . The gamma and neutron source term for fuel stored in the pool (i.e. fuel that has decayed longer than 3 years) was included. 5.2 Acceptance Criteria The calculation will estimate the dose rate from gamma and neutron flux from the spent fuel. With previous industry calculations (Reference 12) indicating very little dose from the n-gamma reaction, no n-gamma dose was determined. This is also indicated by the results of FC08514 (Reference 9) which does not produce that type of source from ORIGEN-ARP. It is expected that gamma dose rate will be the dominate contributor for integrated dose. A 95% confidence interval will be evaluated for the dose rate. The DOE dose at the EAB is expected to be below the 50-mrem in a year for members of the public or% the TEDE dose for member of the public as defined in 10 CFR 20. The dose rate in the Control Room is expected to be below 15-mR/hr as defined in NUREG-0737 (Reference 22). In addition, the 15-mR/hr is consistent with the Control Room emergency action level for the new shutdown EALs (Reference 21 ). FC08513, R0 EC 68969 17

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 18 of 269 5.3 Identification of Computer Code MCNP6' is a general-purpose, continuous-energy, generalized-geometry, time dependent, Monte Carlo radiation-transport code designed to track many particle types over broad ranges of energies. MCNP6 represents the culmination of a multi-year effort to merge the MCNP5' [X-503] and MCNPX' [PEL 11] codes into a single product comprising all features of both. For those familiar with previous versions of MCNP, the code has been expanded to handle a multitude of particles and to include model physics options for energies above the cross-section table range, a material burnup feature, and delayed particle production. Expanded and/or new tally, source, and variance-reduction options are available to the user as well as an improved plotting capability. Although MCNP6 is simply and accurately described as the merger of MCNP5 and MCNPX capabilities, the result is much more than the sum of these two computer codes. MCNP6 is the result of five years of effort by the MCNP5 and MCNPX code development teams. These groups of people, residing in the Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), and Nuclear Engineering and Nonproliferation Division, Systems Design and Analysis Group (NEN-5, formerly D-5}, have combined their code development efforts to produce the next evolution of MCNP. MCNP6.1/MCNP5/MCNPX is maintained and provided by the Radiation Safety Information Computation Center from Los Alamos National Laboratory, Los Alamos, New Mexico for the U.S. DEPARTMENT OF ENERGY. MCNP 6.1 is maintained in OPPD SWIMS, approved by (Software Change Review Committee (SCRC). 5.4 MCNP Benchmark MCNP is widely used and accepted by the nuclear utility industry to perform radiological analysis. The following case from the MCNP Installation files was run for comparison to ensure MCNP6.1 was installed properly. The version utilized in this calculation was MCNP compiled version as delivered by LANL and thus no changes to the code have been made.

              - Hall105a.txt, "a modified hallway problem with gamma source," was used in this calculation as a benchmark since this case uses some similar materials and a gamma source term. The results of the benchmark are shown in Attachment E.

Comparison of the tally results and resulting statistical comparisons show excellent agreement and provide a basic confidence that MCNP6.1 was installed correctly. MCNP5 and MCNP6 can be utilized for radiation dose analysis as intended. The test case was run with both MCNP5 and 6. The case output from the installation was run using MCNP5, and because of computer machine CPU processing differences and updated cross section libraries, the results are not identical. The MCNP5 and MCNP6 test case results, as installed, are identical with exception of runtime. FC08513, R0 EC 68969 18

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 19 of 269 6.0 Model Geometry Overview Fort Calhoun Station is located between Nebraska Highway 75 and the Missouri River at 96 degrees, 4 minutes and 39 seconds west longitude and 41 degrees, 31 minutes and 14 seconds north latitude in Washington County, Nebraska, on the southwest bank of the Missouri River as shown on Figure 6.1 .1-1 (Reference 14.67). The Exclusion Area Boundary (EAB) for FCS is shown in blue. The minimum Distance to the EAB is 910 meters at the 187-degree radial of the containment building (Reference FCS Technical Specification 4.1 ). The actual value used in the model is 904.4 meters. Figure 6.1 .1-1 FCS Site Layout with EAB (USAR Figure 1.2-2, Reference 14.67)

                                                                /
                                                           $ /    ~,1:-,.
                                                          ~   /         *c.
                                                            /              "il~~

I v4tr I "ilcc

                                                .ff' /                                    ~ss -9c
                                              ~(S   /                                            *
                                           &' /
                                           ~ /
                                            /
                                         /
                                      /
                                    /
                                  /
                               /
                             /
                         //       ~z ,....._           SITE BOUNOERY

( ~ ,-------~ I I I

                        ---------~                                                 l FC08513, R0                                       EC 68969                                            19

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 20 of 269 To calculate the limiting dose rates at the EAB, the closest point at the site EAB was chosen. This was determined to be at the at site access from Highway 75, shown on Figure 6.1 .1-2 (Reference 14.67). This was evident, with the spent fuel pool located in the southwest portion of the auxiliary building, which is southwest of containment (shown in blue). Another attribute contributing to the Highway 75 EAB location was the elevation at highway 75 is greater than 1080' above sea level (USAR Section 2.3 and Reference 14.42). The elevation at the Highway 75 EAB location is above the spent fuel pool walkway of 1038' (Reference 14.9) which provides more of a direct line-of-site path to the radiation source. Other features of the model for this calculation, only the containment and the auxiliary building were modeled ignoring the radwaste building and the old warehouse, both of which are west of the auxiliary building. The EAB distance utilized (904.4 meters) is based upon review of Technical Specification 4.1 (EAB distance is 910 meters from Containment centerline). Figure 6.1.1-2 FCS Plant Layout with EAB SAR F ure 1.2-1, Reference 14. r-------~ II ~ 1I I I

                      '      ..,---~

r---~ c~ I

i  ;..I _______ _

1 II r--------~ L ..11 ___ .J '.: II FC08513, R0 EC 68969 20

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 21 of 269 Utilizing plant elevation and architectural drawings, the major components of the model were developed. These include the concrete structure of the Spent Fuel Pool and transfer canal, the Auxiliary building (Rooms 3, 25, 25A, 67 & 69), Containment and the Control Room (room 77). See Figure 6.1 .1-3 and Figure 6.1 .1-4, which show the general arrangement of the structures. Figure 6.1.1-3 FCS Plant Elevation Layout (FCS Drawing 11405-A-13, Reference 14.9, w/ overlay to show Room 77)

                            - -. -  *( --

ELE!Imiltl.. LOOKING a ANT N(){(TII FC08513, R0 EC 68969 21

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 22 of 269 Figure 6.1.1-4 FCS Primary Operating Floor Plan (FCS Drawing 11405-A-8 Reference 14.4) RoomT/ (Control Room) Room69 Rm2S FC08513, R0 EC 68969 22

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 23 of 269 Using an X, Y, Z coordinate system and establishing the origin located at the containment center at the bottom of the concrete basemat, allows a single point of reference to determine all other MCNP6 model dimensions. The containment basemat is at an elevation of 979' above sea level (Reference 14.14. And every plant elevation (dimension "Z") modeled in MCNP is based upon that reference elevation from the 979' value and convert feet to centimeters (30.48 cm/ft). The same approach is utilized for dimensions in the "X" and "Y" dimensions. East to West is -X to +X and South to North is -Y to +Y, utilizing the containment centerline at X and Y =0. 6.1.1 Photon Flux-to-Dose Rate Conversion Factors For the purposes of this calculation the ANSI/ANS-6.1.1-1977 neutron and photon flux-to-dose rate conversion factors from Reference 15 and presented in Reference 1 are used. The values have been converted from Rem/hr/(p/cm 2-s) to mRem/hr/(p/cm 2-s) by multiplying the value by 1.0E+3, to complete the conversion from Rem to mRem. The neutron flux-to-dose rate conversion factors are presented in Table 6.1.1-1 and the photon flux-to-dose rate conversion factors are presented in Table 6. 1. 1-2, noting there are significantly more energy groups for gamma dose conversion, than there are for neutron. Table 6.1.1-1 ANSI/ANS-6.1.1-1977 Neutron Flux-to-Dose Rate Conversion Factors MeV (mRem[hr)[(n[cm2-s) MeV (mRem[hr)l(n[cm2-s) 2.5E-08 3.67E-03 5.0E-01 9.26E-02 1.0E-07 3.67E-03 1.0 1.32E-01 1.0E-06 4.46E-03 2.5 1.25E-01 l.OE-05 4.54E-03 5.0 1.56E-01 l.OE-04 4.18E-03 7.0 1.47E-01 l.OE-03 3.76E-03 10.0 1.47E-01 1.0E-02 3.56E-03 14.0 2.08E-01 1.0E-01 2.17E-02 20.0 2.27E-01 FC08513, R0 EC 68969 23

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 24 of 269 Table 6.1.1-2 ANSI/ANS-6.1.1-1977 Photon Flux-to-Dose Rate Conversion Factors MeV (mRem[hr}l(~[cm2-s) MeV (mRem[hr)l(~lcm2-s) MeV (mRem[hr)lh~lcm2-s) 0.01 3.96E-Q3 0.55 1.27E-Q3 4.25 5.23E-Q3 0.03 5.82E-Q4 0.6 1.36E-Q3 4.75 5.60E-Q3 0.05 2.90E-Q4 0.65 1.44E-Q3 5 5.80E-Q3 0.07 2.58E-Q4 0.7 1.52E-Q3 5.25 6.01E-Q3 0.1 2.83E-Q4 0.8 1.68E-Q3 5.75 6.37E-Q3 0.15 3.79E-Q4 1 1.98E-Q3 6.25 6.74E-Q3 0.2 5.01E-Q4 1.4 2.51E-Q3 6.75 7.11E-Q3 0.25 6.31E-Q4 1.8 2.99E-Q3 7.5 7.66E-Q3 0.3 7.59E-Q4 2.2 3.42E-Q3 9 8.77E-Q3 0.35 8.78E-Q4 2.6 3.82E-Q3 11 1.03E-Q2 0.4 9.85E-Q4 2.8 4.01E-Q3 13 1.18E-Q2 0.45 1.08E-Q3 3.25 4.41E-Q3 15 1.33E-Q2 0.5 1.17E-Q3 3.75 4.83E-Q3 MCNP computes an energy dependent flux to the tally. These conversion factors translate those fluxes to dose rates for the associated energy bins. MCNP ultimately sums all the dose rates for the energy bins to compute a total dose rate for the scenario at a detector at that point. The conversion factors as utilized in MCNP for the neutron and gamma dose are shown in Figure 6.1.1-1 and Figure 6.1 .1-2, respectively. Figure 6.1.1-1 Neutron Flux to Dose Conversion Cards c C Dose Conversion Factors for Neutrons (mRem/hr)/(particle/cm2-sec) c deO 2 . 5E-08 1.0E-07 1.0E-06 1.0E-05 1.0E-04 1 . 0E-03 1.0E-02 1 . 0E-01 S . OE-01 1.0E+00 2.5E+00 5.0E+00 7.0E+00 1 . 0E+01 1.4E+01 2.0E+01 c dfO 3.67E-03 3.67E-03 4.46E-03 4.54E-03 4.18E-03 3.76E-03 3.56E-03 2.17E-02 9 . 26E-02 1.32E-01 1.25E-01 1.56E-01 1.47E-01 1 . 47E-01 2 . 08E-01 2.27E-01 c FC08513, R0 EC 68969 24

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 25 of 269 Figure 6.1.1-2 Gamma Flux to Dose Conversion Cards c C Dose Conversion Factors for Photons (mrem/hr)/(particle/cm2-sec) c deO 0.01 0.03 0.05 0.07 0.1 0 . 15 0 . 2 0 . 25 0.3 0.35 0.4 0.45 0.5 0.55 0.6 0 . 65 0.7 0.8 1.0 1.4 1.8 2 . 2 2.6 2 . 8 3.25 3.75 4 . 25 4 . 75 5.0 5 . 25 5.75 6.25 6.75 7.5 9 . 0 11.0 13.0 15.0 dfO 3.96E-3 5.82E-4 2.90E-4 2 . 58E-4 2.83E-4 3.79E-4 5.01E-4 6 . 31E - 4 7.59E-4 8 . 78E-4 9 . 85E-4 1.08E-3 1.17E-3 1.27E-3 1.36E-3 1.44E-3 1.52E-3 1 . 68E-3 1.98E-3 2.51E-3 2.99E-3 3.42E-3 3.82E-3 4.01E-3 4.41E-3 4.83E-3 5.23E-3 5.60E-3 5 . 80E-3 6.01E-3 6.37E-3 6.74E-3 7 . 11E-3 7 . 66E-3 8.77E-3 1.03E-2 1 . 18E-2 1 . 33E-2 c 6.1.2 Material Specifications used in MCNP The materials utilized in MCNP for modelling the plant is basic ordinary concrete, (Assumption 4.1), dry air (Assumption 4.2), carbon steel, (Assumption 4.3 and 4.4), References 14.60 and 14.65; and foam insulation (Assumption 4.4 and Assumption 4.11), Polyisocyanurate foam, References, 5 and 14.65). The source material M3 is comprised of a homogenized mixture of spent fuel, stainless steel storage racks (with boral absorbing panels) and because the spent fuel pool is drained of all water, dry air. Materials utilized in MCNP are defined using the material card (Mn), where "n" is the number of the material and materials are specified by their constituents using the nuclide identifiers found in the in the MCNP libraries (Reference 1). Reference 5 provides a list of over 300 commonly used materials with the specification of elements by nuclide in either weight percent or atom fraction. This was utilized for materials M1, M2, M4 and MS. Material M1 The construction of the plant (refer to references 14.12 - 14.37) shows that the structures (floors walls and roofs were built with reinforced (rebar) concrete. Using ordinary concrete with a nominal density 2.35 g/cm 3 (Reference 5) is a representative value for structural concrete. FCS Contract 759 (Reference 6) specifies a minimum density of 145 lbs/ft3 (2.32 gl cm 3) for both Class A and B concrete not accounting for the rebar and 225 lbs/ft3 (3.6032 gl cm 3) for Class C "heavy concrete" which was utilized in shielding applications. With the addition of rebar, as a reinforcing component, the concrete density is greatly increased and thus reduces particle transportation by attenuating the photon or neutron by either scatter or absorption. Thus, the density of 2.35 g/cm 3 for all structures represents a nominal value particle transport and easily modeled. The isotopic composition of ordinary concrete in the MCNP model is taken from Reference 5 and listed in Table 6.1.2-1. FC08513, R0 EC 68969 25

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 26 of 269 Table 6.1.2-1 Isotopic Composition Ordinary Concrete, (National Bureau of Standards- NBS 03) Concrete (Reference 5, Material M1) Model Composition Element (MCNP ID) (Atom Density) H (1001) 0.011914 c (6000) 0.005899 0 {8016) 0.041881 Mg (12000) 0.001408 AI (13027) 0.001892 Si (14000) 0.007311 s {16000} 0.000131 K (19000) 0.000061 Ca (20000) 0.008719 Fe (26000) 0.000280 Density (g/cm 2) 2.35 FC08513, R0 EC 68969 26

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 27 of 269 Material M2 The air utilized in mixture M2. Using air with no moisture (Assumption 4.2) simplifies the model as the water molecules increase the density and reduce particle transport. The isotopic composition of dry air in the MCNP model is taken from Reference 5 and listed in Table 6.1.2-2. Table 6.1.2-2 Isotopic Composition of Dry Air (Reference 5, Material M2) Model Composition Element (MCNP ID) (Atom Density) N (7014) 0.000039 0 (8016) 0.000011 Density (g/cm 2 ) 0.001205 Materials M4 and M5 The interior door of the railroad siding, door 1004-1 A, was modeled as a solid sheet of 20-gauge steel (Assumption 4.3). Thus, the rollup door is comprised of interlocking pieces that are shown on OPPD drawing 232552 (Reference 14.60), which indicates material overlap. Since this overlap increases the effective thickness of the door, modeling the door as a single sheet of carbon steel simplifies the calculation. For simplicity, the exterior door of the railroad siding, door 1004-1C, was modeled as one solid panel (Assumption 4.4), utilizing the panel dimensions as shown on drawing G-576, sheet 20 (Reference 14.65). Since this overlap also increases the effective thickness of the door, modeling of both doors as single components was a conservative approach, in that the objective is to determine the best estimate of dose at the EAB. This allows more particles to transport to the site boundary. Door 1004-1 A construction is of 20-guage steel and Door 1004-1 C utilizes insulation wrapped in 14-guage carbon steel (Assumptions 4.4 and 4.11). The isotopic composition of carbon steel used in the MCNP model is taken from Reference 5 and listed in Table

6. 1.2-3.

Table 6.1.2-3 Isotopic Composition of Simple Carbon Steel (Reference 5, Material M4) Model Composition Element (MCNP ID) (Atom Density) c (6000) 0.001960 Fe (26000) 0.083907 2 Density (g/cm ) 7.82 FC08513, R0 EC 68969 27

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 28 of 269 The specification of insulation for Door 1004-1A is found on drawing G-576, sheet 20 (Reference 14.65) as Trymar 9501, which is a rigid insulation with a chemical form of Polyisocyanurate. The exact composition was not available; therefore, the best suitable material found in Reference 5 was utilized. Polyisocyanurate is found in Reference 5. The composition of Polyisocyanurate used in the MCNP model is taken from Reference 5 and listed in Table 6. 1. 2-4. Table 6.1.2-4 Isotopic Composition of Foam Insulation (Polyisocyanurate) (Reference 5, Material M5) Model Composition Element (MCNP ID) (Atom Density) H (1001) 0.001160 c (6000) 0.001740 N (7014) 0.000232 0 (8016) 0.000232 Density (g/cm 2 ) 0.0482 Spent Fuel potion of Material M3 The source material M3 is comprised of a homogenized mixture of spent fuel , stainless steel storage racks (with boral absorbing panels) and because the spent fuel pool is drained of all water, dry air. Therefore, all components of the fuel assembly, storage racks and air are to be weight averaged over the space taken up by the components in the spent fuel pool. Reference 10 provides a breakdown of fuel assembly hardware by axial region of the fuel assembly. This breakdown is shown in Table 6.1.2-5, which lists by mass, the components of fuel fabricated by AREVA. This table was for a different approach to modeling a fuel assembly, as the materials are sub-divided in to regions Upper End Fitting, Plenum, Active Fuel and Bottom End Fitting . For the purposes of this analysis, the combining of masses for all four (4) regions was utilized in order to calculate the total mass of all spent fuel in the FCS spent fuel pool. The sum of each of the materials utilized which are shown in Table 6.1.2-6. FC08513, R0 EC 68969 28

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 29 of 269 Table 6.1.2-5 EPU Parameters for Fuel Assembly Hardware Parameter Values Quantity Weight Component Material

                                                         /Assembly  (kg/Region)

Upper End Fitting Region

1. Center Guide Tube Assembly

a. Center GT Sleeve 304L 1 0.402

b. Center Guide Tube M5 1 0.062

2. Guide Tube Assembly

a. GT Locking Sleeve 304L 4 1.692

b. Guide Tubes M5 4 0.282

3. UTP Assembly

a. Center Nut Alloy X-750 1 0.533

b. Reaction SQ_ring AlloyX-750 5 0.451

c. Upper Tie Plate CF3 1 2.527 Machining

d. Spring Cup 304L 5 0.610

e. Upper Reaction CF3 1 1.730 Plate Mach

f. Locking Nut Alloy X-750 4 1.287

g. Retaining Sleeve Alloy X-750 4 0.816

4. Fuel Rod Assembly

a. Cladding M5 176 0.046

b. Upper End Caps M5 176 0.634

5. Guide Tube Wear 304L 5 0.196 Sleeves Plenum Region

1. Cage Assembly

a. Center Guide Tube M5 1 0.088

b. Guide Tube M5 4 0.352

c. Spacer Grid M5 1 1.046 Assembly HTP

2. Fuel Rod Assembly

a. Cladding M5 176 3.96

b. Plenum Springs Alloy X-750 176 1.724

3. Guide Tube Wear 304L 5 0.180 Sleeves FC08513, R0 EC 68969 29

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 30 of 269 Active Fuel Region

1. Cage Assembly

a. Spacer HMP Alloy 718 1 0.977 Welded

b. Spacer Capture M5 10 0.052 Rings

c. Center Guide Tube M5 1 1.86

d. Guide Tubes M5 4 7.44

e. Spacer Grid M5 7 7.322 Assembly HTP

2. Fuel Rod Assembly

a. Cladding M5 176 83.4

3. Guide Tube Wear 304L 5 0.273 Sleeves Bottom End Fitting Region

1. Cage Assembly

a. Center Guide Tube M5 1 0.019

b. Guide Tubes M5 4 0.018

c. Guide Tube Lower Zirc-4 4 0.179 End Fitting

2. Alignment Pin Alloy X-750 4 2.234

3. Lower Tie Plate CF3/304L 1 7.095 Machining

4. Dowel Pin 304L 4 0.081

5. Fuel Rod Assembly

a. Lower End Cap M5 176 0.070

b. Cladding M5 176 0.046

5. Guide Tube Cap Screw Alloy X-750 4 0.104 FC08513, R0 EC 68969 30

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 31 of 269 Table 6.1.2-6 Summary of Fuel Assembly Weights Material Weights Zircaloy-Region 304L M5 Alloy X-750 CF3 Alloy 718 4 uo2 (kg) (kg) (kg) (kg) (kg) (kg) (kg) Upper End 2.900 1.024 3.087 4.257 0 0 0 Fittino Reoion Plenum Region 0.180 5.446 1.724 0 0 0 0 Active Fuel 0.273 100.074 0 0 0.977 0 440.859 (1) Region Bottom End 0.081 0.153 2.338 7.095 0 0.179 0 Fitting Region Total per 3.434 106.697 7.149 11.352 0.977 0.179 440.859 Assembly 944 Assemblies 3,241.696 100,721.968 6,748.656 10,716.288 922.288 168.976 416,171.248 (kg) 944 Assemblies 3,241,696 100,721 ,968 6,748,656 10,716,288 922,288 168,976 416,171,248 (grams) (1) U02 Mass is calculated from the U/U02 mass ratio. (enrichment of 3.5%)

                        -Assembly U mass is 388.6 kg.
                        - U/U02 ratio is the atomic mass of U (enrichment 3.5%)- 237.9426516/

U02 atomic mass is 237.9414516 + (2 x 15.9994) = 269.9414516.

                        - U/U02 Ratio is 237.9426516/269.9414516 = 0.88146.
                        -Mass of U02 = 388.6/0.88146 = 440.859 kg I assembly To get a representative mass of all fuel in the spent fuel pool, an approximation of the enrichment of the fuel located in the FCS spent fuel pool is necessary. From OPPD calculation FC08514, Attachment 10.2, which lists all FCS fuel on site by serial number and enrichment, the enrichment values were averaged with a value of 3.509 weight percent U-235. See Attachment "U-235" for the calculation of the average enrichment and calculation for atomic weight for Uranium at 3.5 weight percent U-235.

Since the use of the enrichment is to assist in determining weight, an approximate value is adequate. The average enrichment for the fuel in the spent fuel pool used was 3.50% U-235. This provides a nominal value to calculate Uranium and Uranium Oxide masses and volume occupied by the Uranium Oxide, necessary in determining the source term density. FC08513, R0 EC 68969 31

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 32 of 269 Spent Fuel Racks portion of M3 Adding the mass of the Spent Fuel Racks to the fuel components of the M3 composition was accomplished by first determining the volume of the parts of the Region I and Region II racks. This is necessary because the racks are of differing designs. The Region I racks have a much larger cell-to-cell pitch by design than the Region II racks, allowing for a flux trap between cells and therefore, more air between cells when the spent fuel pool is drained. The two different designs are identified on Holtec Drawings 1000 through 1004 and 1007 (References 14.51 through 14.57). Drawing 1000 (Reference 14.51) shows the overall layout of the spent fuel racks. In addition, as shown in the upper right corner of the drawing, each defined rack has the base dimensions and the number of cells within each rack. The source term volume neglects the area below the fuel assembly alignment pins. At the bottom of the racks, a%" plate of Stainless steel was not included in the weight or volume calculations (References 14.53 and 14.55). The racks, comprised of differing thicknesses of stainless steel panels, are the main structure with boral panels used for neutron absorption with regard to maintaining the configuration subcritical. Region I Volume Drawing 1000 (Reference 14.51) shows Region I racks as A1 and A2 with the larger cell-to-cell pitch. The Region I racks, as shown on drawing 1001 (Reference 14.52) identifies a grouping of 4 rows by 5 columns and each cell is identified by a number 1 through 20. Each 4 by 5 grouping of cells is repeated 4 times defining racks A 1 and A2. Utilizing these uniquely designed 20 cells, the panels are catalogued by their individual parts. Table 6.1.2- 7 and Table 6.1.2-8 are utilized to arrive at the total number of parts and dimension for each cell. The dimensions for each box or panel are found on Holtec drawings 1001 and 1002 (References 14.52 and 14.53). The equations for volumes by part in Table 6.1.2-8 are shown in the notes 1 through 5 after the table. FC08513, R0 EC 68969 32

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 33 of 269 Table 6.1.2-7 Region I Rack Component Dimensions Cell Cell Dimensions (SS Box) Width 8.46" 8.46 Cell Dimensions (SS Box) Height 161" 161.00 Cell Dimensions (SS Box) Thickness 0.075" 0.075 Boral sheet -Width 7.25" 7.25 Boral sheet -Height 128" 128.00 Boral sheet -Thickness 0.075" 0.075 Boundary SS Wrapper sheet -Width 7.25" 7.25 Boundary SS Wrapper sheet -Height 128" 130.00 Boundary SS Wrapper sheet -Thickness 0.075" 0.075 Inner SS Wrapper sheet -Thickness 0.0235" 0.0235 Cell Joints 9a Dimensions - Width 1.542" 1.542 9a Dimensions - Height 8" 4 times 8.00 9a Dimensions - Thickness 0.0897" 0.0897 9b Dimensions - Width 1 II 1.00 9b Dimensions - Height 8" 4 times 8.00 9b Dimensions - Thickness 0.0897" 0.0897 9c Dimensions - Width 1.542" 1.542 9c Dimensions - HeJght 3" 3.00 9c Dimensions - Thickness 0.0897" 0.0897 9d Dimensions - Width 1 1.00 9d Dimensions - Heii)ht 3" 3.00 9d Dimensions - Thickness 0.0897" 0.0897 FC08513, R0 EC 68969 33

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 34 of 269 Table 6.1 .2-8 Region I Rack Component Volume by parts (Holtec Drawings 1001 & 1002) Box Number Box 1 Box2 Box3 Box4 Box5 SS Box Volume (in 408 .618 408.618 408.618 408 .618 408.618 "3) Bora I' Volume (in 278.4 278.4 278.4 278.4 278.4 "3) SS Wrapper Volume (in 88.595 88.595 137.13375 88.595 88.595 "3) Box Number Box6 Box7 Box8 Box9 Box 10 SS Box Volume (in 408.618 408.618 408.618 408.618 408.618 "3) BoraiTM Volume (in 278.4 278.4 278.4 278.4 278.4 "3) SS Wrapper Volume (in 88.595 137.13375 137.13375 88.595 137.13375 "3) Box Number Box 11 Box 12 Box 13 Box 14 Box 15 SS Box Volume (in 408.618 408.618 408.618 408 .618 408.618 "3) Bora I' Volume (in 278.4 278.4 278.4 278.4 278.4 "3) SS Wrapper Volume (in 137.13375 137.13375 137.13375 137.13375 137.13375 "3) Box Number Box 16 Box 17 Box 18 Box 19 Box20 SS Box Volume (in 408.618 408.618 408.618 408.618 408.618 "3) BoraiTM Volume (in 278.4 278.4 278.4 278.4 278.4 "3) FC08513, R0 EC 68969 34

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 35 of 269 SS Wrapper Volume (in 137.13375 185.6725 185.6725 185.6725 185.6725 "3) Joint Type 9a 9b 9c 9d SS corners SS Volume 4.4262 2.8704 0.4150 0.2691 35.0625 (in "3) Number of 20 18 12 12 16 Joints Total Volume by 88.523 51.667 4.979 3.229 561 Joint Type (in "3) Notes: (1) SS Box Volume is calculated by multiplying the cell dimensions listed in (2) Table 6.1 .2-7 above: Box Volume = 4 Sides per Cell

• Width
• Height
• Thickness

                    = 4

• 8.46
• 161
• 0.075

                    = 408 .618 in 3 (3) Boral' Volume is calculated by multiplying the sheet dimensions listed in (4) Table 6.1.2-7:

Boral' Volume= 4 Sheets per Cell* Width* Height* Thickness

                    = 4

• 7.25
• 128
• 0.075

                    = 278.4in 3 (5) SS Wrapper Volume is calculated by multiplying the wrapper dimensions listed in (6) Table 6. 1.2-7:

SS Wrapper Volume= (4,3,2) Interior Wrappers per Cell

• Width
• Height
• Thickness+

(0, 1,2) Exterior Wrappers per Cell *Width

• Height
• Thickness +

                    = (4,3,2)

• 7.25
• 130
• 0.0235 + (0, 1,2)
• 7.25
• 130
• 0.075

                    = 88.595 in 3 (Boxes 1, 2, 4, 5, 6 & 9)
                    = 137.13375 in 3 (Boxes 3, 7, 8, 10, 11-16)
                    = 185.6725 in 3 (Boxes 17-20)

(7) SS Joint Volume 9a-9d are calculated by multiplying the joint dimensions listed in (8) Table 6.1.2-7 above: 9a SS Joint Volumes= Number per joint (4)

• Width
• Height
• Thickness

                    = 4

• 1.542
• 8.0
• 0.0897

                    = 4.4262 in 3 9b SS Joint Volumes = Number per joint (4) *Width

• Height* Thickness FC08513, R0 EC 68969 35

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 36 of 269

                   = 4

• 1.0
• 8.0
• 0.0897

                   = 2.8704 in 3 9c SS Joint Volumes= Number per joint (1) *Width* Height* Thickness
                   = 1

• 1.542
• 3.0
• 0.0897

                   = 0.4150 in 3 9d SS Joint Volumes= Number per joint (1) *Width* Height* Thickness
                   = 1 *1.0

• 3.0
• 0.0897

                   = 0.2691 in 3 (9) SS Corner Volumes are calculated by multiplying the corner dimensions listed in (10)       Table 6.1.2-7:

SS Corner Volume = Width

• Height
• Thickness

                   = 8.5

• 22.0
• 0.1875

                   = 35.0625 in 3 ss Total Volume (Boxes 1-20, Joints 9a-9d            46,109.43
                                           & SS comers) in 3 Boral Total Volume (Boxes 1-20, Joints 9a-9d &           22,272.00 SS comers) in 3 FC08513, R0                                     EC 68969                                     36

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 37 of 269 Region I Mass Using the density of stainless steel 304L of 0.29 lbs./ in 3 , (Reference 10) and multiplying the volume of the stainless steel portion of the racks (Table 5.1.3-8), gives the total mass of stainless steel 304L. This is the same approach for determining mass of the 8oral'. The density of 8oral' is 2.70 grams/cm 3 or 0.09754421bs./ in 3 (Reference 16). Using the volume of 8oral' from Table 6.1.2-8, and the 8oral' density determined mass of the 8oral', Table 6.1.2-9 shows the results of multiplying the volume by the density for both SS 304L and 8oralrM. Table 6.1.2-9 Reg1on

                                   . IRac ks C omponen tM ass Mass                 Mass Total SS Volume                  (lbs)              (grams) x0.29               X 453.59 46,109.43    I      in 3            13371.733         6065284.597 Total 8oral Volume              X 0.09754             X 453.59 22,272.00    J      in 3              2172.500         984979.800 Region II Volume Drawing 1000 (Reference 14.51) shows Region II racks labeled as 81, 82, G1 , G2, C, D, E, F1 and F2. The Region II racks, as shown on drawing 1003 (Reference 14.54), identify a generic rack of6 rows by 7 columns. Drawing 1000 (Reference 14.51) defines each of the racks, which is as shown, provides the various combinations of rows and columns.

The volume for the Region II racks was calculated utilizing a similar approach as the Region I racks, but with the multiple combinations of rows and columns, the process wasn't straight forward because the inner and outer sheath thicknesses varied based on the row/column combinations. The overall goal is to still determined the volume of Stainless Steel and 8oral'. Table 6.1 .2-10 shows each rack by name and overall cell count. Table 6.1.2-10 (Holtec Drawin ~ 1000) Definition of Racks Rack 81 12 X 9 108 Rack 82 12 X 9 108 Rack G1 10 X 9 99 Rack G2 10 X 9 88 RackC 11 X 9 100 Rack D 11 X 8 90 Rack E 10 X 10 90 Rack F1 10 X 12 120 Rack F2 10 X 12 120 FC08513, R0 EC 68969 37

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 38 of 269 Each set of racks was characterized by the parts with which each rack was constructed. Table 6.1 .2-11 and Table 6.1.2-12 are utilized to arrive at the total number of parts and dimensions for each set of racks. The dimensions for each box or panel are found on Holtec drawings 1003 and 1004. Table 6.1.2-11 Region II Racks Cell Dimensions (SS Box) Width 8.46" 8.46 Cell Dimensions (SS Box) Height 161" 161.00 Cell Dimensions (SS Box) Thickness 0.075" 0.075 Boral sheet -Width 7.25" 7.25 Boral sheet -Height 128" 128.00 Boral sheet -Thickness 0.075" 0.075 SS Wrapper sheet -Width 8.00" 8.00 SS Wrapper sheet -Height 130" 130.00 SS Wrapper sheet -Thickness 0.035" 0.035 Shim -Thickness 0.075" 0.075 Shim -width 9" 9.000 Corner -Thickness 3/16" 0.1875 Corner -length 22" 22.000 Corner -Width 8.5" 8.500 Table 6.1.2-12 Rack Component Count and Dimensions (Holtec Drawing 1003 & 1004) AI-B4C ss ss ss Rows Rack Bora I Shims X Sheathing Corners Designation Panels length* Columns Rack B1 12 X 9 195 195 70 8 Rack B2 12 X 9 195 195 70 8 Rack G1 10 X 9 178 178 70 8 Rack G2 10 X 9 157 157 70 8 RackC 11 X 9 180 180 80 8 Rack D 11 X 8 161 161 70 8 Rack E 10 X 10 161 161 70 8 Rack F1 10 X 12 218 218 70 8 Rack F2 10 X 12 218 218 70 8 Panel Total 1663 1663 640 72 Panel Volume 69.6 36.4 0.675

• 35.0625 (inA3)

Total Volume 115744.80 60533.20 432.00 2524.50 FC08513, R0 EC 68969 38

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 39 of 269 Table 6.1.2-13 SS Rack Component Dimensions and Volume (Holtec Drawing 1003 & 1004) East-West Panel North-South Panel Volume Volume Rows North- All East-West South Panel All Panels Panel Rack X Panels Length Length Volume Volume Designation (in"3) (inches) (inches) (in"3) (in"3) (in"3) Columns 103.824 77.868 1253.67 16297.77 940.26 9402.56 Rack 81 12 X 9 103.824 77.868 1253.67 16297.77 940.26 9402.56 Rack 82 12 X 9 95.172 77.868 1149.20 12641.22 940.26 9402.56 Rack G1 10 X 9 86.52 77.868 1044.73 11492.02 940.26 9402.56 Rack G2 10 X 9 95.172 77.868 1149.20 13790.42 940.26 9402.56 Rack C 11 X 9 95.172 69.216 1149.20 13790.42 835.78 7522.05 Rack D 11 X 8 86.52 86.52 1044.73 11492.02 1044.73 11492.02 Rack E 10 X 10 103.824 86.52 1253.67 16297.77 1044.73 11492.02 Rack F1 10 X 12 103.824 86.52 1253.67 16297.77 1044.73 11492.02 Rack F2 10 X 12 Notes: ( 1) SS Box Volume is calculated by multiplying the height of the racks by the length of each row (East-West) or North-South) by the thickness if the sheet and the number of rows of sheets in that rack, using the dimensions from Table 6.1 .2-11 and Table 6.1.2-13 above, for example Rack 81: SS East West Volume= East -West Length* Height* Thickness

                      = 103.824

• 161
• 0.075

                      = 1253.67 in 3
                       = 1253.67

• 13 (Panels per Row/Column, Cell Number 12+1)

                      = 16297.77 in 3 SS North-South Volume = North-South Length

• Height
• Thickness

                      = 77.868

• 161
• 0.075

                      =   940.26 in 3
                       = 940.26

• 10 (Panels per Row/Column, Cell Number 10+1)

                      = 9402.56 in 3 (2) Boral' Volume is calculated by multiplying the sheet dimensions listed in Table 6.1.2-11 by the number of Boral' panels per Rack shown in Table 6.1.2-12.

Boral' Volume = *Width

• Height* Thickness

                      = 7.25

• 128
• 0.075

                      =  69 .6 in 3 FC08513, R0                                          EC 68969                                         39

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 40 of 269

                   = 69.6

• 1663 (Panels per Rack summed, 1663)

                   = 115744.80 in 3 (3) SS Wrapper Volume is calculated by multiplying the sheet dimensions listed in Table 6.1 .2-11 by the number of SS Wrapper panels per Rack shown in Table 6.1.2-12.

SS Wrapper Volume = Width

• Height
• Thickness

                   = 8.0

• 130
• 0.035

                   = 36.4 in 3
                   = 69.6

• 1663 (Panels per Rack summed, 1663)

                   = 60533.20 in 3 (4) SS Shim Volume is calculated by multiplying the wrapper dimensions listed in Table 6.1 .2-11 :

SS Shim Volume= Width* Thickness *Total Length of Shims

                   = 9.0

• 0.075

                   = 0.675 in 2
                   = 0.675

• 640 in (Shims total length, 640")

                   = 432.00 in 3 (5) SS Corner Volumes are calculated by multiplying the corner dimensions listed in Table 6.1 .2-11:

SS Corner Volume = Width

• Height
• Thickness

                   = 8.5

• 22.0
• 0.1875

                   = 35.0625 in 3
                   = 35.0625

• 72 (8 Corners per Rack summed, 72)

                   = 2524.50 in 3 SS Total Volume 128397.19                89010.91 (inA3)

SS from 60533.20 432.00 2524.50 Table 5.1.3-10 Total SS Region II (inA3) 280897.80 Total Boral Region II (inA3) 115744.80 Region II Rack Mass Using the density of stainless steel 304L of 0.29 lbs./ in 3 , (Reference 10) and multiplying the volume of the stainless steel portion of the racks (Table 6.1.2-13), gives the total mass of stainless steel 304L. This is the same approach for determining mass of the Bora I'. The density of Boral' is 2.70 grams /cm 3 or 0.0975442 lbs./ in 3 (Reference 10). Using the volume of Boral' from Table 6.1.2-13Table 6.1.2-14, and the Boral' density, determined mass of the Boral'. Table 6.1.2-14 shows the results of multiplying the volume by the FC08513, R0 EC 68969 40

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 41 of 269 density for both SS 304L and Bora I'. And Table 6. 1.2-15 shows the summary of both the Region I and Region II material masses. Table 6.1.2-14 Region II Racks Component Mass Total SS Mass Mass Volume (lbs) (grams) X 0.29 X 453.59 280897.80 in 3 81,460.363 36,949,606.24 Total Bora IrM Volume X 0.09754 x453.59 115744.80 in 3 11,290.234 5,121 '137.204 Table 6.1.2-15 Total Spent Fuel Racks SS Component Mass Boral' Mass SS Mass (grams) (grams) Region I 985,426.260 6,065,284.60 Region II 5,121,137.204 36,949,606.24 Total 6,106,563.360 43,014,890.84 FC08513, R0 EC 68969 41

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 42 of 269 6.1.3 M3 Material Composition Combining the materials from the fuel and the racks into one composite material was put together by tabulating all of the materials and their constituents by weight. The materials utilized here are Stainless Steei304L, CF3 (Stainless steel). lnconel X-750, lnconel A718, M5 (a proprietary Zircaloy-4 based alloy for cladding and grid material manufactured by Areva), Zircaloy-4, Bora I', and Uranium Oxide (U02). The last component of the mixture is Air. Air is specified in Table 6. 1.2-2, which was determined by subtracting the volumes of all materials summed from the total defined volume of cell 300 (spent fuel). Each of the above materials identified are sub-divided into the nuclides. Table 6.1.3-1 through Table 6.1.3-8 present those materials showing each the nuclide for each of the elemental constituents. These percentages were utilized in homogenizing the elements for material group M3. Table 6.1.3-1 SS 304L (ASTM A276, References 7 & 10) Nominal Composition Value Used Element MCNP Nuclide 10 (%) (%) C (Carbon) 6000 0.03 (maximum) 0.03 Co (Cobalt) 27000 0.03 (maximum) 0.03 Cr (Chromium) 24000 18.00 - 20.00 19.00 Fe (Iron) 26000 Balance (2) 67.865 Mn (Manganese) 25000 2.0 (maximum) 2.00 Ni (Nickel) 28000 8.00 - 12.0 (1) 10.00 P (Phosphorus) 15000 0.045 0.045 Si (Silicone) 14000 1.00 1.00 S (Sulfur) 16000 0.03 0.03 Notes: (1) The average over the elemental nominal composition range is used for Cr and Ni. Thus, (18 +20)/2) = 19% for Cr and (8.00 + 12)/2) =1 0.0% for Ni. (2) The percent of iron (Fe) is the difference between 100% and the sum of the percentages of the other elemental constituents (i.e., C, Co. Cr, Mn, Ni, P, Sand Si). The percent of iron (Fe) is: (1 00 10 - 2.0 - 1.0 - 0.045 - 0.03 - 0.03 - 0.03) = 67.865%. FC08513, R0 EC 68969 42

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 43 of 269 Table 6.1.3-2 CF3- SS (ASTM A743, References 7 & 10) Nominal Composition Value Used Element MCNP Nuclide ID (%) (%) C(Carbon) 6000 0.03 _(maximum) 0.03 Co (Cobalt) 27000 0.03 (maximum) 0.03 Cr (Chromium) 24000 18.00 - 20.00 (1) 19.00 Fe (Iron) 26000 Balance (2) 67.25 Mn Manganese) 25000 1.50 (maximum) 1.50 N (Nitrogen) 7000 0.11 (maximum) 0.11 Ni (Nickel) 28000 8.00 - 12.0 (1) 10.00 P (Phosphorus) 15000 0.04 (maximum) 0.04 S (Sulfur) 16000 0.04 (maximum) 0.04 Si (Silicone) 14000 2.0 (maximum) 2.00 Notes: (1) The average over the elemental nominal composition range is used for Cr and Ni. Thus, (18 +20)/2)= 19% for Cr and (8.00 + 12)/2) =10.0% for Ni. (2) The percent of iron (Fe) is the difference between 100% and the sum of the percentages of the other elemental constituents (i.e., C, Co. Cr, Mn, Ni, P, S and Si). The percent of iron (Fe) is: (1 00 10

       - 2.0- 1.50- 0.04 -0.04- 0.03- 0.03- 0.11) = 67.25%.

Table 6.1.3-3 Alloy X-750 (lnconel, References 7 & 10) Element MCNP Nuclide ID Nominal Composition(%) Value Used(%) AI ( Aluminum) 13027 0.40 - 1.00 0.70 C (Carbon) 6000 0.08 (maximum) 0.08 Co (Cobalt) 27000 0.03 (maximum) (1) 0.03 Cr (Chromium) 24000 14.00- 17.00 15.50 Cu (Copper) 29000 0.50 (maximum) 0.500 Fe (Iron) 26000 5.00 - 9.00 7.000 Mn (Manganese) 25000 1.00 (maximum) 1.00 Niobium (Nb) 41093 0.70 - 1.20 (2) 0.950 Ni (Nickel) 28000 Balance GT 70% (1) 71.18 S (Sulfur) 16000 0.01 (maximum) 0.010 Si (Silicone) 14000 0.50 (maximum) 0.500 Ta (Tantalum) 73000 0.05 (maximum) (2) 0.05 Ti (Titanium) 22000 2.25 - 2.75 2.50 Notes: (1) Alloy X-750 in Table 7-1 of Reference 10 has an element defined as (Ni +Co) with a weight percent of 70.0 (min). Since Table 7-1 of Reference 10 includes a separate Co entry for alloy X-750, the element defined as (Ni +Co) will be treated as Ni. FC08513, R0 EC 68969 43

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 44 of 269 (2) Alloy X-750 also has an element defined as (Nb + Ta) with a weight percent of 0.7 to 1.2. Table 7-1 includes separate Nb and Ta elements for alloy X-750 with weight percents of 0.70 to 1.20 for Nb and a weight percent of 0.05 (max) for Ta. In this calculation, the elemental weight percent of Ta in alloy X-750 is specified as 0.05 and the weight percent of Nb in alloy X-750 is specified as the average over the range of 0. 7 to 1.2, i.e., 0.95. Table 6.1.3-4 Alloy -718 (lnconel, References 7 & 10) MCNP Nuclide Nominal Composition(%) Element ID Value Used AI (Aluminum) 13027 0.20 - 0.80 0.500 B (Boron) 5000 0.006 (maximum) 0.006 B-10 (Boron) 5010 Note (1) 0.0011 B-11 (Boron) 5011 Note (1) 0.0049 C (Carbon) 6000 0.08 (maximum) 0.08 Co (Cobalt) 27000 0.03 (maximum) 0.03 Cr (Chromium) 24000 17.00-20.00 19.00 Cu (Copper) 29000 0.30 (maximum) 0.300 Fe (Iron) 26000 Balance (3) 17.729 Mn (Manganese) 25000 0.35 (maximum) 0.35 Mo (Molybdenum) 42000 2.80 - 3.30 3.05 Niobium (Nb) 41093 4.75 - 5.50 5.125 Ni (Nickel) 28000 50.0 - 55.0 52.50 P (Phosphorus) 15000 0.015 (maximum) 0.015 S (Sulfur) 16000 0.015 (maximum) 0.015 Si (Silicone) 14000 0.35 (maximum) 0.350 Ta (Tantalum) 73000 0.05 (maximum) 0.05 Ti (Titanium) 22000 0.65 - 1.15 0.90 Notes: (1) In order to run cases utilizing neutrons, specific nuclides must have the complete ZAID for each material. This is due to a neutron specific library is required for neutron cases. Therefore, all cases will utilize these material specific nuclides. Basic Boron is not in the neutron library, therefore B-10 and B-11 which are in the neutron library and are specified based upon the natural abundance The weight percent for B-1 0 and B-11 is determined as follows using the atomic percentages and isotopic weights (Reference 18): Element A/0 AMU Weight Wt.% B-10 (Boron) 19.61 10.0129388 1.963537 18.158% B-11 (Boron) 80.39 11.0093053 8.850381 81.842% 10.813918 Boron is 0.006%, therefore B-1 0 = 0.18158

• 0.006 = 0.0011

                  =

B-11 0.81842

• 0.006 = 0.0049 FC08513, R0 EC 68969 44

FC08513, R0 EC 68969 45 Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 46 of 269 (1) Assuming natural boron is 18.3% B-10 by weight, the B11 density is: 0.079265 * (1-0.183)/0.183 = 0.353877 g/cm 3 This gives a total natural boron density of 0.079265 + 0.353877 = 0.433143 g/cm 3 Since there is one carbon atom for four boron atoms, and giving the atomic weights of natural boron of 10.811 and carbon 12.011, the density of carbon is: [12.011/(4*1 0.811 )]

• 0.433143 = 0.120305 g/cm 3 Since the boral is largely aluminum, it will be assumed that it has an overall density of aluminum of 2.70 g/cm 3 . The density of aluminum then is:

2.70-0.433143-0.120305 = 2.146552 g/cm 3 Table 6.1.3-8 Uranium Oxide (UOz - 3.5 wt% U-235) Element MCNP Nuclide 10 Nominal Composition(%) Value Used(%) 0 (Oxygen) 8016 N/A 11.8539 U-234 92234 N/A 0.0275 U-235 92234 N/A 3.0851 (*) U-236 92234 N/A 0.0142 U-238 92234 N/A 85.0193 Note:

• The values used are shown for UOz. The nominal value of 3.5 weight percent is reduced when the weight of the 2 Oxygen atoms is included with the mass calculation.

FC08513, R0 EC 68969 46

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 47 of 269 Table 6.1.3-9 Material M3 Percent by Nuclide (Materials: SS-304L, CF3 & Alloy718) MCNP Element SS-304L CF3 Alloy 718 Nuclide ID Density (g/cm 3) 8.027 8.027 8.221 Mass (g) 46,256,586.84 10,716,288.00 922,288.00 AI 13027 - - 0.5000 Ar 18000 - - - B

• 5000 - - 0.0060 B-10 5010 - - 0.0011 B-11 5011 - - 0.0049 c 6000 0.03 0.03 0.080 Cl 17000 - - -

Co 27059 0.03 0.03 0.030 Cr 24000 19.0 19.0 19.000 Cu 29000 - - 0.30 Fe 26000 67.865 67.25 17.729 Mn 25055 2.00 1.50 0.350 Mo 42000 - - 3.050 N 7014 - 0.1100 - Na 11023 - - - Nb 41093 - - 5.1250 Ni 28000 10.00 10.00 52.500 0 8016 - - - p 15031 0.045 0.040 0.015 Pb 82000 - - - s 16000 0.03 0.040 0.015 Si 14000 1.00 2.00 0.350 Sn 50000 - - - Ta 73181 - - 0.050 Ti 22000 - - 0.9000 U-234 92234 - - - U-235 92235 - - - U-236 92236 - - - U-238 92238 - - - v 23000 - - - Zr 40000 - - - FC08513, R0 EC 68969 47

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Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 49 of 269 Table 6.1.3-11 Material M3 Percent by Nuclide (Materials: Boral', U02 & Air) MCNP Element Bora I' U02 Air Nuclide ID 388.6 (KgU) Density (g/cm 3) 2.700 0.001205 10.3965 Mass (g) 6,106,563.36 416,171 ,245.28 194,459.99 AI 13027 79.502 - - Ar 18000 - 1.2827 B

• 5000 - -

B-10 5010 2.936 - - B-11 5011 13.106 - - c 6000 4.456 - 0.0124 Cl 17000 - - - Co 27059 - - - Cr 24000 - - - Cu 29000 - - - Fe 26000 - - - Mn 25055 - - - Mo 42000 - - - N 7014 - - 75.5268 Na 11023 - - - Nb 41093 - - - Ni 28000 - - - 0 8016 - 11.8539 23.1781 p 15031 - - - Pb 82000 - - - s 16000 - - - Si 14000 - - - Sn 50000 - - - Ta 73181 - - - Ti 22000 - - - U-234 92234 - 0.0275 - U-235 92235 - 3.085 - U-236 92236 - 0.0142 - U-238 92238 - 85.0193 - v 23000 - - - Zr 40000 - - - FC08513, R0 EC 68969 49

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 50 of 269 Defining the mixture, the mass of each substance is multiplied by the specified percentage of each nuclide within that substance. The resulting values are then divided by the total mass of all materials combined. The total of the masses is shown in Table 6.1 .3-12. Table 6.1.3-12 Total Mass M3 Materials Material Mass (grams) SS-304L 46,256,586.840 CF3 10,716,288.000 Alloy-X-750 6,748,656.000 Alloy 718 922,288.000 M5 100,721,968.000 Zircaloy-4 168,976.000 Bora I 6,106,563.364 U02 416,171,245.300 Air 194,459.990 Total Mass 588,007,031.474 (grams) With the total weight determined, the individual nuclides weights are calculated for each material. These values are summed by nuclide and divided by the total mass. This fraction is input into MCNP as a weight percent. Table 6.1 .3-13 and Table 6.1.3-14 show the results of the calculation by nuclide for each substance and on Table 6.1.3-14 far right column is the total by nuclide. Figure 6.1.3-1 shows the MNCP input for material M3. FC08513, R0 EC 68969 50

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Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 52 of 269 Table 6.1.3-14 Material M3 Weight Fraction of the Total Weight by Nuclide (Materials: Zicaloy-4, Boral', UOz, Air & M3 Composite) MCNP Material M3 Element Zircaloy-4 Bora I U02 Air Nuclide ID Composite AI 13027 8.2564E-03 8.344615E-03 Ar 18000 4.2420E-06 4.242021 E-06 B

• 5000 O.OOOOOOE+OO B-10 5010 3.0491E-04 3.049262E-04 B-11 5011 1.3611 E-03 1.361160E-03 c 6000 4.6276E-04 4.1008E-08 5.023089E-04 Cl 17000 5.7474E-09 5.747414E-09 Co 27059 3.298115E-05 Cr 24000 3.3048E-07 2.048669E-02 Cu 29000 6.209134E-05 Fe 26000 6.0348E-07 6.672541 E-02 Mn 25055 1.966967E-03 Mo 42000 4. 783920E-05 N 7014 2.4977E-04 2.698222E-04 Na 11023 5.7474E-09 5.747414E-09 Nb 41093 1.902357E-03 Ni 28000 2.8737E-07 1.868235E-02 0 8016 3.5921E-07 8.3898E-02 7.6652E-05 8.421467E-02 p 15031 4.292521 E-05 Pb 82000 3.7358E-08 3.735819E-08 s 16000 3.621267E-05 Si 14000 1.214038E-03 Sn 50000 4.1669E-06 4 .166875E-06 Ta 73181 6.522834E-06 Ti 22000 3.010457E-04 U-234 92234 1.9464E-04 1.946356E-04 U-235 92235 2.1835E-02 2.183528E-02 U-236 92236 1.0050E-04 1.005027E-04 U-238 92238 6.0174E-01 6.017375E-01 v 23000 1.4369E-08 1.436854E-08 Zr 40000 2.8156E-04 1.696187E-01 FC08513, R0 EC 68969 52

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 53 of 269 Figure 6.1.3-1 MCNP Material M3 Weight Fraction by Nuclide c c Fuel Region Mixture = 2.568032 g/cm3 c M3 13027 -8.344615E-03 $ Al Weight Percent 18000 -4.242021E-06 $ Ar Weight Percent 5010 -3.049262E-04 $ B10 Weight Percent 5011 -1.361160E-03 $ B11 Weight Percent 6000 -5.023089E-04 $ c Weight Percent 17000 -5.747414E-09 $ Cl Weight Percent 27059 -3.298115E-05 $ Co Weight Percent 24000 -2.048669E-02 $ Cr Weight Percent 29000 -6.209134E-05 $ Cu Weight Percent 26000 -6.672541E-02 $ Fe Weight Percent 25055 -1.966967E-03 $ Mn Weight Percent 42000 -4.783920E-05 $ Mo Weight Percent 7014 -2.698222E-04 $ N Weight Percent 11023 -5.747414E-09 $ Na Weight Percent 41093 -1.902357E-03 $ Nb Weight Percent 28000 -1.868235E-02 $ Ni Weight Percent 8016 -8.421467E-02 $ 0 Weight Percent 15031 -4.292521E-05 $ p Weight Percent 82000 -3.735819E-08 $ Pb Weight Percent 16000 -3.621267E-05 $ s Weight Percent 14000 -1.214038E-03 $ Si Weight Percent 50000 -4.166875E-06 $ Sn Weight Percent 73181 -6.522834E-06 $ Ta Weight Percent 22000 -3.010457E-04 $ Ti Weight Percent 92234 -1. 946356E-04 $ U-234 Weight Percent 92235 -2.183528E - 02 $ U-235 Weight Percent 92236 -1. 005027E-04 $ U-236 Weight Percent 92238 -6.017375E-01 $ U-238 Weight Percent 23000 -1.436854E-08 $ v Weight Percent 40000 -1.696187E-01 $ Zr Weight Percent c FC08513, R0 EC 68969 53

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 54 of 269 6.1.4 Source Term The ORIGEN-ARP/ORIGEN-S computer programs within the SCALE 6.1 computer code package were used to generate the fuel assembly gamma and neutron source terms for the FCS Combustion Engineering (CE) 14 x 14 type fuel assemblies. This work was documented in FC08514, and a standard CE 14 x 14-assembly type was used versus building a specific AREVA fuel type. The fuel design used in the FCS reactor is a standard CE 14 x 14-fuel type even when manufactured by AREVA or Westinghouse or Exxon. Benchmark calculations for the source term were produced to check against explicit AREVA 14 x 14 ORIGEN output and the source term generation was found to be well within statistical acceptance and documented in FC08514. As noted in the referenced calculation radial or axial peaking factors were not applied to the source term since the assemblies were modeled with plant specific burnups from GARDEL/CECOR online measurement processes. The ORIGEN-ARP/ORIGEN-S calculations were performed to generate a gamma and neutron source for a time point at eighteen months after reactor shutdown. A fourteen-month source term was then scaled from further ORIGEN runs on core offload fuel batches. These two time points are considered relevant for the dose analyses. The explicit operating parameters used to generate the source term can be found in FC08514 and will not be discussed further in this calculation. The source term generated for gamma, (photons) included light elements, actinides and fission product calculations. The gamma radiation that results from neutron interactions (i.e., neutron/gamma reactions) was not included and that discussion is justified in FC08514, and has precedent with the SCE response regarding source term generation. Additionally, FC07586 provided justification for not including the neutron/gamma reactions. Page 28 of FC07586 notes that both the neutron radiation source term in a spent fuel assembly and gamma radiation that results from neutron/gamma reactions, are orders of magnitude less than energy deposition potential from a spent fuel assembly gamma radiation source term. Therefore, FC07586 noted that the spent fuel assembly neutron radiation source terms and gamma radiation that results from neutron/gamma reactions were not further pursued for spent fuel pool concrete gamma heating. The energy deposition and energy strength has direct correlation to dose as well. This calculation will include neutron source term for completeness related to dose consequences. It will not however, include any neutron/gamma interactions that could potentially produce further gamma sources. By far the total gamma source strengths due to fission products, actinides, and light element activation products in the spent fuel are the dominant radiation source term as shown by the source term total strength. The ORIGEN-ARP/ORIGEN-S computer codes were run to obtain time dependent gamma radiation source terms for 18 standard gamma energy groups and to obtain neutron radiation source terms for 44 standard neutron energy groups. The 18 group structure ORIGEN2 was used for the gamma calculations. The 44-group structure ENDF5 was used for the neutron calculations. Details about those calculations are found in FC08514. Gamma and neutron energy bin values (MeV) are provided in the tables noted below. There are 19 gamma energy bin values in descending order and 45 neutron energy bin values in descending order. FC08513, R0 EC 68969 54

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 55 of 269 The radiation source term within the active fuel region of an assembly and stored in the spent fuel racks is homogenized over the active fuel region in the spent fuel pool. The source term generated is uniformly distributed over the surface of the fuel region of the spent fuel pool. No axial power distribution was applied to the source term. This is considered conservative as some fuel cycles have axial power distributions that have a large fraction of power in bottom segments of a spent fuel assembly and as such, the source would be weighted towards bottom of the fuel. Two source terms were generated, one for a fourteen-month post shutdown event, and one for an eighteen month. The source terms generated for these time points were for neutron and photons (gammas). The source term was generated by energy group so that it can be readily input into MCNP calculations. No adjustment to this source term is required as discussed above. The following tables are used to calculate the dose rates at fourteen and eighteen months after reactor shutdown. The first set of tables are for gamma dose calculations and represent the photon source term from the fuel in the spent fuel pool. The energy groups are listed by the output number of the group. The table below shows the energy boundaries for each energy group, so for example output group 1 has an energy bin or boundary of 0 to 0.02 MeV. Output group 18 has an energy bin or boundary of 8 to 11 MeV. Table 6.1.4-1 ORIGEN 0 U[pu t t group ORIGEN2 st ruet ure Gamma photons/sec/basis II ORIGEN2 Group Boundaries MeV ORIGEN grp max gamma MeV 19 MeV eV Output Group 18 0.02 20000 1 O.OOE+OO 2.00E-02 17 0.03 30000 2 2.00E-02 3.00E-02 16 0.045 45000 3 3.00E-02 4.50E-02 15 0.07 70000 4 4.50E-02 7.00E-02 14 0.1 100000 5 7.00E-02 1.00E-01 13 0.15 150000 6 1.00E-01 1.50E-01 12 0.3 300000 7 1.50E-01 3.00E-01 11 0.45 450000 8 3.00E-01 4.50E-01 10 0.7 700000 9 4.50E-01 7.00E-01 9 1 1000000 10 7.00E-01 1.00E+OO 8 1.5 1500000 11 1.00E+OO 1.50E+OO 7 2 2000000 12 1.50E+OO 2.00E+OO 6 2.5 2500000 13 2.00E+OO 2.50E+OO 5 3 3000000 14 2.50E+OO 3.00E+OO 4 4 4000000 15 3.00E+OO 4.00E+OO 3 6 6000000 16 4.00E+OO 6.00E+OO 2 8 8000000 17 6.00E+OO 8.00E+OO 1 11 11000000 18 8.00E+OO 1.10E+01 FC08513, R0 EC 68969 55

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 56 of 269 Table 6.1.4-2 ORIGEN ENDF5 44 Group Max Energy (Mev) eV Output 45 1.00E-11 1.00E-05 44 3.00E-09 3.00E-03 1 43 7.50E-09 7.50E-03 2 42 1.00E-08 1.00E-02 3 41 2.53E-08 2.53E-02 4 40 3.00E-08 3.00E-02 5 39 4.00E-08 4.00E-02 6 38 5.00E-08 5.00E-02 7 37 7.00E-08 7.00E-02 8 36 1.00E-07 1.00E-01 9 35 1.50E-07 1.50E-01 10 34 2.00E-07 2.00E-01 11 33 2.25E-07 2.25E-01 12 32 2.50E-07 2.50E-01 13 31 2.75E-07 2.75E-01 14 30 3.25E-07 3.25E-01 15 29 3.50E-07 3.50E-01 16 28 3.75E-07 3.75E-01 17 27 4.00E-07 4.00E-01 18 26 6.25E-07 6.25E-01 19 25 1.00E-06 1.00E+OO 20 24 1.77E-06 1.77E+OO 21 23 3.00E-06 3.00E+OO 22 22 4.75E-06 4.75E+OO 23 21 6.00E-06 6.00E+OO 24 20 8.10E-06 8.10E+OO 25 19 1.00E-05 1.00E+01 26 18 3.00E-05 3.00E+01 27 17 1.00E-04 1.00E+02 28 16 5.50E-04 5.50E+02 29 15 3.00E-03 3.00E+03 30 14 1.70E-02 1.70E+04 31 13 2.50E-02 2.50E+04 32 12 1.00E-01 1.00E+05 33 11 4.00E-01 4.00E+05 34 10 9.00E-01 9.00E+05 35 9 1.40E+OO 1.40E+06 36 8 1.85E+OO 1.85E+06 37 7 2.35E+OO 2.35E+06 38 6 2.48E+OO 2.48E+06 39 5 3.00E+OO 3.00E+06 40 4 4.80E+OO 4.80E+06 41 3 6.43E+OO 6.43E+06 42 2 8.19E+OO 8.19E+06 43 1 2.00E+01 2.00E+07 44 FC08513, R0 EC 68969 56

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 57 of 269 Table 6.1.4-1 shows that output group 9 has an energy boundary of 0.45 to 0.70 MeV, and group 10 has an energy boundary of 0. 70 to 1.00 MeV. Since majority of fission products and some fuel structural activation products produce gammas within this range, they are distinctly dose relevant for personnel and the source term in these energy groups would be expected to dominate the dose calculation. The output group structure with energy boundaries is important, as they need to correlate to the MCNP input decks as well. The actual table of total photons gamma spectrum as a function of energy output group is provided next. The far right column represents the total core offload and old fuel source term that must be input into MCNP for the eighteen month calculations. Likewise, a similar total gamma spectrum was scaled for the fourteen-month calculation. Table 6.1.4-3 18 Energy Group Gamma Source Term (14 Month Decay Scaled) Total Gamma Spectrum Energy Core Offload and Old Fuel Output Group photons/s/mev 1 2.02E+18 2 4.25E+17 3 5.14E+17 4 3.63E+17 5 2.53E+17 6 3.08E+17 7 2.33E+17 8 1.15E+17 9 2.46E+18 10 4.18E+17 11 7.48E+16 12 5.63E+15 13 5.65E+15 14 1.10E+14 15 9.97E+12 16 1.26E+10 17 1.45E+09 18 1.67E+08 Total 7.20E+18 FC08513, R0 EC 68969 57

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 58 of 269 Table 6.1.4-4 18 Energy Group Gamma Source Term (18 Month Decay) Energy Old Fuel Core Offload Total Gamma Output Group 811 Assemblies 133 Assemblies Spectrum (Core Offload and Old Fuel) photons/s/mev photons/s/mev photons/s/mev 1 8.40E+17 8.29E+17 1.67E+18 2 1.71E+17 1.80E+17 3.51E+17 3 2.21E+17 2.04E+17 4.25E+17 4 1.49E+17 1.51E+17 3.00E+17 5 9.56E+16 1.14E+17 2.10E+17 6 9.78E+16 1.57E+17 2.55E+17 7 8.45E+16 1.08E+17 1.93E+17 8 3.84E+16 5.67E+16 9.51E+16 9 1.43E+18 6.07E+17 2.04E+18 10 1.30E+17 2.16E+17 3.46E+17 11 3.08E+16 3.11E+16 6.19E+16 12 1.40E+15 3.25E+15 4.65E+15 13 4.29E+14 4.24E+15 4.67E+15 14 1.38E+13 7.72E+13 9.10E+13 15 1.28E+12 6.96E+12 8.24E+12 16 8.96E+09 1.46E+09 1.04E+10 17 1.03E+09 1.68E+08 1.20E+09 18 1.19E+08 1.93E+07 1.38E+08 Total 3.29E+18 2.66E+18 5.95E+18 The next set of tables represent the neutron source term for the spent fuel pool at fourteen and eighteen months after reactor shutdown by energy group. The first table shows the neutron energy output group by energy boundaries such that the same binning process has to be preserved for the MCNP calculations. That is the strength of the source for that energy group (and the energy boundary itself) has to be specified so the code can track the particle interactions by energy. For example, the far right hand column represents the Group Output number, the two middle columns represent the energy boundary ranges, the far left column represents the ORIGEN Maximum Neutron Energy group structure. For example, the far left column value of 1 represents the ORIGEN energy boundary from 2E+1 to 2E+ 7 eV, but it's output group is number 44. The energy groups must remain consistent for the dose calculations . That is the number of neutrons for a particular output group must align with the energy for that group structure. FC08513, R0 EC 68969 58

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 59 of 269 Table 6.1.4-5 44 Energy Group Neutron Source Term (14 Month Decay Scaled) Total Energy Neutron Spectrum Out~ut Grou2_ (Core Offload and Old Fuel) neutron s/s/mev 1 3.63E-03 2 2.61E-03 3 2.06E-03 4 6.06E-03 5 2.51E-03 6 8.78E-03 7 6.28E-03 8 1.35E-02 9 2.24E-02 10 4.27E-02 11 5.05E-02 12 3.08E-02 13 4.02E-02 14 3.37E-02 15 5.88E+OO 16 2.03E+OO 17 2.10E+OO 18 2.18E+OO 19 1.92E+01 20 4.57E+01 21 1.28E+02 22 2.53E+02 23 4.69E+02 24 4.05E+02 25 7.62E+02 26 7.81E+02 27 1.21E+04 28 7.66E+04 29 1.09E+06 30 1.39E+07 31 1.87E+08 32 1.58E+08 33 2.45E+09 34 1.77E+10 35 3.87E+10 36 3.86E+10 37 3.09E+10 38 2.91E+10 39 6.31E+09 40 2.25E+10 41 4.09E+10 42 1.16E+10 43 3.69E+09 44 1.27E+09 Total 2.44E+11 FC08513, R0 EC 68969 59

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 60 of 269 Table 6.1.4-6 44 Energy Group Neutron Source Term (18 Month Decay) Total Gamma Spectrum Energy Old Fuel Core Offload (Core Offload and Old Output Group 811 Assemblies 133 Assemblies Fuel) 1 3.17E-03 3.36E-04 3.51E-03 2 2.28E-03 2.42E-04 2.52E-03 3 1.79E-03 1.97E-04 1.99E-03 4 5.30E-03 5.58E-04 5.85E-03 5 2.18E-03 2.42E-04 2.42E-03 6 7.58E-03 9.06E-04 8.48E-03 7 5.45E-03 6.19E-04 6.07E-03 8 1.17E-02 1.34E-03 1.31E-02 9 1.94E-02 2.22E-03 2.16E-02 10 3.70E-02 4.25E-03 4.12E-02 11 4.35E-02 5.24E-03 4.88E-02 12 2.65E-02 3.23E-03 2.97E-02 13 3.45E-02 4.28E-03 3.88E-02 14 2.91E-02 3.54E-03 3.26E-02 15 4.90E+OO 7.81 E-01 5.68E+OO 16 1.69E+OO 2.73E-01 1.96E+OO 17 1.73E+OO 2.94E-01 2.03E+OO 18 1.81E+OO 2.94E-01 2.11E+OO 19 1.60E+01 2.59E+OO 1.85E+01 20 3.80E+01 6.16E+OO 4.42E+01 21 1.06E+02 1.73E+01 1.24E+02 22 2.10E+02 3.41E+01 2.44E+02 23 3.90E+02 6.32E+01 4.53E+02 24 3.37E+02 5.45E+01 3.91E+02 25 6.33E+02 1.03E+02 7.36E+02 26 6.49E+02 1.05E+02 7.54E+02 27 1.01E+04 1.63E+03 1.17E+04 28 6.37E+04 1.03E+04 7.40E+04 29 9.06E+05 1.47E+05 1.05E+06 30 1.15E+07 1.87E+06 1.34E+07 31 1.56E+08 2.52E+07 1.81E+08 32 1.31E+08 2.12E+07 1.52E+08 33 2.04E+09 3.31E+08 2.37E+09 34 1.47E+10 2.39E+09 1.71E+10 35 3.21E+10 5.21E+09 3.73E+10 36 3.21E+10 5.21E+09 3.73E+10 37 2.57E+10 4.18E+09 2.99E+10 38 2.42E+10 3.94E+09 2.82E+10 39 5.24E+09 8.53E+08 6.10E+09 40 1.87E+10 3.05E+09 2.18E+10 41 3.40E+10 5.56E+09 3.95E+10 42 9.66E+09 1.57E+09 1.12E+10 43 3.07E+09 4.97E+08 3.57E+09 44 1.06E+09 1.71E+08 1.23E+09 Total 2.03E+11 3.30E+10 2.36E+11 FC08513, R0 EC 68969 60

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 61 of 269 As can be seen the total number of neutrons is several orders of magnitude less than the gamma source term and would not be expected to contribute to the overall dose calculation significantly. The total source term for gammas and neutrons for the MCNP model is similar to that which was calculated for the SCE source term (Reference 12) and comparable dose predictions would be expected. The gamma and neutron source terms noted above in the previous tables were input directly into the MCNP files for dose calculations related to the control room and exclusion area boundary. 6.1.5 Source Term Output Dose Important Nuclides as a Function of Cooling Time The following output, Figure 6. 1.5-1 , from ORIGEN-ARP (FCS DD1 18 month Curies.out) for the highest enrichment assembly that is being oftloaded from the Cycle 28 core was run to show the measure of curies for the important dose nuclides. The nuclides that were plotted are those nuclides that have relatively high dose conversion factors for external gamma. These nuclides were tracked through the three-cycle fuel irradiation , and then subsequent decay corrected in the ORIGEN calculation process. The output shown is a function of time after the assembly has been discharged from the core (DD1 sub-batch assembly). The output reported is in curies for a single assembly discharged. The units for cooling time are in years. Figure 6.1.5-1 ORIGEN Output 3 Cycles of Operation of Batch DD Cycle 3 Down - DO Assembly units of measure: curie s units of time(lst column): time (years} nuclides t i me br82 br83 il29 il30 i131 il32 il33 il34 il35 kr83m kr8 5 kr85m l.OOOE - 03 1.4 3 1E +03 2.695E+03 1.844E - 02 7.273E+03 3.088E+05 4 . 244E+05 4 . 791E+0 5 2 . 340E+03 2.385E+05 7. 2 42E+03 5.869E+03 1.645E+04

3. 000E-03 1.014E+03 1.703E+01 1.844E-02 2 . 721E+03 2.926E+05 3.629E+05 2.67 1 E+05 2.540E-03 3.751E+04 6 . 635E+01 5 . 869E+03 1.091E+03 2.469E-02 2.425E+01 O.OOOE+OO 1 . 846E-02 6 . 362E - 0 2 1.510E+05 6.726E+04 4.729E+02 O.OOOE+OO 7 . 273E-05 O. OOOE+OO 5 . 860E+03 1.824E-10 7 . 408E-02 4 . 933E - 03 O.OOOE+ OO 1 . 848E - 02 l.BlBE - 12 3.190E+04 1.449E+03 2.565E - 04 O. OOOE+OO 1 . 060E - 24 O. OOOE+OO 5. 842E+03 O.OOOE+OO 2 . 222E - 01 4 . 172E-14 O.OOOE+OO 1.852E-02 O.OOOE+OO 3.005E+02 1.451E-02 4.126E - 23 O.OOOE+OO O.OOOE+OO O.OOOE+OO 5 . 786E+03 O.OOOE+OO 6 . 671E-01 O.OOOE+OO O.OOOE+OO 1.854E-02 O.OOOE+OO 2.480E-04 1 . 410E - 17 O.OOOE+OO O. OOOE+OO O.OOOE+OO O.OOOE+OO 5. 6 2 2E+03 O.OOOE+OO l.SOOE+OO O. OOOE+OO O. OOOE+OO 1 . 854E - 02 O.OOOE+OO 1.004E-15 O.OOOE+OO O. OOOE+OO O. OOOE+OO O. OOOE+OO O. OOOE+OO 5.327E+03 O.OOOE+OO FC08513, R0 EC 68969 61

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 62 of 269 Figure 6.1.5-1 (Cont'd) ORIGEN Output 3 Cycles of Operation of Batch DD units o f measure : curies units of time(lst column): t i me (years) nuclides time kr87 kr88 xel3lm xel33 xel33m xe 1 35 l.OOOE-03 1.034E+03 1 . 885E+04 4.660E+03 6.240E+05 1.945E+04 2.639E +05 3.000E-03 7.333E-02 2.611 E+02 4.602E+03 6 . 006E+05 1.757E+04 1.354E+05 2.469E-02 O.OOOE+OO 1.827 E - 1 8 3.743E+03 2.327E+05 1.833E+03 1 . 265E-01 7.408E-02 O.OOO E+OO O.OOOE+OO 1.782E+03 2.155E+04 6.104E+OO 6.982E - 16 2.222E-01 O.OOOE+OO O.OOOE+OO 1 . 003E+02 1 .687E+01 2.229E-07 O.OOOE+OO 6.67 1E-01 O.OOOE+OO O.OOOE+OO 8.307E - 03 7.912E - 09 O.OOOE+OO O.OOOE+OO 1.500E+00 O.OOOE+OO O.OOOE+OO 1 . 669E-10 2 . 679E-26 O.OOOE+OO O.OOOE+OO The output shows that at 0.001 years after discharge (-9 hours) that the curie content for the nuclides plotted is substantial, total curie count is 2.185E7 curies, and each nuclide shown has relevant measure related to potential gamma dose from internal dose. After 1.5 years of cooling the output for this assembly shows that the only nuclide that has relevant curie measure is Kr85 at 5.327E3 Curies. After 1.5 years of decay, the fuel assembly activity available for release is predominately Kr85 (noble gas). The potential release of Kr85 from an assembly would not be expected to contribute to groundshine since noble gases do not interact with soils or plateout. Therefore, there is no potential for plateout of Kr85 to the soil and a potential groundshine dose nor could it be released from the soil for dose contribution. Based upon the isotopes of dose significance identified, Kr85 is only one available for potential release at this time-point (only if cladding compromised) there would be no expected contribution of dose from a groundshine itself. Therefore, this calculation assumes no clad failure , and as such, even though 1129 and Cesium's would be of significance in total Curies they are not available for potential groundshine contribution. Figure 6.1 .5-2 illustrates the decrease in curie measure for the nuclides tracked above that would be potentially relevant for groundshine dose. This figure shows how quickly the tracked nuclides decay. The remaining nuclides within the assembly are what are modeled and tracked for source term related to direct and scattered shine, (i.e. Cs, Sr and Y). That output is documented as part of the source term generated for direct and scattered gamma and neutron contribution in FC08514. This calculation does not document a boil off or fuel handling accident, and as such inhalation doses are not required to be computed. FC08513, R0 EC 68969 62

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 63 of 269 Figure 6.1 .5-2 Batch DD Curies by Isotope versus Time 1()8 1()4 1()2 total br82 br83 1129 Cycle 3 Down - DD Assembly 1130 1131 1132 1133 1134 1135 kr83m kr85 kr85m kr87 100 10'2 1()-4 1()-6 1o-- ~D-10 iJ1D-12 1D-14 1D-18 1D-18 1Q-20 10'22 10'24 0.00 0.25 0.50 0 . 75 1.00 125 1.50 tlme(yan) FC08513, R0 EC 68969 63

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 64 of 269 6.2 Plant Geometry The analysis geometry covers a very large area, which extends out to the 9-mile radius of the FCS Low Population Zone, which was reduced to 1 Mile after initial scoping cases showed the dose at the EAB to be low enough that the LPZ dose would not be necessary to determine for this calculation. Review of the FCS USAR indicates the minimum LPZ is at nine (9) miles versus nominal 10 miles for emergency planning zones. In order to keep the model simplified, only the structures of interest to the problem are modeled . These include the portions of the FCS auxiliary building (Rooms 3, 25, 25A, 68, and 69) containment, the control room and the topography at the EAB. Containment is modeled because of the amount of area it takes up and provides a source of material that helps scatter particles back to the south and west EAB areas. Figure 6. 1. 5-1 and Figure 6. 1. 5-2 show the overview of the plant areas as modelled in MCNP. Figure 6.1.5-1 shows the overview of the SFP, Containment, the Control Room and the Room 69 adjacent area between the control room and the SFP above the 1025' plant elevation in the X-Y plane. Figure 6.1.5-2 shows the horizontal slice of the SFP, including the fuel, Containment, the and the area above the SFP in the X-Z plane. Figure 6.1.5-3 shows the horizontal slice of the SFP, including the fuel and the and the Room 69 areas north and south of the SFP in the Y-Z plane. Figure 6.1.5-1 MCNP Model Auxiliary Building Overview (X-Y) Figure 6.1.5-2 MCNP Model Auxiliary Building Slice (X-Z)

                                       - Looking towards plant North -

FC08513, R0 EC 68969 64

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 65 of 269 Figure 6.1.5-3 MCNP Model Auxiliary Building Slice (Y-Z)

                                      - Looking towards plant West -

There are many possible slices that could be shown to provide insight on how the model was built, but to ensure the simplicity of the documentation, the three figures provide an adequate overview of the plant. However, Figure 6.1.1-4 depicts the Y-Z cutaway allowing the distance FC08513, R0 EC 68969 65

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 66 of 269 from the plant and the elevation change at the EAB to be shown. This is where the EAB dose detector was located, 900 Meters South-Southwest of the plant. Figure 6.1.5-4 MCNP Model Plant Cutaway Slice (Y-Z)

                                          - Looking towards plant North -

6.2.1 Model Geometry The starting point (model origin) of plant to model the FCS geometry was the containment centerline at the 979' elevation. Each section of the plant modeled in MCNP utilized plant architectural and structural drawings to ascertain wall locations thicknesses, and elevations. In order to simplify the review process, the MCNP cards shown are grouped by structure and do not match the order of cards as needed to execute MCNP. 6.2.1 .1 Containment References 14.12 through 14.17 provide the details of containment. Containment is modeled as a right circular cylinder with the base at the 991' elevation, 12 feet thick (365.76 em), the internal radius of 55 feet, 0.25 inches (1677.35 em) and external radius of 58 feet, 10 % inches (1795.15 em). The containment dome is not spherical; this resulted in the approximation as a right circular cylinder (rcc). This is acceptable since the containment design more closely reflects the geometry of a cylinder with containment extending well above the top of the auxiliary building, thus any particles leaving the east side of the of the auxiliary building are either scattered back, absorbed or otherwise lost. Reference 14.13, was utilized to determine the approximate height of the cylinder at the 1121 .5' elevation (4343.07 em) utilizing the average of the containment arc. Since the vertical rise of the tendon structure is approximately 18 feet, 3.5 inches, the inside of the containment cylinder is 1103.21' (3785.87 em). The containment tendon structure at the top of the containment FC08513, R0 EC 68969 66

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 67 of 269 provides additional concrete material included due to the thick containment walls at that elevation. Defining containment was done utilizing the surface cards shown in Figure 6.2.1-1. The surfaces defined are the top and bottom elevations of the containment basemat, 3230 and 3260, respectively. The right circular cylinder surfaces 5000 and 5100, centered at 0, 0, 0 and 0, 0, 365.76 are the exterior and interior walls of containment, respectively. Figure 6.2.1-1 Containment Geometry Cards c 3230 pz 365.76 $ Top of Containment Basemat (991'0") 3260 pz 0.00 $ Bottom of Containment Basemat (979'0") c c Containment Building c c 5000 rcc 0 . 0. 0. 0. 0. 4343.07 1795.15 $ CAN Outer 5100 rcc 0. 0. 365.76 0. 0. 3785.87 1677.35 $ CAN Inner c The Cell cards then define the relationship of the surfaces as show in Figure 6.2.1-2. Cell 200 is the containment wall, floor and ceiling made up of material M2 with density of 2.35 grams/cm 3 . Cell 220 is the air within containment made up of material M1 with density of 0.001205 grams/cm 3 . Figure 6.2.1-2 Containment Cell Cards c c C Containment Building c 200 1 -2.35 5100 -5000 $ Containment Structure 220 2 -0.001205 -5100 $ Volume in Containment FC08513, R0 EC 68969 67

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 68 of 269 6.2.1.2 Auxiliary Building Spent Fuel Pool, Transfer Canal and surrounding structures. The surfaces that define the structure within the Auxiliary building are based upon References 14.1 through 14.37. Each structure defined will reference the specific drawing utilized. Spent fuel pool dimensions are found on drawing References 14.19, 14.22, 14.31 and 14.32. An image from drawing 14.22 is shown in Figure 6.2.1-3. The main boundaries of the SFP are highlighted in red. The weir gate is not depicted on this drawing, but is modeled because particles have a more direct path to the exterior walls. The weir gate has walls that are angled. These were modeled as vertical walls using the average distance of the opening from top to bottom. Figure 6.2.1-3 Spent Fuel Pool Main Structure The dimensions are identified as follows: FC08513, R0 EC 68969 68

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 69 of 269 Table 6.2 .1.2-1 S;pen t Fue I P00 I 0 verv1ew o*1mens1ons Dimensions Numerical SFP from DWG value DWG outer width =5.5+20'7"+5.5 31.5833 11405-S-51 outer length =5.5+33'3"+4 42.75 11405-S-51 Wall thickness 5'6" 5.5 11405-S-61 Wall thickness Canal end 4'0" 4.0 11405-S-61 Top of Pool 1038'6" 1038.5 11405-S-61 Bottom of Pool 995.5' 995.5 11405-S-61 Pool basemat 10'+989'0" 16.5 11405-S-60 Distance from Containment to AUX =55'6" + 5'0" 60.5 11405-S-48 SFP Weir gate

                   @1008.5' Width                        5'              5.0             11405-S-61
                   @1039' Width                         7'8"            7.667            11405-S-61 Average Width                        6'4"            6.333            11405-S-61 Utilizing the containment centerline, the definition of each wall's surface was determined by that wall's distance from that point. These values show below converted from feet and inches to centimeters. Figure 6.2.1-4 shows the MCNP surface cards for dimensions defined below.

Table 6.2.1.2-2 s;pent Fue I Poo an dT rans ter Cana IStructure o*1mens1ons Numerical Numerical Description of Surface Dimensions from DWG DWG value (ft) value (em) East Exterior Wall of SFP = 60'6" + 5'6" -66.00 -2011.680 11405-S-61 East Interior Wall of SFP = 60'6" + 5'6" + 5'6" -71.50 -2179.320 11405-S-61 West Interior Wall of SFP = 60'6" + 5'6" + 5'6" -20'7" -92.08 -2806.700 11405-S-61 West Exterior Wall of SFP = 60'6" + 5'6" + 5'6" + 20'7" + 5'6" -97.58 -2974.340 11405-S-61 South Exterior Wall of SFP = -1'6" -1.50 -45.720 11405-S-61 South Interior Wall of SFP = + 2'6" 2.50 76.200 11405-S-61 North Interior Wall of SFP = + 2'6" + 33'3" 35.75 1089.660 11405-S-61 North Exterior Wall of SFP = + 2'6" + 33'3" + 5'6" 41.25 1257.300 11405-S-61 South Interior Wall of TC =- 1'6"- 5'0" -6.50 -198.120 11405-S-61 South Exterior Wall of TC =- 1'6" - 5'0" - 5'0" -11.50 -350.520 11405-S-61 East Exterior Wall of TC = 60'6" -60.50 -1844.040 11405-S-61 FC08513, R0 EC 68969 69

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 70 of 269 East Interior Wall of TC at bottom = 60'6" + 2" -62.50 -1905.000 11405-S-61 East Interior Wall of TC overall = 60'6" + 2'0" + 2'0" -64.50 -1965.960 11405-S-61 North Exterior Wall of TC = + 1'6" + 4'3" 5.75 175.260 11405-S-61

                                =- (60'6" + 2'6" + 13'1" + Avg (4' East Edge of Weir Gate           + 2'6")                                -72.83      -2219.960   11405-S-61
                                =- (60'6" + 2'6" + 13'1"- Avg (4' +

West Edge of Weir Gate 2'6") -79.33 -2418.080 11405-S-61 SFP Walkway =+ 1038'6" - 979'0" 59.00 1798.32 11405-S-61 Bottom of Weir Gate =+ 1008'6" - 979'0" - 1'6" 29.50 899.16 11405-S-61 Pool Bottom =+ 995'6" - 979'0" 16.50 502.92 11405-S-61 Aux Bldg Basemat Top Elevation =+ 989'0" - 979'0" 10.00 304.80 11405-S-61 Aux Bldg Basemat Bottom Elevation =+ 983'6"- 979'0" 4.50 137.16 11405-S-61 Cards 2100 through 2900 define surfaces, which are shown with a "PX" or "PY," referring to the aspect of the dimension. For example, surface 2000 (SFP south outer wall) is a plane in "Y" direction at negative 45.72 em from the containment centerline. Surface 2200 (SFP east outer wall) is a plane in "X" direction at negative 2,011.68 em from the containment centerline. Using all of the X-Y planes defines the structures of the spent fuel pool relative to the centerline of containment. Adding cards 3200 through 3250 define the surfaces, which are shown with a "PZ" defining the elevations of the spent fuel pool structure. Surface 3200 is a plane in "Z" direction 1798.32 em from the containment base mat elevation of 979'. Putting all three dimensions together defines the cells and those cards are shown in Figure 6.2.1-4. FC08513, R0 EC 68969 70

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 71 of 269 Figure 6.2.1-4 Spent Fuel Pool Surface Cards c C SFP c 2000 py -45.72 $ SFP South Outer wall 2050 py 76.20 $ SFP South Inner wall 2100 py 1089.66 $ SFP North Inner wall 2150 py 1257.30 $ SFP North Outer wall 2200 px -2011.68 $ SFP East Outer wall 2250 px -2179.32 $ SFP East Inner wall 2300 px -2806.70 $ SFP West Inner wall 2350 px -2974.34 $ SFP West Outer wall 2400 py -198 . 12 $ TC South Inner wall 2450 py -350.52 $ TC South Outer wall 2550 px -1969.77 $ TC East Inner wall 2600 py 175.26 $ TC North Outer wall 2700 py -1211.58 $ Rm 25A South Inner wall 2750 py -1257.30 $ Rm 25A South Outer wall 2800 px -2219.96 $ Weir Gate East Edge 2850 px -2418.08 $ Weir Gate West Edge 2900 px -2928.62 $ Rm 25A West Inner wall c c 3200 pz 1798.32 $ Top of SFP Walkway (1038'6") 3210 pz 899.16 $ Bottom of Weir Gate (1008'6") 3220 pz 502 . 92 $ Bottom of SFP (995'6") 3250 pz 137.16 $ Aux Bldg Basemat(983'6") c c The cell cards define the relationship of the surfaces as shown in Figure 6.2.1-5. For example, cells 70, 80, 90, 100, 105 and 110 define the SFP, made up of material M2 (concrete) with a density of 2.35 grams/cm 3 , and cells 112, 120 and 125 finish defining the transfer canal (TC) all made up of material M2 (concrete) with a density 2.35 grams/cm 3 . Figure 6.2.1-6 shows images from MCNP of the spent fuel pool with cell numbers displayed, showing where each cell is located in the X-Y and X-Z planes. FC08513, R0 EC 68969 71

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 72 of 269 Figure 6.2.1-5 Spent Fuel Pool Cell Cards 70 1 -2.35 2350 -2200 2100 -2150 3220 -3200 $ SFP North wall 80 1 -2.35 2350 -2300 2450 -2100 3220 -3200 $ SFP West wall 90 1 -2.35 2250 -2200 2000 -2100 3220 -3200 $ SFP East wall 100 1 -2.35 2300 -2850 2000 -2050 3220 -3200 $ SFP South wall w 105 1 -2.35 2800 -2250 2000 -2050 3220 -3200 $ SFP South wall E 110 1 -2.35 2850 -2800 2000 -2050 3220 -3210 $ SFP South wall Weir 112 1 -2.35 2200 -1500 2000 -2600 3220 -3200 $ TC Northeast wall 114 1 - 2.35 1550 -1750 2000 -2600 3220 -3060 $ TC Northeast extension

                #200 #220 120   1    -2.35 2550      -1500 2450   -2000     3220   -3200  $ TC East wall 125   1    -2.35    2300   -2550 2450   -2400     3220   -3200  $ TC South wall 130   1    -2.35    2350   -2900 2750   -2450     3220   -3200  $ Rm 25A West wall 135   1    -2 . 35 2900    -1500 2750   -2700     3220   -3200  $ Rm 25A South wall 137  1     -2 . 35 1450    -1500 1300   -2750     3220   -7000  $ RR Siding Floor 140  1     -2.35 1450      -2350 2750   -1050     3110   -3100  $ Rm 67-69W-1025' elevation 142 1      -2.35    2350   -1500 2150   -1050     3110   - 3100 $ Rm 69N-1025' elevation FC08513, R0                              EC 68969                                    72

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 73 of 269 Figure 6.2.1-6 MCNP Images of Spent Fuel Pool Cells FC08513, R0 EC 68969 73

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 74 of 269 Figure 6.2.1-7 MCNP Images of Spent Fuel Pool Cells FC08513, R0 EC 68969 74

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 75 of 269 6.2.1.3 Control Room, Room 69 Adjacent Area and remaining Area of the Auxiliary Building. The surfaces that define the structure within the Auxiliary building are based upon References 14.1 through 14.37. Each structure defined will reference the specific drawing utilized. Figure 6.2.1-9 shows the MCNP surface cards for Rooms 3, 25, 25A, 67, 69 and 77 (Control Room). The auxiliary building, control room and areas between the spent fuel pool are included because the dose to the control room operator was to determined. The dose was calculated by locating an F5 detector (flux at a point) in the at-the-controls area of the control room, where the operators are required to be stationed 24 hours-a-day, 7 days-a-week (Reference 17). Room 69 encompasses a large open area of the Auxiliary building north and east of the spent fuel pool at the 1025' elevation. This area is also abutted up to the containment wall. An image from the MCNP Vised software shows the model which includes adjacent area in a 3-D layout. See Figure 6.2.1-8. The surfaces which define the all of the areas are shown in Figure 6.2.1-9 are listed in Figure 6.2.1-10. The dimensions are found on drawing References 14.19, 14.22, 14.31 and 14.32. Figure 6.2.1-8 3-D model of the FCS Auxiliary Building Showing Rooms 3, 69 and 77 FC08513, R0 EC 68969 75

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 76 of 269 Table 6.2.1.3-3 Auxiliary Building and Control Room Structure Dimensions South Exterior Wall of Aux (Room 3) =- 62' -62.00 -1889.76 11405-S-51 South Interior Wall of Aux (Room 3) =- 62' + 1'6" -60.50 -1844.04 11405-S-51 North Interior Wall of Aux (Room 3) = +68'6" 68.50 2087.88 11405-S-57 North Exterior Wall of Aux (Room 3) = + 68'6" + 1'6" 70.00 2133.60 11405-S-57 North Interior Wall of Aux (Room 69, below 1044') = + 134'0" - 1'6" 132.50 4038.60 11405-S-55 North Exterior Wall of Aux (Room 69, below 1044') = + 134'0" 134.00 4084.32 11405-S-55 South Interior Wall of East of SFP (Room 69, below 1044') = + 21'6"- 4'3" -17.25 -525.78 11405-S-55 South Exterior Wall of East of SFP (Room 69, below 1044') = + 21'6" + 1'6"- 4'3" -18.75 -571.50 11405-S-55 East Exterior Wall (Room 25A) =- 60'6" -60.50 -1844.04 11405-S-51 East Interior Wall (Room 25A) =- 62'- 1'6" -62.00 -1889.76 11405-S-51

                                       = -(60'6" + 5'6" + 5'6" +

West Interior Wall (Room 25A) 20'7" + 5'6" -1.5) -96.08 -2928.62 11405-S-57

                                       =- (60'6" + 5'6" + 5'6" +

West Exterior Wall (Room 25A) 20'7" + 5'6") -97.58 -2974.34 11405-S-57 South Interior Wall (Room 25A) =- (62' + 21'11" + 0'9") -39.75 -1211.58 11405-S-57 South Exterior Wall (Room 25A) =- (62' + 21 '11" - 0'9") -41.25 -1257.30 11405-S-57 South Exterior Wall of CR (Room 77) = 29'0" + 36'0" - 1'0" 64.00 1950.72 11405-S-55 South Interior Wall of CR (Room 77) = 29'0" + 36'0" + 0'3" 65.25 1988.82 11405-S-55 East Exterior Wall of CR (Room 77) = 118'0" 118.00 3596.64 11405-S-55 East Interior Wall of CR (Room 77) = 118'0"- 1'6" 116.50 3550.92 11405-S-55

                                       = (1'0" + 18'6" + 18'6" +

West Interior Wall of CR (Room 77) 0'9") 38.75 1181.10 11405-S-55 West Exterior Wall of CR (Room 77) = (1'0" + 18'6" + 18'6"- 0'9") 37.25 1135.38 11405-S-55 East Exterior Wall of Aux (Room 3) = - 60'6" -60.50 -1844.04 11405-S-51 East Interior Wall of Aux (Room 3) = - 60'6"- 1'6" -62.00 -1889.76 11405-S-51 West Interior Wall of Aux (Room 3) = - 60'6"- 64'6" + 1'6" -123.50 -3764.28 11405-S-51 West Exterior Wall of Aux (Room 3) = - 60'6" - 64'6" -125.00 -3810.00 11405-S-51 Room 67, 68, 69 Floor (1025') = + 1025'0" - 979'0" 46.00 1402.08 11405-S-63 Room 67, 68, 69 Thickness (6") = + 1025'0" - 979'0" - 0'6" 45.50 1386.84 11405-S-59 Room 3 Roof (1083') = + 1083'0" - 979'0" 104.00 3169.92 11405-S-63 Room 3 Thickness (6") = + 1083'0" - 979'0" - 0'6" 103.50 3154.68 11405-S-59 FC08513, R0 EC 68969 76

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 77 of 269 Room 77 Roof (1057') = + 1057'0" - 979'0" 78.00 2377.44 11405-S-63 Room 77 Thickness (1'6") = + 1057'0" - 979'0" - 1'6" 76.50 2331.72 11405-S-59 Room 69 Roof (1044') = + 1044'0" - 979'0" 65.00 1981.20 11405-S-63 Room 69 Thickness (6") = + 1044'0" - 979'0" -0'6" 64.50 1965.96 11405-S-59 Room 77 Floor (1036') = + 1036'0" - 979'0" 57.00 1737.36 11405-S-63 Room 77 Thickness (0'6") = + 1036'0" - 979'0" - 0'6" 56.50 1722.12 11405-S-59 Figure 6.2.1-9 Room 69 and Control Room Surface Cards 1000 py 4084.60 $ North Outer Wall (Rms. 69 & 77) 1050 py 4038.60 $ North Inner Wall (Rms. 69 & 77) 1100 py 2133 600 $ North Outer Wall (Rms. 3 & 69) 1150 py 2087.88 $ North Inner Wall (Rms. 3 & 69) 1300 py -1844.04 $ South Inner Wall (Rms. 3 & 25) 1350 py -1889.76 $ South Outer Wall (Rms. 3 & 25) 1400 px -3810.00 $ West Outer Wall (Rms. 3 & 25) 1450 px -3764.28 $ West Inner Wall (Rms. 3 & 25) 1500 px -1889.76 $ East Inner Wall (Rms. 3 & 25) 1550 px -1844.04 $ East Outer Wall (Rms. 3 & 25) 1600 py 1988.82 $ South Inner CR Wall (Rm . 77) 1650 py 1950.72 $ South Outer CR Wall (Rm. 77) 1700 px 1135 .38 $ West Outer CR Wall (Rm. 77) 1750 px 11 81.10 $ West Inner CR Wall (Rm. 77) 1800 px 3550 . 92 $ East Inner CR Wall (Rm. 77) 1850 px 3596.64 $ East Outer CR Wall (Rm. 77) c C Aux Building Z Planes c 3000 pz 3169.92 $ Aux Bldg top of Roof Room 3 & 25 (1083') 3010 pz 3167.38 $ Aux Bldg roof - one inch, 3 & 25 (1083') 3011 pz 3164.84 $ Aux Bldg roof - two inches, 3 & 25 (1083') 3012 pz 3162.30 $ Aux Bldg roof- three inches, 3 & 25 (1083') 3013 pz 3159.76 $ Aux Bldg roof - four inches, 3 & 25 (1083') 3014 pz 3157.22 $ Aux Bldg roof - five inches, 3 & 2 5 (1083') 3015 pz 3154.68 $ Aux Bldg roof- six inches, 3 & 25 (1083') 3020 pz 2377.44 $ Aux Bldg top of Roof Room 77 (1057') 3030 pz 2331.72 $ Aux Bldg ceiling of Roof Room 77 (1057') 3040 pz 1737.36 $ Aux Bldg floor slab Room 77 (1036') 3050 pz 1722.12 $ Aux Bldg floor slab thickness Room 77 (1036') 3060 pz 1981.20 $ Aux Bldg top of Roof Room 69 (1044') 3070 pz 1965.96 $ Aux Bldg ceiling of Roof Room 69 (1044') c 3100 pz 1402.08 $ Aux Bldg Room 69 Walkway Floor (1025') 3110 pz 1386.84 $ Aux Bldg Room 69 Walkway Ceiling (1025') c FC08513, R0 EC 68969 77

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 78 of 269 The cell cards define the relationsh ip of the surfaces as shown in Figure 5.2.1.3-3. For example, cells 25 , 30 , 32 , 45 , 68 and 69 frame the Room 77 (control room) , cells 15 20 , 35 , 40 , 401 ' 402 , 403 , 404 , 45 , 50 , 55 , 501 ' 65 , 70 , 80 , 115, 116, 117, 130, 135, 137 140, 142,143 and 604 frame rooms 3, 25 , 67 and 69. Additionally, Room 25A is framed on the bottom or floor by cell 8000 and the vertical space by cells 50 , 80, 120, 125 and 135, made up of material M2 (concrete) with a density of 2.35 grams/cm 3 , and cells 112, 120 and 125 define the transfer canal (TC) all made up of material M2 (concrete) with a density of 2.35 grams/cm 3 . Figure 6.2.1-10 Room 69 and Control Room Cell Cards 10 1 -2.35 1400 -1550 1 350 - 1 000 3250 -3220 $ Aux Bldg Basemat 15 1 -2.35 1400 -1550 1150 -1100 3060 -3015 $ North wall Rm 3 20 1 -2 . 35 1 400 -1550 1050 - 1 000 3220 -30 7 0 $ North wall Rm 69 25 1 -2.35 1 7 00 -1850 1 050 -1000 3100 -3030 $ North wa l l Rm 7 7 30 1 -2 . 35 1 700 - 1850 1 650 - 1 600 3100 -3030 $ South wall Rm 7 7 32 1 -2 . 35 1 800 -1850 1600 - 1 050 3100 -3030 $ East wall Rm 7 7 35 1 -2.35 1400 - 1550 1350 -1300 3220 -3015 $ South wall Rm 3 40 1 -2.35 1400 - 1450 1150 - 1 050 3220 -30 7 0 $ West wall rm 69 401 1 -2.35 1400 - 1450 3310 -1150 3220 -3015 $ West N wall rms 3 & 69 402 1 -2 . 35 1400 - 1450 3320 -33 1 0 3300 -3015 $ West wall above door rm 25 403 1 -2.35 1400 - 1450 3320 -3310 3220 -7000 $ West wall below door rm 25 404 1 -2.35 1400 -1450 1300 -3320 3220 -3015 $ West S wall rm 25 45 1 -2.35 1700 -1750 1600 -1050 3100 -3030 $ West wall rm 77 50 1 -2 . 35 1500 -1550 1300 -2600 3220 -30 1 5 $ East wall Rm 3 501 1 -2.35 1500 - 1 550 2600 -1150 3070 -3015 $ East wa l l Rm 69 above 1025' 55 1 -2.35 1500 - 1 550 2600 -1050 3220 -3100 $ East wall Rm 69 below 1025' 600 1 -2 . 35 1400 - 1 550 1350 - 11 00 3010 -3000 $ Roof Segment Rm 3 1 083 ' -1 601 1 -2 . 35 1400 -1550 1350 -1100 3011 -3010 $ Roof Segment Rm 3 1 083'-2 602 1 -2 . 35 1400 - 1550 1350 -1100 3012 - 3011 $ Roof Segment Rm 3 1083'-3 603 1 -2 . 35 1 400 - 1550 1350 -1100 3013 - 3012 $ Roof Segment Rm 3 1083'-4 604 1 -2 . 35 1400 - 1550 1350 -1100 3014 - 3013 $ Roof Segment Rm 3 1 083'-5 605 1 -2.35 1400 -1550 1350 -1100 3015 -3014 $ Roof Segment Rm 3 1083'-6 65 1 -2.35 1400 -1550 1150 -1000 3070 -3060 $ Roof Segment 1044A' 68 1 -2 . 35 1750 -1800 1 600 - 1 050 3050 -3040 $ F l oor Segment 1036 ' 69 1 -2 . 35 1 7 00 -1850 1650 -1 000 3030 -3020 $ Roof Segment 1 05 7' 115 1 - 2.35 1 550 - 1750 2600 - 1 000 3110 -3100 $ Rm 69 Floor

                #200 # 220 116 1     -2.35 1 5 50      - 1 7 00 2600    - 1 000  3070    -3060    $ Rm 69 Ce i ling
                # 200 #220 11 7 1    -2.35 1550        -1700    1050    - 1 000  3100    - 30 7 0 $ Rm 69 North Wa ll CR 118 1     -2.35 1 700       - 1 7 50 2600    -1 650   3100    -3070    $ Rm 69 East Wal l CR
                # 200 #220 130   1   -2.35 2350        - 2900   2 7 50  - 2450   3220    -3200    $ Rm   25A West wal l 135   1   -2.35 2900        -1500    275 0   -2 7 00  322 0   -3200    $ Rm   25A South wall 137   1   -2 . 35 14 5 0 -1500       1300    -2 7 50  3220    - 70 00  $ RR   Siding Floor 140   1   -2.35 14 5 0      -2350    2 7 50  - 1 050  3110    -3 1 00  $ Rm   67-69W-1025' elevation 142   1   -2.3 5 2350       -1500    2150    -1050    3 11 0  -310 0   $ Rm   69N-1025' elev at i o n 143   1   -2 . 3 5 2200     -15 0 0  2600    -2 1 50  3 11 0  -3100    $ Rm   69N- 1025 ' elevat i o n FC08513, R0                                       EC 68969                                              78

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 79 of 269 Room 25 also has the railroad siding doors that have been accounted for in the model. The doors, made of carbon steel have an effective density of less that concrete, due to the thickness of the two doors is very thin. The security gate outside is not modeled due to the wire mesh design. The doors are modeled using right parallel piped surfaces, using the dimensional information from Table 6.2.1 .3-4. Table 6.2.1.3-4 Railroad Siding Doors Dimensions and References Railroad Siding Doors Elevation at bottom (1004'0", Use Plant Grade level1004'6") = + 1004'6" - 979'0" 25.50 777.24 11405-S-51 Top of Door Opening (22' above 11405-S-51 & 1004'0") = + 1004'0" + 22'0" - 979'0" 47.00 1432.56 11405-S-74 Door Opening width including Frame 11405-S-51 & (16'6") = 16' 16.00 487.68 11405-S-74 South Door Opening from CL of 11405-S-51 & Containment (16'6") =- 62' + 1'6" + 2'11.5" -57.54 -1753.87 11405-S-74 North Door Opening from CL of 11405-S-51 & Containment (16'6") =- 62' + 1'6" + 2'11.5" 16' -41.54 -1266.19 11405-S-74 Door 1004-1A design thickness (20 ga) = 20 gauge steel (.0375") 0.03750 0.0953 232552 Door 1004-lC Sheathing design thickness (14 ga) = 14 gauge steel (.078125") 0.07813 0.1984 G-576 Door 1004-1C design insulation =Foam insulation (3 .75") 3.84375 9.7631 G-576 thickness Use remainder of door thickness ( 4.0-0.078125

• 2)=3 .84375 Door "B" Frame Dimension (Ignore Security Gate) N/A The rollup door was cell 405 made up of surface 3330, a single sheet of 20-gauge carbon steel. The vertical lift door was the door cells 407 and 408, made up of surfaces 3340 and 3350 utilizing 14-gauge carbon steel and polyisocyanurate foam, respectively. The doors surface and cell cards described are shown in Figure 6.2.1-11 and an MCNP image of an overhead view of the doors is shown in Figure 6.2.1-12.

FC08513, R0 EC 68969 79

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 80 of 269 Figure 6.2.1-11 Railroad Siding Doors Surface and Cell Cards c c Rollup Door 1004-1C c 3330 rpp -3764.37525 -3764.28 -1753 . 87 -1266.190 777.24 1432.56 c c c Vertical lift Door 1004-1A c 3340 rpp -3789 . 90 -3779.74 -1753 . 87 -1266.190 777 . 24 1432.56 3350 rpp -3789.70156 -3779.93844 -1753.67156 -1266.38844 777.43844 1432.36156 c 405 4 -7.82 -3330 $ Railroad Siding Door 1004-1C 407 4 -7.82 -3340 3350 $ Railroad Siding Door 1004-1A 408 5 -0.0482 -3350 $ Railroad Siding Door 1004-1A insul Figure 6.2.1-12 Railroad Siding Doors cutaway at Concrete Wall FC08513, R0 EC 68969 80

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 81 of 269 6.2.1.4 Spent Fuel and Source Term Definition (Material M3). The fuel source is the simplest surface and cell to define from a geometry perspective. Because the fuel and spent fuel racks have been homogenized, defining the source was modeled using a single right parallel piped, fitting into the shape of the spent fuel pool. These surfaces utilize the exact elevations of the top and bottom of the spent fuel racks and the perimeter of the spent fuel pool. Table 6.2.1.4-5 provides the dimensional information utilized to define the fuel source that includes the specific drawing utilized. The fuel source door was cell 300 made up of surfaces 3330, 3340 and 3350 utilizing material M3 with a density of2.568032 grams/cm 3 as shown in Table 6.2.1.4-6. The surface and cell cards described are shown in Figure 6.2.1-13. Table 6.2.1.4-5 Spent Fuel Source Dimensions and References Spent Fuel Racks & Fuel Numerical Numerical Description of Surface Dimensions from DWG DWG value (ft) value (em)

                               =  (995'6" -979'0") + 161" +                             11405-S-61 + Holtec Top of Fuel Racks              7.125"                              30.510    929.94480  Dwgs 1002, 1004 11405-S-61 + Holtec Bottom of Fuel Racks           = (995'6" -979'0") + 7.125"         17.094    521.02512  Dwgs 1002, 1004
                               =-(60'6" + 5'6" + 5'6" +                               - 11405-S-61 + Holtec East Edge of Fuel Racks        1.578"                             -71.632  2183.34336   Dwg 1000 + TDB-111.35
                               = -(60'6" + 5'6" + 5'6" +                              - 11405-S-61 + Holtec West Edge of Fuel Racks        20'7") + 0.698"                    -92.025  2804.93206   Dwg 1000 + TDB-111.35 11405-S-61 + Holtec North Edge of Fuel Racks       =+ 2'6" + 33'3"- 3.203"             35.483  1081.52184   Dwg 1000 + TDB-111.35 11405-S-61 + Holtec South Edge of Fuel Racks       =+ 2'6" + 3.042"                     2.753     83.91144  Dwg 1000 + TDB-111.35 11405-S-61 + Holtec South Edge of CPA              =+ 2'6" + 33'3"- 96"- 3.813"        27.432    836.12736  Dwg 1000 + TDB-111.35
                               =-(60'6" + 5'6" + 5'6" +                               - 11405-S-61 + Holtec East Edge of CPA               20'7") + 1.313"                    -83.974  2559.53758   Dwg 1000 + TDB-111.35 Alternate Perimeter dimensions (Used in MCNP and for the Source material density)

East Edge of Fuel Racks =-(60'6" + 5'6" + 5'6" -71.500 2179.32000 11405-S-61

                               =-(60'6" + 5'6"  + 5'6" +                              -

West Edge of Fuel Racks 19.38) -90.880 2770.02240 11405-S-61 North Edge of Fuel Racks =+ 2'6" + 33'3"- 3.203" 33.600 1024.12800 11405-S-61 South Edge of Fuel Racks =+ 2'6" + 3.042" 2.500 76.20000 11405-S-61 FC08513, R0 EC 68969 81

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 82 of 269 Table 6.2.1.4-6 Spent Fuel Source Density Total Mass of Source Volume of Source (Table 5.1.4-10) (Table 5.2.1.4-1) grams cm 3 X*Y*Z

                                                   = [(- 2179.32)- (-2770.0224)]

• 588,007,031.474 [(1024.128)- (76.20)] *

[(929.9448)- (521 .02512)]

                                                   = 228,971 ,853.30 Density (gmlcm 3)                       2.568032 grams I cm 3 Fuel Source Density =Total Mass of Source I Volume of Source Figure 6.2.1-13 Spent Fuel Surface and Cell Cards 2050        py         76.20           $ SFP South Inner wall 2250       px     -2179.32            $ SFP East Inner wall 6000       pz        929 . 945        $ Top of Fuel Racks 6050       pz        521.025          $ Bottom of Fuel Racks 6200       px     -2770 . 022         $ West Boundary of Fuel Racks 6300       py       1024.128          $ North Boundary of Fuel Racks c

C Spent Fuel Pool c 300 3 -2.568032 6200 -2250 2050 -6300 6050 -6000 $ Fuel Mixture volume c In order to utilize geometry and the results of the ORIGEN cases shown in Table 6.1.4-3 through Table 6.1.4-6, the source term was define using the MCNP "SDEF card. The SDEF defines the functionality of the source. Since the sources for these cases are 3-dimensional and energy dependent (as determined by the ORIGEN cases), these sources must be defined using, the dimensions of the fuel surface geometry, which are shown on the Sl1 (x), Sl2 (y), Sl3(z) cards matching the definitions of fuel geometry. The SP1, SP2, and SP3 cards accompanying the Sl cards indicates the functionality of the source is based upon a distribution of 0 to 1, which reflects a probability of the source (1) is in effect at all times. The energy dependence of the photons and neutrons is identified similarly using the Sl4 and SP4 cards. The Sl4 card lays out the energy dependence of the source and the SP4 card the number of photons or neutrons in that energy bin, thus providing a probability distribution for MCNP to utilize. For the photon source, there are 18 energy groups as provided in Table 6. 1.4-3 and Table 6.1.4-4, while there are 44 energy groups for the FC08513, R0 EC 68969 82

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 83 of 269 neutrons as provided in Table 6.2.1.4-5 and Table 6.2.1.4-6. Figure 6.2.1-14 shows the MCNP cards as utilized for the 18-month neutron source. Figure 6.2.1-14 Spent Fuel Source Term Definition (18 Month 44 Energy Group Neutron) c c Source Definition c sdef X=d1 y=d2 z=d3 par=1 erg=d4 si1 H -2770.022 -2179 . 320 $ Surface 6200 to 2250 sp1 D 0 1 si2 H 76.200 1024.128 $ surface 6400 to 6300 sp2 D 0 1 si3 H 521.025 929.945 $ Surface 6050 to 6000 sp3 D 0 1 c c C Neutron Spectrum -Total Source Strength is 2.36E+l1 C Table 8.7 Neutron Source Term (Eighteen Month) C by 44 Energy Group (FC08514, page 46) c si4 H 1.00E-11 3 . 00E-09 7.50E-09 1.00E-08 2.53E-08 3.00E-08 4.00E-08 5 . 00E-08 7.00E-08 1. OOE- 07

1. 50E- 07 2.00E-07 2.25E-07 2 . 50E-07 2.75E-07 3.25E-07 3.50E-07 3.75E-07 4 . 00E-07 6 . 25E-07 1.00E-06 1.77E-06 3.00E-06 4 . 75E-06 6.00E-06 8.10E-06 1.00E-05 3.00E-05 1.00E-04 5.50E-04 3.00E-03 1.70E-02 2 . 50E-02 1.00E-01 4.00E-01 9.00E-01 1 . 40E+00 1 . 85E+00 2.35E+00 2.48E+00 3.00E+00 4.80E+00 6 . 43E+00 8.19E+00 2.00E+01 sp4 D O.OOE+OO 3.51E-03 2.52E-03 1. 99E- 03 5.85E-03 2.42E-03 8 . 48E-03 6.07E-03 1.31E-02 2.16E-02 4.12E-02 4.88E-02 2.97E-02 3 . 88E-02 3.26E-02 5.68E+00 1.96E+00 2.03E+00 2.11E+00 1.85E+01 4 . 42E+01 1.24E+02 2.44E+02 4.53E+02 3.91E+02 7.36E+02 7.54E+02 1.17E+04 7 . 40E+04 1.05E+06 1.34E+07 1.81E+08 1.52E+08 2.37E+09 1 . 71E+10 3.73E+10 3.73E+10 2 . 99E+10 2.82E+10 6.10E+09 2 . 18E+10 3.95E+l0 1.12E+10 3 . 57E+09 1 . 23E+09 c

FC08513, R0 EC 68969 83

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 84 of 269 7.0 Conclusions Gamma and neutron dose rates for the FCS EAB and Control Room were determined using MCNP6.1. The dose rate (neutron and gamma) at the EAB 14-months after shutdown with the Spent Fuel Pool completely drained is 1.66E-03 mRem/hr. Therefore, the DOE dose to the public from this event is 14.6 mrem, which is less than the 10CFR20 limit of 50 mRem. The dose rate (neutron and gamma) from this event in the Control Room is 2.32E-03 mRem/hr. which is_below 15-mR/hr as defined in NUREG-0737 (Reference 22). The dose rates (neutron and gamma) 18-months after shutdown at the EAB with the Spent Fuel Pool completely drained is 1.38E-03 mRem/hr and the Control Room is 2.23E-03 mRem/hr. The reported dose rates above are determined by averaging MCNP tally output for Gamma (Table 7-1) and Neutron (Table 7-2). Because the number of histories for the Neutron results were run at different times (and required considerably more computer execution time than the gamma cases), the 14-month neutron value reported was determined by a ratio of the tally result at the last point (nps=1.0E+08) to the 18-month tally value (nps=1.006633E E+08). The tally trend on both cases were similar, ensuring the ratio was valid. The full tally analysis are shown in Attachment D spreadsheets ("tallies_gamm20161017.xlsx" and "tallies_neut20161 017). FC08513, R0 EC 68969 84

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 85 of 269 Table 7-1 G am rna TIIR a1y esu Its EAB Dose CR Dose EAB Dose CR Dose Histories (14mog) (14mog) (18mog) (18mog) (nps) mRem/hr mRem/hr mRem/hr mRem/hr 4.0265318E+08 6.5911E-04 9.5316E-05 5.4485E-04 7.8906E-05 8.053063 7E+08 4.8715E-04 2.7896E-04 4.0254E-04 2.3080E-04 1.2079596E+09 4.2242E-04 1.9971E-04 3.4907E-04 1.6521E-04 1.6106127E+09 3.9640E-04 1.8288E-04 3.2760E-04 1.5128E-04 2.0132659E+09 1.4730E-03 1.5396E-04 1.2176E-03 1.2733E-04 2.4159191E+09 1.2988E-03 1.3851E-04 1.0736E-03 1.1457E-04 2.8185723E+09 1.1754E-03 1.2464E-04 9.7174E-04 1.0312E-04 3.2212255E+09 1.1539E-03 1.2634E-04 9.5482E-04 1.0345E-04 3.6238787E+09 1.0903E-03 1.1738E-04 9.0215E-04 9.6153E-05 4.0265318E+09 1.2041E-03 1.3338E-04 9.9591E-04 1.0938E-04 4.4291850E+09 1.1089E-03 1.2901E-04 9.1718E-04 1.0581E-04 4.8318382E+09 1.0650E-03 1.2521E-04 8.8082E-04 1.0275E-04 S.OOOOOOOE+09 1.0343E-03 1.2267E-04 8.5547E-04 1.0068E-04 5.6371446E+09 1.1640E-03 1.1937E-04 9.6275E-04 9.8168E-05 6.4424509E+09 1.3709E-03 1.1075E-04 1.1354E-03 9.1104E-05 7.2477573E+09 1.2599E-03 1.1823E-04 1.0420E-03 9.7243E-05 8.0530637E+09 1.1613E-03 1.1826E-04 9.6048E-04 9.7341E-05 8.8583700E+09 1.1584E-03 1.1429E-04 9.5782E-04 9.4124E-05 9.6636764E+09 1.0900E-03 1.0969E-04 9.0125E-04 9.0353E-05 l.OOOOOOOE+ 10 1.0590E-03 1.0735E-04 8.7565E-04 8.8403E-05 Average Tally 1.0416E-03 1.3630E-04 8.6144E-04 1.1231E-04 Table 7-2 Neutron Tally Results EAB Dose CR Dose EAB Dose CR Dose Histories Histories (14mon) (14mon) (18mon) (18mon) (nps) (nps) mRem/hr mRem/hr mRem/hr mRem/hr 3.1457280E+06 5.5506E-06 1.946E-03 5.0331648E+07 1.0360E-05 1.5929E-03 6.2914560E+06 5.5102E-06 1.879E-03 1.0066330E+08 1.1342E-05 1.4876E-03 6.2914560E+06 5.5102E-06 1.879E-03 1.5099494E+08 1.3747E-05 1.6460E-03 9.4371840E+06 5.3999E-06 1.760E-03 2.0132659E+08 1.3273E-05 1.6499E-03 1.2582912E+07 6.0547E-06 1.554E-03 2.5165824E+08 1.3310E-05 1.7010E-03 1.5728640E+07 1.1182E-05 1.834E-03 3.0198989E+08 1.3783E-05 1.7342E-03 1.8874368E+07 1.0251E-05 1.733E-03 3.5232154E+08 1.4124E-05 1.7089E-03 FC08513, R0 EC 68969 85

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 86 of 269 2.2020096E+07 1.1553E-05 1.623E-03 4.0265318E+08 1.7055E-05 1.9534E-03 2.5165824E+07 1.1146E-05 l .SSSE-03 4.5298483E+08 1.7312E-05 1.9429E-03 2.8311552E+07 1.0529E-05 l.SOGE-03 5.0331648E+08 1.7236E-05 1.9028E-03 3.1457280E+07 1.0482E-05 1.432E-03 5.5364813E+08 1.7087E-05 1.8580E-03 3.4603008E+07 1.0182E-05 1.378E-03 6.0397978E+08 1.6951E-05 1.8157E-03 3.5000000E+07 1.0176E-05 1.364E-03 6.5431142E+08 1.6618E-05 1.7817E-03 3.7748736E+07 1.0632E-05 1.348E-03 7.0464307E+08 1.6364E-05 1.7739E-03 4.4040192E+07 l.OSSOE-05 1.435E-03 7. 2500000E+08 1.6178E-05 1.7783E-03 5.0331648E+07 1.0192E-05 1.425E-03 Average Tally 1.4983E-05 1.7551E-03 5.6623104E+07 1.0035E-05 1.406E-03 6.2914560E+07 1.0128E-05 1.366E-03 6.9206016E+07 1.0046E-05 1.547E-03 7.5497472E+07 1.1073E-05 1.519E-03 8.1788928E+07 1.1576E-05 1.482E-03 8.8080384E+07 1.1358E-05 1.456E-03 9.4371840E+07 1.1220E-05 1.384E-03 l.OOOOOOOE+08 1.0897E-05 1.4437E-03 7 .2500000E+08 1.5595E-05 1.8085E-03 The EAB and CR dose rates for the 14-month neutron dose rate was determined by using the ratio of the mean values at 1.0E+08 histories for the 14-month and 18-month runs. This is due to the 14-month run did not utilize the same total number of histories (7.25E+08). To calculate the upper boundary, the error for the 18-month case was applied. Based upon the results of the MCNP cases , dose rates at the EAB and control room, have larger relative errors than expected. Notwithstanding, the dose rates including a two-sigma (standard deviation) error, which did meet the objective of this calculation. See Table 7-3 for gamma and neutron dose rate results with the error applied. Dose rates are well below the acceptance criteria of less than 15-mR/hr as defined in NUREG-0737 (Reference 22) in the Control Room. In addition, the 15-mR/hr is consistent with the Control Room emergency action level for the new shutdown EALs (Reference 21 ). In addition, the DOE dose is less than the 50-mrem in a year for members of the public or % the TEDE dose for member of the public as defined in 10 CFR 20. FC08513, R0 EC 68969 86

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 87 of 269 Table 7-3 Point Detector Gamma Dose Rates (b) Detector Location Mean Dose Upper Boundary Decay I Dose Type I Relative Error (EAB or Control Rate 95% Confidence Tally Number Room) (fraction) (mRemlhr) Interval (a) 14 Mo./ Gamma 115 EAB 1.0416E-03 0.2935 1.64E-03 14 Mo./ Gamma 125 Control Room 1.3630E-04 0.1950 1.88E-04 18 Mo./ Gamma 115 EAB 8.614E-04 0.2934 1.36E-03 18 Mo./ Gamma I 25 Control Room 1.1231 E-04 0.1958 1.55E-04 14 Mo./ Neutron 115 1.91 E-05 EAB 1.5595E-05 0.1158 (c) 14 Mo./ Neutron I 25 2.13E-03 Control Room 1.8085E-03 0.0907 (c) 18 Mo./ Neutron 115 EAB 1.4983E-05 0.1158 1.84E-05 18 Mo./ Neutron I 25 Control Room 1.7551 E-03 0.0907 2.07E-03 NOTES: (a) The upper boundary of the 95% confidence interval is calculated by adding the mean value to the product of 1.96, the mean value and the relative error. For the14-month gamma dose rate at the EAB, the upper boundary is 1.0416E-03 mRemlhr + (1.96 x 1.0416E-03 mRemlhr x 0.2935)

    =1.64E-03 mRemlhr.

(b) The relative errors for these dose locations did not pass all the MCNP 10 statistical checks and the errors are greater than 0.05. Table 1.1 : "Guidelines for Interpreting the Relative Error" of the MCNP6 User's Manual indicates that relative errors between 0.05 and 0.50 imply that the tally results could be off by a factor of a few. Due to the dose rates being significantly low, the results indicate a large margin to the acceptance criteria. (c) The 14-month neutron dose rate was determined by using the ratio of the mean values at 1.0E+08 histories for the 14-month and 18-month runs. This is due to the 14-month run did not utilize the same total number of histories (7.25E+08). To calculated the upper boundary, the error for the 18-motnh case was applied. This is a valid approach since the histories show identical trends with respect to mean value, error, variance, pdf and figure of merit. FC08513, R0 EC 68969 87

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 88 of 269 Attachment A: MCNP Input Files Four different sets of MCNP models were run: Table A-1 presents the run log for this analysis.

1) FCS_14MOg_EABCR.inp 14 Month decay using Gamma source term.

2) FCS_18M0g_EABCR.inp 18 Month decay using Gamma source term.

3) FCS_14MOn_EABCR.inp 14 Month decay using Neutron source term.

4) FCS_18M0n_EABCR.inp 18 Month decay using Neutron source term.

Table A-1. MCNP Run Log CASE Input Output (*) Tape 14-Month Gamma FCS_14MOg_EABCR.inp FCS_14MOg_EABCR.out FCS_14MOgEABCR.tap FCS_ 14MOg_EABCRca.out FCS_ 14MOg_EABCRcb . out FCS_ 14MOg_EABCRcc.out FCS_14MOg_EABCRcd.out FCS_14MOg_EABCRce.out FCS_ 14MOg_EABCRcf . out FCS_14MOg_EABCRcg.out FCS_14MOg_EABCRch.out FCS_14MOg_EABCRci.out 18-Month Gamma FCS_18MOg_EABCR . inp FCS_18MOg_EABCR.out FCS_18MOg_EABCR.tap FCS_18MOg_EABCRca.out 14-Month Neutron FCS_14MOn_EABCR . inp FCS - 14MOn- EABCR.out FCS_14MOn_EABCR.tap FCS - 14MOn- EABCRca.out FCS - 14MOn- EABCRcb.out FCS - 14MOn- EABCRcc.out FCS - 14MOn- EABCRcd.out FCS - 14MOn- EABCRce.out 18 - Month Neutron FCS_18MOn_EABCR . inp FCS - 18MOn- EABCR.out FCS_18MOn_EAB . tap FCS_18MOg_EABCRca.out FCS - 18MOn- EABCRcb . out FCS - 18MOn- EABCRcc . out

• Multiple output files are the results of the cases being restarted because of the Windows Operating system reboots or the cases were extended for the purposes of bettering the statistics. Restarting the cases has no effect on the final result.

FC08513, R0 EC 68969 88

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 89 of 269 FCS 14MOg EABCR. inp Fort Calhoun Station SFP Gamma Dose Rates Following Drain Down c c cell cards c --------------------------------------------------------- c Fuel Handling Building c 10 1 -2.35 1400 -1550 1350 -1000 3250 -3220 $ Aux Bldg Basemat 15 1 -2.35 1400 -1550 1150 -1100 3060 -3015 $ North wall Rm 3 20 1 -2.35 1400 -1550 1050 -1000 3220 -3070 $ North wall Rm 69 25 1 -2 . 35 1700 -1850 1050 -1000 3100 -3030 $ North wall Rm 77 30 1 -2.35 1700 -1850 1650 -1600 3100 -3030 $ South wall Rm 77 32 1 -2.35 1800 -1850 1600 -1050 3100 -3030 $ East wall Rm 77 35 1 -2.35 1400 -1550 1350 -1300 3220 -3015 $ South wall Rm 3 40 1 -2.35 1400 -1450 1150 -1050 3220 -3070 $ West wall rm 69 401 1 -2.35 1400 -1450 3310 -1150 3220 -3015 $ West N wall rms 3 & 69 402 1 -2.35 1400 -1450 3320 -3310 3300 -3015 $ West wall above door rm 25 403 1 -2.35 1400 -1450 3320 -3310 3220 -7000 $ West wall below door rm 25 404 1 -2.35 1400 -1450 1300 -3320 3220 -3015 $ West S wall rm 25 405 4 -7.82 -3330 $ Railroad Siding Door 1004-1C 407 4 -7.82 -3340 3350 $ Railroad Siding Door 1004-1A 408 5 -0.0482 -3350 $ Railroad Siding Door 1004-1A insul 45 1 -2.35 1700 -1750 1600 -1050 3100 -3030 $ West wall rm 77 50 1 -2.35 1500 -1550 1300 -2600 3220 -3015 $ East wall Rm 3 501 1 -2.35 1500 -1550 2600 -1150 3070 -3015 $ East wall Rm 69 above 1025' 55 1 -2.35 1500 -1550 2600 -1050 3220 -3100 $ East wall Rm 69 below 1025' 600 1 -2.35 1400 -1550 1350 -1100 3010 -3000 $Roof Segment Rm 3 1083'-1 601 1 -2.35 1400 -1550 1350 -1100 3011 -3010 $Roof Segment Rm 3 1083'-2 602 1 -2.35 1400 -1550 1350 -1100 3012 -3011 $Roof Segment Rm 3 1083'-3 603 1 -2.35 1400 -1550 1350 -1100 3013 -3012 $Roof Segment Rm 3 1083'-4 604 1 -2.35 1400 -1550 1350 -1100 3014 -3013 $Roof Segment Rm 3 1083'-5 605 1 -2.35 1400 -1550 1350 -1100 3015 -3014 $Roof Segment Rm 3 1083'-6 65 1 -2.35 1400 -1550 1150 -1000 3070 -3060 $ Roof Segment 1044A' 68 1 -2.35 1750 -1800 1600 -1050 3050 -3040 $ Floor Segment 1036' 69 1 -2.35 1700 -1850 1650 -1000 3030 -3020 $ Roof Segment 1057' 70 1 -2.35 2350 -2200 2100 -2150 3220 -3200 $ SFP North wall 80 1 -2.35 2350 -2300 2450 -2100 3220 -3200 $ SFP West wall 90 1 -2.35 2250 -2200 2000 -2100 3220 -3200 $ SFP East wall 100 1 -2 . 35 2300 -2850 2000 -2050 3220 -3200 $ SFP South wall W 105 1 -2.35 2800 -2250 2000 -2050 3220 -3200 $ SFP South wall E 110 1 -2.35 2850 -2800 2000 -2050 3220 -3210 $ SFP South wall Weir 112 1 -2.35 2200 -1500 2000 -2600 3220 -3200 $ TC Northeast wall 114 1 -2.35 1550 -1750 2000 -2600 3220 -3060 $ TC Northeast extension

              #200 #220 115 1    -2.35 1550 -1750      2600 -1000 3110   -3100 $ Rm 69 Floor
              #200 #220 116 1    -2.35 1550      -1700 2600 -1000 3070   -3060 $ Rm 69 Ceiling
              #200 #220 117 1    -2.35 1550 -1700      1050 -1000 3100   -3070 $ Rm 69 North Wall CR 118 1    -2.35 1700 -1750      2600 -1650 3100   -3070 $ Rm 69 East Wall CR
              #200 #220 120 1    -2.35 2550      -1500 2450 -2000 3220   -3200 $ TC East wall 125 1    -2.35 2300      -2550 2450 -2400 3220   -3200 $ TC South wall 130 1    -2.35 2350      -2900 2750 -2450 3220   -3200 $ Rm 25A West wall FC08513, R0                            EC 68969                                    89

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 90 of 269 135 1 -2.35 2900 -1500 2750 -2700 3220 -3200 $ Rm 25A South wall 137 1 -2.35 1450 -1500 1300 -2750 3220 -7000 $ RR Siding Floor 140 1 -2.35 1450 -2350 2750 -1050 3110 -3100 $ Rm 67-69W-1025' elevation 142 1 -2.35 2350 -1500 2150 -1050 3110 - 3 100 $ Rm 69N-1025' elevation 143 1 -2.35 2200 -1500 2600 -2150 3110 -3100 $ Rm 69N-1025' elevation 150 2 -0 . 001205 1450 -2350 2750 -1050 7000 -3110 $ Area West of SFP below 1025' 151 2 -0.001205 1450 -2350 2750 -1050 3100 -3200 $Area West of SFP above 1025' 155 2 - 0 . 001205 1450 -1500 2750 -1150 3200 -3015 $Area Above SFP 1038'6" 156 2 -0 . 001205 1500 -1550 2600 - 1050 3100 -3070 $Area East of SFP below 1025' 160 2 -0 . 001205 2350 -1500 2150 -1050 7000 -3110 $Area North of SFP below 1025' 161 2 -0 . 001205 2350 -1500 2150 -1050 3100 -3200 $Area North of SFP above 1025' 162 2 -0 . 001205 1450 -1500 1150 -1050 3 2 00 -3070 $ Upper Area North of SFP 164 2 -0.001205 2200 -1500 2600 -2150 3220 -3110 $Area East of SFP below 1025' 165 2 -0.001205 1550 -1700 2600 -1050 3100 -3070 $ Area in Rm 69 East

                #200 #220 166  2    -0.001205 2200 -1500 2600 -2150             3100   -3200     $Area East of SFP above 1025' 170  2    -0.001205 1450 -1500 1300 -2750             7000   -3015     $ RR Siding South of Rm 25A 175  2    -0.001205 2900 -1500 2700 -2450             7000   -3200     $ Area in Rm 25A 185  2    -0.001205 2850 -2800 2000 -2050             3210   -3200     $ Area in Weir Gate 190  2    -0.001205 2300 -2550 2400 -2000             3220   -3200     $ Area in TC Gate c

195 2 - 0 . 001205 1750 -1800 1600 -1050 3040 -3030 $Area in Room77 c c C Containment Building c 200 1 -2.35 5100 -5000 $ Containment Structure 220 2 -0.001205 -5100 $ Volume in Containment c C Spent Fuel Pool c 300 3 -2 . 568032 6200 -2250 2050 -6300 6050 -6000 $ Fuel Mixture volume 310 2 -0 . 001205 2300 - 2 250 2050 -2100 6000 -3200 $ Area Above Fuel Racks 315 2 -0 . 001205 2300 -2250 2050 -2100 3220 -6050 $ Area Below Fuel Racks 320 2 -0.001205 2300 -6200 2050 -2100 6050 -6000 $ Area West of Fuel Racks 325 2 -0.001205 6200 -2250 6300 -2100 6050 -6000 $ Area North of Fuel Racks c c Ground Plane Region c 8000 1 -2.35 (-7000 3260 -10000) $ Plant Grade Region

            #10 #20 #35 #40 #50 #55 #70 #80 #90 #100 #105 #110 #112
           #114 #120 #125 #130 #135 #137 #150 #160 #164 #170 #175 #190 #200
           #220 #300 #315 #320 #325 #401 #403 #404 8100      2      -0.001205 7100 7000 -7200 -10000                      $ 1004'6" Elevation Region
            #15 #20 #25 #30 #32 #35 #40 #45 #50 #55 #501 #65 #68
            #69 #70 #80 #90 #100 #105 #110 #112 #114 #115 #116 #117 #118 #120
           #125 #130 #135 #140 #142 #143 #150 #151 #155 #156 #160 #161 #162 #164
           #165 #166 #170 #175 #185 #190 #195 #200 #220 # 3 00 #310 #320 #325
           #501 #401 #402 #404 #405 #407 #408 8200      1      -2 . 35 7000      -10000 -7100    -7200 -10000        $ 1080' Elevation Region 8300      2      -0 . 001205 7200 -10000                     $Air Above 1080'
            #15 #35 #50 #600 #601 #602 #603 #604 #605 #155 #170
            #200 #220 #501 #401 #402 #404 c

9100 0 -3260 -10000 $ Universe Below Basemat c FC08513, R0 EC 68969 90

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 91 of 269 c void c 9999 0 10000 c C surface Cards c C Aux Building XY Planes c 1000 py 4084.60 $ North Outer Wall (Rms. 69 & 77) 1050 py 4038.60 $ North Inner Wall (Rms. 69 & 77) 1100 py 2133 . 60 $ North Outer Wall (Rms. 3 & 69) 1150 py 2087 . 88 $North Inner Wall (Rms. 3 & 69) 1300 py -1844 . 04 $ South Inner Wall (Rms. 3 & 25) 1350 py -1889.76 $ South Outer Wall (Rms . 3 & 25) 1400 px -3810.00 $ West Outer Wall (Rms. 3 & 25) 1450 px -3764.28 $ West Inner Wall (Rms . 3 & 25) 1500 px -1889 . 76 $ East Inner Wall (Rms. 3 & 25) 1550 px -1844.04 $ East Outer Wall (Rms. 3 & 25) 1600 py 1988 . 82 $ South Inner CR Wall (Rm. 77) 1650 py 1950.72 $ South Outer CR Wall (Rm. 77) 1700 px 1135.38 $ West Outer CR Wall (Rm. 77) 1750 px 1181.10 $ West Inner CR Wall (Rm. 77) 1800 px 3550.92 $ East Inner CR Wall (Rm. 77) 1850 px 3596.64 $ East Outer CR Wall (Rm . 77) c C SFP c 2000 py -45.72 $ SFP South Outer wall 2050 py 76.20 $ SFP South Inner wall 2100 py 1089.66 $ SFP North Inner wall 2150 py 1257.30 $ SFP North Outer wall 2200 px -2011.68 $ SFP East Outer wall 2250 px -2179.32 $ SFP East Inner wall 2300 px -2806 . 70 $ SFP West Inner wall 2350 px -2974 . 34 $ SFP West Outer wall 2400 py -198 . 12 $ TC South Inner wall 2450 py -350 .52 $ TC South Outer wall 2550 px -1969 . 77 $ TC East Inner wall 2600 py 175 .26 $ TC North Outer wall 2700 py -1211.58 $ Rm 25A South Inner wall 2 750 py -1257.30 $ Rm 25A South Outer wall 2800 px -2219.96 $ Weir Gate East Edge 2850 px -2418.08 $ Weir Gate West Edge 2900 px -2928.62 $ Rm 25A West Inner wall c C Aux Building z Planes c 3000 pz 3169.92 $ Aux Bldg top of Roof Room 3 & 25 (1083') 3010 pz 3167.38 $ Aux Bldg roof - one inch, 3 & 25 (1083') 3011 pz 3164.84 $ Aux Bldg roof - two inches, 3 & 25 (1083') 3012 pz 3162.30 $ Aux Bldg roof- three inches, 3 & 25 (1083') 3013 pz 3159.76 $ Aux Bldg roof- four inches, 3 & 25 (1083') 3014 pz 3157.22 $ Aux Bldg roof- five inches, 3 & 25 (1083') 3015 pz 3154.68 $ Aux Bldg roof - six inches, 3 & 25 (1083') 3020 pz 2377.44 $ Aux Bldg top of Roof Room 77 (1057') FC08513, R0 EC 68969 91

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 92 of 269 3030 pz 2331.72 $ Aux Bldg ceiling of Roof Room 77 (1057') 3040 pz 1737.36 $ Aux Bldg floor slab Room 77 (1036') 3050 pz 1722.12 $ Aux Bldg floor slab thickness Room 77 (1036') 3060 pz 1981 . 20 $ Aux Bldg top of Roof Room 69 (1044') 3070 pz 1965.96 $ Aux Bldg ceiling of Roof Room 69 (1044') c 3100 pz 1402.08 $ Aux Bldg Room 69 Walkway Floor (1025') 3110 pz 1386.84 $ Aux Bldg Room 69 Walkway Ceiling (1025') c 3200 pz 1798.32 $Top of SFP Walkway (1038'6") 3210 pz 899 . 16 $ Bottom of Weir Gate (1008'6") 3220 pz 502.92 $ Bottom of SFP (995'6") 3250 pz 137 . 16 $ Aux Bldg Basemat(983'6") c c 3300 pz 1432 . 56 $ Aux Bldg Door Top of Opening 3310 py -1266 . 19 $ Aux Bldg Door North side of Opening 3320 py -1753 . 87 $ Aux Bldg Door South side of Opening c c Rollup Door 1004-1C c 3330 rpp -3764.37525 -3764.28 -1753.87 -1266.190 777.24 1432.56 c c c Vertical lift Door 1004-1A c 3340 rpp -3789.90 -3779 . 74 -1753.87 -1266.190 777.24 1432.56 3350 rpp -3789.70156 -3779.93844 -1753.67156 -1266.38844 777.43844 1432.36156 c 3230 pz 365.76 $ Top of Containment Basemat (991'0") 3260 pz 0.00 $ Bottom of Containment Basemat (979'0") c C Containment Building c c 5000 rcc 0. 0. 0. 0. 0. 4343.07 1795.15 $ CAN Outer 5100 rcc 0 . 0 . 365 . 76 0 . 0. 3785.87 1677.35 $ CAN Inner c c C Fuel Racks c c 6000 pz 929 . 945 $ Top of Fuel Racks 6050 pz 521 . 025 $ Bottom of Fuel Racks 6200 px -2770.022 $ West Boundary of Fuel Racks 6300 py 1024 . 128 $ North Boundary of Fuel Racks c c Optional planes of more detial of the CPA c c 6100 py -2183 . 343 $ East Boundary of Fuel Racks c 6200 py -2804.932 $ West Boundary of Fuel Racks c 6300 px 1081.522 $ North Boundary of Fuel Racks c 6400 px 83.911 $ South Boundary of Fuel Racks c 6500 px -2559.538 $ East Boundary of CPA FC08513, R0 EC 68969 92

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 93 of 269 c 6600 px 836.127 $ South Boundary of CPA c c Ground Plane c 7000 pz 777.24 $ Plant Grade Elevation 1004'6" 7100 px -46918.96 $ Distance to West Side of Bridge c 7100 px -46939.20 $ Distance to EAB 7200 pz 3078.48 $ 1080' Elevation c c Universe c c 10000 so 1448410.0 $ LPZ at 9 miles 10000 so 161000.0 $ LPZ at 1 mile C Data cards c c Physics Definition c mode p c c c Material Definition c C ordinary concrete = 2.35 g/cm3 C 95 Concrete, Ordinary (NBS 03) c m1 1001 0.011914 $ H atoms/barn-em 6000 0.005899 $ c atoms/barn-em 8016 0.041881 $ 0 atoms/barn-em 12000 0.001408 $ Mg atoms/barn-em 13027 0.001892 $ Al atoms/barn-em 14000 0.007311 $ Si atoms/barn-em 16000 0.000131 $ S atoms/barn-em 19000 0.000061 $ K atoms/barn-em 20000 0.008719 $ Ca atoms/barn-em 26000 0 . 000280 $ Fe atoms/barn-em c c c c c Air 0 . 001205 g/cm3 c c m2 7014 0.000039 $ N atoms/barn-em 8016 0.000011 $ 0 atoms/barn-em c C Fuel Region Mixture 2 . 568032 g/cm3 c M3 13027 -8.344615E-03 $ Al Weight Percent 18000 -4 . 242021E-06 $ Ar Weight Percent 5010 -3.049262E-04 $ B10 Weight Percent 5011 -1.361160E-03 $ B11 Weight Percent 6000 -5.023089E-04 $ c Weight Percent 17000 -5.747414E-09 $ Cl Weight Percent 27059 -3.298115E-05 $ Co Weight Percent 24000 -2.048669E-02 $ Cr Weight Percent FC08513, R0 EC 68969 93

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 94 of 269 29000 -6 . 209134E-05 $ Cu Weight Percent 26000 -6 . 672541E-02 $ Fe Weight Percent 25055 -1. 966967E-03 $ Mn Weight Percent 42000 -4.783920E-05 $ Mo Weight Percent 7014 -2.698222E-04 $ N Weight Percent 11023 -5.747414E-09 $ Na Weight Percent 41093 -1.902357E-03 $ Nb Weight Percent 28000 -1.868235E-02 $ Ni Weight Percent 8016 - 8.421467E-02 $ 0 Weight Percent 15031 -4.292521E-05 $ p Weight Percent 82000 -3 . 735819E-08 $ Pb Weight Percent 16000 -3 . 621267E-05 $ s Weight Percent 14000 -1 . 214038E-03 $ Si Weight Percent 50000 -4.166875E-06 $ Sn Weight Percent 73181 -6.522834E-06 $ Ta Weight Percent 22000 -3.010457E-04 $ Ti Weight Percent 92234 -1 . 946356E-04 $ U-234 Weight Percent 92235 -2.1835 2 8E-02 $ U-235 Weight Percent 92236 -1.005027E-04 $ U-236 Weight Percent 92238 -6.017375E-01 $ U-238 Weight Percent 23000 -1 . 436854E-08 $ v Weight Percent 40000 -1 . 696187E-01 $ Zr Weight Percent c C carbon steel Door (density: 7.82 g/cm3) c M4 06000 0.001960 26000 0.083907 c C Door 1004-1 Insulation C From PNNLpnnl-15870 Rev 1: Material 249 c Polyisocyanurate (PIR, density: 0.048200 g/cm3) c M5 1001 0.00116 6000 0.00174 7014 0.000232 8016 0.000232 c C Variance Reduction c c imp:p 1 73r 0 0 c c c Source Definition c sdef X=d1 y=d2 z=d3 par=2 erg=d4 si1 H -2770 . 022 -2179 . 320 $ Surface 6200 to 2250 sp1 D 0 1 si2 H 76 . 200 1024.128 $ surface 6400 to 6300 sp2 D 0 1 si3 H 521.025 929.945 $ Surface 6050 to 6000 sp3 D 0 1 c c C Gamma Spectrum -Total Source Strength is 7.20E+18 C Table 8.6 Gamma Source Term (Fourteen Month) FC08513, R0 EC 68969 94

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 95 of 269 C by 18 Energy Group (FC08514, Page 45) c si4 H O.OOE+OO 2.00E-02 3.00E-02 4.50E-02 7 .00 E-02 1.00E-01 1.50E -0 1 3.00E-01 4.50E-01 7.00E-01 1.00E+00 1.50E+00 2.00E+00 2.50E+00 3.00E+00 4.00E+00 6.00E+00 8.00E+00 1.10E+01 sp4 D O.OOE+OO 2.02E+18 4.25E+17 5.14E+17 3.63E+17 2.53E+17 3.08E+17 2.33E+17 1.15E+17 2.46E+18 4.18E+17 7.48E+16 5.63E+15 5 . 65E+15 1 . 10E+14 9.97E+12 1 . 26E+10 1 .45 E+09 1 .6 7E+08 c c c Tally Definition c c c Point Detector Particle Fluence Tallies c c f15 :p -90000.00 -8950.00 3078.00 50 $ EAB SSWest f25 :p 2964.18 2758.44 1920.24 50 $ Control Room Proper c fc15 Gamma Dose Rate at EAB SSWest fc25 Gamma Dose Rate at interior of CR (Rm 77) c fm15 7 .20 E+18 fm25 7.20E+18 c c C Dose Conversion Factors for Photons (mrem/hr)/(particle/cm2-sec) c c deO 0.01 0.03 0.05 0.07 0.1 0.15 0 .2 0.25 0.3 0.35 0 . 4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1 .0 1 . 4 1 .8 2 . 2 2 .6 2.8 3.25 3 . 75 4.25 4.75 5.0 5.25 5 . 75 6 . 25 6.75 7.5 9.0 11.0 13 .0 15.0 dfO 3.96E-3 5.82E-4 2.90E-4 2.58E-4 2.83E-4 3.79E-4 5.01E-4 6.31E-4 7.59E-4 8.78E-4 9.85E-4 1 .08 E-3 1.17E-3 1 .2 7E-3 1.36E-3 1. 44E-3 1.52E-3 1.68E-3 1.98E-3 2.51E-3 2.99E-3 3.42E-3 3.82E-3 4 .0 1E-3 4.41E-3 4 .83 E-3 5.23E-3 5.60E-3 5.80E-3 6.01E-3 6.37E-3 6.74E-3 7.11E-3 7.66E-3 8.77E-3 1.03E- 2 1 . 18E-2 1.33E-2 c c C Peripheral cards c c PRDMP 4J 3 c c c Problem cutoff c c nps 5.0E09 c c FC08513, R0 EC 68969 95

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 96 of 269 FCS 18MOq EABCR. inp c Fort Calhoun Station SFP Gamma Dose Rates Following Drain Down c c cell cards c --------------------------------------------------------- c Fuel Handling Building c 10 1 -2.35 1400 -1550 1350 -1000 3250 -3220 $ Aux Bldg Basemat 15 1 -2.35 1400 -1550 1150 -1100 3060 -3015 $ North wall Rm 3 20 1 -2.35 1400 -1550 1050 -1000 3220 -3070 $ North wall Rm 69 25 1 -2.35 1700 -1850 1050 -1000 3100 -3030 $ North wall Rm 77 30 1 -2.35 1700 -1850 1650 -1600 3100 -3030 $ South wall Rm 77 32 1 -2.35 1800 -1850 1600 -1050 3100 -3030 $ East wall Rm 77 35 1 -2.35 1400 -1550 1350 -1300 3220 -3015 $ South wall Rm 3 40 1 -2.35 1400 -1450 1150 -1050 3220 -3070 $ West wall rm 69 401 1 -2.35 1400 -1450 3310 -1150 3220 -3015 $ West N wall rms 3 & 69 402 1 -2.35 1400 -1450 3320 -3310 3300 -3015 $ West wall above door rm 25 403 1 -2.35 1400 -1450 3320 -3310 3220 -7000 $ West wall below door rm 25 404 1 -2.35 1400 -1450 1300 -3320 3220 -3015 $ West S wall rm 25 405 4 -7.82 -3330 $ Railroad Siding Door 1004-lC 407 4 -7.82 -3340 3350 $ Railroad Siding Door 1004-lA 408 5 -0.0482 -3350 $ Railroad Siding Door 1004-lA insul 45 1 -2.35 1700 -1750 1600 -1050 3100 -3030 $ West wall rm 77 50 1 -2.35 1500 -1550 1300 -2600 3220 -3015 $ East wall Rm 3 501 1 -2.35 1500 -1550 2600 -1150 3070 -3015 $East wall Rm 69 above 1025' 55 1 -2.35 1500 -1550 2600 -1050 3220 -3100 $ East wall Rm 69 below 1025' 600 1 -2.35 1400 -1550 1350 -1100 3010 -3000 $Roof Segment Rm 3 1083'-1 601 1 -2.35 1400 -1550 1350 -1100 3011 -3010 $Roof Segment Rm 3 1083'-2 602 1 -2.35 1400 -1550 1350 -1100 3012 -3011 $Roof Segment Rm 3 1083'-3 603 1 -2.35 1400 -1550 1350 -1100 3013 -3012 $Roof Segment Rm 3 1083'-4 604 1 -2.35 1400 -1550 1350 -1100 3014 -3013 $Roof Segment Rm 3 1083'-S 605 1 -2.35 1400 -1550 1350 -1100 3015 -3014 $Roof Segment Rm 3 1083'-6 65 1 -2.35 1400 -1550 1150 -1000 3070 -3060 $ Roof Segment 1044A' 68 1 -2.35 1750 -1800 1600 -1050 3050 -3040 $ Floor Segment 1036' 69 1 -2.35 1700 -1850 1650 -1000 3030 -3020 $Roof Segment 1057' 70 1 -2.35 2350 -2200 2100 -2150 3220 -3200 $ SFP North wall 80 1 -2.35 2350 -2300 2450 -2100 3220 -3200 $ SFP West wall 90 1 -2.35 2250 -2200 2000 -2100 3220 -3200 $ SFP East wall 100 1 -2.35 2300 -2850 2000 -2050 3220 -3200 $ SFP South wall W 105 1 -2.35 2800 -2250 2000 -2050 3220 -3200 $ SFP South wall E 110 1 -2.35 2850 -2800 2000 -2050 3220 -3210 $ SFP South wall Weir 112 1 -2.35 2200 -1500 2000 -2600 3220 -3200 $ TC Northeast wall 114 1 -2.35 1550 -1750 2000 -2600 3220 -3060 $ TC Northeast extension

             #200 #220 115 1   -2.35 1550 -1750      2600 -1000 3110   -3100 $ Rm 69 Floor
             #200 #220 116 1   -2.35 1550 -1700      2600 -1000 3070   -3060 $ Rm 69 Ceiling
             #200 #220 117 1   -2.35 1550 -1700      1050 -1000 3100   -3070 $ Rm 69 North Wall CR 118 1   -2.35 1700      -1750 2600 -1650 3100   -3070 $ Rm 69 East Wall CR
             #200 #220 120 1   -2.35 2550      -1500 2450 -2000 3220   -3200 $ TC East wall 125 1   -2.35 2300      -2550 2450 -2400 3220   -3200 $ TC South wall 130 1   -2.35 2350      -2900 2750 -2450 3220   -3200 $ Rm 25A West wall 135 1   -2.35 2900      -1500 2750 -2700 3220   -3200 $ Rm 25A South wall FC08513, R0                           EC 68969                                    96

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 97 of 269 137 1 -2 . 35 1450 -1500 1300 -2750 3220 -7000 $ RR Siding Floor 140 1 -2.35 1450 -2350 2750 -1050 3110 -3100 $ Rm 67-69W-1025' elevation 142 1 -2 . 35 2350 -1500 2150 -1050 3110 -3100 $ Rm 69N-1025' elevation 143 1 -2 . 35 2200 -1500 2600 -2150 3110 -3100 $ Rm 69N-1025' elevation 150 2 -0.001205 1450 -2350 2750 -1050 7000 -3110 $ Area West of SFP below 1025' 151 2 -0 . 001205 1450 -2350 2750 -1050 3100 -3200 $ Area West of SFP above 1025' 155 2 -0 . 001205 1450 -1500 2750 - 1150 3200 -3015 $Area Above SFP 1038'6" 156 2 -0 . 001205 1500 -1550 2600 -1050 3100 -3070 $Area East of SFP below 1025' 160 2 -0 . 001205 2350 -1500 2150 -1050 7000 -3110 $Area North of SFP below 1025' 161 2 -0 . 001205 2350 -1500 2150 -1050 3100 -3200 $Area North of SFP above 1025' 162 2 -0.001205 1450 -1500 1150 -1050 3200 -3070 $ Upper Area North of SFP 164 2 -0 . 001 2 05 2200 -1500 2600 -2150 3220 -3110 $ Area East of SFP below 1025' 165 2 -0 . 001205 1550 - 1700 2600 -1050 3100 -3070 $ Area in Rm 69 East

               #200 #220 166  2    -0 . 001205 2200 -1500 2600 -2150          3100  -3200    $  Area East of SFP above 1025' 170  2    -0 . 001 2 05 1450 -1500 1300 -2750        7000  -3015    $  RR Siding South of Rm 25A 175  2    -0 . 001205 2900 -1500 2700 -2450          7000  -3200    $  Area in Rm 2 5A 185  2    -0 . 001205 2850 -2800 2000 -2050          3210  - 3200   $  Area in Weir Gate 190  2    -0 . 001205 2300 -2550 2400 -2000          3220  - 3200   $  Area in TC Gate c

195 2 -0 . 001205 1750 -1800 1600 -1050 3040 -3030 $Area in Room77 c c C Containment Building c 200 1 -2 .35 5100 -5000 $ Containment Structure 220 2 - 0.001205 -5100 $ Volume in Containment c C Spent Fuel Pool c 300 3 -2.568032 6 2 00 -2250 2050 -6300 6050 -6000 $ Fuel Mixture volume 310 2 -0.001205 2300 -2250 2050 -2100 6000 -3200 $ Area Above Fuel Racks 315 2 -0 . 001205 2300 -2250 2050 -2100 3220 -6050 $ Area Below Fuel Racks 320 2 -0 . 001205 2 300 -6200 2050 -2100 6050 -6000 $ Area West of Fuel Racks 325 2 -0.001205 6200 -2250 6300 -2100 6050 -6000 $ Area North of Fuel Racks c C Ground Plane Region c 8000 1 -2.35 (-7000 3260 -10000) $ Plant Grade Region

            #10 #20 #35 #40 #50 #55 #70 #80 #90 #100 #105 #110 #112
           #114 #1 2 0 #125 #130 #135 #137 #150 #160 #164 #170 #175 #190 #200
           #220 #300 #315 #320 #325 #401 #403 #404 8100      2     -0.001205 7100 7000 -7200 -10000                    $ 1004'6" Elevation Region
            #15 #20 #25 #30 #32 #35 #40 #45 #50 #55 #501 #65 #68
            #69 #70 #80 #90 #100 #105 #110 #112 #114 #115 #116 #117 #118 #120
           #125 #130 #135 #140 #142 #143 #150 #151 #155 #156 #160 #161 #162 #164
           #165 #166 #170 #175 #185 #190 #195 #200 #220 #300 #310 #320 #325
           #501 #401 #402 #404 #405 #407 #408 8200      1     -2.35 7000        -10000 - 7100   -7200 -10000      $ 1080' Elevation Region 8300      2     -0.001205 7200 -10000                      $Air Above 1080'
            #15 #35 #50 #600 #601 #602 #603 #604 #605 #155 #170
            #200 #220 #501 #401 #402 #404 c

9100 0 -3260 -10000 $ Universe Below Basemat c C void FC08513, R0 EC 68969 97

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 98 of 269 c 9999 0 10000 c C surface Cards c C Aux Building XY Planes c 1000 py 4084.60 $ North Outer Wall (Rms. 69 & 77) 1050 py 4038.60 $ North Inner Wall (Rms. 69 & 77) 1100 py 2133.60 $ North Outer Wall (Rms. 3 & 69) 1150 py 2087.88 $ North Inner Wall (Rms. 3 & 69) 1300 py -1844 . 04 $ South Inner Wall (Rms. 3 & 25) 1350 py -1889.76 $ South Outer Wall (Rms. 3 & 25) 1400 px -3810.00 $ West Outer Wall (Rms. 3 & 25) 1450 px -3764.28 $ West Inner Wall (Rms. 3 & 25) 1500 px -1889.76 $ East Inner Wall (Rms. 3 & 25) 1550 px -1844.04 $ East Outer Wall (Rms. 3 & 25) 1600 py 1988.82 $ South Inner CR Wall (Rm. 77) 1650 py 1950 . 72 $ South Outer CR Wall (Rm. 77) 1700 px 1135 . 38 $ West Outer CR Wall (Rm. 77) 1750 px 1181.10 $ West Inner CR Wall (Rm. 77) 1800 px 3550 . 92 $ East Inner CR Wall (Rm. 77) 1850 px 3596.64 $ East Outer CR Wall (Rm. 77) c C SFP c 2000 py -45.72 $ SFP South Outer wall 2050 py 76.20 $ SFP South Inner wall 2100 py 1089.66 $ SFP North Inner wall 2150 py 1257 . 30 $ SFP North Outer wall 2200 px -2011 . 68 $ SFP East Outer wall 2250 px -2179.32 $ SFP East Inner wall 2300 px -2806.70 $ SFP West Inner wall 2350 px -2974.34 $ SFP West Outer wall 2400 py -198.12 $ TC South Inner wall 2450 py -350.52 $ TC South Outer wall 2550 px -1969 . 77 $ TC East Inner wall 2600 py 175.26 $ TC North Outer wall 2700 py -1211.58 $ Rm 25A South Inner wall 2750 py -1257.30 $ Rm 25A South Outer wall 2800 px -2219.96 $ Weir Gate East Edge 2850 px -2418.08 $ Weir Gate West Edge 2900 px -2928.62 $ Rm 25A West Inner wall c C Aux Building Z Planes c 3000 pz 3169.92 $ Aux Bldg top of Roof Room 3 & 25 (1083') 3010 pz 3167.38 $ Aux Bldg roof- one inch, 3 & 25 (1083') 3011 pz 3164.84 $ Aux Bldg roof - two inches, 3 & 25 (1083') 3012 pz 3162.30 $ Aux Bldg roof - three inches, 3 & 25 (1083') 3013 pz 3159 . 76 $ Aux Bldg roof- four inches, 3 & 25 (1083') 3014 pz 3157 . 22 $ Aux Bldg roof- five inches, 3 & 25 (1083') 3015 pz 3154.68 $ Aux Bldg roof - six inches, 3 & 25 (1083') 3020 pz 2377.44 $ Aux Bldg top of Roof Room 77 (1057') 3030 pz 2331.72 $ Aux Bldg ceiling of Roof Room 77 (1057') FC08513, R0 EC 68969 98

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 99 of 269 3040 pz 1737 . 36 $ Aux Bldg floor slab Room 77 (1036') 3050 pz 1722 . 12 $ Aux Bldg floor slab thickness Room 77 (1036') 3060 pz 1981 . 20 $ Aux Bldg top of Roof Room 69 (1044') 3070 pz 1965.96 $ Aux Bldg ceiling of Roof Room 69 (1044') c 3100 pz 1402.08 $ Aux Bldg Room 69 Walkway Floor (1025') 3110 pz 1386.84 $ Aux Bldg Room 69 Walkway Ceiling (1025') c 3200 pz 1798.32 $Top of SFP Walkway (1038'6") 3210 pz 899.16 $Bottom of Weir Gate (1008'6") 3220 pz 502.92 $ Bottom of SFP (995'6") 3250 pz 137.16 $ Aux Bldg Basemat(983'6") c c 3300 pz 1432.56 $ Aux Bldg Door Top of Opening 3310 py -1266.19 $ Aux Bldg Door North side of Opening 3320 py -1753.87 $ Aux Bldg Door South side of Opening c c Rollup Door 1004-1C c 3330 rpp -3764 .3 7525 -3764.28 -1753 . 87 -1266.190 777.24 1432.56 c c c Vertical lift Door 1004-1A c 3340 rpp -3789.90 -3779 . 74 -1753 . 87 -1266.190 777.24 1432 . 56 3350 rpp -3789.70156 -3779 . 93844 -1753.67156 -1266.38844 777.43844 1432.36156 c 3230 pz 365 . 76 $ Top of Containment Basemat (991'0") 3260 pz 0.00 $ Bottom of Containment Basemat (979'0") c C Containment Building c c 5000 rcc 0. 0. 0. 0. 0 . 4343 . 07 1795.15 $ CAN Outer 5100 rcc 0. 0. 365.76 o. o. 3785 . 87 1677.35 $CAN Inner c c C Fuel Racks c c 6000 pz 929.945 $ Top of Fuel Racks 6050 pz 521.025 $ Bottom of Fuel Racks 6200 px -2770.022 $ West Boundary of Fuel Racks 6300 py 1024 . 128 $ North Boundary of Fuel Racks c c Optional planes of more detial of the CPA c c 6100 py -2183.343 $ East Boundary of Fuel Racks c 6200 py -2804.932 $ West Boundary of Fuel Racks c 6300 px 1081.522 $ North Boundary of Fuel Racks c 6400 px 83.911 $ South Boundary of Fuel Racks c 6500 px -2559.538 $ East Boundary of CPA c 6600 px 836.127 $ South Boundary of CPA FC08513, R0 EC 68969 99

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 100 of 269 c c Ground Pl ane c 7000 pz 777 . 24 $ Plant Grade Elevation 1 004'6 " 7100 px -46918.96 $ Distance to West Side of Bri d ge c 7100 px -46939.20 $ Distance to EAB 7 200 pz 3078.48 $ 1080 ' El evation c c Univ erse c c 1 0000 so 1448410.0 $ LP Z at 9 miles 10000 so 161000. 0 $ LPZ a t 1 mile c Data cards c c Physics De fi nition c mode p c c c Material Definition c c ordinary con crete = 2 . 35 g / cm3 c 95 Concrete, Ordinary (NBS 03) c m1 10 01 0 . 01 1914 $ H atoms/barn - em 6000 0 . 005899 $ c atoms/barn-em 80 1 6 0 . 041881 $ 0 atoms/barn - em 12000 0.001408 $ Mg atoms / barn - e m 13027 0 . 00189 2 $ Al atoms / barn - em 14000 0. 007311 $ Si atoms/barn - e m 16000 0 . 000131 $ S atoms/barn-em 19000 0 . 000061 $ K atoms/barn - em 20000 0 . 008719 $ Ca atoms/barn - em 260 0 0 0 . 000280 $ Fe atoms/barn - e m c c c c c Air 0 . 001205 g/cm3 c c m2 7 0 14 0.000039 $ N atoms/barn-em 80 1 6 0. 0000 11 $ 0 atoms/barn-em c C Fuel Reg i on Mi xtu re 2 . 568032 g / cm3 c M3 1302 7 -8.344615E- 03 $ Al Weight Percen t 1 8000 - 4 . 242021E-06 $ Ar Weight Percen t 5010 - 3 . 049262E-04 $ B10 Weight Percent 50 11 - 1. 361160E-03 $ B11 Weigh t Percen t 6 000 - 5.023089E-04 $ c We ight Pe rcent 17 000 - 5.747414 E-09 $ Cl Weight Percent 2 7 05 9 -3. 29 8 115E-05 $ Co Wei ght Pe rcen t 2 4 0 00 -2 . 0486 6 9E- 02 $ Cr We i ght Perc ent 290 0 0 -6.2 091 34 E-0 5 $ Cu Weight Pe rcen t FC08513, R0 EC 68969 100

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 101 of 269 26000 -6.672541E-02 $ Fe Weight Percent 25055 -1 . 966967E-03 $ Mn Weight Percent 42000 -4 . 783920E-05 $ Mo Weight Percent 7014 -2 . 698222E-04 $ N Weight Percent 11023 -5 . 747414E-09 $ Na Weight Percent 41093 -1 . 902357E-03 $ Nb Weight Percent 28000 -1 . 868235E-02 $ Ni Weight Percent 8016 -8.421467E-02 $ 0 Weight Percent 15031 -4. 292521E-05 $ p Weight Percent 82000 -3 . 735819E-08 $ Pb Weight Percent 16000 -3 . 621267E-05 $ s Weight Percent 14000 -1.214038E-03 $ Si Weight Percent 50000 -4 .166875E-06 $ Sn Weight Percent 73181 -6.522834E - 06 $ Ta Weight Percent 22000 -3.010457E-04 $ Ti Weight Percent 92234 -1.946356E-04 $ U-234 Weight Percent 92235 -2.183528E-02 $ U-235 Weight Percent 92236 -1.005027E-04 $ U-236 Weight Percent 92238 -6.017375E-01 $ U-238 Weight Percent 23000 -1.436854E-08 $ v Weight Percent 40000 -1.696187E-01 $ Zr Weight Percent c C carbon steel Door (density: 7.82 g/cm3) c M4 06000 0.001960 26000 0.083907 c C Door 1004-1 Insulation C From PNNL-15870 Rev 1: Material 249 C Polyisocyanurate (PIR, density: 0 . 048200 g/cm3) c M5 1001 0 . 00116 6000 0.00174 7014 0.000232 8016 0.000232 c C Variance Reduction c c imp:p 1 73r 0 0 c c c Source Definition c sdef X=d1 y=d2 z=d3 par=2 erg=d4 si1 H -2770.022 -2179 . 320 $ Surface 6200 to 2250 sp1 D 0 1 si2 H 76.200 1024.128 $ surface 6400 to 6300 sp2 D 0 1 si3 H 521.025 929.945 $ Surface 6050 to 6000 sp3 D 0 1 c c C Gamma Spectrum -Total Source Strength is 5 . 95E+18 C Table 8.5 Gamma Source Term (Eighteen Month) C by 18 Energy Group (FC08514, Page 44) FC08513, R0 EC 68969 101

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 102 of 269 c si4 H O.OOE+OO 2.00E-02 3.00E-02 4.50E-02 7.00E-02 1.00E-01 1 . 50E-01 3.00E-01 4 .50 E-01 7.00E-01 1.00E+00 1.50E+00 2.00E+00 2.50E+00 3.00E+00 4.00E+00 6.00E+00 8.00E+00 1 . 10E+01 sp4 D O.OOE+OO 1.67E+18 3.51E+17 4.25E+17 3.00E+17 2.10E+17 2.55E+17 1.93E+17 9 . 51E+16 2 . 04E+18 3.46E+17 6.19E+16 4 .65 E+15 4.67E+15 9.10E+13 8.24E+12 1.04E+10 1 .20 E+09 1.38E+08 c c c Tally Definition c c c Point Detector Particle Fluence Tallies c c f15:p -90000.00 -8950.00 3078 . 00 50 $ EAB SSWest f25:p 2964.18 2758.44 1920.24 50 $ Control Room Proper c fc15 Gamma Dose Rate at EAB SSWest fc25 Gamma Dose Rate at interior of CR (Rm 77) c fm15 5.95E+18 fm25 5.95E+18 c c C Dose Conversion Factors for Photons (mrem/hr)/(particle/cm2-sec) c c deO 0.01 0.03 0.05 0.07 0.1 0.15 0.2 0.25 0 .3 0.35 0.4 0.45 0.5 0.55 0.6 0.65 0.7 0.8 1.0 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 4.75 5.0 5.25 5.75 6.25 6.75 7 . 5 9.0 11.0 13 . 0 15 . 0 dfO 3.96E-3 5.82E-4 2.90E-4 2.58E-4 2.83E-4 3.79E-4 5.01E-4 6 .3 1E-4 7 .59 E-4 8.78E-4 9.85E-4 1.08E-3 1.17E-3 1 . 27E-3 1.36E-3 1.44E-3

1. 52E-3 1. 68E-3 1.98E-3 2.51E-3 2.99E-3 3.42E-3 3.82E-3 4.01E-3 4.41E-3 4.83E-3 5.23E-3 5.60E-3 5.80E-3 6.01E-3 6.37E-3 6.74E-3 7 . 11E-3 7.66E-3 8.77E-3 1.03E-2 1.18E-2 1. 33E-2 c

c C Peripheral cards c c PRDMP 4J 3 c c C Problem cutoff c c nps 5.0E09 c cc FC08513, R0 EC 68969 102

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 103 of 269 FCS 14MOn EABCR.inp c Fort Calhoun Station SFP Gamma Dose Rates Following Drain Down c c cell cards c C Fuel Handling Building c 10 1 -2. 3 5 1400 -1550 1350 -1000 3250 -3220 $ Aux Bldg Basemat 15 1 -2.35 1400 -1550 1150 -1100 3060 -3015 $ North wall Rm 3 20 1 -2. 3 5 1400 -1550 1050 -1000 3220 -3070 $ North wall Rm 69 25 1 -2.35 1700 -1850 1050 -1000 3100 -3030 $ North wall Rm 77 30 1 - 2.35 1700 -1850 1650 -1600 3100 -3030 $ South wall Rm 77 32 1 -2.35 1800 -1850 1600 -1050 3100 -3030 $ East wall Rm 77 35 1 -2 . 35 1400 -1550 1350 -1300 3220 -3015 $ South wall Rm 3 40 1 -2.35 1400 -1450 1150 -1050 3220 -3070 $ West wall rm 69 401 1 -2 . 35 1400 -1450 3310 -1150 3220 -3015 $ West N wall rms 3 & 69 402 1 -2 . 35 1400 -1450 3320 - 3310 3300 -3015 $ West wall above door rm 25 403 1 -2 . 35 1400 -1450 3320 -3310 3220 -7000 $ West wall below door rm 25 404 1 -2 . 35 1400 -1450 1300 -3320 3220 -3015 $ West S wall rm 25 405 4 -7.82 - 3330 $ Railroad Siding Door 1004-1C 407 4 -7.82 -3340 3350 $ Railroad Siding Door 1004-1A 408 5 -0.0482 -3350 $ Railroad Siding Door 1004-1A insul 45 1 - 2 .35 1700 -1750 1600 -1050 3100 -3030 $ West wall rm 77 50 1 -2.35 1500 -1550 1300 -2600 3220 -3015 $ East wall Rm 3 501 1 -2.35 1500 - 1550 2600 -1150 3070 - 3015 $East wall Rm 69 above 1025' 55 1 -2.35 1500 - 1550 2600 -1050 3220 -3100 $East wall Rm 69 below 1025' 600 1 -2 . 35 1400 -1550 1350 -1100 3010 -3000 $Roof Segment Rm 3 1083'-1 601 1 -2.35 1400 -1550 1350 -1100 3011 -3010 $Roof Segment Rm 3 1083'-2 602 1 -2 . 35 1400 - 1550 1350 -1100 3012 -3011 $Roof Segment Rm 3 1083'-3 603 1 -2 . 35 1400 -1550 1350 -1100 3013 -3012 $Roof Segment Rm 3 1083'-4 604 1 -2 . 35 1400 -1550 1350 -1100 3014 -3013 $Roof Segment Rm 3 1083'-5 605 1 -2 . 35 1400 -1550 1350 -1100 3015 -3014 $Roof Segment Rm 3 1083'-6 65 1 -2.35 1400 -1550 1150 - 1000 3070 -3060 $ Roof Segment 1044A' 68 1 -2.35 1750 -1800 1600 -1050 3050 -3040 $ Floor Segment 1036' 69 1 -2.35 1700 -1850 1650 -1000 3030 -3020 $Roof Segment 1057' 70 1 -2.35 2350 -2200 2100 - 2150 3220 -3 2 00 $ SFP North wall 80 1 -2.35 2350 -2300 2450 -2100 3220 -3200 $ SFP West wall 90 1 -2 . 35 2250 -2200 2000 -2100 3220 -3200 $ SFP East wall 100 1 -2 . 35 2300 - 2850 2000 -2050 3220 -3200 $ SFP South wall W 105 1 -2 . 35 2800 -2250 2000 -2050 3220 -3200 $ SFP South wall E 110 1 -2.35 2850 -2800 2000 -2050 3220 -3210 $ SFP South wall Weir 112 1 -2.35 2200 -1500 2000 -2600 3220 -3200 $ TC Northeast wall 114 1 -2.35 1550 -1750 2000 -2600 3220 -3060 $ TC Northeast extension

             #200 #220 115 1   -2.35 1550 -1750        2600 -1000  3110   -3100   $ Rm 69 Floor
             #200 #220 116 1   -2.35 1550 -1700        2600 -1000  3070   -3060   $ Rm 69 Ceiling
             #200 #220 117 1   -2.35 1550       -1700  1050 -1000  3100   -3070   $ Rm 69 North Wall CR 118 1   -2.35 1700       -1750  2600 -1650  3100   -3070   $ Rm 69 East Wall CR
             #200 #220 120 1   -2. 3 5 2550     -1500  2450 -2000  3220   -3200   $ TC East wall 125 1   -2 . 35 2300     -2550  2450 -2400  3220   -3200   $ TC South wall 130 1   -2 . 35 2350     -2900  2750 -2450  3220   -3200   $ Rm 25A West wall 135 1   -2 . 35 2900     -1500  2750 -2700  3220   -3200   $ Rm 25A South wall FC08513, R0                              EC 68969                                     103

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 104 of 269 137 1 -2.35 1450 -1500 1300 -2750 3220 -7000 $ RR Siding Floor 140 1 -2.35 1450 -2350 2750 -1050 3110 -3100 $ Rm 67-69W-1025' elevation 142 1 -2.35 2350 -1500 2150 -1050 3110 -3100 $ Rm 69N-1025' elevation 143 1 -2.35 2200 -1500 2600 -2150 3110 -3100 $ Rm 69N-1025' elevation 150 2 -0.001205 1450 -2350 2750 -1050 7000 -3110 $Area West of SFP below 1025' 151 2 -0 . 001205 1450 -2350 2750 -1050 3100 -3200 $Area West of SFP above 1025' 155 2 -0.001205 1450 -1500 2750 -1150 3200 -3015 $Area Above SFP 1038'6" 156 2 -0.001205 1500 -1550 2600 -1050 3100 -3070 $Area East of SFP below 1025' 160 2 -0.001205 2350 -1500 2150 -1050 7000 -3110 $Area North of SFP below 1025' 161 2 -0.001205 2350 -1500 2150 -1050 3100 -3200 $Area North of SFP above 1025' 162 2 -0 . 001205 1450 -1500 1150 -1050 3200 -3070 $ Upper Area North of SFP 164 2 -0.001205 2200 -1500 2600 - 2150 3220 -3110 $ Area East of SFP below 1025' 165 2 -0 . 001205 1550 -1700 2600 -1050 3100 -3070 $ Area in Rm 69 East

               #200 #220 166  2    -0.001205 2200 -1500 2600 -2150          3100  -3200   $Area East of SFP above 1025' 170  2    -0.001205 1450 -1500 1300 -2750          7000  -3015   $ RR Siding South of Rm 25A 175  2    -0.001205 2900 -1500 2700 -2450          7000  -3200   $ Area in Rm 25A 185  2    -0.001205 2850 -2800 2000 -2050          3210  -3200   $ Area in Weir Gate 190  2    -0.001205 2300 -2550 2400 -2000          3220  -3200   $ Area in TC Gate c

195 2 -0.001205 1750 -1800 1600 -1050 3040 -3030 $ Area in Room77 c C Containment Building c 200 1 -2.35 5100 -5000 $ Containment Structure 220 2 -0.001205 -5100 $ Volume in Containment c c Spent Fuel Pool c 300 3 -2.568032 6200 -2250 2050 -6300 6050 -6000 $ Fuel Mixture volume 310 2 -0.001205 2300 -2250 2050 -2100 6000 -3200 $ Area Above Fuel Racks 315 2 -0 . 001205 2300 -2250 2050 -2100 3220 -6050 $ Area Below Fuel Racks 320 2 -0.001205 2300 -6200 2050 -2100 6050 -6000 $ Area West of Fuel Racks 325 2 -0.001205 6200 -2250 6300 -2100 6050 -6000 $ Area North of Fuel Racks c C Ground Plane Region c 8000 1 -2 . 35 (-7000 3260 -10000) $ Plant Grade Region

            #10 #20 #35 #40 #50 #55 #70 #80 #90 #100 #105 #110 #112
           #114 #120 #125 #130 #135 #137 #150 #160 #164 #170 #175 #190 #200
           #220 #300 #315 #320 #325 #401 #403 #404 8100      2     -0.001205 7100 7000 -7200 -10000                 $ 1004'6" Elevation Region
            #15 #20 #25 #30 #32 #35 #40 #45 #50 #55 #501 #65 #68
            #69 #70 #80 #90 #100 #105 #110 #112 #114 #115 #116 #117 #118 #120
           #125 #130 #135 #140 #142 #143 #150 #151 #155 #156 #160 #161 #162 #164
           #165 #166 #170 #175 #185 #190 #195 #200 #220 #300 #310 #320 #325
           #501 #401 #402 #404 #405 #407 #408 8200      1     -2.35 7000      -10000 -7100    -7200 -10000     $ 1080' Elevation Region 8300      2     -0.001205 7200 -10000                    $Air Above 1080'
            #15 #35 #50 #600 #601 #602 #603 #604 #605 #155 #170
            #200 #220 #501 #401 #402 #404 c

9100 0 -3260 -10000 $ Universe Below Basemat c C void c FC08513, R0 EC 68969 104

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 105 of 269 9999 0 10000 c C surface Cards c C Aux Building XY Planes c 1000 py 4084.60 $ North Outer Wall (Rms. 69 & 77) 1050 py 4038.60 $ North Inner Wall (Rms. 69 & 77) 1100 py 2133.60 $ North Outer Wall (Rms. 3 & 69) 1150 py 2087 . 88 $ North Inner Wall (Rms. 3 & 69) 1300 py -1844.04 $ South Inner Wall (Rms. 3 & 25) 1350 py -1889.76 $ South Outer Wall (Rms. 3 & 25) 1400 px -3810.00 $ West Outer Wall (Rms . 3 & 25) 1450 px -3764.28 $ West Inner Wall (Rms . 3 & 25) 1500 px -1889.76 $ East Inner Wall (Rms. 3 & 25) 1550 px -1844.04 $ East Outer Wall (Rms. 3 & 25) 1600 py 1988.82 $ South Inner CR Wall (Rm. 77) 1650 py 1950.72 $ South Outer CR Wall (Rm. 77) 1700 px 1135.38 $ West Outer CR Wall (Rm. 77) 1750 px 1181 . 10 $ West Inner CR Wall (Rm. 77) 1800 px 3550 . 92 $ East Inner CR Wall (Rm. 77) 1850 px 3596 . 64 $ East Outer CR Wall (Rm. 77) c C SFP c 2000 py -45.72 $ SFP South Outer wall 2050 py 76.20 $ SFP South Inner wall 2100 py 1089.66 $ SFP North Inner wall 2150 py 1257.30 $ SFP North Outer wall 2200 px -2011.68 $ SFP East Outer wall 2250 px -2179.32 $ SFP East Inner wall 2300 px -2806.70 $ SFP West Inner wall 2350 px -2974.34 $ SFP West Outer wall 2400 py -198.12 $ TC South Inner wall 2450 py -350.52 $ TC South Outer wall 2550 px -1969.77 $ TC East Inner wall 2600 py 175.26 $ TC North Outer wall 2700 py -1211.58 $ Rm 25A South Inner wall 2750 py -1257.30 $ Rm 25A South Outer wall 2800 px -2219.96 $ Weir Gate East Edge 2850 px -2418.08 $ Weir Gate West Edge 2900 px -2928.62 $ Rm 25A West Inner wall c C Aux Building Z Planes c 3000 pz 3169 . 92 $ Aux Bldg top of Roof Room 3 & 25 (1083') 3010 pz 3167.38 $ Aux Bldg roof - one inch, 3 & 25 (1083') 3011 pz 3164.84 $ Aux Bldg roof - two inches, 3 & 25 (1083') 3012 pz 3162 . 30 $ Aux Bldg roof - three inches, 3 & 25 (1083') 3013 pz 3159.76 $ Aux Bldg roof - four inches, 3 & 25 (1083') 3014 pz 3157.22 $ Aux Bldg roof- five inches, 3 & 25 (1083') 3015 pz 3154.68 $ Aux Bldg roof - six inches, 3 & 25 (1083') 3020 pz 2377.44 $ Aux Bldg top of Roof Room 77 (1057') 3030 pz 2331.72 $ Aux Bldg ceiling of Roof Room 77 (1057') 3040 pz 1737.36 $ Aux Bldg floor slab Room 77 (1036') FC08513, R0 EC 68969 105

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 106 of 269 3050 pz 1722.12 $ Aux Bldg floor slab thickness Room 77 (1036') 3060 pz 1981.20 $ Aux Bldg top of Roof Room 69 (1044') 3070 pz 1965.96 $ Aux Bldg ceiling of Roof Room 69 (1044') c 3100 pz 1402 . 08 $ Aux Bldg Room 69 Walkway Floor (1025') 3110 pz 1386.84 $ Aux Bldg Room 69 Walkway Ceiling (1025') c 3200 pz 1798.32 $Top of SFP Walkway (1038'6") 3210 pz 899.16 $Bottom of Weir Gate (1008'6") 3220 pz 502.92 $Bottom of SFP (995'6") 3250 pz 137.16 $ Aux Bldg Basemat(983'6") c c 3300 pz 1432.56 $ Aux Bldg Door Top of Opening 3310 py -1266.19 $ Aux Bldg Door North side of Opening 3320 py -1753.87 $ Aux Bldg Door South side of Opening c c Rollup Door 1004-1C c 3330 rpp -3764 .3 7525 -3764.28 -1753.87 -1266.190 777.24 1432.56 c c c Vertical lift Door 1004-1A c 3340 rpp -3789.90 -3779.74 -1753.87 -1266.190 777.24 1432.56 3350 rpp -3789.70156 -3779.93844 -1753 . 67156 -1266.38844 777 . 43844 1432.36156 c 3230 pz 365.76 $ Top of Containment Basemat (991'0") 3260 pz 0.00 $ Bottom of Containment Basemat (979'0") c C Containment Building c c 5000 rcc 0. 0 . 0. 0. 0. 4343 . 07 1795.15 $ CAN Outer 5100 rcc 0 . 0 . 365.76 0. 0. 3785.87 1677.35 $ CAN Inner c C Fuel Racks c c 6000 pz 929.945 $ Top of Fuel Racks 6050 pz 521 . 025 $ Bottom of Fuel Racks 6200 px -2770.022 $ West Boundary of Fuel Racks 6300 py 1024 . 128 $ North Boundary of Fuel Racks c c Optional planes of more detial of the CPA c c 6100 py -2 183.343 $ East Boundary of Fuel Racks c 6200 py -2804.932 $ West Boundary of Fuel Racks c 6300 px 1081 . 522 $ North Boundary of Fuel Racks c 6400 px 83.911 $ South Boundary of Fuel Racks c 6500 px -2559.538 $ East Boundary of CPA c 6600 px 836.127 $ South Boundary of CPA c c Ground Plane FC08513, R0 EC 68969 106

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 107 of 269 c 7000 pz 777024 $ Plant Grade Elevation 1004'6" 7100 px -46918096 $ Distance to West Side of Bridge c 7100 px -46939020 $ Distance to EAB 7200 pz 3 07 8 048 $ 1080' Elevation c C Universe c c 10000 so 144841000 $ LPZ at 9 miles 10000 so 161000 00 $ LPZ at 1 mile c Data cards c c Physics Definition c mode n c c Material Definition c C ordinary concrete = 2 035 g/cm3 C 95 Concrete, Ordinary (NBS 03) c m1 1001 0 0011914 $ H atoms/barn-em 6000 Oo005899 $ c atoms/barn-em 8016 Oo041881 $ 0 atoms/barn-em 12000 00001408 $ Mg atoms/barn-em 13027 00001892 $ Al atoms/barn-em 14000 0 0007311 $ Si atoms/barn-em 16000 Oo000131 $ S atoms/barn-em 19000 Oo000061 $ K atoms/barn-em 20000 Oo008719 $ Ca atoms/barn-em 26000 00000280 $ Fe atoms/barn-em c c c c c Air Oo001205 g/ c m3 c c m2 7014 Oo000039 $ N atoms/barn-em 8016 Oo000011 $ 0 atoms/barn-em c c Fuel Region Mixture 2 0568032 g/cm3 c M3 13027 -8o344615E-03 $ Al Weight Percent 18000 -4 o242021E-06 $ Ar Weight Percent 5010 -3 0049262E-04 $ B10 Weight Percent 5011 -1 0361160E-03 $ B11 Weight Percent 6000 -5 o023089E-04 $ c Weight Percent 17000 -5o747414E-09 $ Cl Weight Percent 27059 -3 0298115E-05 $ Co Weight Percent 24000 -2o048669E-02 $ Cr Weight Percent 29000 -6 o209134E-05 $ Cu Weight Percent 26000 -6o672541E-02 $ Fe Weight Percent 25055 -10966967E-03 $ Mn Weight Percent 42000 -4 o783920E-05 $ Mo Weight Percent FC08513, R0 EC 68969 107

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 108 of 269 7014 -2 . 698222E-04 $ N Weight Percent 11023 -5.747414E-09 $ Na Weight Percent 41093 -1.902357E-03 $ Nb Weight Percent 28000 -1.868235E-02 $ Ni Weight Percent 8016 -8.421467E-02 $ 0 Weight Percent 15031 -4. 292521E-05 $ p Weight Percent 82000 -3.735819E-08 $ Pb Weight Percent 16000 -3.621267E-05 $ s Weight Percent 14000 -1.214038E-03 $ Si Weight Percent 50000 -4.166875E-06 $ Sn Weight Percent 73181 -6 . 522834E-06 $ Ta Weight Percent 22000 -3 . 010457E-04 $ Ti Weight Percent 92234 -1 . 946356E-04 $ U-234 Weight Percent 92235 -2 . 183528E-02 $ U-235 Weight Percent 92236 -1.005027E-04 $ U-236 Weight Percent 92238 -6.017375E-01 $ U-238 Weight Percent 23000 -1.436854E-08 $ v Weight Percent 40000 -1.696187E-01 $ Zr Weight Percent c C carbon steel Door (density : 7.82 g/cm3) c M4 06000 0.001960 26000 0.083907 c c Door 1004-1 Insulation C From PNNLpnnl-15870 Rev 1: Material 249 C Polyisocyanurate (PIR, density : 0.048200 g/cm3) c M5 1001 0.00116 6000 0.00174 7014 0.000232 8016 0 . 000232 c C Variance Reduction c c imp:n 1 73r 0 0 c c c Source Definition c sdef X=d1 y=d2 z=d3 par=1 erg=d4 si1 H -2770.022 -2179.320 $ Surface 6200 to 2250 sp1 D 0 1 si2 H 76.200 1024.128 $ surface 6400 to 6300 sp2 D 0 1 si3 H 521.025 929 . 945 $ Surface 6050 to 6000 sp3 D 0 1 c c C Neutron Spectrum -Total Source Strength is 2.36E+11 C Table 8.7 Neutron Source Term (Eighteen Month) C by 44 Energy Group (FC08514, page 46) c si4 H 1 . 00E-11 3 . 00E-09 7.50E-09 1.00E-08 2.53E-08 3 . 00E-08 4 . 00E-08 5.00E-08 7.00E-08 1.00E-07 FC08513, R0 EC 68969 108

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 109 of 269 1.50E-07 2.00E-07 2.25E-07 2 . 50E-07 2.75E-07 3.25E-07 3.50E-07 3.75E-07 4.00E-07 6.25E-07 1.00E-06 1.77E-06 3.00E-06 4.75E-06 6 . 00E-06 8.10E-06 1.00E-05 3.00E-05 1.00E-04 5.50E-04 3.00E-03 1.70E-02 2.50E-02 1.00E-01 4 . 00E-01 9.00E-01 1.40E+00 1 . 85E+00 2.35E+00 2 . 48E+00 3.00E+00 4.80E+00 6 . 43E+00 8.19E+00 2.00E+01 sp4 D O.OOE+OO 3.63E-03 2.61E-03 2.06E-03 6.06E-03 2.51E-03 8.78E-03 6.28E-03 1. 35E-02 2.24E-02 4.27E-02 5.05E-02 3.08E-02 4.02E-02 3.37E-02 5.88E+00 2.03E+00 2.10E+00 2.18E+00 1 . 92E+01 4.57E+01 1.28E+02 2.53E+02 4.69E+02 4 . 05E+02 7.62E+02 7 . 81E+02 1.21E+04 7.66E+04 1 . 09E+06 1 . 39E+07 1.87E+08 1.58E+08 2.45E+09 1 . 77E+10 3 . 87E+10 3 . 86E+10 3 . 09E+10 2.91E+10 6.31E+09 2.25E+10 4 . 09E+10 1 . 16E+10 3.69E+09 1.27E+09 c c c Tally Definition c c c Point Detector Particle Fluence Tallies c c f15 : n - 90000.00 -8950 . 00 3078.00 50 $ EAB SSWest f25 : n 2964 . 18 2758.44 1920.24 50 $ Point X Control Room Proper c fc15 Gamma Dose Rate at EAB SSWest fc25 Gamma Dose Rate at interior of CR (Rm 77) c fm15 2.44E+11 fm25 2.44E+11 c C Dose Conversion Factors for Neutrons (mRem/hr)/(particle/cm2-sec) c c deO 2 . 5E-08 1 . 0E-07 1.0E-06 1.0E-05 1.0E-04 1.0E-03 1.0E-02 1 . 0E-01 5 . 0E-01 1.0E+00 2.5E+00 5 . 0E+00 7.0E+00 1.0E+01 1 . 4E+01 2 . 0E+01 c dfO 3.67E-03 3.67E-03 4 . 46E-03 4.54E-03 4.18E-03 3 . 76E-03 3.56E-03 2.17E-02 9.26E-02 1.32E-01 1.25E-01 1.56E-01 1.47E-01 1.47E-01 2.08E-01 2.27E-01 c c C Peripheral cards c c PRDMP 4J 3 c c C Problem cutoff c c nps 2.0E06 c FC08513, R0 EC 68969 109

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 11 0 of 269 FCS 18MOn EABCR.inp c Fort Calhoun Station SFP Gamma Dose Rates Following Drain Down c c cell cards c --------- - ----- - ---------------- - --- - ------ - -------- ---- - c Fuel Handling Building c 10 1 -2.35 1400 -1550 1350 -1000 3250 -3220 $ Aux Bldg Basemat 15 1 -2.35 1400 -1550 1150 -1100 3060 -3015 $ North wall Rm 3 20 1 -2.35 1400 -1550 1050 -1000 3220 -3070 $ North wall Rm 69 25 1 -2.35 1700 -1850 1050 -1000 3100 -3030 $ North wall Rm 77 30 1 -2.35 1700 -1850 1650 -1600 3100 -3030 $ South wall Rm 77 32 1 -2.35 1800 -1850 1600 -1050 3100 -3030 $ East wall Rm 77 35 1 -2.35 1400 -1550 1350 -1300 3220 -3015 $ South wall Rm 3 40 1 -2.35 1400 -1450 1150 -1050 3220 -3070 $ West wall rm 69 401 1 -2.35 1400 -1450 3310 -1150 3220 -3015 $ West N wall rms 3 & 69 402 1 -2 . 35 1400 -1450 3320 -3310 3300 -3015 $ West wall above door rm 25 403 1 -2.35 1400 -1450 3320 -3310 3220 -7000 $ West wall below door rm 25 404 1 - 2. 35 1400 -1450 1300 -3320 3220 -3015 $ West S wall rm 25 405 4 -7 . 82 -3330 $ Railroad Siding Door 1004-1C 407 4 -7.8 2 - 3340 3 350 $ Railroad Siding Door 1004-1A 408 5 -0.0482 -3350 $ Railroad Siding Door 1004-1A insul 45 1 -2.35 1700 -1750 1600 -1050 3100 -3030 $ West wall rm 77 50 1 -2.35 1500 -1550 1300 -2600 3220 -3015 $ East wall Rm 3 501 1 -2.35 1500 -1550 2600 -1150 3070 -3015 $ East wall Rm 69 above 1025' 55 1 -2.35 1500 -1550 2600 -1050 3220 -3100 $ East wall Rm 69 below 1025' 600 1 -2.35 1400 -1550 1350 -1100 3010 -3000 $ Roof Segment Rm 3 1083'-1 601 1 -2.35 1400 -1550 1350 -1100 3011 -3010 $ Roof Segment Rm 3 1083'-2 602 1 -2.35 1400 -1550 1350 -1100 3012 -3011 $ Roof Segment Rm 3 1083'-3 603 1 - 2.35 1400 - 1550 1350 -1100 3013 -3012 $ Roof Segment Rm 3 1083'-4 604 1 -2.35 1400 -1550 1350 - 1100 3014 -3013 $ Roof Segment Rm 3 1083'-5 605 1 -2. 3 5 1400 -1550 1350 -1100 3015 -3014 $ Roof Segment Rm 3 1083'-6 65 1 -2 . 35 1400 -1550 1150 -1000 3070 - 3060 $ Roof Segment 1044A' 68 1 -2 . 35 1750 -1800 1600 -1050 3050 - 3040 $ Floor Segment 1036' 69 1 -2 . 35 1700 -1850 1650 -1000 3030 -3020 $ Roof Segment 1057' 70 1 -2 . 35 2350 -2200 2100 -2150 3220 -3200 $ SFP North wall 80 1 -2 . 35 2350 -2300 2450 -2100 3220 - 3200 $ SFP West wall 90 1 -2 . 35 2250 -2200 2000 -2100 3220 -3200 $ SFP East wall 100 1 -2 . 35 2300 -2850 2000 -2050 3220 -3200 $ SFP South wall W 105 1 - 2 .35 2800 -2250 2000 -2050 3220 -3200 $ SFP South wall E 110 1 -2.35 2850 -2800 2000 -2050 3220 -3210 $ SFP South wall Weir 112 1 -2.35 2200 -1500 2000 -2600 3220 -3200 $ TC Northeast wall 114 1 -2.35 1550 -1750 2000 -2600 3220 -3060 $ TC Northeast extension

               #200 #220 115 1    -2.35 1550       -1750  2600  -1000    3110    -3100    $ Rm 69 Floor
               #200 #220 116 1    -2.35 1550       -1700  2600  -1000    3070    -3060    $ Rm 69 Ceiling
               #200 #220 117 1    -2 . 35 1550 -1700      1050  -1000    3100    -3070    $ Rm 69 North Wall CR 118 1    - 2 . 35 1700 -1750     2600  -1650    3100    -3070    $ Rm 69 East Wall CR
               #200 #220 120  1   -2.35 2550       -1500  2450  -2000    3220    -3200    $ TC  East wall 125  1   -2.35 2300       - 2550 2450  -2400    3220    -3200    $ TC  South wall 130  1   -2.35 2350       -2900  2750  -2450    3220    -3200    $ Rm  25A West wall 135  1   -2.35 2900       -1500  2750  -2700    3220    -3200    $ Rm  25A South wall FC08513, R0                                  EC 68969                                      110

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 111 of 269 137 1 -2.35 1450 -1500 1300 -2750 3220 -7000 $ RR Siding Floor 140 1 -2.35 1450 -2350 2750 -1050 3110 -3100 $ Rm 67-69W-1025' elevation 142 1 -2.35 2350 -1500 2150 -1050 3110 -3100 $ Rm 69N-1025' elevation 143 1 -2.35 2200 -1500 2600 -2150 3110 -3100 $ Rm 69N-1025' elevation 150 2 -0.001205 1450 -2350 2750 -1050 7000 -3110 $Area West of SFP below 1025' 151 2 -0.001205 1450 -2350 2750 -1050 3100 -3200 $Area West of SFP above 1025' 155 2 -0.001205 1450 -1500 2750 -1150 3200 -3015 $Area Above SFP 1038'6" 156 2 -0.001205 1500 -1550 2600 -1050 3100 -3070 $Area East of SFP below 1025' 160 2 -0.001205 2350 -1500 2150 -1050 7000 -3110 $Area North of SFP below 1025' 161 2 -0.001205 2350 -1500 2150 -1050 3100 -3200 $ Area North of SFP above 1025' 162 2 -0.001205 1450 -1500 1150 -1050 3200 -3070 $ Upper Area North of SFP 164 2 -0.001205 2200 -1500 2600 -2150 3220 -3110 $Area East of SFP below 1025' 165 2 -0.001205 1550 -1700 2600 -1050 3100 -3070 $ Area in Rm 69 East

              #200 #220 166  2    -0.001205 2200 -1500 2600 -2150        3100   -3200   $Area East of SFP above 1025' 170  2    -0.001205 1450 -1500 1300 -2750        7000   -3015   $ RR Siding South of Rm 25A 175  2    -0.001205 2900 -1500 2700 -2450        7000   -3200   $ Area in Rm 25A 185  2    -0.001205 2850 -2800 2000 -2050        3210   -3200   $ Area in Weir Gate 190  2    -0.001205 2300 -2550 2400 -2000        3220   -3200   $ Area in TC Gate c

195 2 -0.001205 1750 -1800 1600 -1050 3040 -3030 $ Area in Room77 c c C Containment Building c 200 1 -2.35 5100 -5000 $ Containment Structure 220 2 -0.001205 -5100 $ Volume in Containment c C Spent Fuel Pool c 300 3 -2.568032 6200 -2250 2050 -6300 6050 -6000 $ Fuel Mixture volume 310 2 -0.001205 2300 -2250 2050 -2100 6000 -3200 $ Area Above Fuel Racks 315 2 -0.001205 2300 -2250 2050 -2100 3220 -6050 $ Area Below Fuel Racks 320 2 -0.001205 2300 -6200 2050 -2100 6050 -6000 $ Area West of Fuel Racks 325 2 -0.001205 6200 -2250 6300 -2100 6050 -6000 $ Area North of Fuel Racks c C Ground Plane Region c 8000 1 -2.35 (-7000 3260 -10000) $ Plant Grade Region

            #10 #20 #35 #40 #50 #55 #70 #80 #90 #100 #105 #110 #112
           #114 #120 #125 #130 #135 #137 #150 #160 #164 #170 #175 #190 #200
           #220 #300 #315 #320 #325 #401 #403 #404 8100      2    -0.001205 7100 7000 -7200 -10000                 $ 1004'6" Elevation Region
            #15 #20 #25 #30 #32 #35 #40 #45 #50 #55 #501 #65 #68
            #69 #70 #80 #90 #100 #105 #110 #112 #114 #115 #116 #117 #118 #120
           #125 #130 #135 #140 #142 #143 #150 #151 #155 #156 #160 #161 #162 #164
           #165 #166 #170 #175 #185 #190 #195 #200 #220 #300 #310 #320 #325
           #501 #401 #402 #404 #405 #407 #408 8200      1    -2.35    7000   -10000 -7100   -7200 -10000      $ 1080' Elevation Region 8300      2    -0 . 001205 7200 -10000                  $Air Above 1080'
            #15 #35 #SO #600 #601 #602 #603 #604 #605 #155 #170
            #200 #220 #501 #401 #402 #404 c

9100 0 -3260 -10000 $ Universe Below Basemat c C void FC08513, R0 EC 68969 111

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 112 of 269 c 9999 0 10000 c C surface Cards c C Aux Building XY Planes c 1000 py 4084.60 $ North Outer Wall (Rms . 69 & 77) 1050 py 4038.60 $ North Inner Wall (Rms . 69 & 77) llOO py 2133.60 $ North Outer Wall (Rms . 3 & 69) l150 py 2087.88 $ North Inner Wall (Rms. 3 & 69) 1300 py -1844.04 $ South Inner Wall (Rms. 3 & 25) 1350 py -1889.76 $ South Outer Wall (Rms. 3 & 25) 1400 px -3810.00 $ West Outer Wall (Rms. 3 & 25) 1450 px -3764.28 $ West Inner Wall (Rms. 3 & 25) 1500 px -1889.76 $ East Inner Wall (Rms . 3 & 25) 1550 px -1844.04 $ East Outer Wall (Rms. 3 & 25) 1600 py 1988.82 $ South Inner CR Wall (Rm . 77) 1650 py 1950.72 $ South Outer CR Wall (Rm. 77) 1700 px l135. 38 $ West Outer CR Wall (Rm. 77) 1750 px l181. 10 $ West Inner CR Wall (Rm. 77) 1800 px 3550.92 $ East Inner CR Wall (Rm. 77) 1850 px 3596.64 $ East Outer CR Wall (Rm. 77) c C SFP c 2000 py -45.72 $ SFP South Outer wall 2050 py 76.20 $ SFP South Inner wall 2100 py 1089.66 $ SFP North Inner wall 2150 py 1257.30 $ SFP North Outer wall 2200 px -2011 . 68 $ SFP East Outer wall 2250 px -2179.32 $ SFP East Inner wall 2300 px -2806.70 $ SFP West Inner wall 2350 px -2974.34 $ SFP West Outer wall 2400 py -198.12 $ TC South Inner wall 2450 py -350.52 $ TC South Outer wall 2550 px -1969.77 $ TC East Inner wall 2600 py 175.26 $ TC North Outer wall 2700 py -12l1.58 $ Rm 25A South Inner wall 2750 py -1257.30 $ Rm 25A South Outer wall 2800 px -2219.96 $ Weir Gate East Edge 2850 px -2418.08 $ Weir Gate West Edge 2900 px -2928.62 $ Rm 25A West Inner wall c C Aux Building Z Planes c 3000 pz 3169.92 $ Aux Bldg top of Roof Room 3 & 25 (1083') 3010 pz 3167.38 $ Aux Bldg roof- one inch, 3 & 25 (1083') 3011 pz 3164.84 $ Aux Bldg roof - two inches, 3 & 25 (1083') 3012 pz 3162 . 30 $ Aux Bldg roof - three inches, 3 & 25 (1083') 3013 pz 3159.76 $ Aux Bldg roof - four inches, 3 & 25 (1083') 3014 pz 3157.22 $ Aux Bldg roof - five inches, 3 & 25 (1083') 3015 pz 3154 . 68 $ Aux Bldg roof- six inches, 3 & 25 (1083') 3020 pz 2377.44 $ Aux Bldg top of Roof Room 77 (1057') 3030 pz 2331 . 72 $ Aux Bldg ceiling of Roof Room 77 (1057') FC08513, R0 EC 68969 112

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 113 of 269 3040 pz 1737.36 $ Aux Bldg floor slab Room 77 (1036 1 ) 3050 pz 1722.12 $ Aux Bldg floor slab thickness Room 77 (1036 1

                                                                                    )

3060 pz 1981.20 $ Aux Bldg top of Roof Room 69 (1044 1 ) 3070 pz 1965.96 $ Aux Bldg ceiling of Roof Room 69 (1044 1 ) c 3100 pz 1402.08 $ Aux Bldg Room 69 Walkway Floor (1025 1 ) 3110 pz 1386.84 $ Aux Bldg Room 69 Walkway Ceiling (1025 1

                                                                               )

c 3200 pz 1798.32 $Top of SFP Walkway (1038 6") 1 3210 pz 899.16 $Bottom of Weir Gate (1008 1 6") 3220 pz 502 . 92 $ Bottom of SFP (995 1 6") 3250 pz 137.16 $ Aux Bldg Basemat(983 1 6") c c 3300 pz 1432 . 56 $ Aux Bldg Door Top of Opening 3310 py -1266 . 19 $ Aux Bldg Door North side of Opening 3320 py -1753 . 87 $ Aux Bldg Door South side of Opening c c Rollup Door 1004-1C c 3330 rpp -3764 . 37525 -3764.28 -1753.87 -1266.190 777.24 1432.56 c c c Vertical lift Door 1004-1A c 3340 rpp -3789.90 -3779 . 74 -1753 . 87 -1266 . 190 777.24 1432 . 56 3350 rpp - 3789.70156 -3779 . 93844 -1753.67156 -1266.38844 777.43844 1432 . 36156 c 3230 pz 365.76 $ Top of Containment Basemat (991 I 0 11 ) 3260 pz 0.00 $ Bottom of Containment Basemat (979 I 0 10 ) c c Containment Building c c 5000 rcc 0. 0. 0 . 0. 0. 4343.07 1795.15 $ CAN Outer 5100 rcc 0. 0 . 365.76 0. 0. 3785.87 1677.35 $ CAN Inner c c c Fuel Racks c c 6000 pz 929.945 $ Top of Fuel Racks 6050 pz 521.025 $ Bottom of Fuel Racks 6200 px -2770.022 $ West Boundary of Fuel Racks 6300 py 1024.128 $ North Boundary of Fuel Racks c c Optional planes of more detial of the CPA c c 6100 PY -2183.343 $ East Boundary of Fuel Racks c 6200 py -2804.932 $ West Boundary of Fuel Racks c 6300 px 1081.522 $ North Boundary of Fuel Racks c 6400 px 83.911 $ South Boundary of Fuel Racks c 6500 px -2559.538 $ East Boundary of CPA c 6600 px 836.127 $ South Boundary of CPA FC08513, R0 EC 68969 113

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 114 of 269 c c Ground Plane c 7000 pz 777.24 $ Plant Grade Elevation 1004'6" 7100 px -46918.96 $ Distance to West Side of Bridge c 7100 px -46939.20 $ Distance to EAB 7200 pz 3078.48 $ 1080' Elevation c C Universe c c 10000 so 1448410 . 0 $ LPZ at 9 miles 10000 so 161000 . 0 $ LPZ at 1 mile c Data cards c c Physics Definition c mode n c c c Material Definition c c ordinary concrete = 2.35 g/cm3 c 95 Concrete, Ordinary (NBS 03) c m1 1001 0 . 011914 $ H atoms/barn-em 6000 0.005899 $ c atoms/barn-em 8016 0 . 041881 $ 0 atoms/barn-em 12000 0.001408 $ Mg atoms/barn-em 13027 0.001892 $ Al atoms/barn-em 14000 0. 007311 $ Si atoms/barn-em 16000 0.000131 $ S atoms/barn-em 19000 0.000061 $ K atoms/barn-em 20000 0.008719 $ Ca atoms/barn-em 26000 0.000280 $ Fe atoms/barn-em c c C Air 0.001205 g/cm3 c c m2 7014 0.000039 $ N atoms/barn-em 8016 0 . 000011 $ 0 atoms/barn-em c C Fuel Region Mixture 2 . 568032 g/cm3 c M3 13027 -8.344615E-03 $ Al Weight Percent 18000 -4.242021E-06 $ Ar Weight Percent 5010 -3.049262E-04 $ B10 Weight Percent 5011 -1.361160E-03 $ B11 Weight Percent 6000 -5.023089E-04 $ c Weight Percent 17000 -5 . 747414E-09 $ Cl Weight Percent 27059 -3 . 298115E-05 $ Co Weight Percent 24000 -2 . 048669E-02 $ Cr Weight Percent 29000 -6. 209134E-05 $ Cu Weight Percent 26000 -6. 672541E-02 $ Fe Weight Percent 25055 -1 . 966967E-03 $ Mn Weight Percent FC08513, R0 EC 68969 114

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 115 of 269 42000 -4.783920E-05 $ Mo Weight Percent 7014 -2.698222E-04 $ N Weight Percent 11023 -5 . 747414E-09 $ Na Weight Percent 41093 -1.902357E-03 $ Nb Weight Percent 28000 -1.868235E-02 $ Ni Weight Percent 8016 -8.421467E-02 $ 0 Weight Percent 15031 -4.292521E-05 $ p Weight Percent 82000 -3.735819E-08 $ Pb Weight Percent 16000 -3.621267E-05 $ s Weight Percent 14000 -1.214038E-03 $ Si Weight Percent 50000 -4.166875E-06 $ Sn Weight Percent 73181 -6.522834E-06 $ Ta Weight Percent 2 2 000 -3.010457E-04 $ Ti Weight Percent 92234 -1 . 946356E-04 $ U-234 Weight Percent 92235 -2.183528E-02 $ U-235 Weight Percent 92236 -1.005027E-04 $ U-236 Weight Percent 92238 -6.017375E-01 $ U-238 Weight Percent 23000 - 1 . 436854E - 08 $ v Weight Percent 40000 - 1.696187E-01 $ Zr Weight Percent c C carbon steel Door (density : 7.82 g/cm3) c M4 06000 0.001960 26000 0 . 083907 c C Door 1004-1 Insulation C From PNNLpnnl-15870 Rev 1 : Material 249 c Polyisocyanurate (PIR, density : 0 . 048200 g/cm3) c M5 1001 0.00116 6000 0 . 00174 7014 0.000232 8016 0.000232 c c Variance Reduction c c imp : n 1 73r 0 0 c c c Source Definition c sdef X=d1 y=d2 z=d3 par=1 erg=d4 si1 H -2770.022 -2179 . 320 $ Surface 6200 to 2250 sp1 D 0 1 si2 H 76 . 200 1024 . 128 $ surface 6400 to 6300 sp2 D 0 1 si3 H 521.025 929.945 $ Surface 6050 to 6000 sp3 D 0 1 c c C Neutron Spectrum -Total Source Strength is 2.36E+11 C Table 8 . 7 Neutron Source Term (Eighteen Month) C by 44 Energy Group (FC08514, page 46) c si4 H 1.00E-11 3 . 00E-09 7.50E-09 1.00E-08 2.53E-08 FC08513, R0 EC 68969 115

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 116 of 269 3.00E - 08 4.00E-08 5.00E-08 7.00E-08 1.00E-07 1.50E - 07 2.00E-0 7 2.25E-07 2.50E-07 2.75E-07 3.25E - 07 3.50E-07 3. 7 5E-0 7 4.00E-07 6.25E-07 1 . 00E-06 1.77E-06 3.00E-06 4. 7 5E - 06 6 . 00E-06 8.10E-06 1.00E-05 3.00E-05 1.00E - 04 5 . 50E-04 3.00 E -03 1.70E-02 2.50E-02 1. 00E-01 4 . 00E-0 1 9 . 00 E -01 1. 40E+00 1.85E+00 2 . 35E+00 2.48E+00 3.00E+00 4.80E+00 6.43 E+00 8.19E+00 2 . 00E+01 sp4 D O. OOE+OO 3.51E-03 2 . 52E-03 1 .99E - 03 5.85E-03 2.42 E -03 8 . 48E-03 6 . 07 E -03 1.3 1E-02 2.16E-02

4. 1 2 E -02 4.88E-02 2 . 9 7E -02 3.88E - 02 3.26E-02 5.68 E+00 1 . 96E+00 2 . 03 E+ 00 2.11E+00 1.85E+01 4.42 E+01 1 . 24E+02 2 . 44E+02 4.53E+02 3.91E+02 7.36E+02 7.54E+02 1 . 17 E+04 7 .4 0E +04 1. 05E+ 0 6 1.34E+07 1 . 81E+08 1.52 E+08 2 .3 7 E+09 1. 7 1E+10 3.73E+10 3 . 73E+ 1 0 2 . 99 E+1 0 2.82E+10 6. 1 0E+09 2 . 18E+10 3.95E +10 1. 1 2 E+1 0 3.57E+09 1.23E+09 c

c c Tal l y Defini tion c c c Point Detector Particle F l uence Tallies c c f15:n -90000 . 00 -8950 . 00 30 7 8.00 50 $ EAB SSWes t f25:n 2964.18 2758.44 1920.24 50 $ Point X Control Room Proper c fc 1 5 Gamma Dose Rate a t EAB S SWest fc25 Gamma Dose Rate at interior of CR (Rm 77) c f m15 2 . 36 E+ll f m25 2 . 36 E+ ll c C Dose Convers i on Factors for Neu t rons (mRem/hr ) /(partic l e/cm2-sec) c c deO 2 . 5 E -08 1 . 0E-0 7 1.0E-06 1 . 0E-05 1.0E- 0 4 1 .0 E -03 1.0E - 02 1.0E -01 5 . 0E -0 1 1 . 0E+00 2.5E+0 0 5 .0E+00 7 .0E+00 1 . 0E+01 1 . 4E+01 2 . 0E+01 c df O 3 . 6 7E - 03 3.6 7E-03 4.46 E -03 4.54 E -03 4.18E-03 3 .7 6E-03 3 . 56 E- 03 2.1 7E-02 9 . 26E-02 1 . 32 E -01 1 . 25 E -01 1 .56E-01 1. 4 7E - 0 1 1 . 4 7E-01 2 . 0 8E- 01 2 . 2 7E -01 c c C Periphera l cards c c PRDMP 4 J 3 c C Prob lem c u to f f c c n p s 1 .0E09 c FC08513, R0 EC 68969 116

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 117 of 269 Attachment B: MCNP Tally Output

1) FCS_14M0g_EABCRci.txt 14 Month decay using Gamma source term.

tally 15 tally 25 nps mean error vov slope fom mean error vov slope fom 402653184 6. 5911E-04 0.5942 0.9264 1.7 9.5E-03 9.5316E-05 0.4717 0.3369 1.6 1.5E-02 805306368 4. 8715E-04 0.4395 0.6738 1.8 8.6E-03 2.7896E-04 0.5776 0.5716 1.6 5 . 0E-03 1207959552 4.2242E-04 0.3470 0.6077 1.7 9.2E-03 1. 9971E-04 0.5385 0.5692 1.5 3.8E-03 1610612736 3.9640E-04 0.2858 0. 5411 1.7 1. OE-02 1.8288E-04 0.4575 0.4943 1.5 4.0E-03 2013265920 1.4730E-03 0.7517 0.9861 1.7 1.2E-03 1. 5396E-04 0.4354 0.4912 1.6 3.6E-03 2415919104 1.2988E-03 0. 7111 0.9824 1.8 1.1E-03 1.3851E-04 0.4058 0.4795 1.6 3.5E-03 2818572288 1.1754E-03 0.6738 0.9805 1.7 1.1E-03 1. 2464E-04 0. 3871 0.4764 1.6 3.3E-03 3221225472 1.1539E-03 0.6040 0 . 9587 1.7 1.2E-03 1. 2634E-04 0.3440 0. 4271 1.5 3.7E-03 3623878656 1. 0903E-03 0.5698 0.9478 1.8 1. 2E-03 1.1738E-04 0.3296 0.4244 1.5 3.6E-03 4026531840 1.2041E-03 0.4944 0.7516 1.8 1.4E-03 1. 3338E-04 0.3214 0.3000 1.5 3.4E-03 4429185024 1.1089E-03 0.4880 0.7516 1.8 1. 4E-03 1.2901E-04 0.3046 0.2907 1.5 3.5E-03 4831838208 1. 0650E-03 0.4662 0.7490 1.8 1. 4E-03 1.2521E-04 0 . 2913 0.2769 1.6 3.5E-03 5000000000 1.0343E-03 0.4639 0.7490 1.8 1.3E-03 1.2267E-04 0.2875 0.2764 1.6 3.5E-03 805306368 4.8715E-04 0.4395 0.6738 1.8 8.6E-03 2.7896E-04 0.5776 0.5716 1.6 5.0E-03 1610612736 3.9640E-04 0.2858 0. 5411 1.7 1.0E-02 1. 8288E-04 0.4575 0.4943 1.5 4.0E-03 2415919104 1.2988E-03 0. 7111 0.9824 1.8 1.1E-03 1.3851E-04 0 . 4058 0.4795 1.6 3.5E-03 3221225472 1.1539E-03 0.6040 0.9587 1.7 1.2E-03 1.2634E-04 0.3440 0.4271 1.5 3.7E-03 4026531840 1.2041E-03 0.4944 0.7516 1.8 1.4E-03 1.3338E -04 0.3214 0.3000 1.5 3.4E-03 4831838208 1. 0650E-03 0 . 4662 0.7490 1.8 1.4E-03 1.2521E-04 0.2913 0.2769 1.6 3.5E-03 5637144576 1.1640E-03 0.3876 0.6028 1.7 1. 7E-03 1.1937E-04 0.2660 0.2608 1 .6 3.6E-03 6442450944 1.3709E-03 0.3440 0.3791 1.7 1. 9E- 03 1.1075E -04 0.2520 0.2561 1.6 3 . 5E-03 7247757312 1.2599E-03 0.3329 0.3785 1.7 1. BE- 03 1.1823E-04 0.2324 0.1975 1.6 3.7E-03 8053063680 1.1613E-03 0.3251 0.3782 1.7 1.7E -03 1 . 1826E-04 0.2158 0.1769 1.6 3.9E-03 8858370048 1.1584E-03 0.3029 0.3482 1.7 1.8E-03 1 . 1429E-04 0.2038 0.1741 1.6 4 .0 E-03 9663676416 1.0900E-03 0.2951 0.3478 1.7 1.7E-03 1. 0969E-04 0.1974 0 . 1655 1.6 3.9E-03 10000000000 1.0590E-03 0.2935 0.3478 1.7 1.7E -03 1 . 0735E-04 0.1950 0.1652 1.6 3.8E-03 FC08513, R0 EC 68969 117

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 118 of 269

2) FCS_18MOg_EABCRca.txt 18 Month decay using Gamma source term.

tally 15 tally 25 nps mean error VOV slope fom mean error vov slope fom 402653184 5.4485E-04 0.5941 0.9262 1.7 6.7E-04 7.8906E-05 0.4716 0.3374 1.6 1.1E-03 805306368 4.0254E-04 0.4395 0.6738 1.8 6.0E-04 2.3080E-04 0.5777 0.5722 1.6 3 . 5E-04 1207959552 3.4907E-04 0.3470 0.6077 1.7 6.4E-04 1.6521E-04 0.5386 0.5698 1.5 2 . 7E-04 1610612736 3.2760E-04 0.2857 0.5411 1.7 7 . 1E-04 1 .5 1 28 E -04 0.4576 0 . 4949 1.5 2.8E-04 2013265920 1.2176E-03 0 . 7516 0.9861 1.7 8 . 2E-05 1.2733E-04 0.4356 0.4917 1.5 2.5E-04 2415919104 1 .0 736E-03 0 . 7110 0 . 9824 1.8 7 . 7E-05 1.1457E-04 0.4059 0.4800 1.6 2.4E-04 2818572288 9.7174E-04 0.6737 0.9805 1.7 7.4E-05 1. 0312E-04 0.3872 0 . 4769 1.6 2.2 E-04 322 1225472 9.5482E-04 0 . 6033 0.9588 1.7 8.0E-05 1. 0345E-04 0.3475 0.4283 1.5 2.4E-04 3623878656 9.0215E-04 0 . 5692 0 . 9478 1.8 8.0E-05 9.6153E-05 0.3328 0.4256 1.5 2.4E-04 4026531840 9.9591E-04 0.4940 0.7519 1.8 9.6E-05 1.0938E-04 0.3237 0.3003 1.5 2.2E-04 4429185024 9.1718E-04 0.4876 0.7519 1.8 9.0E-05 1. 0581E-04 0.3066 0.2910 1.5 2.3E-04 4831838208 8.8082E-04 0.4658 0.7493 1.8 9.0E-05 1 .02 75E-04 0.2932 0.2771 1.6 2.3E-04 5000000000 8.5547E-04 0.4635 0.7493 1.8 8.8E-05 1.0068E-04 0.2893 0.2766 1.6 2.3E-04 tally 15 tally 25 nps mean error vov slope fom mean error VOV slope fom 805306368 4.0254E-04 0.4395 0.6738 1.8 6.0E-04 2.3080E-04 0.5777 0.5722 1.6 3.5E-04 1610612736 3.2760E-04 0.2857 0. 5411 1.7 7.1E-04 1. 5128E-04 0.4576 0.4949 1.5 2.8E-04 2415919104 1.0736E-03 0. 7110 0.9824 1.8 7.7E-05 1.1457E-04 0.4059 0.4800 1.6 2.4E-04 3221225472 9.5482E-04 0.6033 0.9588 1.7 8.0E-05 1. 0345E-04 0.3475 0.4283 1.5 2.4E-04 4026531840 9. 9591E-04 0.4940 0.7519 1.8 9.6E-05 1. 0938E-04 0.3237 0.3003 1.5 2.2E-04 4831838208 8.8082E-04 0.4658 0.7493 1.8 9.0E-05 1.0275E-04 0.2932 0.2771 1.6 2.3E-04 5637144576 9 . 6275E-04 0.3873 0.6029 1.7 1.1E-04 9.8168E-05 0.2674 0.2602 1.6 2.3E-04 6442450944 1.1354E-03 0 . 3433 0 . 3792 1.6 1. 2E-04 9.1104E-05 0.2532 0.2555 1.6 2.3E-04 7247757312 1 .0 420E-03 0.3326 0.3785 1.7 1. 2E-04 9.7243E-05 0.2335 0 .1971 1.6 2.4E-04 8053063680 9.6048E-04 0.3248 0.3783 1.7 1.1E-04 9.7341E-05 0.2168 0.1763 1.6 2 . 5E-04 8858370048 9.5782E-04 0.3027 0.3484 1.7 1. 2E-04 9.4124E-05 0.2046 0.1736 1.6 2.6E-04 9663676416 9.0125E-04 0.2949 0.3480 1.7 1.1E-04 9.0353E-05 0 . 1981 0.1650 1.6 2.5E-04 10000000000 8.7565E-04 0.2934 0.3480 1.7 1.1E-04 8.8403E-05 0.1958 0.1647 1.6 2.5E-04 FC08513, R0 EC 68969 118

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 119 of 269

3) FCS_14MOn_EABCRci.txt 14 Month decay using Neutron source term.

tally 15 tally 25 nps mean error vov slope fom mean error vov slope fom 3145728 5.5506E-06 0 . 1782 0.2265 1.8 5.1E-02 1.9457E-03 0.3880 0 . 2959 1.2 1.1E-02 6291456 5.5102E-06 0.1345 0.2032 2 .0 5.1E-02 1.8790E-03 0 . 2889 0 . 1686 1. 3 1.1E-02 9437184 5.3999E - 06 0.1133 0 . 1195 2 .2 4 . 8E-02 1.7601E-03 0.2381 0.1173 1.3 1.1E-02 12582912 6.0547E-06 0.1287 0.1876 2 .1 2.8E-02 1.5535E-03 0 . 2275 0.1117 1.3 8.8E-03 15728640 1.1182E-05 0.2539 0.4204 1.9 5.7E-03 1 . 8337E-03 0.2185 0.1641 1.3 7 . 7E-03 18874368 1.0251E - 05 0.2316 0.4149 2.0 5 . 7E-03 1 . 7333E-03 0.2030 0.1407 1.4 7.4E-03 22020096 1.1553E-05 0.2057 0.2612 1. 9 6. 2E-03 1 . 6227E-03 0.1889 0.1321 1.5 7.3E-03 25165824 1 . 1146E-05 0.1876 0 . 2556 2.0 6.5E - 03 1.5546E-03 0.1776 0.1194 1.6 7.2E-03 28311552 1.0529E-05 0.1771 0 . 2523 2.0 6.4E-03 1.5055E-03 0 . 1679 0.1080 1.6 7.2E-03 31457280 1.0482E-05 0.1629 0 . 2361 2.0 6 . 7E-03 1 . 4315E-03 0.1610 0.1031 1.7 6.9E-03 34603008 1.0182E-05 0 . 1531 0.2316 2.1 6.8E-03 1.3783E-03 0 . 15 3 4 0.0994 1.8 6.8E-03 35000000 1 . 0176E-05 0 . 1517 0 . 2306 2.1 6.9E - 03 1.3642E-03 0 . 1532 0.0994 1.8 6.7E-03 tally 15 tally 25 nps mean error vov slope fom mean error vov slope fom 6291456 5.5102E-06 0.1345 0.2032 2.0 5.1E-02 1.8790E-03 0.2889 0.1686 1.3 1.1E-02 12582912 6.0547E-06 0.1287 0.1876 2.1 2.8E-02 1.5535E-03 0.2275 0.1117 1.3 8.8E-03 18874368 1.0251E-05 0.2316 0 . 4149 2.0 5.7E-03 1.7333E-03 0 . 2030 0 . 1407 1.4 7 .4E-03 25165824 1.1146E-05 0.1876 0.2556 2.0 6.5E-03 1.5546E-03 0.1776 0.1194 1.6 7.2E-03 31457280 1.0482E-05 0.1629 0.2361 2.0 6.7E-03 1.4315E-03 0 . 1610 0.1031 1.7 6.9E-03 37748736 1.0632E-05 0.1410 0.1952 2 .1 5 . 1E-03 1.3475E-03 0.1467 0.0925 1.8 4 . 8E - 03 44040192 1.0550E-05 0.1259 0.1720 2 .1 3.6E-03 1.4353E-03 0.1362 0.0762 1.9 3.1E-03 50331648 1.0192E-05 0.1162 0.1599 2.1 2 . 9E-03 1.4246E-03 0.1302 0 . 0697 2 .0 2.3E-03 56623104 1 . 0035E - 05 0.1093 0 . 1388 2 .1 2.5E-03 1.4061E-03 0.1216 0.0623 2.2 2.0E-03 62914560 1 . 0128E-05 0 . 1010 0.1219 2.1 2.4E-03 1 . 3655E-03 0.1155 0.0570 2.3 1.8E-03 69206016 1 . 0046E - 05 0.0955 0.1099 2.1 2.2E-03 1.5473E-03 0.1415 0.2487 2 .3 l.OE-03 75497472 1.1073E-05 0.1085 0.186 2 2.0 1 . 5E-03 1.5188E-03 0.1335 0 . 2390 2.3 9.9E-04 81788928 1.1576E-05 0.1098 0.1423 2.0 1. 3E-03 1.4821E-03 0.1278 0 . 2281 2.5 9.5E-04 88080384 1.1358E - 05 0.1043 0.1397 2 .1 1. 3E-03 1.4562E-03 0.1221 0.2187 2.6 9.3E-04 94371840 1.1220E - 05 0 . 0992 0.1362 2.3 1.3E-0 3 1.3839E-03 0.1200 0.2179 2.7 8 . 7E-04 100000000 1.0897E - 05 0.0966 0 . 1354 2.3 1.2E - 03 1.4437E-03 0 . 1161 0.1751 2.8 8.5E-04

4) FCS_18MOn_EABCRcc.txt 18 Month decay using Neutron source term.

tally 15 tally 25 nps mean error vov slope fom mean error vov slope fom 50331648 1.0360E-05 0.1388 0.1777 2.1 8.4E-04 1.59 2 9E-03 0.1891 0.2831 2.0 4.5E-04 100663296 1.1342E-05 0.1126 0 . 2250 2.1 6.3E-04 1.4876E-03 0 . 1208 0.1464 2.7 5 . 5E-04 150994944 1.3747E-05 0.1038 0.1065 2.1 5 . 0E-04 1.6460E-03 0 . 0955 0.0781 2.6 5.9E-04 201326592 1.3273E-05 0.0871 0.0872 2.3 5 . 3E-04 1 . 6499E-03 0.0794 0 . 0546 2.8 6.4E-04 251658240 1.3310E-05 0.0761 0.0675 2.7 5 . 6E-04 1 . 7010E-03 0 . 0709 0.0421 3.2 6 . 4E-04 301989888 1.3783E-05 0.0693 0.0569 2.9 5 . 6E-04 1 . 7342E-03 0.0640 0.0326 3.5 6 . 6E-04 352321536 1 . 4124E-05 0.0646 0.0438 2.9 5.5E-04 1.7089E-03 0.0581 0.0278 3.9 6.8E-04 402653184 1.7055E-05 0.1888 0.8688 2.6 5.6E-05 1 . 9534E-03 0.1418 0 . 7906 3.2 l.OE-04 452984832 1.7312E-05 0.1679 0 . 8186 2.5 6.3E-05 1.9429E-03 0.1281 0 . 7568 3.3 1.1E-04 503316480 1.7236E-05 0.1529 0 . 7933 2.5 6.9E-05 1.9028E-03 0.1184 0 . 7387 3.1 1.1E-04 553648128 1.7087E-05 0.1410 0 . 7764 2.5 7.4E-05 1.8580E-03 0 . 1106 0.7281 3.0 1.2E-04 603979776 1.6951E-05 0.1313 0 . 7520 2.6 7 . 8E-05 1.8157E-03 0 . 1041 0.7187 3.2 1.2E-04 654311424 1.6618E-05 0 . 1241 0.7412 2.7 8 . 0E-05 1 . 7817E-03 0 . 0983 0.7095 3.1 1.3E-04 704643072 1 . 6364E-05 0 . 1177 0.7246 2.8 8 . 3E-05 1.7739E-03 0.0934 0.6596 3.0 1.3E-04 725000000 1 . 6178E-05 0 . 1158 0.7235 2.8 8 . 3E-05 1.7783E-03 0 . 0907 0.6531 3.0 1.4E-04 FC08513, R0 EC 68969 119

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 FC08513, R0 Page 120 of 269 Attachment C: Review of NRC RAis as they pertain to SFP shine dose. Request for Additional Information (RAis) and Applicability October 2, 2014 SCE to NRC, Docket No. 50-206, 50-361, 50-362, and 72-041 Response to Request for Additional Information Regarding Emergency Planning Exemption Request San Onofre Nuclear Generating Station, Units 1, 2, 3 and ISFSI Document/RAJ# RAI Question Applicable Justification (Yes or No, Y or N) ARCB-RAI-001 Please provide additional information for each of these N DBA scenarios are not addressed in FC08513 remaining UFSAR Chapter 15 design basis accident (DBA) or FC08514. Only Beyond design basis event EC 68969 scenarios that apply to a permanently defueled facility that is addressed, i.e. pool drain down and decay has the potential to result in a radiological release. Provide a heat load. table for each relevant DBA that lists: (1) The parameters used in the dose analysis, (2) the original modeled value for each parameter listed in the SONGS, Units 2 and 3 Updated Final Safety Analysis Report (UFSAR) , (3) the new modeled value for each parameter in the revised DBAs; and, (4) a description as to why the new modeled value for each parameter is justified (if changed). ARCB-RAI-002 Enclosure 1, Section 4.2, of the exemption request proposes y The calculation FC08513 documents the a beyond-design-basis event concerning a loss of water modeling assumptions for Monte Carlo N-inventory from the SONGS spent fuel pools (SFPs) as of Particle Transport Code (MCNP6). The June 12, 2013, the date on which SCE certified permanent calculations performed were to document and cession of power operations of SONGS, Units 2 and 3. SCE show estimated doses would be well below the stated that the purpose of this calculation is to determine the acceptance criterion for exemption from potential radiological impact due to loss of shielding to the requiring offsite emergency planning zones. public at the Exclusion Area Boundary (EAB) for the event in Details regarding assumptions are which the spent fuel assemblies are uncovered following documented in FC08513. These assumptions drain down. include geometry, fuel source, material composition , dose conversion. The licensee concluded , "Based on calculated direct and scattered dose rates from spent fuel assemblies in a SONGS 120 SFP following drain down. it is concluded that the maximum

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 FC08513, R0 Page 121 of 269 dose at the EAB would be well below the acceptance criteria." Therefore, the U. S. Nuclear Regulatory Commission (NRC) staff needs additional information on the modeling assumptions for such an analysis and a Monte Carlo N Particle Transport Code, version 5 (MCNP5) input deck. The purpose of this request is to confirm that these calculations are acceptable and that the estimated doses are well below the acceptance criterion for exemption from requiring offsite emergency planning zones is less than 1 rem projected dose for a four-day period . ARCB-RAI-003 Enclosure 1, Section 4.2, page 13 of 22 of the exemption y The gamma and neutron source strength in request states: units of photons and neutrons per second respectively and by energy group spectrum "The source terms for neutron and gamma radiation in spent EC 68969 were documented in FC08514. This source fuel pools were calculated with consideration of plant term was determined using the ORIGEN-shutdown dates as outlined earlier." ARP/ORIGEN-S modules of the SCALE 6.1 Please provide both a summary of the source term analysis computer code package. Details regarding the that calculated the neutron and gamma source terms for the modeling and assumptions were documented spent fuel pools and the source term used to perform the in FC08514. An 18 group gamma and 44 MCNP5 calculation. group neutron energy structure was used versus defaults in order to perform benchmarks. An 18-month decay source term was generated , and a scaled 14 month source term. ARCB-RAI-004 The dose criteria specified in 10 CFR 50.67 is in terms of y As shown in FC08513 after 18 months of total effective dose equivalent (TEDE). As such, confirmatory decay in an assembly (i.e. the highest burnup calculations are performed in terms of TEDE to compare assembly from Cycle 28 offload) the activity against the dose acceptance criteria specified in NUREG- available for release is predominantly Kr85 . 0800, Section 15.0.1. This analysis differs from the (noble gas) that does not contribute to the Environmental Protection Agency (EPA) Protective Action groundshine. These calculations only address Guide [PAG] and Planning Guidance for Radiological beyond design basis event (i.e. loss of pool Incidents (a.k.a. , the EPA PAG Manual) recommendation for water) . This calculation determines the a projected whole body dose of I to 5 Rem total effective external EDE only or the DOE component of dose (TED) over four days. The TEDE is defined as the sum the TEDE dose. There is no internal 121 of the effective dose equivalent for external exposures and component for the body calculated from a re-the committed effective dose equivalent for internal suspended radioisotope in ground exj:>osures. The projected whole body dose calculation for contamination or direct exposure from _ground

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 FC08513, R0 Page 122 of 269 TED is defined as the sum of the effective dose from external contamination. This is only a DOE calculation. radiation exposure (i.e., groundshine and cloudshine) and the This calculation does not document a release committed effective dose from inhaled radioactive material. fraction of Cs137 or Cs1234 (i.e. NUREG 1910 Dose conversion factors (DCF) applied in both analyses or few other isotopes potentially released from convert the estimated environmental exposure to dose in the the pool. The actual design basis accidents units of concern. Dose conversion factors acceptable to the use different assumptions and fractions of NRC staff are derived from data provided in International isotopes for releases to address CEDE Commission on Radiological Protection Publication 30, components. Once fuel has cooled such that "Limits for Intakes of Radionuclides by Workers" and can be cladding is intact even with a water boil off and found in Federal Guidance Report 11, "Limiting Values of the concentration of Cs137, Co60 that was in Radionuclide Intake and Air Concentration and Dose the pool water would not exceed the PAG. This Conversion Factors for Inhalation, Submersion, and assumption is outside this calculation. Ingestion," and Federal Guidance Report No. 12, "External Exposure to Radionuclides in Air, Water, and Soil," for exposure to radionuclides in air, water, and soil." EC 68969 While both TEDE and TED calculate dose for both external and internal exposure, the underlying dosimetry models used to develop the DCFs are not the same. Please provide a summary explaining why the use of the EPA PAG dose criteria TED is acceptable in addition to the dose criteria specified in 10 CFR 50.67, which is calculated in terms of TEDE. ARCB-RAI-005 Section 3.0, Page 3 of 21 of Enclosure 1 to the exemption N The purpose of these calculations FC08513 request states that the UFSAR has been updated accordingly and FC08514 is to evaluate a beyond design in accordance with 10 CFR 50.71(e), to reflect the possible basis event. The revised USAR descriptions of accident/transients scenarios pertinent to the reactor being Chapter 14 (DBAs) for a defueled or shutdown permanently shut down and defueled . Please provide the condition are not part of this scope and would revised UFSAR descriptions of Chapter 15. be required to be addressed elsewhere. 122

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 FC08513, R0 Page 123 of 269 September 9, 2014, SCE to NRC, Docket No. 50-206, 50-361, 50-362, and 72-041 Response to Request for Additional Information Regarding Emergency Planning Exemption Request San Onofre Nuclear Generating Station, Units 1, 2, 3 and ISFSI Document/RAI # RAI Question Applicable Justification (Yes or No , Y or N) (MF3835) RAI-001 The basis for exemption of Item 1 in Table 1 (Enclosure 2) is N Calculations FC08513 and FC08514 do not generic and does not state specifically why SONGS should include exemptions to rules in the calculation be considered for exemption. Similarly, the following item itself related to emergency planning. The numbers also contain only generic information in the basis for actual exemptions to the rules would be exemption : 3, 4, 5, 9, 11, 17, 18, 20, 23, 25, 26, 32, 43 , 48 documented in a submittal to the NRC based and 49. Please provide justifications specific to SONGS for upon results from these calculations noted granting the exemptions listed above. above. EC 68969 (MF3835) RAI-002 The basis for exemption of Item 1 in Table 1 (Enclosure 2) N Calculations FC08513 and FC08514 do not does not address design-basis accidents (DBAs). Please address design basis accidents (DBAs). provide a discussion justifying that no currently applicable Therefore, th is RAI is not applicable. DBA will exceed U.S. Environmental Protection Agency (EPA) Protective Action Guides. (MF3835) RAI-003 A licensee shall have the capability to notify responsible N FC08513 and FC08514 do not propose to State and local governmental agencies within 15 minutes change any FCS emergency plan documents. promptly (within 60 minutes) after declaring an emergency." The NRC staff cannot approve the addition of rule language via an exemption , only by issuing rulemaking. Please provide the site-specific justification for extending the notification time beyond 15 minutes, including the notification time to which SONGS will be committed to. (MF3835) RAI-004 Item 40 in Table 1 (Enclosure 2) contains no justification for N FC08513 and FC08514 do not propose to deletion of "Civil Defense" and "local news media persons". change any FCS emergency plan documents. Please provide site-specific justification for exempting these requirements. (MF3835) RAI-005 Justifications for Items 46 and 47 in Table 1 (Enclosure 2) N FC08513 and FC08514 do not propose to state: "see basis for section IV.2." The basis for section IV.2 change any FCS emergency plan documents. 123 states "see basis for 50.47(b)(1 0)." Please provide specific

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 FC08513, R0 Page 124 of 269 835justification for these two items or reference the correct section. (MF3835) RAI-006 Although formal offsite radiological EP plans have typically N FC08513 and FC08514 do not propose to been exempted for decommissioning sites, offsite change any FCS emergency plan documents. organizations continue to be relied upon for firefighting , law enforcement, ambulance and medical services in support of the licensee's (onsite) emergency plan . Additionally, the licensee is responsible for control of activities in the Exclusion Area , including public access. Please provide further justification as to why this requirement would not be applicable based on the context described above. (MF3835) RAI-007 Although the NRC has previously exempted N FC08513 and FC08514 do not propose to decommissioning reactors from "hostile action" change any FCS emergency plan documents. enhancements based on the applicability of the new EP Rule EC 68969 (as stated in the Statement of Considerations), some EP requirements for events are maintained , such as the classification of security-based events, notification of offsite authorities, and coordination for the response of offsite response organizations (i.e., firefighting , medical assistance) onsite. Please revise the requested exemption accordingly or provide further justification for exemption. (MF3835) RAI-008 State and local jurisdictions may take actions as part of their N FC08513 and FC08514 do not propose to comprehensive emergency response (all-hazard) planning. change any FCS emergency plan documents. Licensee actions shall not impede State and local authorities to respond to emergencies as they determine the need. Please provide specific justification for exempting this requirement or restore language consistent with revised wording proposed . (MF3835) RAI-009 It appears to the NRC staff that 10 CFR 50 Appendix N FC08513 and FC08514 do not propose to E.IV.E.9.c as exempted would be redundant to 10 CFR 50 change any FCS emergency plan documents. Appendix E.IV.E.9.a. Please explain what different organizations would be contacted and what different communication systems would be tested for compliance with 10 CFR 50 Appendix E.IVE.9.c, as exempted , as opposed to the ones in 10 CFR 50, Appendix E.IV.E.9.a, as exempted. 124 (MF3835) RAI-01 0 Exemption of requirements to emergency planning N This RAI relates to AOP and FLEX mitigation requirements, as requested , partially depends on the ability strategies. The purpose of FC08513 and

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 FC08513, R0 Page 125 of 269 of the licensee to mitigate the consequences of design basis FC08514 was only to provide a source term and beyond DBAs. Please describe the actions SONGS and the estimated dose from drain down event. could take to mitigate the consequences of an event Actual mitigating strategies would be involving the spent fuel pool (SFP). addressed in other documents. (MF3835) RAI-011 The Executive Summary in NUREG-1 738 states, in part, .. . N FC08513 and FC08514 only provide the staffs analyses and conclusions apply to calculated dose information. They do not decommissioning facilities with SFPs that meet the design propose to change the emergency plan. and operational characteristics assumed in the risk analysis. Emergency plan changes proposed would These characteristics are identified in the study as industry have to address seismic checklist, and decommissioning commitments (IDCs) and staff assumptions related to the IDSs and SDAs. decommissioning assumptions (SDAs). Provisions for Other calculations address natural circulation confirmation of these characteristics would need to be an and indefinite cooling by air of fuel. FC08514 integral part of rulemaking . The IDCs and SDAs are listed in does provide the decay heat output for NUREG-1 738, Tables 4.1-1 and 4.1-2, respectively. Please individual assemblies which is then used in explain the extent each of these IDCs and SDAs will be other subsequent analyses. EC 68969 satisfied at SONGS during the decommissioning phase, considering proposed exemptions from portions of the emergency planning requirements of 10 CFR 50.47(b) , 10 CFR 50.47(c)(2) and 10 CFR Part 50, Appendix E. With respect to the seismic checklist discussed in SDA No. 6, the explanation may focus on the reduced potential for a zirconium fire due to the delay in seeking emergency plan changes (i.e., a demonstration that the fuel has decayed such that it can be indefinitely cooled by natural circulation of air in its design storage configuration) , as described in Item 10 of the checklist. (MF3835) RAI-012 The NRC staff determined that the description of the analysis y Fuel management records were used for of the adiabatic heatup of the hottest fuel assembly was determining the hottest assembly and incomplete as presented in Section 4.1 of Enclosure 1 to the modeling . This is described in FC08514. An exemption request letter dated March 31 , 2014. Please attachment to that calculation provides the fuel provide the following additional information regarding this management records along with the projected analysis: core burnup for cycle 28. Axial and core radial power distributions were not explicitly used ,

                                *the information in the fuel management records (mentioned because the core radial power distribution is in Section 4.1.2) implicitly considered via selecting and
                                " the process used to determine the limiting assembly from              modeling the highest burnup assembly. The 125 these records , including how assembly axial and core radial            assembly axial power distribution was not power distributions were considered                                     modeled for the source term. It was found that

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 FC08513, R0 Page 126 of 269

                           " the decay heat model used to determine the decay heat          a uniform distribution over the entire length rate as a function of decay time                                 was conservative. The decay heat generation rates for the limiting fuel assembly were
                           " the source and value used for the specific heat of the documented in FC08514 using the ORIGEN-uranium dioxide in the limiting assembly ARP/ORIGEN-S modules within SCALE 6.1.
                           " since two different values of uranium dioxide density are      To be more conservative the highest burnup provided (by fuel vendor), specify which value was used for      for the limiting assembly was used. The the hottest fuel assembly and why that application is limiting   specific heat of the fuel was not part of the dose consequence calculation and as such would be expected to be documented elsewhere. The density of the uranium dioxide theoretical density was used in the analysis for dose. This is appropriate because a lower uranium density percentage results in a lower mass and thus, a lower effective density of the EC 68969 fuel source. The value of% TD is the lowest stack density shown in vendor specifications.

126

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 FC08513, R0 Page 127 of 269 October 21, 2014, SCE to NRC, Docket No. 50-206, 50-361, 50-362, and 72-041 Response to Request for Additional Information and Supplement Regarding Permanently Defueled Emergency Action Levels Amendment Application Numbers 224, 268, and 253 San Onofre Nuclear Generating Station, Units 1, 2, and 3 and ISFSI Document/RAI # RAI Question Applicable Justification (Yes or No, Y or N) SONGS/ Please annotate in Section 1, "Purpose," that this document N Related to Emergency Action Levels will be maintained in accordance with 10 CFR 50.54(q). Technical Bases Manual. FC08513 and EAL-RAI-01 FC08514 do not alter the actual basis documents within the calculation itself, but will be required to be maintained in accordance with 10 CFR 50.54 (q) since they would EC 68969 support a further change to the basis documents. SONGS/ Please explain why the definitions for the following terms are N Related to Emergency Action Levels not included , as stated in the endorsed guidance, or revise Technical Bases Manual. FC08513 and EAL-RAI-02 accordingly: FC08514 do not alter the basis document.

• Explosion,
• Fire, and *Visible Damage.

SONGS/ Under Initiating Condition PD-AU1 , please explain how EAL N Related to Emergency Action Levels

                                 #2 is declared in a timely fashion and whether the capability                   Technical Bases Manual. FC08513 and EAL-RAI-03          to perform this evaluation is maintained on-site 24-hours per                   FC08514 do not alter the basis document.

day, seven days per week (24/7) . SONGS/ Under Initiating Condition PD-AU2, please clarify whether N Related to Emergency Action Levels there are any remote reading alarms associated with the Technical Bases Manual. FC08513 and EAL-RAI-04 decrease in spent fuel pool water level, and if not, what FC08514 do not alter the basis document. means will be in place to ensure timely classification if SFPLI is addressed by FLEX mitigating warranted. strategies. SONGS/ Under Initiating Condition PD-SU1 , please clarify whether N Related to Emergency Action Levels there are any remote reading alarms associated with the Technical Bases Manual. FC08513 and EAL-RAI-05 increase in spent fuel pool water temperature, and if not, FC08514 do not alter the basis document. what means will be in place to ensure timely classification, if SFPLI is addressed by FLEX mitigating 127 warranted. strategies.

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 FC08513, R0 Page 128 of 269 Kewaunee Document, December 11, 2013, DOMINION ENERGY KEWAUNEE, INC. KEWAUNEE POWER STATION SUPPLEMENT I AND RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING REQUEST FOR EXEMPTIONS FROM PORTIONS OF 10 CFR 50.47 AND 10 CFR 50, APPENDIX E Document/RAJ # RAI Question Applicable Justification (Yes or No, Y or N) NRC Question Page 43 of 55 references a site-specific adiabatic heat up to N FC08513 and FC08514 provide decay heat MF2567-RAII- address a partial drain down of the SFP (identified as and dose calculations. They do not address ORLOB-Norris-0 13 Reference 12), assuming no air-cooling it states that the time the adiabatic heatup itself. Further additional necessary for the hottest fuel assembly to reach the critical calculations are required to address the EC 68969 temperature of 5650C is six hours after the fuel rods have beyond design basis loss of spent fuel pool become uncovered. Some previous exemptions were cooling and assembly heat up. granted based [on] time to reach the cladding auto-ignition temperature of 9000C. Based on 17 months of decay time, at what time after the fuel is uncovered, assuming an adiabatic heatup, would the hottest fuel assembly reach 900°C? Please provide a copy of this analysis and the existing analysis for the Partial Loss of Cooling Water Inventory with No Air Cooling. 128

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 129 of 269 Attachment D: Excel Spreadsheets. Attachment 01 M3 Calculation20161014.xlsx Material M3 Calculation Attachment 02 FCS MCNP Source lnputs.xlsx Source Density Calculation Attachment 03 FCS Geo Inputs 20161017.xlsx MCNP Geometry Inputs Attachment 04 FC08514 ext enr.xlsx U-235 Average Enrichment Attachment 05 tallies_gamm20161017.xlsx Gamma Tally Results Attachment 06 tallies neut20131017.xlsx Neutron Tally Results FC08513, R0 EC 68969 129

M3 Calculation20161014.xlsx Attachment D SpreadSheets FC08513 Material Calculation MCNP Nuclide FRACTION of Element Page 130 of 269 ID TOTAL MASS Weight Fractions AI 13027 8.344615E-03 $ AI weight percent Ar 18000 4.242021E-06 $ Ar weight percent B 5000 - $ B weight percent B-10 5010 3.049262E-04 $ BlO weight percent B-11 5011 1.361160E-03 $ B11 weight percent c 6000 5.023089E-04 $ C weight percent Cl 17000 5.747414E-09 $ Cl weight percent Co 27059 3.298115E-05 $ Co weight percent

     '        Cr           24000   2.048669E-02 $ Cr weight percent Cu           29000   6.209134E-05 $ Cu weight percent Fe           26000   6.672541E-02 $ Fe weight percent Mn            25055   1.966967E-03   $  Mn weight percent Mo            42000   4. 783920E-05  $  Mo weight percent N            7014   2.698222E-04   $  N weight percent Na           11023   5.747414E-09   $  Na weight percent Nb            41093   1.902357E-03   $  Nb weight percent Ni           28000   1.868235E-02   $  Ni weight percent 0            8016   8.421467E-02   $  0 weight percent p          15031   4.292521E-05   $  P weight percent Pb           82000   3.735819E-08   $  Pb weight percent s          16000   3.621267E-05   $ S weight percent Si          14000   1.214038E-03   $ Si weight percent Sn           50000   4.166875E-06   $ Sn weight percent Ta           73181   6.522834E-06   $ Ta weight percent Ti          22000   3.010457E-04   $ Ti weight percent U-234            92234   1.946356E-04   $  U-234 weight percent U-235            92235   2.183528E-02   $  U-235 weight percent U-236            92236   1.005027E-04   $ U-236 weight percent U-238            92238   6.017375E-01   $ U-238 weight percent v           23000   1.436854E-08   $ V weight percent Zr           40000   1.696187E-01   $ Zr weight percent l.OOOOOOE+OO FC08513, R0                                   EC 68969                          130

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation FC08513, R0 Reference FC07586 Page 131 of 269 Total Volume One Assembly Volume (cc) 304l One Assemb TOP Region 304l 361 .27481 427.799 TOP Region M5 157.42358 X-750 375.50684 M5 CF3 530.32650 1424.53 18479.190 A718 Fuel Plenum 304l 22.42395 118.844 Fuel Plenum M5 837.23515 X-750 is internal to X-750 ------------ the plenum X-750 Clad Volume 2076.23760 2935.90 659.904 CF3 Fuel Region U02 42404.46089 1414.204 Fuel Region EC 68969 304l 34.00966 M5 15384.77238 Zircaloy-4 A718 118.84360 57942 .09 27 .286 U02 BOT Region 42404.461 BOT Region 304l 10.09078 Total Assembly 3 M5 23 .52130 Volume (cm ) X-750 284.39747 63531 .688 CF3 883 .87750 Zircaloy-4 27.28616 1229.17 944 Assembly 3 Volume (cm ) Rack Material Volume Tab FCS FUEL (Cells A62-P124) 3 59,973,913 .63 2,261,690.13 Boral (cm ) 3 3 131 5,358,688.42 SS 304 (cm ) Total Volume (cm )

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation 228,971,853.30 I FC08513, R0 Total Volume Page 132 of 269 (Materials) 67,594,292 .19 Rema ining Volume (air) 161,377,561.12 Volume of Source 228,971,853 .30 cm3 2.568032 gm/cc I Mass U02 416,171,245 .28 g ss 304 46,256,586.84 g Bora I 6,106,563.36 g M5 100,721,968.00 g A-718 922,288.00 g X-750 6,748,656.00 g EC 68969 CF3 10,716,288.00 g Zircaloy-4 168,976.00 g Air 194,459.96 g 132

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation

Reference:

FC08513, R0 Document No. 51- Page 133 of 269 944 Assembly Mass (g) Mass (g) 9124707-000 ly Mass (kg) Mass (kg) Mass (g) I 304L 304L 2.900 2,737,600 I 3,241,696 MS 1.024 966,656 X-750 3.087 2,914,128 MS Fuel U02 CF3 4.257 4,018,608 100,721,968 Volume pins d iameter Stack Height 2587.679 176 0.3805 129.3 304L 0.180 169,920 A718 922,28~ I 42404.46 MS 5.446 5,141,024 X-750 0 .224 1.724 1,627,456 X-750 0.205 6,748,656 13.5 w/o CF3 U02 440.859 416,171,245 10,716,288 EC 68969 304L 0.273 257,712 MS 100.074 94,469,856 Zircaloy-4 A718 0.977 922,288 168,976 U02 416,171,245 304L 0.081 76,464 Total Assembly MS 0.153 144,432 mass (gms) X-750 2.338 2,207,072 538691117.280 CF3 7.095 6,697,680 Zircaloy-4 0.179 168,976 133

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Volume based on Defined MCNP source ('FCS Fuei'!X3) FC08513, R0 Page 134 of 269 Air Mass 3 0.001205 g/cm Reference : PNNL-15870 Rev. 1 194,459.961grams SFP Racks Mass Mass (g) SS 304 r---4-3,-0-14-,-89-0-.8-4.,1 Sum of "FCS Racks" cells (G59 + J59)

• Density of SS 304 8.03 gm/cm 3 3

Mass (g) Boral 6,106,563.361 Sum of "FCS Racks" cells (G62 + J62)

• Density of Bora I - 2.70 gm/cm 364,972.691 1,896. 717.45 1 EC 68969 134

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation 5.0 CALCULATIONS Uranium and Oxygen Weight Percents and U02 Densities Page 135 of 269 The weight percents of U-234, U-236, and U-238 are calculated as a function of the initial enrichment of U-235 using the relationships presented in Section D LA in the ORIGEN-ARP manual of Reference 8.1 .1. Specifically, U-234 wt% = 0.0089 x U-235 wt% U-236 wt% = 0.0046 X U-235 wt% U-238 wt% = 100- U-234 wt%- U-235 wt%- U-236 wt% U-234, U-236, and U-238 weight percents for U-235 enrichments of 2.0% through 5% are presented in the following table. U-234 U-235 U-236 U-238 234.040947 235.043924 236.045563 238.050785 Table 5-2 Uranium Isotope Weight Percents in a Uranium Blend, as a Function of U-235 Enrichment U-235 Enrichment Uranium Weight Percentages(%) Wt% U-234 U-235 U-236 U-238 2.0 0 .01780 2.000 0 .00920 97 .97300 2.5 0.02225 2.500 0 .01150 97.46625 3.0 0 .02670 3.000 0.01380 96.95950 3.5 0.03115 3.500 0.01610 96.45275 4 .0 0.03560 4 .000 0 .01840 95.94600 4.5 0.04005 4.500 0 .02070 95.43925 5.0 0.04450 5.000 0.02300 94.93250 FC08513, R0 EC 68969 135

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation INotes: Page 136 of 269 The atomic masses of U-234, U-235, U-236, and U-238 are 234.040947, 235.043924, 236.045563, and 238.050785, respectively (Design Input Section 2.9). The atomic mass of oxygen is 15.9994 (Design Input Section 2.8). The fraction of a uranium isotope in a uranium blend is equal to the product of atomic mass of the uranium blend and the ratio of the weight percent of the uranium isotope in the uranium blend to the atomic mass of the uranium isotope. The sum of the fractions of the uranium isotopes in the uranium blend is equal to 1. Specifically, Where: AM Us is the atomic mass of the urani urn blend at a specific U-235 enrichment, U-234 wt%, U-235 wt%, U-236 wt%, and U-238 wt% are the uranium isotope weight percents as a function of U-235 enrichment (from Table 5-1), and AMU-234, AM U-235, AM U-236, and AM U-238 are the atomic masses of U-234, U-235, U-236, and U-238 (per Design Input Section 2.9) . Equation 1 can be rearranged as : Equation 3 is solved to obtain atomic masses of the uranium blend in the fuel region for uranium enrichments of 4.5% and 5% (refer to Table 5-2). The atomic mass of U02 in the uranium blend, as a function of U-235 enrichment, is equal to the atomic mass of uranium in the uranium blend plus 2 times the atomic mass of oxygen. The atomic mass of U02 in the uranium blend is calculated in Table 5-2. FC08513, R0 EC 68969 136

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Table 5-1 Uranium Blend Atomic Masses as a Function of U-235 Enrichment Fage 137 of 269 Atom1c Mass ot UU:l 1n Uranium U-235 Enrichment (wt%) AMU B(1) Blend (2) 2.00 237.988982 269.987782 2.50 237.973537 269.972337 3.00 237.958093 269.956893 3.50 237.942652 269.941452 4.00 237.927212 269.926012 4.50 237.911774 269.910574 5.00 237.896339 269.895139 (1} The AMUb is calculated using Equation 3. The atomic masses of U-234, U-235, U-236, and U- 238 are 234.040947, 235.043924, 236.045563, and 238.050785, respectively (Design Input Section i 2.9}. Weight percents of the uranium isotopes are from Table 5-1. For a U-235 enrichment of 3.5 wt%, the atomic mass of the uranium blend (AMU8} [100/{(0.04005/234.040947} + (4.5/235.043924} + (0.02070/236.045563} + {95.43925/238.050785}] = 237.911774. (2} The atomic mass of Oxygen is 15.9994 (Design Input 2.8}. The atomic mass of U02 for a U-235 enrichment of 4.5%, is 237.911774 + (2 x 15.9994} = 269.910574. FC08513, R0 EC 68969 137

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Page 138 of 269 Uranium and Oxvgen Masses (per Assembly) The masses of u ra ni um i sotopes, the mass of U 02, and the mass of oxygen i n an assem bly of the uranium blend are calculated in Table 5-3 for U-235 enrichments of 4.5 wt% and 5.0 wt%. The masses of the uranium isotopes are calculated by multiplying the mass of uranium in anassembly (i.e., 388.6 kg per Design Input Table 2-1) by the uranium isotope weight percents. The mass of U02 in the uranium blend is calculated by multi pi ying the mass of uranium in an assembly by the ratio of the atomic mass of U02 i n the uran ium blend to the atomic mass of the uranium blend. The mass of oxygen in the uranium blend is the mass of U02 in the uranium blend minus the assembly mass of uranium in the blend. Table 5-3 Uranium Mass as a Function of U-235 Enrichment U-235 Enrichment Uranium Weight Properites U-234 U-235 U-236 U-238 Wt% U02 {glassy} {glassy} {glassy} {glassy} {glassy} 2.0 440849.1989 69.1708 7772.0000 35.7512 380723.0780 2.5 440852.5901 86.4635 9715.0000 44.6890 378753 .8475 3.0 440855.9813 103.7562 11658.0000 53.6268 376784.6170 3.5 440859.3726 121.0489 13601.0000 62.5646 374815.3865 4.0 440862.7638 138.3416 15544.0000 71.5024 372846.1560 4.5 440866.1550 155.6343 17487.0000 80.4402 370876.9255 5.0 440869.5462 172.9270 19430.0000 89.3780 368907.6950 FC08513, R0 EC 68969 138

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Table 5-4 Page 139 of 2 59 Weight Percents of Uranium Isotopes and Oxygen as a Function of U-235 Enrichment U-235 Enrichment Uranium Oxide Weight Properites U-235 U-236 U-238 Wt% U-234 (wLol (gLass~) (gLass~) (gLass~) 0 2.0 0.0157% 1.7630% 0.0081% 86.3613% 11.8519% 2.5 0.0196% 2.2037% 0.0101% 85.9139% 11.8526% 3.0 0.0235% 2.6444% 0.0122% 85.4666% 11.8533% 3.5 0.0275% 3.0851% 0.0142% 85.0193% 11.8540% 4.0 0.0314% 3.5258% 0.0162% 84.5719% 11.8547% 4.5 0.0353% 3.9665% 0.0182% 84.1246% 11.8553% 5.0 0.0392% 4.4072% 0.0203% 83.6773% 11.8560% Table~ - ~ U02 Density as a function U-235 Enrichment U-235 Enrichment Uranium Weight Properites Wt% U02 (gLass~) gill._ 2.0 440,849.20 10.3963 2.5 440,852.59 10.3964 3.0 440,855.98 10.3965 3.5 440,859.37 10.3965 4.0 440,862.76 10.3966 4.5 440,866.16 10.3967 5.0 440,869.55 10.3968 FC08513, R0 EC 68969 139

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Holtec Drawin FC08513, R0 Region I Racks Box Number Box 1 Box 2 Page 140 of 269 SS Box Volume Cell Dimensions (SS Box) Width 8.46" 8.46 4 (in"3) 408.618 408.618 Boral Box Volume Cell Dimensions (SS Box) Height 161" 161.00 3 (in"3) 278.4 278.4 55 Wrapper Cell Dimensions (SS Box) Thickness 0.075" 0.075 2 Volume (in"3) 88.595 88.595 Boral sheet -Width 7.25" 7.25 Boral sheet - Height 128" 128.00 Boral sheet -Thickness 0.075" 0.075 Box 10 Box 11 Boundary SS Wrapper sheet -Width 7.25" 7.25 408.618 408.618 Boundary SS Wrapper sheet -Height 128" 130.00 278.4 278.4 Boundary 55 Wrapper sheet - Thickness 0.075" 0.075 137.13375 137.13375 Inner 55 Wrapper sheet -Thickness 0.0235" 0.0235 9a Dimensions- Width 1.542" 1.542 Box 19 Box 20 9a Dimensions - Height 8" 4 times 8.00 408.618 408.618 9a Dimensions- Thickness 0.0897" 0.0897 278.4 278.4 185.6725 185.6725 9b Dimensions- Width 1" 1.00 EC 68969 9b Dimensions - Height 8" 4 times 8.00 9b Dimensions - Thickness 0.0897" 0.0897 9c Dimensions- Width 1.542" 1.542 Total 55 9c Dimensions- Height 3" 3.00 46,109.43 in"3 9c Dimensions- Thickness 0.0897" 0.0897 Total Boral 9d Dimensions - Width 1 1.00 22,272.00 in"3 9d Dimensions- Height 3" 3.00 9d Dimensions - Thickness 0.0897" 0.0897 Table 5.1.3-10 Region II Racks AI-B4C ss Rack Cell Dimensions (SS Box) Width 8.46" 8.46 Designation RxC Boral Panels Sheating Cell Dimensions (SS Box) Height 161" 161.00 Rack B1 12 X 9 195 195 Cell Dimensions (SS Box) Thickness 0.075" 0.075 Rack B2 12 X 9 195 195 Boral sheet -Width 7.25" 7.25 Rack G1 10 X 9 178 178 140 Bora I sheet -Height 128" 128.00 Rack G2 10x9 157 157

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Boral sheet -Thickness 0.075" 0.075 Rack C 11 x9 180 180 FC08513, R0 Boundary 55 Wrapper sheet -Width 8.00" 8.00 Rack D 11 X 8 161 161 Page 141 of 269 Boundary 55 Wrapper sheet - Height 130" 130.00 Rack E 10 X 10 161 161 Boundary 55 Wrapper sheet -Thickness 0.075" 0.075 Rack F1 10 X 12 218 218 55 Wrapper sheet -Thickness 0.035" 0.035 Rack F2 10 X 12 218 218 Shim -Thickness 0.075" 0.075 Shim -width 9" 9.000 I Panel Total 1663 1663 Corner -Thickness 3/16" 0.1875 Panel Volume (in"3 69.6 36.4 Corner -length 22" 22.000 !Total Volume 115744.80 60533.20 Corner -Width 8.5" 8.500 Holtec Drawing 1000 Rack B1 12 X 9 108 Rack B2 12 X 9 108 Rack G1 10x9 99 Rack G2 10 X 9 88 Rack C llx 9 100 Rack D 11 X 8 90 Rack E 10 X 10 90 EC 68969 Rack F1 10 X 12 120 Region I Rack F2 10 X 12 120 923 Region I Total 160 1083 141

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation 1gs 1001 & 1002 FC08513, R0 Box3 Box4 Box5 Box 6 Box 7 Box 8 Box9 Page 142 of 269 408.618 408.618 408.618 408.618 408.618 408.618 408.618 278.4 278.4 278.4 278.4 278.4 278.4 278.4 137.13375 88.595 88.595 88.595 137.13375 137.13375 88.595 Box 12 Box 13 Box 14 Box 15 Box 16 Box 17 Box 18 408.618 408.618 408.618 408.618 408.618 408.618 408.618 278.4 278.4 278.4 278.4 278.4 278.4 278.4 137.13375 137.13375 137.13375 137.13375 137.13375 185.6725 185.6725 9a 9b 9c 9d SS corners 4.4262 2.8704 0.4150 0.2691 35 .0625 20 18 12 12 16 88.523 51.667 4.979 3.229 561 EC 68969 Mass 13371.733 6065284.6 2171.520 984979 .8 ss ss E-W N-S shims length corners Rack Dimensions overhang 16ths volume panel all panels volume panel all panels 70 8 103.824 77.868 1253.67 16297.77 940.26 9402.56 70 8 103.824 77.868 1253.67 16297.77 940.26 9402.56 70 8 95.172 77.868 1149.20 12641.22 940.26 9402.56 142 70 8 86.52 77.868 1044.73 11492.02 940.26 9402.56

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation 80 8 95.172 77.868 1149.20 13790.42 940.26 9402.56 FC08513, R0 70 8 95.172 69.216 1149.20 13790.42 835 .78 PageM:S.<Df 269 70 8 86.52 86.52 1044.73 11492.02 1044.73 11492.02 70 8 103.824 86.52 1253.67 16297.77 1044.73 11492.02 70 8 103.824 86.52 1253.67 16297.77 1044.73 11492.02 128397.19 89010.91 640 72 0.675 35.0625 432.00 2524.50 0.097544198 Total volume 55 Total volume 55 Total volume 55 EC 68969 3 3 3 cm Region II cm Overall cm 755,598.11 4,603,090.31 5,358,688.42 Total volume AI- Total volume AI- Total volume AI - 3 3 3 B4C cm B4C cm Overall B4C cm 364,972.69 1,896,717.45 2,261,690.13 143

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Reference FC08514, Reference FC08514, Page 144 of 269 Page 44 Page 45 Origen Inputs to MCNP Origen Inputs to MCNP FCS Source FCS Source photons/sf photons/sf Mev mev Mev mev O.OOE+OO O.OOOE+OO O.OOE+OO O.OOOE+OO 2.00E-02 1.670E+18 2.00E-02 2.020E+18 3.00E-02 3.510E+17 3.00E-02 4.250E+l7 4.50E-02 4.250E+17 4 .50E-02 5.140E+17 7.00E-02 3.000E+17 7.00E-02 3.630E+17 l.OOE-01 2.100E+17 l.OOE-01 2.530E+17 1.50E-01 2.550E+17 l.SOE-01 3.080E+17 3.00E-01 1.930E+17 3.00E-01 2.330E+17 4 .50E-01 9.510E+16 4 .50E-01 1.150E+17 7.00E-01 2.040E+18 7.00E-01 2.460E+l8 l.OOE+OO 3.460E+17 l.OOE+OO 4.180E+l7 l.SOE+OO 6.190E+16 l.SOE+OO 7.480E+16 2.00E+OO 4.650E+l5 2.00E+OO 5.630E+15 2.50E+OO 4.670E+15 2.50E+OO 5.650E+15 3.00E+OO 9.100E+13 3.00E+OO 1.100E+14 4.00E+OO 8.240E+12 4.00E+OO 9.970E+12 6.00E+OO 1.040E+10 6.00E+OO 1.260E+10 8.00E+OO 1.200E+09 8.00E+OO 1.450E+09 1.10E+01 1.380E+08 1.10E+Ol 1.670E+08 Total 5.95E+18 Total 7.20E+18 FC08513, R0 EC 68969 144

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Reference FC08514, Reference FC08514, Page 145 of 269 Page 46 Page47 Origen Inputs to MCNP Origen Inputs to MCNP FCS Source FCS Source 18 Month Decay 14 Month Decay neutrons/sf neutrons/sf Mev mev Mev mev 1.00E-11 O.OOE+OO l.OOE-11 O.OOE+OO 1 3.00E-09 3.51E-03 1 3.00E-09 3.63E-03 2 7.50E-09 2.52E-03 2 7.50E-09 2.61E-03 3 l.OOE-08 1.99E-03 3 1.00E-08 2.06E-03 4 2.53E-08 5.85E-03 4 2.53E-08 6.06E-03 5 3.00E-08 2.42E-03 5 3.00E-08 2.51E-03 6 4.00E-08 8.48E-03 6 4.00E-08 8.78E-03 7 5.00E-08 6.07E-03 7 5.00E-08 6.28E-03 8 7.00E-08 1.31E-02 8 7.00E-08 1.35E-02 9 1.00E-07 2.16E-02 9 l.OOE-07 2.24E-02 10 1.50E-07 4.12E-02 10 1.50E-07 4.27E-02 11 2.00E-07 4.88E-02 11 2.00E-07 5.05E-02 12 2.25E-07 2.97E-02 12 2.25E-07 3.08E-02 13 2.50E-07 3.88E-02 13 2.50E-07 4.02E-02 14 2.75E-07 3.26E-02 14 2.75E-07 3.37E-02 15 3.25E-07 5.68E+OO 15 3.25E-07 5.88E+OO 16 3.50E-07 1.96E+OO 16 3.50E-07 2.03E+OO 17 3.75E-07 2.03E+OO 17 3.75E-07 2.10E+OO 18 4.00E-07 2.11E+OO 18 4.00E-07 2.18E+OO 19 6.25E-07 1.85E+01 19 6.25E-07 1.92E+01 20 l.OOE-06 4.42E+01 20 1.00E-06 4.57E+01 21 1.77E-06 1.24E+02 21 1.77E-06 1.28E+02 22 3.00E-06 2.44E+02 22 3.00E-06 2.53E+02 23 4.75E-06 4.53E+02 23 4.75E-06 4.69E+02 24 6.00E-06 3.91E+02 24 6.00E-06 4.05E+02 25 8.10E-06 7.36E+02 25 8.10E-06 7.62E+02 26 l.OOE-05 7.54E+02 26 l.OOE-05 7.81E+02 27 3.00E-05 1.17E+04 27 3.00E-05 1.21E+04 28 l.OOE-04 7.40E+04 28 l.OOE-04 7.66E+04 29 5.50E-04 1.05E+06 29 5.50E-04 1.09E+06 30 3.00E-03 1.34E+07 30 3.00E-03 1.39E+07 31 1.70E-02 1.81E+08 31 1.70E-02 1.87E+08 32 2.50E-02 1.52E+08 32 2.50E-02 1.58E+08 33 l.OOE-01 2.37E+09 33 l.OOE-01 2.45E+09 34 4.00E-01 1.71E+10 34 4.00E-01 1.77E+10 35 9.00E-01 3.73E+10 35 9.00E-01 3.87E+10 36 1.40E+OO 3.73E+10 36 1.40E+OO 3.86E+10 37 1.85E+OO 2.99E+10 37 1.85E+OO 3.09E+10 38 2.35E+00 2.82E+10 38 2.35E+OO 2.91E+10 39 2.48E+00 6.10E+09 39 2.48E+OO 6.31E+09 FC08513, R0 EC 68969 145

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Reference FC08514, Reference FC08514, Page 146 of 269 Page46 Page47 Origen Inputs to MCNP Origen Inputs to MCNP FCS Source FCS Source 18 Month Decay 14 Month Decay neutrons/sf neutrons/sf Mev mev Mev mev 40 3.00E+OO 2.18E+10 40 3.00E+OO 2.25E+10 41 4.80E+OO 3.95E+10 41 4.80E+OO 4.09E+10 42 6.43E+OO 1.12E+10 42 6.43E+OO 1.16E+l0 43 8.19E+OO 3.57E+09 43 8.19E+OO 3.69E+09 44 2.00E+Ol 1.23E+09 44 2.00E+Ol 1.27E+09 Total 2.36E+ll Total FC08513, R0 EC 68969 146

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Surface Number MCNP Fuel Racks Plus Fuel FC08513, R0 47 of 269 6000 pz 929.945 $Top of Fuel Racks Top of Fuel Racks = (995'6" -979'0") + 161" + 7.125" 35.~'0 6050 pz 521.025 $ Bottom of Fuel Rae Bottom of Fuel Racks = (995'6" -979'0") + 7.125" 17.094 Not Used East Edge of Fuel Racks =-(60'6" + 5'6" + 5'6" + 1.578" -71.632 Not Used West Edge of Fuel Racks =-(60'6" + 5'6" + 5'6" + 20'7") + 0.698" -92.025 Not Used North Edge of Fuel Racks =+ 2'6" + 33'3"- 3.203" 35.483 Not Used South Edge of Fuel Racks =+ 2'6" + 3.042" 2.753 Not Used South Edge of CPA =+ 2'6" + 33'3"- 96"- 3.813" 27.432 Not Used East Edge of CPA =-(60'6" + 5'6" + 5'6" + 20'7") + 1.313" -83.974 Alternate dimensions 2250 px -2179.32 $ SFP East Inner wal East Edge of Fuel Racks =-(60'6" + 5'6" + 5'6" -71.500 6200 px -2770.022 $West Boundary of West Edge of Fuel Racks =-(60'6" + 5'6" + 5'6" + 19.38) -90.880 6300 py 1024.128 $ North Boundary o North Edge of Fuel Racks =+ 2'6" + 33'3"- 3.203" 33 .600 2050 py 76.20 $ SFP South Inner w South Edge of Fuel Racks =+ 2'6" + 3.042" 2.500 Fuel + Rack SQ Ft EC 68969 667.47 602.66 19.38 31.10 FCS Fuel Source 944

                                                                                   -2179.3200                                          71.50
                                                                                   -2770.0220                                          90.88               19.38 76.2000                                           2.50               31.10 1024.1280                                           33.60               13.42 929.9450                                           30.51 521.0250                                           17.09 147

FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation FC08513, R0 Page 148 of 269 929.94480 11405-S-61 + 1002, 1004 z 168.125 14.010 521.02512 11405-S-61 + 1002, 1004 z 7.125 0.594

                       -2183.34336    11405-S-61              X                    1.578        0.132
                       -2804.93206    11405-S-61              X                    0.698        0.058 1081.52184    11405-S-61              y                    3.203        0.267 83.91144    11405-S-61              y                    3.042        0.253 836.12736    11405-S-61              y                    3.813        0.318
                       -2559.53758    TDB-                    X                    1.313        0.109
                       -2179.32000    11405-S-61             X
                       -2770.02240    11405-S-61             X 1024.12800    11405-S-61             y 76.20000    11405-S-61             y EC 68969 0.95021 602.65509 Fuel Volume (Fuel+ Racks+ Air) 228,971,853.30 cm3 8086.1 ft3 148

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 149 of 269 SFP feet OPPD Drawing Material Elevation outer width =5 .5+20'7"+5 .5 31.5833 11405-S-51 Concrete 995.5- 1007 outer length =5.5+33 '3"+4 42.75 11405-S-51 Wall thinckness 5'6" 5.5 11405-S-61 Wall thinckness Canal end 4'0" 4.0 11405-S-61 Top of Pool 1038'6" 1038.5 11405-S-61 Bottom of Pool 995.5' 43.0 11405-S-61 Pool basemat 10'+989'0" 16.5 11405-S-60 Distance from Containment toAUX =55'6" + 5'0" 60.5 11405-S-48 SFP Weir gate

                          @1008.5' Width                    5'                     11405-S-61
                          @1039' Width                     7'8"                    11405-S-61 EC 68969 149

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 150 of 269 Surface Number MCNP Dimensional feet em OPPD Drawing Attribute 2200 px -2011. East Exterior Wall of SFP = 60'6" + 5'6" -66.00 -2011.680 11405-S-61 X 2250 px -2179. East Interior Wall of SFP = 60'6" + 5'6" + 5'6" -71.50 -2179.320 11405-S-61 X 2300 px -2806. West Interior Wall of SFP = 60'6" + 5'6" + 5'6" -20'7" -92.08 -2806.700 11405-S-61 X 2350 px -2974.3 West Exterior Wall of SFP = 60'6" + 5'6" + 5'6" + 20'7" + 5'6" -97.58 -2974.340 11405-S-61 X 2000 py -45.7..! South Exterior Wall of SFP =-1'6" -1.50 -45.720 11405-S-61 y 2050 py 76.20 South Interior Wall of SFP = + 2'6" 2.50 76.200 11405-S-61 y 2100 py 1089.6 North Interior Wall of SFP = + 2'6" + 33'3" 35.75 1089.660 11405-S-61 y 2150 py 1257.3 North Exterior Wall of SFP = + 2'6" + 33'3" + 5'6" 41.25 1257.300 11405-S-61 y 2400 py -198.1 South Interior Wall ofTC =- 1'6"- 5'0" -6.50 -198.120 11405-S-61 y 2450 py -350.5 South Exterior Wall ofTC =- 1'6 11 - 5'0" - 5'0" -11.50 -350.520 11405-S-61 y EC 68969 1550 px -1844.C East Exterior Wall of TC = 60'6" -60.50 -1844.040 11405-S-61 X Not Used East Interior Wall of TC at bottom = 60'6" + 2" -62.50 -1905.000 11405-S-61 X 2550 px -1969. East Interior Wall of TC overall = 60'6" + 2'0" + 2'0" + 1.5" -64.63 -1969.770 11405-S-61 X 2600 py 175.2 North Exterior Wall of TC = + 1'6" + 4'3" 5.75 175.260 11405-S-61 y 2800 px -2219.S East Edge of Weir Gate =- (60'6" + 2'6" + 13'1" + Avg (4' + 2'6") -72.83 -2219.960 11405-S-61 X 2850 px -2418.C West Edge of Weir Gate =- {60'6" + 2'6" + 13'1" - Avg (4' + 2'6") -79.33 -2418.080 11405-S-61 X 150

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 151 of 269 MCNP cartesian coordinate Dimensi Surface on Number Perspect MCNP feet em DWG ive 1500 px - South Exterior Wall of Aux (Room 3) =- 62' -62.00 -1889.76 11405-S-51 y 1550 px - South Interior Wall of Aux (Room 3) =- 62' + 1'6" -60.50 -1844.04 11405-S-51 y 1150 PY 2 North Interior Wall of Aux (Room 3) = +68'6" 68.50 2087.88 11405-S-57 y 1100 py 2 North Exterior Wall of Aux (Room 3) = + 68'6" + 1'6" 70.00 2133.60 11405-S-57 y 1050 py 4 North Interior Wall of Aux (Room 69, below 1044') = + 134'0" - 1'6" 132.50 4038.60 11405-S-55 y 1000 py 4 North Exterior Wall of Aux (Room 69, below 1044') = + 134'0" 134.00 4084.32 11405-S-55 y Not Used South Interior Wall of East of SFP (Room 69, below 1044') = + 21'6"- 4'3" -17.25 -525.78 11405-S-55 y Not Used South Exterior Wall of East of SFP (Room 69, below 1044') = + 21'6" + 1'6"- 4'3" -18.75 -571.50 11405-S-55 y 1300 py - East Exterior Wall (Room 25A) =- 60'6" -60.50 -1844.04 11405-S-51 X 1350 py - East Interior Wall (Room 25A) = - 62' - 1'6" -62.00 -1889.76 11405-S-51 X 2900 px - West Interior Wall (Room 25A) = -(60'6" + 5'6" + 5'6" + 20'7" + 5'6" -1.5) -96.08 -2928.62 11405-S-57 X 2350 px - West Exterior Wall (Room 25A) = - (60'6" + 5'6" + 5'6" + 20'7" + 5'6") -97.58 -2974.34 11405-S-57 X 2700 py - South Interior Wall (Room 25A) = - (62' + 21'11" + 0'9") -39.75 -1211.58 11405-S-57 y 2750 py - South Exterior Wall (Room 25A) = - (62' + 21'11" - 0'9") -41.25 -1257.30 11405-S-57 y EC 68969 1650 py 1 South Exterior Wall of CR (Room 77) = 29'0" + 36'0" - 1'0" 64.00 1950.72 11405-S-55 y 1600 py 1 South Interior Wall of CR (Room 77) = 29 '0" + 36'0" + 0'3" 65 .25 1988.82 11405-S-55 y 1850 px 3 East Exterior Wall of CR (Room 77) = 118'0" 118.00 3596.64 11405-S-55 X 1800 px 3 East Interior Wall of CR (Room 77) = 118'0"- 1'6" 116.50 3550.92 11405-S-55 X 1750 px 1 West Interior Wall of CR (Room 77) = (1'0" + 18'6" + 18'6" + 0'9") 38.75 1181.10 11405-S-55 X 1700 px 1 West Exterior Wall of CR (Room 77) = (1 '0" + 18'6" + 18'6"- 0'9" ) 37.25 1135.38 11405-S-55 X 1550 px - East Exterior Wall of Aux (Room 3) =- 60'6" -60.50 -1844.04 11405-S-51 X 1500 px -1 East Interior Wall of Aux (Room 3) = - 60'6" - 1'6" -62.00 -1889.76 11405-S-51 X 1450 px -3 West Interior Wall of Aux (Room 3) = - 60'6" - 64'6" + 1'6" -123.50 -3764.28 11405-S-51 X 1400 px -3 West Exterior Wall of Aux (Room 3) = - 60'6" - 64'6" -125 .00 -3810.00 11405-S-51 X 3100 pz 1 Room 67, 68, 69 Floor (1025') = + 1025'0" - 979'0" 46.00 1402.08 11405-S-63 3110 pz 1 Room 67, 68, 69 Thickness (6") = + 1025'0"- 979'0"- 0'6" 45 .50 1386.84 11405-S-59 3000 pz 3 Room 3 Roof (1083') = + 1083'0" - 979'0" 104.00 3169.92 11405-S-63 3015 pz 3 Room 3 Thickness (6") = + 1083'0" - 979'0"- 0'6" 103.50 3154.68 11405-S-59 151 3020 pz 2 Room 77 Roof (1057') = + 1057'0"- 979'0" 78.00 2377.44 11405-S-63 z 3030 pz 2 Room 77 Thickness (1'6") = + 1057'0" - 979'0" -1'6" 76.50 2331.72 11405-S-59

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 o.,.,...,. 11:;.' of269 3060 pz 1 Room 69 Roof {1044') = + 1044'0" - 979'0" 65.00 1981.20 11405-S-63 "' z 3070 pz 1 Room 69 Thickness {6") = + 1044'0"- 979'0" -0'6" 64.50 1965.96 11405-S-59 3040 pz 1 Room 77 Floor {1036') = + 1036'0" - 979'0" 57.00 1737.36 11405-S-63 z 3050 pz 1 Room 77 Thickness {0'6") = + 1036'0" - 979'0" - 0'6" 56.50 1722.12 11405-S-59 z Not Used Containment top = + 1128'4.5"- 979 '0"+ 0'3" 149.63 4560.57 11405-S-63 Not Used Containment Dome Inside elevation = + 1128'4.5"- 979 '0" 149.38 4552.95 11405-S-59 Not Used Top of Cylinder 1099' Elevation = + 1099'0" - 979'0" 120.00 3657.60 11405-S-59 Not Used Containment Basemat Top Elevation = + 991'0" - 979'0" 12.00 365 .76 11405-S-63 Not Used Containment Basemat Bottom Elevation = + 979'0" - 979'0" 0.00 0.00 11405-S-12 3200 pz 1 SFP Walkway = + 1038'6" - 979 '0" 59.00 1798.32 11405-S-61 3210 pz ! Bottom of Weir Gate = + 1008'6" - 979'0" - 1'6" 29.50 899.16 11405-S-61 3220 pz ' Pool Bottom = + 995'6"- 979'0" 16.50 502.92 11405-S-61 Not Used Aux Bldg Basemat Top Elevation = + 989'0" - 979'0" 10.00 304.80 11405-S-61 3250 pz Aux Bldg Basemat Bottom Elevation = + 983'6" - 979'0" 4.50 137.16 11405-S-61 EC 68969 7000 pz Plant Grade Elevation = + 1004'6" - 979'0" 25.50 777.24 11405-S-59 z 7200 pz 3 EAB Elevation Hwy 75 = + 1080'0" - 979'0" 101.00 3078.48 11405-S-251-1:12 7100 px -4 EAB Elevation Hwy 75 Step = -1540 -1540.00 -46939.20 11405-S-270 X 10000 so LPZ = 1 Mile (5280) 5280.00 160934.40 so Not Used LPZ = 9 Miles {9*5280) 47520.00 1448409.60 so Containment Bu1ldmg Not Used Liner Interior Radius 55'0" 55.00 1676.40 11405-S-55 cz Not Used Liner Thickness 0'0.25" 0.021 0.63 11405-S-55 cz 5100 rcc 0 Containment Interior Radius 1677.035 5000 rcc 0 Containment Exterior Radius 58'10.75" 58.90 1795.145 11405-S-55 cz Not Used Containment Dome Thickness 3' 3.00 11405-S-2 s Not Used Containment Dome radius 90'0" 90.00 2743.20 11405-S-2 s Not Used Containment Dome Inside elevation 1128'4.5" 149.38 4552.95 11405-S-2 Not Used Containment Dome radius at 1099' elevation 20'0" 20.00 11405-S-2 s Not Used Containment radius at 1099' elevation for dome shape 35'0" 35.00 11405-S-2 s Railroad Siding Doors Not Used Elevation at botton {1004 '0", Use Plant Grade level 1004'6") = + 1004'6" - 979'0" 25.50 777.24 11405-S-51 z 3330,3340, Top of Door Opening {22' above 1004'0") = + 1004'0" + 22'0" - 979'0" 47.00 1432.56 11405-S-51 & 11405-S-74 z 3330, 3340, Door Opening w idth including Frame{16'6") = 16' 16.00 487.68 11405-S-51 & 11405-S-74 y 152 3330, 3340, South Door Opening from CL of Containemt {16'6 ") = - 62' + 1'6" + 2'11.5" -57.54 -1753.87 11405-S-51 & 11405-S-74 y

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 o .... ,.,,.,.

                                                                                                                                                                   ....       11:.   ,...f'>ao 3330,3340,: North Door Opening from CL of Containemt (16'6")   =- 62' + 1'6" + 2'11.5" 16'                -41 .54   -1266.19   11405-S-51 & 11405-S-74                     y 3330  rpp - Door 1004-1A design thickness (20 ga)              = 20 guage steel (.0375")                0.03750       0.0953   232552                                      X 3340  rpp - Door 1004-1C Sheathing design thickness (14 ga)    = 14 guage steel (.078125")              0.07813       0.1984   G-576                                       X 3350  rpp - Door 1004-1C design insulation thickness           =Foam insulation (3.75")                 3.84375       9.7631   G-576                                       X Use remainder of door thickness   (4.0 0.078125*2)=3.84375 Not Used     Door "B" Frame Dimension (Ignore Security Gate)   N/A EC 68969 153

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 154 of 269 AUX Building feet OPPD Drawin Dwg Loc Elevation North Outerwall to Containment Centerline 134' 134 11405-S-51 995.5 - 1007 North Wall Thickness 1.5' 1.5 11405-S-67 West Wall Thickness 2.0' 2 11405-S-67 South Wall Thickness 2.0' 2 11405-S-67 East Wall Thickness 2.0' 2 11405-S-67 East Outerwall to Containment Centerline 60.5' 60.5 11405-S-67 Roof Elevation 1083' 11405-S-60 Roof Thickness West Wall to East Wall 64.5' 64.5 11405-S-60 North Wall to South Wall 132.0' 132 11405-S-60 EC 68969 Containment Building Liner Interior Rad ius 55'0" 55.000 11405-S-55 Li ner Thickness 0'0.25 " 0.021 11405-S-55 Exterior Radius 58'10.75 " 58.896 11405-S-55 Basemat Top Elevation 989'0" 989.0 11405-S-60 Containment Basemat Bottom Elevation 979'0" 979.0 11405-S-64 Aux Bldg Basemat Bottom Elevation 983'6" 983.5 11405-S-64 SFP Bottom Elevation 995 '6" 995.5 11405-S-64 Containment Dome Thickness 3' 3.00 11405-5-2 Containment Dome radius 90'0" 90.00 11405-S-2 Containment Dome Inside elevat ion 1128'4.5" 1128.375 11405-S-2 Containment Dome radius at 1099' elevation 20'0" 20.00 11405-S-2 Containment radius at 1099' elevation for dome shape 35 '0" 35.00 11405-S-2 Containment Cylinder Elevation 1099'0" 1099.00 11405-S-2 Top of Dome to 1099' Elevation 29'4.5" 29.375 11405-S-2 154 Railroad Siding Doors Insulation Thickness

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 x-direction Steel y-direction 14 gauge Steel y-direction Steel z-direction 14 gauge Steel z-direction x-directionl2o gauge Steel 3764.375251 3764.280001 0.095251 Gauge Measure (inches) numerical Centimeters 14 5/64 0.078125 1.984375 0.078125 20 3/80 0.0375 0.9525 EC 68969 0.037500 155

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 156 of 269 Scatch pad Area Scatch pad Area c Rol lup Door 1 004-1C c Volume of Door Area 3330 r pp -3764 . 37 5 2 5 - 3764.28 - 1 7 5 3 . 87 -1 266. 1 90 777 . 24 1432 . 56 c 30440.6 238045.571 c c Ver ti c al lift Door 1004-1A c .:5L46998.409 3340 r pp -3789 . 90 -3779.74 -1753 . 87 -1266.190 3115""6) 3 4 7 7 7.24 1432 . 56 131236.0951 1026266.26 33 5 0 rpp -3789.70156 - 3 7 79 . 93844 -1753.67156 - 1266.38844 777.43844 1432.36 1 56 3115762.314 24365261.3 c 3230 pz 365 . 76 $ Top of Conta i n me n t Basemat (991' 0") 3260 pz 0.00 $ Bot t om o f Con tainme nt Basemat (979'0") 14611492.84 c 3 11334053 p 13657.5349 1.755004 g/cm EC 68969 4 05 4 - 7 . 82 -3330 $ Railroad Siding Door 1004-1C 407 4 -7 . 82 - 3340 3350 $ Railroad Siding Door 1004-1A 408 5 -0.0482 - 3350 $ Rai l road Siding Door 1004 - 1A i nsul 1400 px -3810.00 $West Outer Wall (Rms. 3 & 25) 1450 px -3764.28 $West Inner Wall (Rms . 3 & 25) 3300 pz 1432.56 $ Aux Bldg Door Top of Open ing 3310 py -1266.19 $ Aux Bldg Door North side of Opening 3320 py -1753.87 $ Aux Bldg Door South side of Opening 7000 pz 777.24 $ Plant Grade Elevation 1004'6" 156

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 157 of 269 OPPD Drawings: 11405-S-02, 11405-S-08, 11405-S-08 degrees radians X dome Y dome degrees radians edge. edge. (Not used) (Not used) 120.042 60.667 X y X y 0 0.000 55.00 120.04 0 0.000 58.90 120.04 0 0.000 90.00 59.38 1 0.017 55.00 120.39 1 0.017 58.90 120.44 1 0.017 89.99 60.95 2 0.035 54.99 120.74 2 0.035 58.89 120.84 2 0.035 89.95 62.52 3 0.052 54.97 121.09 3 0.052 58.87 121.25 3 0.052 89.88 64.09 4 0.070 54.95 121.44 4 0.070 58.84 121.65 4 0.070 89.78 65.65 5 0.087 54.92 121.79 5 0.087 58.81 122.05 5 0.087 89.66 67.22 6 0.105 54.89 122.13 6 0.105 58.77 122.45 6 0.105 89.51 68.78 7 0.122 54.85 122.48 7 0.122 58.73 122.84 7 0.122 89.33 70.34 8 0.140 54.81 122.83 8 0.140 58.68 123.24 8 0.140 89.12 71.90 9 0.157 54.75 123.17 9 0.157 58.62 123.64 9 0.157 88.89 73.45 EC 68969 10 0.175 54.70 123.51 10 0.175 58.55 124.04 10 0.175 88.63 75.00 11 0.192 54.63 123.86 11 0.192 58.48 124.43 11 0.192 88.35 76.55 12 0.209 54.56 124.20 12 0.209 58.40 124.82 12 0.209 88.03 78.09 13 0.227 54.49 124.54 13 0.227 58.31 125.22 13 0.227 87.69 79.62 14 0.244 54.41 124.88 14 0.244 58.22 125.61 14 0.244 87.33 81.15 15 0.262 54.32 125.22 15 0.262 58.12 125.99 15 0.262 86.93 82.67 16 0.279 54.23 125.55 16 0.279 58.01 126.38 16 0.279 86.51 84.18 17 0.297 54.13 125.89 17 0.297 57.90 126.77 17 0.297 86.07 85.69 18 0.314 54.02 126.22 18 0.314 57.77 127.15 18 0.314 85.60 87.19 19 0.332 53.91 126.55 19 0.332 57.65 127.53 19 0.332 85.10 88.68 20 0.349 53.79 126.88 20 0.349 57.51 127.91 20 0.349 84.57 90.16 21 0.367 53.67 127.21 21 0.367 57.37 128.28 21 0.367 84.02 91.63 22 0.384 53.54 127.53 22 0.384 57.23 128.66 22 0.384 83.45 93.09 23 0.401 53.41 127.86 23 0.401 57.07 129.03 23 0.401 82.85 94.54 24 0.419 53.27 128.18 24 0.419 56.91 129.40 24 0.419 82.22 95.98 25 0.436 53.13 128.49 25 0.436 56.75 129.76 25 0.436 81.57 97.41 26 0.454 52.98 128.81 26 0.454 56.57 130.12 26 0.454 80.89 98.83 27 0.471 52.82 129.12 27 0.471 56.39 130.48 27 0.471 80.19 100.23 28 0.489 52.66 129.43 28 0.489 56.21 130.84 28 0.489 79.47 101.63 157 29 0.506 52.49 129.74 29 0.506 56.02 131.19 29 0.506 78.72 103.01 30 0.524 52.32 130.04 30 0.524 55.82 131.54 30 0.524 77.94 104.37

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 158 of 269 31 0.541 52.14 130.34 31 0.541 55.61 131.89 31 0.541 77.15 105.73 32 0.559 51.96 130.64 32 0.559 55.41 132.23 32 0.559 76.32 107.07 33 0.576 51.77 130.93 33 0.576 55.19 132.57 33 0.576 75.48 108.39 34 0.593 51.58 131.23 34 0.593 54.97 132.90 34 0.593 74.61 109.70 35 0.611 51.38 131.51 35 0.611 54.74 133.23 35 0.611 73.72 111.00 36 0.628 51.18 131.80 36 0.628 54.51 133.56 36 0.628 72.81 112.28 37 0.646 50.97 132.08 37 0.646 54.27 133.88 37 0.646 71.88 113.54 38 0.663 50.76 132.36 38 0.663 54.02 134.20 38 0.663 70.92 114.78 39 0.681 50.54 132.63 39 0.681 53.77 134.52 39 0.681 69.94 116.01 40 0.698 50.32 132.90 40 0.698 53.52 134.83 40 0.698 68.94 117.23 41 0.716 50.09 133.16 41 0.716 53.26 135.13 41 0.716 67.92 118.42 42 0.733 49.86 133.42 42 0.733 52.99 135.43 42 0.733 66.88 119.60 43 0.750 49.63 133.68 43 0.750 52.72 135.73 43 0.750 65.82 120.75 44 0.768 49.39 133.94 44 0.768 52.44 136.02 44 0.768 64.74 121.89 45 0.785 49.14 134.18 45 0.785 52.16 136.31 45 0.785 63.64 123.01 46 0.803 48.89 134.43 46 0.803 51.88 136.59 46 0.803 62.52 124.12 47 0.820 48.64 134.67 47 0.820 51.59 136.86 47 0.820 61.38 125.20 EC 68969 48 0.838 48.38 134.90 48 0.838 51.29 137.13 48 0.838 60.22 126.26 49 0.855 48.12 135.14 49 0.855 50.99 137.40 49 0.855 59.05 127.30 50 0.873 47.86 135.36 50 0.873 50.68 137.66 50 0.873 57.85 128.32 51 0.890 47.59 135.58 51 0.890 50.37 137.92 51 0.890 56.64 129.32 52 0.908 47.31 135.80 52 0.908 50.06 138.17 52 0.908 55.41 130.30 53 0.925 47.04 136.01 53 0.925 49.74 138.41 53 0.925 54.16 131.25 54 0.942 46.76 136.22 54 0.942 49.42 138.65 54 0.942 52.90 132.19 55 0.960 46.47 136.43 55 0.960 49.09 138.88 55 0.960 51.62 133.10 56 0.977 46.18 136.62 56 0.977 48.76 139.11 56 0.977 50.33 133.99 57 0.995 45.89 136.82 57 0.995 48.43 139.33 57 0.995 49.02 134.86 58 1.012 45.60 137.00 58 1.012 48.09 139.55 58 1.012 47.69 135.70 59 1.030 45.30 137.19 59 1.030 47.75 139.76 59 1.030 46.35 136.52 60 1.047 45.00 137.36 60 1.047 47.40 139.96 60 1.047 45.00 137.32 61 1.065 44.70 137.53 61 1.065 47.05 140.16 61 1.065 43.63 138.09 62 1.082 44.39 137.70 62 1.082 46.70 140.35 62 1.082 42.25 138.84 63 1.100 44.08 137.86 63 1.100 46.34 140.54 63 1.100 40.86 139.57 64 1.117 43.77 138.02 64 1.117 45.98 140.71 64 1.117 39.45 140.27 65 1.134 43.45 138.17 65 1.134 45.62 140.89 65 1.134 38.04 140.94 66 1.152 43.13 138.31 66 1.152 45.25 141.05 66 1.152 36.61 141.59 158 67 1.169 42.81 138.45 67 1.169 44.89 141.21 67 1.169 35.17 142.22 68 1.187 42.49 138.59 68 1.187 44.52 141.37 68 1.187 33.71 142.82

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 159 of 269 69 1.204 42.17 138.71 69 1.204 44.14 141.51 69 1.204 32.25 143.40 70 1.222 41.84 138.84 70 1.222 43.77 141.65 70 1.222 30.78 143.95 71 1.239 41.51 138.95 71 1.239 43.39 141.79 71 1.239 29.30 144.47 72 1.257 41.18 139.06 72 1.257 43.01 141.92 72 1.257 27.81 144.97 73 1.274 40.85 139.17 73 1.274 42.62 142.04 73 1.274 26.31 145.44 74 1.292 40.51 139.27 74 1.292 42.24 142.15 74 1.292 24.81 145.89 75 1.309 40.18 139.36 75 1.309 41.85 142.26 75 1.309 23.29 146.31 76 1.326 39.84 139.45 76 1.326 41.46 142.36 76 1.326 21.77 146.70 77 1.344 39.50 139.53 77 1.344 41.07 142.45 77 1.344 20.25 147.07 78 1.361 39.16 139.60 78 1.361 40.68 142.54 78 1.361 18.71 147.41 79 1.379 38.82 139.67 79 1.379 40.29 142.62 79 1.379 17.17 147.72 80 1.396 38.47 139.74 80 1.396 39.89 142.69 80 1.396 15.63 148.01 81 1.414 38.13 139.80 81 1.414 39.50 142.76 81 1.414 14.08 148.27 82 1.431 37.78 139.85 82 1.431 39.10 142.82 82 1.431 12.53 148.50 83 1.449 37.44 139.89 83 1.449 38.70 142.87 83 1.449 10.97 148.70 84 1.466 37.09 139.93 84 1.466 38.30 142.92 84 1.466 9.41 148.88 85 1.484 36.74 139.97 85 1.484 37.90 142.95 85 1.484 7.84 149.03 EC 68969 86 1.501 36.40 139.99 86 1.501 37.50 142.99 86 1.501 6.28 149.16 87 1.518 36.05 140.01 87 1.518 37.10 143.01 87 1.518 4.71 149.25 88 1.536 35.70 140.03 88 1.536 36.70 143.03 88 1.536 3.14 149.32 89 1.553 35.35 140.04 89 1.553 36.30 143.04 89 1.553 1.57 149.36 90 1.571 35.00 140.04 90 1.571 35.90 143.04 90 1.571 0.00 149.37 91 1.588 -35.35 140.04 91 1.588 35.50 143.04 91 1.588 -1.57 149.36 92 1.606 -35.70 140.03 92 1.606 35.10 143.03 92 1.606 -3.14 149.32 93 1.623 -36.05 140.Q1 93 1.623 34.70 143.01 93 1.623 -4.71 149.25 94 1.641 -36.40 139.99 94 1.641 34.30 142.99 94 1.641 -6.28 149.16 95 1.658 -36.74 139.97 95 1.658 33.90 142.95 95 1.658 -7.84 149.03 96 1.676 -37.09 139.93 96 1.676 33.50 142.92 96 1.676 -9.41 148.88 97 1.693 -37.44 139.89 97 1.693 33.10 142.87 97 1.693 -10.97 148.70 98 1.710 -37.78 139.85 98 1.710 32.70 142.82 98 1.710 -12.53 148.50 99 1.728 -38.13 139.80 99 1.728 32.30 142.76 99 1.728 -14.08 148.27 100 1.745 -38.47 139.74 100 1.745 31.91 142.69 100 1.745 -15.63 148.01 101 1.763 -38.82 139.67 101 1.763 31.51 142.62 101 1.763 -17.17 147.72 102 1.780 -39.16 139.60 102 1.780 31.12 142.54 102 1.780 -18.71 147.41 103 1.798 -39.50 139.53 103 1.798 30.73 142.45 103 1.798 -20.25 147.07 159 104 1.815 -39.84 139.45 104 1.815 30.34 142.36 104 1.815 -21 .77 146.70 105 1.833 -40.18 139.36 105 1.833 29.95 142.26 105 1.833 -23.29 146.31 106 1.850 -40.51 139.27 106 1.850 29.56 142.15 106 1.850 -24.81 145.89

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 160 of 269 107 1.868 -40.85 139.17 107 1.868 29.18 142.04 107 1.868 -26.31 145.44 108 1.885 -41.18 139.06 108 1.885 28.79 141.92 108 1.885 -27.81 144.97 109 1.902 -41.51 138.95 109 1.902 28.41 141.79 109 1.902 -29.30 144.47 110 1.920 -41.84 138.84 110 1.920 28.03 141.65 110 1.920 -30.78 143.95 111 1.937 -42.17 138.71 111 1.937 27.66 141.51 111 1.937 -32.25 143.40 112 1.955 -42.49 138.59 112 1.955 27.28 141.37 112 1.955 -33.71 142.82 113 1.972 -42.81 138.45 113 1.972 26.91 141.21 113 1.972 -35.17 142.22 114 1.990 -43.13 138.31 114 1.990 26.55 141.05 114 1.990 -36.61 141.59 115 2.007 -43.45 138.17 115 2.007 26.18 140.89 115 2.007 -38.04 140.94 116 2.025 -43.77 138.02 116 2.025 25.82 140.71 116 2.025 -39.45 140.27 117 2.042 -44.08 137.86 117 2.042 25.46 140.54 117 2.042 -40.86 139.57 118 2.059 -44.39 137.70 118 2.059 25.10 140.35 118 2.059 -42.25 138.84 119 2.077 -44.70 137.53 119 2.077 24.75 140.16 119 2.077 -43.63 138.09 120 2.094 -45.00 137.36 120 2.094 24.40 139.96 120 2.094 -45.00 137.32 121 2.112 -45.30 137.19 121 2.112 24.05 139.76 121 2.112 -46.35 136.52 122 2.129 -45.60 137.00 122 2.129 23.71 139.55 122 2.129 -47.69 135.70 123 2.147 -45.89 136.82 123 2.147 23.37 139.33 123 2.147 -49.02 134.86 EC 68969 124 2.164 -46.18 136.62 124 2.164 23.04 139.11 124 2.164 -50.33 133.99 125 2.182 -46.47 136.43 125 2.182 22.71 138.88 125 2.182 -51.62 133.10 126 2.199 -46.76 136.22 126 2.199 22.38 138.65 126 2.199 -52.90 132.19 127 2.217 -47.04 136.01 127 2.217 22.06 138.41 127 2.217 -54.16 131.25 128 2.234 -47.31 135.80 128 2.234 21.74 138.17 128 2.234 -55.41 130.30 129 2.251 -47.59 135.58 129 2.251 21.43 137.92 129 2.251 -56.64 129.32 130 2.269 -47.86 135.36 130 2.269 21.12 137.66 130 2.269 -57.85 128.32 131 2.286 -48.12 135.14 131 2.286 20.81 137.40 131 2.286 -59.05 127.30 132 2.304 -48.38 134.90 132 2.304 20.51 137.13 132 2.304 -60.22 126.26 133 2.321 -48.64 134.67 133 2.321 20.21 136.86 133 2.321 -61.38 125.20 134 2.339 -48.89 134.43 134 2.339 19.92 136.59 134 2.339 -62.52 124.12 135 2.356 -49.14 134.18 135 2.356 19.64 136.31 135 2.356 -63.64 123.01 136 2.374 -49.39 133.94 136 2.374 19.36 136.02 136 2.374 -64.74 121.89 137 2.391 -49.63 133.68 137 2.391 19.08 135.73 137 2.391 -65.82 120.75 138 2.409 -49.86 133.42 138 2.409 18.81 135.43 138 2.409 -66.88 119.60 139 2.426 -50.09 133.16 139 2.426 18.54 135.13 139 2.426 -67.92 118.42 140 2.443 -50.32 132.90 140 2.443 18.28 134.83 140 2.443 -68.94 117.23 141 2.461 -50.54 132.63 141 2.461 18.03 134.52 141 2.461 -69.94 116.01 142 2.478 -50.76 132.36 142 2.478 17.78 134.20 142 2.478 -70.92 114.78 160 143 2.496 -50.97 132.08 143 2.496 17.53 133.88 143 2.496 -71.88 113.54 144 2.513 -51.18 131.80 144 2.513 17.29 133.56 144 2.513 -72.81 112.28

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 161 of 269 145 2.531 -51.38 131.51 145 2.531 17.06 133.23 145 2.531 -73.72 111.00 146 2.548 -51.58 131.23 146 2.548 16.83 132.90 146 2.548 -74.61 109.70 147 2.566 -51.77 130.93 147 2.566 16.61 132.57 147 2.566 -75.48 108.39 148 2.583 -51.96 130.64 148 2.583 16.39 132.23 148 2.583 -76.32 107.07 149 2.601 -52.14 130.34 149 2.601 16.19 131.89 149 2.601 -77.15 105.73 150 2.618 -52.32 130.04 150 2.618 15.98 131.54 150 2.618 -77.94 104.38 151 2.635 -52.49 129.74 151 2.635 15.78 131.19 151 2.635 -78.72 103.01 152 2.653 -52.66 129.43 152 2.653 15.59 130.84 152 2.653 -79.47 101.63 153 2.670 -52.82 129.12 153 2.670 15.41 130.48 153 2.670 -80.19 100.23 154 2.688 -52.98 128.81 154 2.688 15.23 130.12 154 2.688 -80.89 98.83 155 2.705 -53.13 128.49 155 2.705 15.05 129.76 155 2.705 -81.57 97.41 156 2.723 -53.27 128.18 156 2.723 14.89 129.40 156 2.723 -82.22 95.98 157 2.740 -53.41 127.86 157 2.740 14.73 129.03 157 2.740 -82.85 94.54 158 2.758 -53.54 127.53 158 2.758 14.57 128.66 158 2.758 -83.45 93.09 159 2.775 -53.67 127.21 159 2.775 14.43 128.28 159 2.775 -84.02 91.63 160 2.793 -53.79 126.88 160 2.793 14.29 127.91 160 2.793 -84.57 90.16 161 2.810 -53.91 126.55 161 2.810 14.15 127.53 161 2.810 -85.10 88.68 EC 68969 162 2.827 -54.02 126.22 162 2.827 14.03 127.15 162 2.827 -85.60 87.19 163 2.845 -54.13 125.89 163 2.845 13.90 126.77 163 2.845 -86.07 85.69 164 2.862 -54.23 125.55 164 2.862 13.79 126.38 164 2.862 -86.51 84.18 165 2.880 -54.32 125.22 165 2.880 13.68 125.99 165 2.880 -86.93 82.67 166 2.897 -54.41 124.88 166 2.897 13.58 125.61 166 2.897 -87.33 81.15 167 2.915 -54.49 124.54 167 2.915 13.49 125.22 167 2.915 -87.69 79.62 168 2.932 -54.56 124.20 168 2.932 13.40 124.82 168 2.932 -88.03 78.09 169 2.950 -54.63 123.86 169 2.950 13.32 124.43 169 2.950 -88.35 76.55 170 2.967 -54.70 123.51 170 2.967 13.25 124.04 170 2.967 -88.63 75.00 171 2.985 -54.75 123.17 171 2.985 13.18 123.64 171 2.985 -88.89 73.45 172 3.002 -54.81 122.83 172 3.002 13.12 123.24 172 3.002 -89.12 71.90 173 3.019 -54.85 122.48 173 3.019 13.07 122.85 173 3.019 -89.33 70.34 174 3.037 -54.89 122.13 174 3.037 13.03 122.45 174 3.037 -89.51 68.78 175 3.054 -54.92 121.79 175 3.054 12.99 122.05 175 3.054 -89.66 67.22 176 3.072 -54.95 121.44 176 3.072 12.96 121.65 176 3.072 -89.78 65.65 177 3.089 -54.97 121.09 177 3.089 12.93 121.25 177 3.089 -89.88 64.09 178 3.107 -54.99 120.74 178 3.107 12.91 120.84 178 3.107 -89.95 62.52 179 3.124 -55.00 120.39 179 3.124 12.90 120.44 179 3.124 -89.99 60.95 180 3.142 -55.00 120.04 180 3.142 12.90 120.04 180 3.142 -90.00 59.38 161

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 162 of 269 OPPD Drawings: 11405-S-02, 11405-S-08, 11405-S-08 Angle from Containment Centerline at feet above Y Position basemat See drawing 11405-S- X Position from 02 from centerline X y 120.042 60.667 Degrees radians centerline 93' radius 55.00 0.00 0 0.000 93 .50 59.38 55.00 120.04 1 0.017 93.49 61.00 2 0.035 93.44 62.62 3 0.052 93.37 64.24 X y 4 0.070 93.27 65.86

                 -55.00         0.00                                   5                   0.087        93.14          67.48
                 -55.00       120.04                                   6                   0.105        92.99          69.10 7                   0.122        92.80          70.71 Height above Basemat                                      8                   0.140        92.59          72.32 1128'4.5"- 90' - 979'                                     9                   0.157        92.35          73.92 EC 68969 10                   0.175        92.08          75.52 59.375                                               11                   0.192        91.78          77.12 12                   0.209        91.46          78.71 13                   0.227        91.10          80.30 14                   0.244        90.72          81.87 15                   0.262        90.31          83.45 16                   0.279        89.88          85.01 17                   0.297        89.41          86.57 18                   0.314        88.92          88.11 19                   0.332        88.41          89.65 20                   0.349        87.86          91.18 21                   0.367        87.29          92.70 22                   0.384        86.69          94.21 23                   0.401        86.07          95.71 24                   0.419        85.42          97.20 25                   0.436        84.74          98.68 26                   0.454        84.04         100.14 27                   0.471        83.31         101.60 28                   0.489        82.56         103.04 162 29                   0.506        81.78         104.46 30                   0.524        80.97         105.87

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 163 of 269 31 0.541 80.15 107.27 32 0.559 79.29 108.66 33 0.576 78.42 110.03 34 0.593 77.52 111.38 35 0.611 76.59 112.72 36 0.628 75.64 114.04 37 0.646 74.67 115.34 38 0.663 73.68 116.63 39 0.681 72.66 117.90 40 0.698 71.63 119.15 41 0.716 70.57 120.39 42 0.733 69.48 121.60 43 0.750 68.38 122.80 44 0.768 67.26 123.98 45 0.785 66.11 125.14 46 0.803 64.95 126.27 47 0.820 63.77 127.39 EC 68969 48 0.838 62.56 128.49 49 0.855 61.34 129.56 50 0.873 60.10 130.62 51 0.890 58.84 131.65 52 0.908 57.56 132.66 53 0.925 56.27 133.65 54 0.942 54.96 134.61 55 0.960 53 .63 135.56 56 0.977 52 .28 136.48 57 0.995 50.92 137.37 58 1.012 49.55 138.24 59 1.030 48.16 139.09 60 1.047 46.75 139.92 61 1.065 45 .33 140.71 62 1.082 43 .90 141.49 63 1.100 42.45 142.24 64 1.117 40.99 142.96 65 1.134 39.51 143.66 66 1.152 38.03 144.33 163 67 1.169 36.53 144.98 68 1.187 35.03 145.60

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 164 of 269 69 1.204 33.51 146.20 70 1.222 31 .98 146.77 71 1.239 30.44 147.31 72 1.257 28.89 147.82 73 1.274 27.34 148.31 74 1.292 25 .77 148.77 75 1.309 24.20 149.21 76 1.326 22 .62 149.61 77 1.344 21.03 149.99 78 1.361 19.44 150.34 79 1.379 17.84 150.67 80 1.396 16.24 150.96 81 1.414 14.63 151.23 82 1.431 13 .01 151.47 83 1.449 11.39 151.68 84 1.466 9.77 151.87 85 1.484 8.15 152.02 EC 68969 86 1.501 6.52 152.15 87 1.518 4.89 152.25 88 1.536 3.26 152.32 89 1.553 1.63 152.36 90 1.571 0.00 152.37 91 1.588 -1.63 152.36 92 1.606 -3.26 152.32 93 1.623 -4.89 152.25 94 1.641 -6.52 152.15 95 1.658 -8.15 152.02 96 1.676 -9.77 151.87 97 1.693 -11.39 151.68 98 1.710 -13 .01 151.47 99 1.728 -14.63 151.23 100 1.745 -16.24 150.96 101 1.763 -17.84 150.67 102 1.780 -19.44 150.34 103 1.798 -21.03 149.99 104 1.815 -22 .62 149.61 164 105 1.833 -24.20 149.21 106 1.850 -25.77 148.77

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 165 of 269 107 1.868 -27.34 148.31 108 1.885 -28.89 147.82 109 1.902 -30.44 147.31 110 1.920 -31.98 146.77 111 1.937 -33.51 146.20 112 1.955 -35.03 145.60 113 1.972 -36.53 144.98 114 1.990 -38.03 144.33 115 2.007 -39.51 143.66 116 2.025 -40.99 142.96 117 2.042 -42.45 142.24 118 2.059 -43 .90 141.49 119 2.077 -45.33 140.71 120 2.094 -46.75 139.92 121 2.112 -48.16 139.09 122 2.129 -49.55 138.24 123 2.147 -50.92 137.37 EC 68969 124 2.164 -52.28 136.48 125 2.182 -53.63 135.56 126 2.199 -54.96 134.61 127 2.217 -56.27 133.65 128 2.234 -57.56 132.66 129 2.251 -58.84 131.65 130 2.269 -60.10 130.62 131 2.286 -61.34 129.56 132 2.304 -62.56 128.49 133 2.321 -63.77 127.39 134 2.339 -64.95 126.27 135 2.356 -66.11 125.14 136 2.374 -67.26 123.98 137 2.391 -68.38 122.80 138 2.409 -69.48 121.60 139 2.426 -70.57 120.39 140 2.443 -71.63 119.15 141 2.461 -72.66 117.90 142 2.478 -73.68 116.63 165 143 2.496 -74.67 115.34 144 2.513 -75.64 114.04

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 166 of 269 145 2.531 -76.59 112.72 146 2.548 -77.51 111.38 147 2.566 -78.42 110.03 148 2.583 -79.29 108.66 149 2.601 -80.15 107.27 150 2.618 -80.97 105.88 151 2.635 -81.78 104.46 152 2.653 -82.56 103.04 153 2.670 -83.31 101.60 154 2.688 -84.04 100.14 155 2.705 -84.74 98.68 156 2.723 -85.42 97.20 157 2.740 -86.07 95 .71 158 2.758 -86.69 94.21 159 2.775 -87.29 92.70 160 2.793 -87.86 91.18 161 2.810 -88.41 89.65 EC 68969 162 2.827 -88.92 88.11 163 2.845 -89.41 86.57 164 2.862 -89.88 85 .01 165 2.880 -90.31 83.45 166 2.897 -90.72 81.87 167 2.915 -91.10 80.30 168 2.932 -91.46 78.71 169 2.950 -91.78 77.12 170 2.967 -92.08 75 .52 171 2.985 -92.35 73.92 172 3.002 -92.59 72.32 173 3.019 -92.80 70.71 174 3.037 -92.99 69.10 175 3.054 -93 .14 67.48 176 3.072 -93.27 65 .86 177 3.089 -93.37 64.24 178 3.107 -93.44 62 .62 179 3.124 -93.49 61.00 180 3.142 -93.50 59.38 166

FCS Geo Inputs 20161017 .xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 167 of 269 X y

                 -63 .13                58 .90         0.00                     1795.144           0 58.90        120.04                     1795.144   3658.880 61.35        122.50                     1870.073   3733 .809 61.35        138.50                     1870.073   4221.489 46.50        139.92                     1417.320   4264.609 Containment effective height approximation based on equivalent cylinder.

EC 68969 uses shape of containment dome- coarsely integrated X Y 145.53 1124.53

                                      -58.90         0.00
                                      -58.90      120.04                          142.53    1121.53 Effective Height
                                      -61 .35     122.50                         4344.43
                                      -61.35      138.50
                                      -46.50      139.92 167

0> <0 N 0 CX> <0 0 r-- Q) C> ro a.. 0 0.0 0 Lf'l 0 m 0 N 0 c QJ E c ro

 .....                                            0 c

0 u Vl u LL. 0 0

                                                  'i' 0
                                                  'f
                           - ---    -             0 t;-

0 0 0 0 0 0 0 0 0 0.0 N 0 00 0.0

            ..-< ..-< ..-<                "'" N FC08513, R0                   EC 68969             168

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 169 of 269 Alternate calculation for Rack weight (Not USED in MCNP) OPPD Drawing 1000 (Holtec Racks) Rack Al 10x8 16000 12800 2159.11 10640.89 396.71 Rack A2 10x8 16000 12800 2169.02 10630.98 Rack Bl 12 X 9 15200 12160 2182.96 9977.04 Rack B2 12 x9 15200 12160 2181.60 9978.40 Rack Gl 10 X 9 12600 10080 1886.33 8193 .67 Rack G2 10 x9 12600 10080 1885.20 8194.80 Rack C 11 X 9 13900 11120 2035.67 9084.33 Rack D 11 x8 12400 9920 1854.59 8065.41 Rack E lOx 10 14000 11200 2052.06 9147.94 Rack Fl lOx 12 16800 13440 2183.99 11256.01 Rack F2 10 X 12 16800 13440 2380.32 11059.68 48176.49 EC 68969 lsubracting out weight below rack base plate 181.5325 89.231 19.47542 51.75 169

FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs FC08513, R0 Page 170 of 269 Information used as an initial point for EAB distance 900 M, USAR 2.2 and T.S. 4.1 for distance from containment centerline and highway elevation of 1080' 101.25 feet I second Latitude 6.074 75 .83 feet I second Latitude 4.550 Containment 12+90N 20+00W 1305.9 1516.7 0.016667 16.66667 101.23 longitude 75 .83 latitude Bridge End 10+30N 40+00W 1042.7 3033 .3 feet meters em 263 .2 1516.7 1539 469.19 46918.96 Distance 1539.3 feet 0.292 miles EC 68969 Hwy 75 Access 10+30N 52+00W feet meters 1042.7 3943 .3 2440.9 743.99 74398.61 263 .2 2426.7 0.462 Property Corner O+OON 51+69.24V feet meters 0.0 3920.0 2735 .2 833 .70 83369 .6 1305.9 2403 .3 0.518 170

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 171 of 269 Assembly In itial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID A001 1.400 7913 14654 0.37461 2/8/1975 1.400 4/10/1975 A002 1.380 8638 14654 0.37165 2/8/1975 1.380 4/10/1975 A003 1.380 8638 14654 0.37135 2/8/1975 1.380 4/10/1975 A004 1.380 15310 13310 0.37155 10/15/1978 1.380 12/10/1978 AOOS 1.390 8494 14654 0.37595 2/8/1975 1.390 4/10/1975 A006 1.390 8494 14654 0.37684 2/8/1975 1.390 4/10/1975 A007 1.400 8494 14654 0.37643 2/8/1975 1.400 4/10/1975 A008 1.400 8252 14654 0.37611 2/8/1975 1.400 4/10/1975 A009 1.400 8494 14654 0.37581 2/8/1975 1.400 4/10/1975 A010 1.390 7913 14654 0.37517 2/8/1975 1.390 4/10/1975 A011 1.390 7913 14654 0.37468 2/8/1975 1.390 4/10/1975 A012 1.390 15310 13310 0.37314 10/15/1978 1.390 12/10/1978 A013 1.390 15667 13310 0.37173 10/15/1978 1.390 12/10/1978 A014 1.390 9026 14654 0.37247 2/8/1975 1.390 4/10/1975 A015 1.390 15310 13310 0.37186 10/15/1978 1.390 12/10/1978 A016 1.400 7842 14654 0.37169 2/8/1975 1.400 4/10/1975 A017 1.400 17050 14054 0.37166 10/1/1976 1.400 12/8/1976 A018 1.400 7842 14654 0.37192 2/8/1975 1.400 4/10/1975 A019 1.390 15310 13310 0.37195 10/15/1978 1.390 12/10/1978 A020 1.390 9026 14654 0.37055 2/8/1975 1.390 4/10/1975 A021 1.390 15667 13310 0.37105 10/15/1978 1.390 12/10/1978 A022 1.390 7842 14654 0.37279 2/8/1975 1.390 4/10/1975 A023 1.390 8768 14654 0.37138 2/8/1975 1.390 4/10/1975 A024 1.390 8439 14654 0.37217 2/8/1975 1.390 4/10/1975 A025 1.390 15667 13310 0.37123 10/15/1978 1.390 12/10/1978 A026 1.390 9026 14654 0.37017 2/8/1975 1.390 4/10/1975 A027 1.380 15310 13310 0.37110 10/15/1978 1.380 12/10/1978 A028 1.380 15310 13310 0.37082 10/15/1978 1.380 12/10/1978 A029 1.380 15310 13310 0.37027 10/15/1978 1.380 12/10/1978 A030 1.380 15667 13310 0.37163 10/15/1978 1.380 12/10/1978 A031 1.380 9026 14654 0.37231 2/8/1975 1.380 4/10/1975 A032 1.390 15310 13310 0.37090 10/15/1978 1.390 12/10/1978 A033 1.390 15667 13310 0.37097 10/15/1978 1.390 12/10/1978 A034 1.390 8439 14654 0.37143 2/8/1975 1.390 4/10/1975 A035 1.390 15667 13310 0.37121 10/15/1978 1.390 12/10/1978 A036 1.400 8439 14654 0.37080 2/8/1975 1.400 4/10/1975 A037 1.400 17050 14054 0.37036 10/1/1976 1.400 12/8/1976 A038 1.400 17050 14054 0.37034 10/1/1976 1.400 12/8/1976 A039 1.400 7913 14654 0.37132 2/8/1975 1.400 4/10/1975 A040 1.400 17050 14054 0.37207 10/1/1976 1.400 12/8/1976 A041 1.390 8439 14654 0.37225 2/8/1975 1.390 4/10/1975 A042 1.390 7842 14654 0.37298 2/8/1975 1.390 4/10/1975 A043 1.380 8439 14654 0.37688 2/8/1975 1.380 4/10/1975 A044 1.390 15667 13310 0.37051 10/15/1978 1.390 12/10/1978 A045 1.390 15667 13310 0.37144 10/15/1978 1.390 12/10/1978 B001 2.370 27041 13690 0.37214 9/30/1977 2.370 11/16/1977 B002 2.370 27041 13690 0.35354 9/30/1977 2.370 11/16/1977 B003 2.370 27041 13690 0.37214 9/30/1977 2.370 11/16/1977 B004 2.370 27041 13690 0.37214 9/30/1977 2.370 11/16/1977 BOOS 2.370 27041 13690 0.35448 9/30/1977 2.370 11/16/1977 B006 2.390 27041 13690 0.37288 9/30/1977 2.390 11/16/1977 FC08513, R0 EC 68969 171

FCOB514 ext enr.xlsx Attachment D SpreadSheets FCOB513 U-235 Average Enrichment Page 172 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID B007 2.3BO 27041 13690 0.372Bl 9/30/1977 2.3BO 11/16/1977 BOOB 2.3BO 27041 13690 0.37071 9/30/1977 2.3BO 11/16/1977 B009 2.3BO 2B040 13310 0.37057 10/15/197B 2.3BO 12/10/197B BlOl 2.370 22090 14054 0.35420 10/1/1976 2.370 12/B/1976 B102 2.370 2B564 13690 0.35354 9/30/1977 2.370 11/16/1977 B103 2.370 22090 14054 0.354B6 10/1/1976 2.370 12/B/1976 B104 2.370 200BO 14054 0.35570 10/1/1976 2.370 12/B/1976 B105 2.370 2B564 13690 0.3544B 9/30/1977 2.370 11/16/1977 B106 2.410 2B661 13690 0.36146 9/30/1977 2.410 11/16/1977 B107 2.3BO 200BO 14054 0.35431 10/1/1976 2.3BO 12/B/1976 BlOB 2.3BO 2B564 13690 0.35225 9/30/1977 2.3BO 11/16/1977 B109 2.3BO 22090 14054 0.35332 10/1/1976 2.3BO 12/B/1976 B110 2.3BO 22090 14054 0.35364 10/1/1976 2.3BO 12/B/1976 B111 2.370 22030 14054 0.353B7 10/1/1976 2.370 12/B/1976 B112 2.370 2B564 13690 0.353BO 9/30/1977 2.370 11/16/1977 B113 2.370 22030 14054 0.35400 10/1/1976 2.370 12/B/1976 B114 2.370 22090 14054 0.35451 10/1/1976 2.370 12/B/1976 B115 2.3BO 22090 14054 0.35552 10/1/1976 2.3BO 12/B/1976 B116 2.3BO 22030 14054 0.35395 10/1/1976 2.3BO 12/B/1976 B117 2.3BO 2B564 13690 0.35345 9/30/1977 2.3BO 11/16/1977 B11B 2.3BO 200BO 14054 0.35520 10/1/1976 2.3BO 12/B/1976 B119 2.370 2B564 13690 0.35347 9/30/1977 2.370 11/16/1977 B120 2.370 2B564 13690 0.35393 9/30/1977 2.370 11/16/1977 B121 2.400 200BO 14054 0.35329 10/1/1976 2.400 12/B/1976 B122 2.370 22090 14054 0.35473 10/1/1976 2.370 12/B/1976 B123 2.370 22030 14054 0.3543B 10/1/1976 2.370 12/B/1976 B124 2.370 22090 14054 0.35479 10/1/1976 2.370 12/B/1976 B201 2.410 22190 14054 0 .35333 10/1/1976 2.410 12/B/1976 B202 2.370 29014 13690 0.354B2 9/30/1977 2.370 11/16/1977 B203 2.370 29014 13690 0 .35409 9/30/1977 2.370 11/16/1977 B204 2.370 22B10 14054 0.3542B 10/1/1976 2.370 12/B/1976 B205 2.370 29014 13690 0.35427 9/30/1977 2.370 11/16/1977 B206 2.370 29014 13690 0.35445 9/30/1977 2.370 11/16/1977 B207 2.370 29014 13690 0.35376 9/30/1977 2.370 11/16/1977 B20B 2.3BO 29014 13690 0.353B9 9/30/1977 2.3BO 11/16/1977 B209 2.3BO 22190 14054 0.35501 10/1/1976 2.3BO 12/B/1976 B210 2.3BO 22190 14054 0.35479 10/1/1976 2.3BO 12/B/1976 B211 2.3BO 22190 14054 0.35466 10/1/1976 2.3BO 12/B/1976 B212 2.3BO 22B10 14054 0.35490 10/1/1976 2.3BO 12/B/1976 B213 2.3BO 29014 13690 0.35423 9/30/1977 2.3BO 11/16/1977 B214 2.3BO 22B10 14054 0.35453 10/1/1976 2.3BO 12/B/1976 B215 2.3BO 22B10 14054 0.35425 10/1/1976 2.3BO 12/B/1976 B216 2.390 29014 13690 0.35252 9/30/1977 2.390 11/16/1977 B217 2.3BO 22B10 14054 0.35219 10/1/1976 2.3BO 12/B/1976 B21B 2.3BO 22B10 14054 0.35250 10/1/1976 2.3BO 12/B/1976 B219 2.3BO 22B10 14054 0.3533B 10/1/1976 2.3BO 12/B/1976 B220 2.410 22660 14054 0.3544B 10/1/1976 2.410 12/B/1976 B221 2.410 22660 14054 0.35374 10/1/1976 2.410 12/B/1976 B222 2.410 22660 14054 0.352BO 10/1/1976 2.410 12/B/1976 B223 2.400 22660 14054 0.35360 10/1/1976 2.400 12/B/1976 B224 2.390 22B10 14054 0.353BO 10/1/1976 2.390 12/B/1976 FC08513, R0 EC 68969 172

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 173 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID COOl 3.180 35491 13310 0.36999 10/15/1978 3.180 12/10/1978 C002 3.190 25442 13690 0.37182 9/30/1977 3.190 11/16/1977 C003 3.190 25442 13690 0.37283 9/30/1977 3.190 11/16/1977 C004 3.190 25442 13690 0.37192 9/30/1977 3.190 11/16/1977 coos 3.190 27396 13690 0.37279 9/30/1977 3.190 11/16/1977 C006 3.190 27396 13690 0 .37259 9/30/1977 3.190 11/16/1977 C007 3.190 27396 13690 0 .37338 9/30/1977 3.190 11/16/1977 coos 3.180 27642 13690 0.37232 9/30/1977 3.180 11/16/1977 C009 3.190 27642 13690 0 .37041 9/30/1977 3.190 11/16/1977 COlO 3.180 35491 13310 0.37002 10/15/1978 3.180 12/10/1978 C011 3.190 35491 13310 0 .37138 10/15/1978 3.190 12/10/1978 C012 3.180 25442 13690 0.37271 9/30/1977 3.180 11/16/1977 C013 3.190 35491 13310 0 .37046 10/15/1978 3.190 12/10/1978 C014 3.180 27642 13690 0 .37032 9/30/1977 3.180 11/16/1977 C015 3.180 27642 13690 0 .37119 9/30/1977 3.180 11/16/1977 C016 3.190 27396 13690 0 .37265 9/30/1977 3.190 11/16/1977 ClOl 3.180 29445 13690 0.36389 9/30/1977 3.180 11/16/1977 Cl02 3.180 29445 13690 0 .36372 9/30/1977 3.180 11/16/1977 Cl03 3.190 29445 13690 0.36322 9/30/1977 3.190 11/16/1977 Cl04 3.190 29445 13690 0.36205 9/30/1977 3.190 11/16/1977 ClOS 3.190 29445 13690 0.36194 9/30/1977 3.190 11/16/1977 Cl06 3.190 29445 13690 0.36227 9/30/1977 3.190 11/16/1977 C107 3.190 29445 13690 0 .36274 9/30/1977 3.190 11/16/1977 Cl08 3.180 29245 13690 0.36280 9/30/1977 3.180 11/16/1977 Cl09 3.180 29245 13690 0.36253 9/30/1977 3.180 11/16/1977 C110 3.170 29245 13690 0.36288 9/30/1977 3.170 11/16/1977 Cl11 3.170 29245 13690 0.36408 9/30/1977 3.170 11/16/1977 C112 3.160 29445 13690 0.36355 9/30/1977 3.160 11/16/1977 C113 3.170 29245 13690 0.36323 9/30/1977 3.170 11/16/1977 C114 3.160 29245 13690 0.36443 9/30/1977 3.160 11/16/1977 C115 3.160 29245 13690 0.36479 9/30/1977 3.160 11/16/1977 C116 3.160 29245 13690 0.36284 9/30/1977 3.160 11/16/1977 DOOl 2.960 28394 13310 0.37157 10/15/1978 2.960 12/10/1978 D002 2.960 27388 13310 0.37227 10/15/1978 2.960 12/10/1978 D003 2.960 27618 13310 0.37879 10/15/1978 2.960 12/10/1978 D004 2.950 28394 13310 0.37225 10/15/1978 2.950 12/10/1978 DOOS 2.950 51504 11799 0.37199 12/3/1982 2.950 2/18/1983 D006 2.940 30218 12849 0.37184 1/18/1980 2.940 4/16/1980 D007 2.960 28394 13310 0.37083 10/15/1978 2.960 12/10/1978 DOOS 2.960 28394 13310 0.37171 10/15/1978 2.960 12/10/1978 D009 2.960 30218 12849 0.37125 1/18/1980 2.960 4/16/1980 DOlO 2.960 27608 13310 0.37276 10/15/1978 2.960 12/10/1978 DOll 2.960 27388 13310 0.37310 10/15/1978 2.960 12/10/1978 D012 2.960 27388 13310 0.37307 10/15/1978 2.960 12/10/1978 D013 2.960 27388 13310 0.37310 10/15/1978 2.960 12/10/1978 D014 2.960 27608 13310 0.37184 10/15/1978 2.960 12/10/1978 D015 2.960 27608 13310 0.37191 10/15/1978 2.960 12/10/1978 D016 2.950 28394 13310 0.37215 10/15/1978 2.950 12/10/1978 D017 2.960 28394 13310 0.37159 10/15/1978 2.960 12/10/1978 D018 2.950 28394 13310 0.37211 10/15/1978 2.950 12/10/1978 D019 2.960 27388 13310 0.37283 10/15/1978 2.960 12/10/1978 FC08513, R0 EC 68969 173

FC08514 ext enr.xlsx Attachment 0 SpreadSheets FC08513 U-235 Average Enrichment Page 174 of 269 Assembly Initial U- Burn up Assembly 10 235 Enr MWO/T 10 Assembly 10 Assembly 10 Assembly 10 Assembly 10 0020 2.960 27388 13310 0.37303 10/15/1978 2.960 12/10/1978 0021 2.960 27608 13310 0.37278 10/15/1978 2.960 12/10/1978 0022 2.960 27388 13310 0 .37309 10/15/1978 2.960 12/10/1978 0023 2.960 27608 13310 0 .37276 10/15/1978 2.960 12/10/1978 0024 2.960 27388 13310 0.37288 10/15/1978 2.960 12/10/1978 0025 2.960 27608 13310 0 .37266 10/15/1978 2.960 12/10/1978 0026 2.960 27608 13310 0 .37265 10/15/1978 2.960 12/10/1978 0027 2.960 33748 12849 0 .37272 1/18/1980 2.960 4/16/1980 0028 2.970 33346 12849 0 .37285 1/18/1980 2.970 4/16/1980 0029 2.970 33748 12849 0 .37181 1/18/1980 2.970 4/16/1980 0030 2.970 33748 12849 0.37216 1/18/1980 2.970 4/16/1980 0031 2.960 33748 12849 0.37277 1/18/1980 2.960 4/16/1980 0032 2.970 33748 12849 0.37205 1/18/1980 2.970 4/16/1980 0033 2.970 33346 12849 0.37241 1/18/1980 2.970 4/16/1980 0034 2.970 33748 12849 0.37214 1/18/1980 2.970 4/16/1980 0035 2.970 33346 12849 0.37225 1/18/1980 2.970 4/16/1980 0036 2.970 33346 12849 0 .37273 1/18/1980 2.970 4/16/1980 0037 2.970 33748 12849 0.37191 1/18/1980 2.970 4/16/1980 0038 2.970 33748 12849 0.37177 1/18/1980 2.970 4/16/1980 0039 2.970 33346 12849 0.37277 1/18/1980 2.970 4/16/1980 0040 2.970 33346 12849 0.37223 1/18/1980 2.970 4/16/1980 0041 2.970 30218 12849 0.37238 1/18/1980 2.970 4/16/1980 0042 2.970 33346 12849 0.37217 1/18/1980 2.970 4/16/1980 0043 2.990 30218 12849 0.37321 1/18/1980 2.990 4/16/1980 0044 2.970 33346 12849 0.37251 1/18/1980 2.970 4/16/1980 E001 3.030 32830 12241 0.36474 9/18/1981 3.030 12/1/1981 E002 3.030 29083 12241 0.36477 9/18/1981 3.030 12/1/1981 E003 3.030 36880 12241 0.36434 9/18/1981 3.030 12/1/1981 E004 3.050 30315 12241 0.37324 9/18/1981 3.050 12/1/1981 E005 3.030 32830 12241 0.36474 9/18/1981 3.030 12/1/1981 E006 3.030 32830 12241 0.36524 9/18/1981 3.030 12/1/1981 E007 3.030 32830 12241 0.36507 9/18/1981 3.030 12/1/1981 E008 3.030 32830 12241 0.36472 9/18/1981 3.030 12/1/1981 E009 3.030 30324 12241 0.36464 9/18/1981 3.030 12/1/1981 E010 3.030 32830 12241 0.36478 9/18/1981 3.030 12/1/1981 E011 3.030 32830 12241 0.36504 9/18/1981 3.030 12/1/1981 E012 3.030 26442 12849 0.36395 1/18/1980 3.030 4/16/1980 E013 3.040 39667 11799 0.36384 12/3/1982 3.040 2/18/1983 E014 3.040 36880 12241 0.36313 9/18/1981 3.040 12/1/1981 E015 3.040 36880 12241 0.36342 9/18/1981 3.040 12/1/1981 E016 3.040 26442 12849 0.36378 1/18/1980 3.040 4/16/1980 E017 3.040 36880 12241 0.36364 9/18/1981 3.040 12/1/1981 E018 3.030 39667 11799 0.36448 12/3/1982 3.030 2/18/1983 E019 3.030 30324 12241 0.36449 9/18/1981 3.030 12/1/1981 E020 3.030 26442 12849 0.36425 1/18/1980 3.030 4/16/1980 E021 3.030 39667 11799 0.36433 12/3/1982 3.030 2/18/1983 E022 3.030 32830 12241 0.36472 9/18/1981 3.030 12/1/1981 E023 3.040 26442 12849 0.36284 1/18/1980 3.040 4/16/1980 E024 3.030 29083 12241 0.36467 9/18/1981 3.030 12/1/1981 E101 2.720 27677 12849 0.36451 1/18/1980 2.720 4/16/1980 E102 2.720 27677 12849 0.36451 1/18/1980 2.720 4/16/1980 FC08513, R0 EC 68969 174

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 175 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID El03 2.720 27677 12849 0.36498 1/18/1980 2.720 4/16/1980 El04 2.720 30618 12241 0.36457 9/18/1981 2.720 12/1/1981 ElOS 2.720 27932 12849 0.36356 1/18/1980 2.720 4/16/1980 El06 2.720 27932 12849 0.36334 1/18/1980 2.720 4/16/1980 El07 2.720 27932 12849 0.36316 1/18/1980 2.720 4/16/1980 El08 2.720 27932 12849 0.36419 1/18/1980 2.720 4/16/1980 El09 2.720 30618 12241 0.36431 9/18/1981 2.720 12/1/1981 E110 2.720 27932 12849 0.36416 1/18/1980 2.720 4/16/1980 E111 2.720 27677 12849 0.36438 1/18/1980 2.720 4/16/1980 E112 2.720 30618 12241 0.36393 9/18/1981 2.720 12/1/1981 El13 2.720 27932 12849 0.36358 1/18/1980 2.720 4/16/1980 E114 2.720 27677 12849 0.36472 1/18/1980 2.720 4/16/1980 EllS 2.720 30618 12241 0.36417 9/18/1981 2.720 12/1/1981 E116 2.720 27677 12849 0.36451 1/18/1980 2.720 4/16/1980 Ell7 2.720 27677 12849 0.36439 1/18/1980 2.720 4/16/1980 E118 2.720 27677 12849 0 .36478 1/18/1980 2.720 4/16/1980 E119 2.720 27932 12849 0 .36393 1/18/1980 2.720 4/16/1980 El20 2.720 27932 12849 0.36312 1/18/1980 2.720 4/16/1980 FOOl 3.020 39667 11799 0.36450 12/3/1982 3.020 2/18/1983 F002 3.010 32398 12241 0 .36468 9/18/1981 3.010 12/1/1981 F003 3.010 32398 12241 0.36471 9/18/1981 3.010 12/1/1981 F004 3.010 32398 12241 0.36427 9/18/1981 3.010 12/1/1981 FOOS 3.020 39667 11799 0.36523 12/3/1982 3.020 2/18/1983 F006 3.030 39667 11799 0.36497 12/3/1982 3.030 2/18/1983 F007 3.030 30324 12241 0.36577 9/18/1981 3.030 12/1/1981 F008 3.030 29083 12241 0.36660 9/18/1981 3.030 12/1/1981 F009 3.020 29083 12241 0.36632 9/18/1981 3.020 12/1/1981 FOlO 3.020 39667 11799 0.36437 12/3/1982 3.020 2/18/1983 FOll 3.020 39667 11799 0.36442 12/3/1982 3.020 2/18/1983 F012 3.020 32398 12241 0.36418 9/18/1981 3.020 12/1/1981 FlOl 2.740 30618 12241 0.36441 9/18/1981 2.740 12/1/1981 Fl02 2.740 30618 12241 0.36443 9/18/1981 2.740 12/1/1981 Fl03 2.740 29637 12241 0.36444 9/18/1981 2.740 12/1/1981 Fl04 2.740 29637 12241 0.36506 9/18/1981 2.740 12/1/1981 FlOS 2.740 29637 12241 0.36495 9/18/1981 2.740 12/1/1981 Fl06 2.740 29637 12241 0.36568 9/18/1981 2.740 12/1/1981 Fl07 2.740 29637 12241 0.36522 9/18/1981 2.740 12/1/1981 Fl08 2.740 29637 12241 0.36493 9/18/1981 2.740 12/1/1981 Fl09 2.740 29637 12241 0.36538 9/18/1981 2.740 12/1/1981 F110 2.740 29637 12241 0 .36531 9/18/1981 2.740 12/1/1981 Fl11 2.740 30618 12241 0.36441 9/18/1981 2.740 12/1/1981 F112 2.740 30618 12241 0 .36392 9/18/1981 2.740 12/1/1981 GOOl 3.030 38488 10244 0.36349 3/7/1987 3.030 3/24/1987 G002 3.030 38294 11344 0.36334 3/2/1984 3.030 6/18/1984 G003 3.030 38294 11344 0.36325 3/2/1984 3.030 6/18/1984 G004 3.020 33253 11799 0.36317 12/3/1982 3.020 2/18/1983 GOOS 3.020 31973 11799 0.36341 12/3/1982 3.020 2/18/1983 G006 3.020 42169 11344 0.36305 3/2/1984 3.020 6/18/1984 G007 3.020 39051 11344 0.36392 3/2/1984 3.020 6/18/1984 G008 3.020 38577 11344 0.36429 3/2/1984 3.020 6/18/1984 G009 3.030 38577 11344 0.36332 3/2/1984 3.030 6/18/1984 FC08513, R0 EC 68969 175

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 176 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID GOlD 3.020 42169 11344 0.36285 3/2/1984 3.020 6/18/1984 GOll 3.030 39985 11344 0.36277 3/2/1984 3.030 6/18/1984 G012 3.020 33253 11799 0.36285 12/3/1982 3.020 2/18/1983 G013 3.030 41606 10244 0.36282 3/7/1987 3.030 3/24/1987 G014 3.030 39051 11344 0.36293 3/2/1984 3.030 6/18/1984 G015 3.020 31973 11799 0.36267 12/3/1982 3.020 2/18/1983 G016 3.020 31071 11799 0.36346 12/3/1982 3.020 2/18/1983 G017 3.020 42169 11344 0.36283 3/2/1984 3.020 6/18/1984 G018 3.020 42169 11344 0.36286 3/2/1984 3.020 6/18/1984 G019 3.020 31973 11799 0.36202 12/3/1982 3.020 2/18/1983 G020 3.020 41606 10244 0.36346 3/7/1987 3.020 3/24/1987 G021 3.020 33253 11799 0.36311 12/3/1982 3.020 2/18/1983 G022 3.030 38294 11344 0.36352 3/2/1984 3.030 6/18/1984 G023 3.020 41606 10244 0.36335 3/7/1987 3.020 3/24/1987 G024 3.020 31973 11799 0.36319 12/3/1982 3.020 2/18/1983 G025 3.020 39051 11344 0.36337 3/2/1984 3.020 6/18/1984 G026 3.020 31071 11799 0.36397 12/3/1982 3.020 2/18/1983 G027 3.020 33253 11799 0.36347 12/3/1982 3.020 2/18/1983 G028 3.020 31071 11799 0.36382 12/3/1982 3.020 2/18/1983 G029 3.020 39051 11344 0.36371 3/2/1984 3.020 6/18/1984 G030 3.020 41606 10244 0.36351 3/7/1987 3.020 3/24/1987 G031 3.010 38488 10244 0.36540 3/7/1987 3.010 3/24/1987 G032 3.010 38294 11344 0.36494 3/2/1984 3.010 6/18/1984 G033 3.010 38488 10244 0.36507 3/7/1987 3.010 3/24/1987 G034 3.010 38577 11344 0.36488 3/2/1984 3.010 6/18/1984 G035 3.020 38577 11344 0.36468 3/2/1984 3.020 6/18/1984 G036 3.020 36183 10244 0.36470 3/7/1987 3.020 3/24/1987 G037 3.010 39815 10769 0.36444 9/28/1985 3.010 12/8/1985 G038 3.010 39343 10769 0.36479 9/28/1985 3.010 12/8/1985 G039 3.030 36183 10244 0.36416 3/7/1987 3.030 3/24/1987 G040 3.010 39343 10769 0.36510 9/28/1985 3.010 12/8/1985 G041 3.010 39343 10769 0.36414 9/28/1985 3.010 12/8/1985 G042 3.030 36183 10244 0.36435 3/7/1987 3.030 3/24/1987 G043 3.030 36183 10244 0.36446 3/7/1987 3.030 3/24/1987 G044 3.030 38488 10244 0.36403 3/7/1987 3.030 3/24/1987 HAOl 3.480 23264 11344 0.35633 3/2/1984 3.480 6/18/1984 HA02 3.480 23264 11344 0.35704 3/2/1984 3.480 6/18/1984 HA03 3.480 23264 11344 0.35743 3/2/1984 3.480 6/18/1984 HA04 3.480 23264 11344 0.35675 3/2/1984 3.480 6/18/1984 HA05 3.480 33270 11344 0.35726 3/2/1984 3.480 6/18/1984 HA06 3.480 33270 11344 0.35674 3/2/1984 3.480 6/18/1984 HA07 3.480 33270 11344 0.35742 3/2/1984 3.480 6/18/1984 HA08 3.480 39859 10769 0.35810 9/28/1985 3.480 12/8/1985 HA09 3.480 39859 10769 0.35749 9/28/1985 3.480 12/8/1985 HAlO 3.480 36953 9674 0.35751 9/27/1988 3.480 10/30/1988 HAll 3.480 39859 10769 0.35679 9/28/1985 3.480 12/8/1985 HA12 3.480 33270 11344 0.35679 3/2/1984 3.480 6/18/1984 HA13 3.480 36266 10244 0.35836 3/7/1987 3.480 3/24/1987 HA14 3.480 25270 11344 0.35769 3/2/1984 3.480 6/18/1984 HA15 3.480 39859 10769 0.35757 9/28/1985 3.480 12/8/1985 HA16 3.480 37338 10769 0.35759 9/28/1985 3.480 12/8/1985 FC08513, R0 EC 68969 176

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 177 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/f ID Assembly 10 Assembly ID Assembly ID Assembly ID HA17 3.480 37219 10769 0.35684 9/28/1985 3.480 12/8/1985 HA18 3.480 37219 10769 0.35700 9/28/1985 3.480 12/8/1985 HA19 3.430 37229 10769 0.36486 9/28/1985 3.430 12/8/1985 HA20 3.480 39859 10769 0.35635 9/28/1985 3.480 12/8/1985 HA21 3.480 33061 10244 0.35758 3/7/1987 3.480 3/24/1987 HA22 3.490 39487 10769 0.36640 9/28/1985 3.490 12/8/1985 HA23 3.480 39859 10769 0.35767 9/28/1985 3.480 12/8/1985 HA24 3.480 39859 10769 0.35702 9/28/1985 3.480 12/8/1985 HA25 3.480 37219 10769 0.35724 9/28/1985 3.480 12/8/1985 HA26 3.480 37219 10769 0.35759 9/28/1985 3.480 12/8/1985 HA27 3.480 37219 10769 0.35792 9/28/1985 3.480 12/8/1985 HA28 3.480 33061 10244 0.35848 3/7/1987 3.480 3/24/1987 HA29 3.480 33061 10244 0.35816 3/7/1987 3.480 3/24/1987 HA30 3.480 37219 10769 0.35657 9/28/1985 3.480 12/8/1985 HA31 3.480 33061 10244 0.35823 3/7/1987 3.480 3/24/1987 HA32 3.480 33387 10244 0.35830 3/7/1987 3.480 3/24/1987 HA33 3.480 33387 10244 0.35798 3/7/1987 3.480 3/24/1987 HA34 3.480 33387 10244 0 .35797 3/7/1987 3.480 3/24/1987 HA35 3.480 38933 10769 0.35755 9/28/1985 3.480 12/8/1985 HA36 3.480 38933 10769 0 .35755 9/28/1985 3.480 12/8/1985 HA37 3.480 37219 10769 0.35698 9/28/1985 3.480 12/8/1985 HA38 3.480 38933 10769 0.35824 9/28/1985 3.480 12/8/1985 HA39 3.480 33387 10244 0.35729 3/7/1987 3.480 3/24/1987 HA40 3.480 38933 10769 0.35728 9/28/1985 3.480 12/8/1985 IA01 3.510 27167 10769 0.35686 9/28/1985 3.510 12/8/1985 IA02 3.510 27167 10769 0.35731 9/28/1985 3.510 12/8/1985 IA03 3.510 38557 9674 0.35735 9/27/1988 3.510 10/30/1988 IA04 3.510 27167 10769 0.35747 9/28/1985 3.510 12/8/1985 lAOS 3.510 27167 10769 0 .35769 9/28/1985 3.510 12/8/1985 IA06 3.510 27167 10769 0.35735 9/28/1985 3.510 12/8/1985 IA07 3.510 27167 10769 0.35776 9/28/1985 3.510 12/8/1985 IA08 3.510 27167 10769 0.35767 9/28/1985 3.510 12/8/1985 IA09 3.510 34980 10769 0 .35758 9/28/1985 3.510 12/8/1985 IA10 3.470 34917 10769 0.36555 9/28/1985 3.470 12/8/1985 IAll 3.510 34980 10769 0.35782 9/28/1985 3.510 12/8/1985 IA12 3.510 34980 10769 0.35689 9/28/1985 3.510 12/8/1985 IA13 3.510 34980 10769 0.35754 9/28/1985 3.510 12/8/1985 IA14 3.470 34897 10769 0.36565 9/28/1985 3.470 12/8/1985 IA15 3.510 34980 10769 0 .35736 9/28/1985 3.510 12/8/1985 IA16 3.510 34980 10769 0.35695 9/28/1985 3.510 12/8/1985 IA17 3.510 32189 10769 0.35748 9/28/1985 3.510 12/8/1985 IA18 3.510 32189 10769 0.35721 9/28/1985 3.510 12/8/1985 IA19 3.510 32189 10769 0.35745 9/28/1985 3.510 12/8/1985 IA20 3.510 32189 10769 0.35698 9/28/1985 3.510 12/8/1985 IA21 3.510 23835 10769 0.35773 9/28/1985 3.510 12/8/1985 IA22 3.510 23835 10769 0.35730 9/28/1985 3.510 12/8/1985 IA23 3.510 23835 10769 0.35730 9/28/1985 3.510 12/8/1985 IA24 3.510 23835 10769 0.35676 9/28/1985 3.510 12/8/1985 IA25 3.510 36088 10769 0.35702 9/28/1985 3.510 12/8/1985 IA26 3.510 36088 10769 0.35725 9/28/1985 3.510 12/8/1985 IA27 3.510 36088 10769 0.35695 9/28/1985 3.510 12/8/1985 FC08513, R0 EC 68969 177

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 178 of 269 Assembly Initial U- Burn up Assembly 10 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID IA28 3.510 36088 10769 0.35736 9/28/1985 3.510 12/8/1985 IA29 3.510 36124 10769 0.35722 9/28/1985 3.510 12/8/1985 IA30 3.510 36124 10769 0.35786 9/28/1985 3.510 12/8/1985 IA31 3.510 36124 10769 0.35738 9/28/1985 3.510 12/8/1985 IA32 3.510 36124 10769 0.35710 9/28/1985 3.510 12/8/1985 IA33 3.510 38091 10769 0.35711 9/28/1985 3.510 12/8/1985 IA34 3.510 38091 10769 0.35786 9/28/1985 3.510 12/8/1985 IA35 3.510 38091 10769 0.35695 9/28/1985 3.510 12/8/1985 IA36 3.510 38091 10769 0 .35721 9/28/1985 3.510 12/8/1985 IA37 3.510 38091 10769 0.35740 9/28/1985 3.510 12/8/1985 IA38 3.510 38091 10769 0 .35677 9/28/1985 3.510 12/8/1985 IA39 3.510 38091 10769 0 .35764 9/28/1985 3.510 12/8/1985 IA40 3.510 38091 10769 0 .35777 9/28/1985 3.510 12/8/1985 JA01 3.510 34439 10244 0 .35533 3/7/1987 3.510 3/24/1987 JA02 3.510 34439 10244 0.35477 3/7/1987 3.510 3/24/1987 JA03 3.510 39620 9674 0 .35616 9/27/1988 3.510 10/30/1988 JA04 3.510 39620 9674 0.35581 9/27/1988 3.510 10/30/1988 JA05 3.510 39570 9674 0 .35538 9/27/1988 3.510 10/30/1988 JA06 3.510 39570 9674 0.35492 9/27/1988 3.510 10/30/1988 JA07 3.510 39570 9674 0.35524 9/27/1988 3.510 10/30/1988 JA08 3.510 39570 9674 0.35492 9/27/1988 3.510 10/30/1988 JA09 3.510 38792 10244 0.35513 3/7/1987 3.510 3/24/1987 JA10 3.510 38792 10244 0.35596 3/7/1987 3.510 3/24/1987 JAll 3.510 38972 9166 0.35592 2/17/1990 3.510 3/16/1990 JA12 3.510 38792 10244 0.35559 3/7/1987 3.510 3/24/1987 JA13 3.510 39669 10244 0.35565 3/7/1987 3.510 3/24/1987 JA14 3.510 39669 10244 0.35565 3/7/1987 3.510 3/24/1987 JA15 3.510 39669 10244 0.35570 3/7/1987 3.510 3/24/1987 JA16 3.510 39669 10244 0.35535 3/7/1987 3.510 3/24/1987 JA17 3.510 38879 10244 0.35586 3/7/1987 3.510 3/24/1987 JA18 3.510 38879 10244 0.35596 3/7/1987 3.510 3/24/1987 JA19 3.510 38879 10244 0.35599 3/7/1987 3.510 3/24/1987 JA20 3.510 38879 10244 0.35548 3/7/1987 3.510 3/24/1987 JA21 3.510 35805 10244 0.35564 3/7/1987 3.510 3/24/1987 JA22 3.510 35805 10244 0.35464 3/7/1987 3.510 3/24/1987 JA23 3.510 35805 10244 0.35514 3/7/1987 3.510 3/24/1987 JA24 3.510 35805 10244 0.35585 3/7/1987 3.510 3/24/1987 JA25 3.510 36464 10244 0.35575 3/7/1987 3.510 3/24/1987 JA26 3.510 36464 10244 0.35544 3/7/1987 3.510 3/24/1987 JA27 3.510 36464 10244 0.35540 3/7/1987 3.510 3/24/1987 JA28 3.510 36464 10244 0.35555 3/7/1987 3.510 3/24/1987 JA29 3.510 39834 9166 0.35582 2/17/1990 3.510 3/12/1990 JA30 3.510 38738 9166 0.35515 2/17/1990 3.510 3/12/1990 JA31 3.510 38792 10244 0.35592 3/7/1987 3.510 3/24/1987 JA32 3.510 39278 9166 0.35566 2/17/1990 3.510 3/16/1990 JA33 3.510 39106 9166 0.35513 2/17/1990 3.510 3/16/1990 JA34 3.510 39611 9166 0.35597 2/17/1990 3.510 3/12/1990 JA35 3.510 38788 9166 0.35509 2/17/1990 3.510 3/12/1990 JA36 3.510 38659 9166 0.35537 2/17/1990 3.510 3/16/1990 KA01 3.520 43910 9166 0.35459 2/17/1990 3.520 3/14/1990 KA02 3.490 42296 9674 0.35594 9/27/1988 3.490 10/30/1988 FC08513, R0 EC 68969 178

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 179 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID KA03 3.490 42296 9674 0.35578 9/27/1988 3.490 10/30/1988 KA04 3.510 42747 9166 0.35450 2/17/1990 3.510 3/14/1990 KAOS 3.520 44202 9166 0.35466 2/17/1990 3.520 3/16/1990 KA06 3.500 44304 9166 0.35644 2/17/1990 3.500 3/14/1990 KA07 3.440 43206 9166 0.36481 2/17/1990 3.440 3/16/1990 KA08 3.490 42296 9674 0.35558 9/27/1988 3.490 10/30/1988 KA09 3.490 42296 9674 0.35541 9/27/1988 3.490 10/30/1988 KA10 3.490 41976 9166 0.35615 2/17/1990 3.490 3/16/1990 KAll 3.480 42174 9166 0.36314 2/17/1990 3.480 3/16/1990 KA12 3.500 42892 9166 0.35616 2/17/1990 3.500 3/14/1990 KA13 3.300 39620 9674 0.35964 9/27/1988 3.300 10/30/1988 KA14 3.500 45692 9674 0.33982 9/27/1988 3.500 10/30/1988 KA15 3.500 44281 9674 0 .33974 9/27/1988 3.500 10/30/1988 KA16 3.500 45048 9674 0.33999 9/27/1988 3.500 10/30/1988 KA17 3.500 38557 9674 0 .33928 9/27/1988 3.500 10/30/1988 KA18 3.490 25716 10244 0 .33977 3/7/1987 3.490 3/24/1987 KA19 3.450 44254 9674 0.34879 9/27/1988 3.450 10/30/1988 KA20 3.490 33684 10244 0.33940 3/7/1987 3.490 3/24/1987 KA21 3.510 45048 9674 0.33992 9/27/1988 3.510 10/30/1988 KA22 3.490 39620 9674 0.33939 9/27/1988 3.490 10/30/1988 KA23 3.500 44281 9674 0.33988 9/27/1988 3.500 10/30/1988 KA24 3.490 33684 10244 0.33970 3/7/1987 3.490 3/24/1987 KA25 3.510 45564 9674 0 .34169 9/27/1988 3.510 10/30/1988 KA26 3.420 45144 9674 0 .35795 9/27/1988 3.420 10/30/1988 KA27 3.500 45692 9674 0 .34052 9/27/1988 3.500 10/30/1988 KA28 3.520 44281 9674 0.33741 9/27/1988 3.520 10/30/1988 KA29 3.510 38557 9674 0 .33856 9/27/1988 3.510 10/30/1988 KA30 3.500 44281 9674 0.33972 9/27/1988 3.500 10/30/1988 KA31 3.500 38557 9674 0 .33915 9/27/1988 3.500 10/30/1988 KA32 3.510 45048 9674 0 .33988 9/27/1988 3.510 10/30/1988 LA01 3.800 38724 9166 0 .35881 2/17/1990 3.800 3/16/1990 LA02 3.810 38541 9166 0 .35804 2/17/1990 3.810 3/12/1990 LA03 3.800 38555 9166 0.35865 2/17/1990 3.800 3/14/1990 LA04 3.810 38832 9166 0 .35817 2/17/1990 3.810 3/16/1990 LAOS 3.800 37296 8452 0.35859 2/1/1992 3.800 2/20/1992 LA06 3.810 37868 8452 0.35758 2/1/1992 3.810 2/22/1992 LA07 3.810 37075 8452 0 .35814 2/1/1992 3.810 2/22/1992 LAOS 3.810 37411 8452 0.35776 2/1/1992 3.810 2/20/1992 LA09 3.810 38089 8452 0 .35793 2/1/1992 3.810 2/22/1992 LA10 3.810 37773 8452 0.35758 2/1/1992 3.810 2/20/1992 LAll 3.810 38033 8452 0.35763 2/1/1992 3.810 2/20/1992 LA12 3.810 38089 8452 0.35745 2/1/1992 3.810 2/22/1992 LA13 3.810 37793 9166 0.35736 2/17/1990 3.810 3/12/1990 LA14 3.810 37633 9166 0.35758 2/17/1990 3.810 3/12/1990 LA15 3.810 38000 9166 0.35686 2/17/1990 3.810 3/12/1990 LA16 3.810 38255 9166 0.35753 2/17/1990 3.810 3/12/1990 LA17 3.810 29735 9674 0.35779 9/27/1988 3.810 10/30/1988 LA18 3.810 29735 9674 0.35775 9/27/1988 3.810 10/30/1988 LA19 3.800 29735 9674 0.35696 9/27/1988 3.800 10/30/1988 LA20 3.810 38770 9166 0.35652 2/17/1990 3.810 3/12/1990 LA21 3.800 39159 9166 0.35614 2/17/1990 3.800 3/12/1990 FC08513, R0 EC 68969 179

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 180 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly 10 Assembly ID Assembly ID Assembly ID LA22 3.810 39441 9166 0.35662 2/17/1990 3.810 3/16/1990 LA23 3.810 39504 9166 0.35664 2/17/1990 3.810 3/14/1990 LA24 3.800 39199 9166 0.35629 2/17/1990 3.800 3/16/1990 LA25 3.800 43394 9166 0.34066 2/17/1990 3.800 3/16/1990 LA26 3.800 43235 9166 0.33994 2/17/1990 3.800 3/12/1990 LA27 3.800 33078 9674 0.33987 9/27/1988 3.800 10/30/1988 LA28 3.800 33078 9674 0.34053 9/27/1988 3.800 10/30/1988 LA29 3.800 43276 9166 0.34115 2/17/1990 3.800 3/12/1990 LA30 3.800 43016 9166 0.33987 2/17/1990 3.800 3/12/1990 LA31 3.800 43655 9166 0.34166 2/17/1990 3.800 3/16/1990 LA32 3.800 43739 9166 0 .34081 2/17/1990 3.800 3/14/1990 LA33 3.800 44194 9166 0.34070 2/17/1990 3.800 3/16/1990 LA34 3.800 43914 9166 0.34174 2/17/1990 3.800 3/12/1990 LA35 3.800 33078 9674 0.34118 9/27/1988 3.800 10/30/1988 LA36 3.800 33078 9674 0 .34125 9/27/1988 3.800 10/30/1988 LA37 3.800 32681 9674 0.34055 9/27/1988 3.800 10/30/1988 LA38 3.800 32681 9674 0 .34100 9/27/1988 3.800 10/30/1988 LA39 3.800 32681 9674 0.34051 9/27/1988 3.800 10/30/1988 LA40 3.800 32681 9674 0.34080 9/27/1988 3.800 10/30/1988 LA41 3.800 33512 9674 0 .34045 9/27/1988 3.800 10/30/1988 LA42 3.800 33512 9674 0 .34070 9/27/1988 3.800 10/30/1988 LA43 3.800 33512 9674 0 .34073 9/27/1988 3.800 10/30/1988 LA44 3.800 33512 9674 0.34026 9/27/1988 3.800 10/30/1988 MOOl 3.810 38167 8452 0.36404 2/1/1992 3.810 2/20/1992 M002 3.810 44172 8452 0.36472 2/1/1992 3.810 2/20/1992 M003 3.810 44041 8452 0.36453 2/1/1992 3.810 2/20/1992 M004 3.810 44063 8452 0.36448 2/1/1992 3.810 2/20/1992 MOOS 3.810 37978 8452 0.36411 2/1/1992 3.810 2/20/1992 M006 3.810 37860 8452 0.36443 2/1/1992 3.810 2/20/1992 M007 3.810 37671 8452 0.36426 2/1/1992 3.810 2/20/1992 MOOS 3.810 43960 8452 0.36467 2/1/1992 3.810 2/20/1992 M009 3.810 37397 8452 0 .36413 2/1/1992 3.810 2/20/1992 MOlO 3.800 38176 8452 0 .36413 2/1/1992 3.800 2/20/1992 MOll 3.780 42819 8452 0 .36520 2/1/1992 3.780 2/20/1992 M012 3.770 42993 8452 0 .36520 2/1/1992 3.770 2/20/1992 M013 3.800 38264 8452 0 .36449 2/1/1992 3.800 2/20/1992 M014 3.770 42710 8452 0 .36461 2/1/1992 3.770 2/20/1992 MOlS 3.780 42816 8452 0.36418 2/1/1992 3.780 2/20/1992 M016 3.780 42851 8452 0 .36407 2/1/1992 3.780 2/20/1992 M017 3.770 42671 8452 0.36413 2/1/1992 3.770 2/22/1992 M018 3.770 42737 8452 0 .36523 2/1/1992 3.770 2/20/1992 M019 3.790 38530 8452 0.36512 2/1/1992 3.790 2/20/1992 M020 3.790 42892 8452 0.36458 2/1/1992 3.790 2/22/1992 MlOl 3.800 40557 5654 0.34872 10/1/1999 3.800 10/12/1999 Ml 02 3.800 40577 5654 0.34905 10/1/1999 3.800 10/12/1999 Ml03 3.810 39214 5654 0.34823 10/1/1999 3.810 10/13/1999 Ml04 3.800 39249 5654 0.34857 10/1/1999 3.800 10/12/1999 MlOS 3.810 40593 5654 0.34831 10/1/1999 3.810 10/13/1999 Ml06 3.800 39140 5654 0.34889 10/1/1999 3.800 10/12/1999 Ml07 3.800 39613 5122 0.34825 3/15/2001 3.800 3/31/2001 Ml08 3.800 45005 8452 0.34749 2/1/1992 3.800 2/20/1992 FC08513, R0 EC 68969 180

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 181 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID M109 3.800 39535 5122 0.34842 3/15/2001 3.800 3/28/2001 MllO 3.800 38905 5654 0.34847 10/1/1999 3.800 10/12/1999 Mlll 3.790 45650 7851 0.34891 9/25/1993 3.790 10/8/1993 M112 3.790 48373 5654 0 .34880 10/1/1999 3.790 10/11/1999 Mll3 3.810 45904 8452 0.34907 2/1/1992 3.810 2/20/1992 M114 3.790 39155 5654 0.34937 10/1/1999 3.790 10/11/1999 M115 3.790 39628 5122 0.34906 3/15/2001 3.790 3/29/2001 M116 3.800 38925 5654 0.34815 10/1/1999 3.800 10/13/1999 Mll7 3.790 38914 5654 0 .34926 10/1/1999 3.790 10/12/1999 M118 3.800 39541 5122 0 .34883 3/15/2001 3.800 3/28/2001 M119 3.810 45901 8452 0 .34879 2/1/1992 3.810 2/20/1992 M120 3.800 40552 5654 0 .34895 10/1/1999 3.800 10/13/1999 M121 3.800 45926 8452 0.34885 2/1/1992 3.800 2/20/1992 M122 3.800 46097 8452 0.34913 2/1/1992 3.800 2/20/1992 M123 3.800 38919 5654 0.34832 10/1/1999 3.800 10/13/1999 M124 3.800 46307 5122 0.34664 3/15/2001 3.800 3/28/2001 NOOl 3.700 41449 5654 0.36399 10/1/1999 3.700 10/13/1999 N002 3.690 44707 7851 0.36373 9/25/1993 3.690 10/8/1993 N003 3.690 45098 7851 0.36431 9/25/1993 3.690 10/8/1993 N004 3.700 45302 5654 0.36472 10/1/1999 3.700 10/11/1999 NODS 3.690 41417 5654 0.36463 10/1/1999 3.690 10/11/1999 N006 3.690 45116 7851 0.36449 9/25/1993 3.690 10/8/1993 N007 3.700 41469 5654 0.36456 10/1/1999 3.700 10/12/1999 N008 3.700 45535 5654 0.36506 10/1/1999 3.700 10/11/1999 N009 3.690 44813 7851 0.36419 9/25/1993 3.690 10/8/1993 NOlO 3.690 45018 7851 0.36415 9/25/1993 3.690 10/8/1993 NOll 3.690 45105 7851 0.36456 9/25/1993 3.690 10/8/1993 N012 3.700 45420 5654 0.36494 10/1/1999 3.700 10/13/1999 NOB 3.690 44797 7851 0.36397 9/25/1993 3.690 10/8/1993 N014 3.690 45167 7851 0.36405 9/25/1993 3.690 10/8/1993 N015 3.690 41404 5654 0.36473 10/1/1999 3.690 10/11/1999 N016 3.700 45459 5654 0 .36518 10/1/1999 3.700 10/12/1999 N017 3.700 46040 7851 0.36557 9/25/1993 3.700 10/6/1993 N018 3.700 46097 7851 0 .36522 9/25/1993 3.700 10/8/1993 N019 3.700 45864 7851 0.36561 9/25/1993 3.700 10/8/1993 N020 3.700 45898 7851 0.36577 9/25/1993 3.700 10/8/1993 NlOl 3.710 45737 7851 0.34837 9/25/1993 3.710 10/6/1993 N102 3.700 41158 7337 0.34891 2/20/1995 3.700 3/4/1995 N103 3.710 45727 7851 0.34858 9/25/1993 3.710 10/8/1993 N104 3.710 45728 7851 0.34826 9/25/1993 3.710 10/8/1993 NlOS 3.710 45750 7851 0.34833 9/25/1993 3.710 10/8/1993 N106 3.690 38438 5654 0.34778 10/1/1999 3.690 10/12/1999 N107 3.700 41292 7337 0.34874 2/20/1995 3.700 3/4/1995 N108 3.700 41302 5654 0.34820 10/1/1999 3.700 10/12/1999 N109 3.700 41334 5654 0.34835 10/1/1999 3.700 10/13/1999 N110 3.700 41161 7337 0.34887 2/20/1995 3.700 3/4/1995 N111 3.700 48376 5654 0 .34978 10/1/1999 3.700 10/12/1999 N112 3.700 48425 5654 0.34966 10/1/1999 3.700 10/12/1999 N113 3.700 48430 5654 0.34953 10/1/1999 3.700 10/11/1999 N114 3.700 48402 5654 0.34937 10/1/1999 3.700 10/11/1999 N115 3.700 41341 5654 0.34837 10/1/1999 3.700 10/13/1999 FC08513, R0 EC 68969 181

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 182 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID N116 3.700 38424 5654 0 .34743 10/1/1999 3.700 10/12/1999 N117 3.700 38446 5654 0 .34759 10/1/1999 3.700 10/13/1999 N118 3.710 41170 5654 0.34782 10/1/1999 3.710 10/12/1999 N119 3.710 41318 7337 0.34786 2/20/1995 3.710 3/4/1995 N120 3.700 38435 5654 0.34763 10/1/1999 3.700 10/13/1999 N121 3.690 40267 5654 0.34748 10/1/1999 3.690 10/13/1999 N122 3.690 40240 5654 0.34743 10/1/1999 3.690 10/11/1999 N123 3.690 40306 5654 0.34721 10/1/1999 3.690 10/12/1999 N124 3.690 40231 5654 0.34709 10/1/1999 3.690 10/11/1999 POOl 3.940 45060 7337 0.36528 2/20/1995 3.940 3/4/1995 P002 3.940 44521 7337 0.36548 2/20/1995 3.940 3/4/1995 P003 3.940 44413 7337 0.36551 2/20/1995 3.940 3/4/1995 P004 3.930 44350 7337 0.36585 2/20/1995 3.930 3/4/1995 PODS 3.930 44448 7337 0.36596 2/20/1995 3.930 3/4/1995 P006 3.940 45229 7337 0.36619 2/20/1995 3.940 3/4/1995 P007 3.940 45181 7337 0.36596 2/20/1995 3.940 3/4/1995 P008 3.940 45145 7337 0.36640 2/20/1995 3.940 3/4/1995 PlOl 3.600 37400 7851 0.34975 9/25/1993 3.600 10/8/1993 P102 3.600 37744 7851 0.34981 9/25/1993 3.600 10/8/1993 P103 3.590 41842 7337 0.34980 2/20/1995 3.590 3/4/1995 P104 3.580 41566 7337 0 .34997 2/20/1995 3.580 3/4/1995 PlOS 3.600 37316 7851 0 .34972 9/25/1993 3.600 10/8/1993 P106 3.600 37809 7851 0 .34981 9/25/1993 3.600 10/6/1993 P107 3.600 37854 7851 0 .34999 9/25/1993 3.600 10/8/1993 P108 3.600 37750 7851 0.34996 9/25/1993 3.600 10/8/1993 P109 3.580 41477 7337 0.35030 2/20/1995 3.580 3/4/1995 P110 3.580 42487 5654 0.34989 10/1/1999 3.580 10/12/1999 P111 3.590 42524 5654 0.34975 10/1/1999 3.590 10/13/1999 P112 3.580 43425 7337 0.34962 2/20/1995 3.580 3/4/1995 P113 3.590 41729 7337 0.35006 2/20/1995 3.590 3/4/1995 P114 3.600 37346 7851 0.34993 9/25/1993 3.600 10/8/1993 P115 3.600 37229 7851 0.34975 9/25/1993 3.600 10/6/1993 P116 3.580 47520 5654 0.34977 10/1/1999 3.580 10/12/1999 P117 3.580 47445 5654 0.35002 10/1/1999 3.580 10/11/1999 P118 3.580 41552 7337 0.34984 2/20/1995 3.580 3/4/1995 P119 3.590 41795 7337 0.35009 2/20/1995 3.590 3/4/1995 P120 3.580 44724 5654 0.35020 10/1/1999 3.580 10/12/1999 P121 3.580 41572 7337 0.34990 2/20/1995 3.580 3/4/1995 P122 3.590 41923 7337 0.34975 2/20/1995 3.590 3/4/1995 P123 3.590 41502 7337 0.34975 2/20/1995 3.590 3/4/1995 P124 3.580 47536 5654 0.34973 10/1/1999 3.580 10/12/1999 P125 3.590 41475 7337 0.34962 2/20/1995 3.590 3/4/1995 P126 3.590 41587 7337 0.34944 2/20/1995 3.590 3/4/1995 P127 3.590 44837 5654 0.34962 10/1/1999 3.590 10/13/1999 P128 3.590 41979 7337 0.34986 2/20/1995 3.590 3/4/1995 P129 3.580 43410 7337 0.34996 2/20/1995 3.580 3/4/1995 P130 3.580 43466 7337 0.34985 2/20/1995 3.580 3/4/1995 P131 3.580 43229 7337 0.34964 2/20/1995 3.580 3/4/1995 P132 3.580 47473 5654 0.34974 10/1/1999 3.580 10/11/1999 ROOl 0.740 13047 6745 0.37585 10/5/1996 0.740 10/20/1996 R002 0.750 12952 6745 0.37604 10/5/1996 0.750 10/18/1996 FC08513, R0 EC 68969 182

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 183 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID R003 0.740 13078 6745 0.37594 10/5/1996 0.740 10/19/1996 R004 0.750 12980 6745 0.37581 10/5/1996 0.750 10/19/1996 RODS 3.860 42402 6745 0.37420 10/5/1996 3.860 10/20/1996 R006 3.850 42250 6745 0.37437 10/5/1996 3.850 10/19/1996 R007 3.850 42348 6745 0.37442 10/5/1996 3.850 10/19/1996 R008 3.850 42298 6745 0.37436 10/5/1996 3.850 10/20/1996 R009 3.850 42111 6745 0.37412 10/5/1996 3.850 10/19/1996 ROlO 3.850 41710 6745 0.37337 10/5/1996 3.850 10/18/1996 ROll 3.850 41911 6745 0.37347 10/5/1996 3.850 10/18/1996 R012 3.850 41988 6745 0.37308 10/5/1996 3.850 10/18/1996 R013 3.850 45732 6745 0.37421 10/5/1996 3.850 10/19/1996 R014 3.850 45845 6745 0.37428 10/5/1996 3.850 10/18/1996 ROlS 3.860 45857 6745 0.37408 10/5/1996 3.860 10/20/1996 R016 3.850 45799 6745 0.37399 10/5/1996 3.850 10/19/1996 R017 3.860 41381 6745 0.37380 10/5/1996 3.860 10/20/1996 R018 3.850 38552 6745 0.37493 10/5/1996 3.850 10/19/1996 R019 3.860 41447 6745 0.37449 10/5/1996 3.860 10/19/1996 R020 3.860 38602 6745 0.37458 10/5/1996 3.860 10/20/1996 R021 3.860 51770 6201 0.37467 4/1/1998 3.860 4/19/1998 R022 3.860 51611 6201 0.37391 4/1/1998 3.860 4/19/1998 R023 3.860 51751 6201 0.37444 4/1/1998 3.860 4/18/1998 R024 3.860 51657 6201 0.37348 4/1/1998 3.860 4/18/1998 R025 3.850 45998 6201 0.37476 4/1/1998 3.850 4/18/1998 R026 3.850 45979 6201 0.37452 4/1/1998 3.850 4/18/1998 R027 3.850 45957 6201 0.37483 4/1/1998 3.850 4/19/1998 R028 3.850 46013 6201 0.37470 4/1/1998 3.850 4/19/1998 R029 3.850 42479 6201 0.37516 4/1/1998 3.850 4/20/1998 R030 3.850 42480 6201 0.37470 4/1/1998 3.850 4/20/1998 R031 3.850 42492 6201 0.37453 4/1/1998 3.850 4/19/1998 R032 3.850 42484 6201 0.37471 4/1/1998 3.850 4/19/1998 R033 3.850 38477 6201 0.37454 4/1/1998 3.850 4/19/1998 R034 3.850 38459 6201 0.37423 4/1/1998 3.850 4/17/1998 R035 3.850 38507 6201 0.37438 4/1/1998 3.850 4/19/1998 R036 3.850 38465 6201 0 .37414 4/1/1998 3.850 4/20/1998 R037 3.860 42224 6745 0 .37352 10/5/1996 3.860 10/20/1996 R038 3.860 42039 6745 0.37323 10/5/1996 3.860 10/19/1996 R039 3.860 42216 6745 0.37432 10/5/1996 3.860 10/19/1996 R040 3.850 42215 6745 0.37385 10/5/1996 3.850 10/20/1996 R041 3.850 41768 6745 0 .37391 10/5/1996 3.850 10/20/1996 R042 3.860 41705 6745 0.37440 10/5/1996 3.860 10/17/1996 R043 3.850 41757 6745 0 .37449 10/5/1996 3.850 10/19/1996 R044 3.850 41706 6745 0.37478 10/5/1996 3.850 10/18/1996 R045 3.600 38374 6745 0.37501 10/5/1996 3.600 10/20/1996 R046 3.600 38369 6745 0.37490 10/5/1996 3.600 10/20/1996 R047 3.600 38359 6745 0.37539 10/5/1996 3.600 10/19/1996 R048 3.600 38357 6745 0.37542 10/5/1996 3.600 10/19/1996 R049 3.600 48573 6201 0.37509 4/1/1998 3.600 4/18/1998 ROSO 3.600 48577 6201 0.37465 4/1/1998 3.600 4/19/1998 ROSl 3.600 48643 6201 0.37509 4/1/1998 3.600 4/19/1998 ROS2 3.600 48625 6201 0.37495 4/1/1998 3.600 4/18/1998 SOOl 3.850 33851 6201 0.37324 4/1/1998 3.850 4/20/1998 FC08513, R0 EC 68969 183

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 184 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID S002 3.850 33872 6201 0.37275 4/1/1998 3.850 4/18/1998 S003 3.850 33847 6201 0.37450 4/1/1998 3.850 4/18/1998 S004 3.850 33780 6201 0.37303 4/1/1998 3.850 4/20/1998 soos 3.840 46261 6201 0.37381 4/1/1998 3.840 4/18/1998 S006 3.850 46217 6201 0.37386 4/1/1998 3.850 4/19/1998 S007 3.850 46244 6201 0.37296 4/1/1998 3.850 4/18/1998 S008 3.840 46195 6201 0.37351 4/1/1998 3.840 4/19/1998 S009 3.840 30264 6745 0.37370 10/5/1996 3.840 10/19/1996 SOlO 3.840 30141 6745 0.37412 10/5/1996 3.840 10/20/1996 SOll 3.840 37446 6745 0.37409 10/5/1996 3.840 10/18/1996 S012 3.840 29991 6745 0.37442 10/5/1996 3.840 10/19/1996 S013 3.840 51971 6201 0.37392 4/1/1998 3.840 4/18/1998 S014 3.840 30136 6745 0.37349 10/5/1996 3.840 10/20/1996 SOlS 3.350 36858 6745 0.37391 10/5/1996 3.350 10/20/1996 S016 3.360 36702 6745 0 .37378 10/5/1996 3.360 10/18/1996 S017 3.360 25321 6745 0.37422 10/5/1996 3.360 10/20/1996 S018 3.360 25264 6745 0 .37397 10/5/1996 3.360 10/19/1996 S019 3.360 36777 6745 0 .37390 10/5/1996 3.360 10/19/1996 S020 3.360 38841 6201 0 .37430 4/1/1998 3.360 4/20/1998 S021 3.360 25244 6745 0 .37460 10/5/1996 3.360 10/18/1996 S022 3.350 25263 6745 0 .37484 10/5/1996 3.350 10/19/1996 S023 3.360 36056 6745 0 .37438 10/5/1996 3.360 10/19/1996 S024 3.350 36733 6745 0.37399 10/5/1996 3.350 10/19/1996 S025 3.350 38839 6201 0 .37472 4/1/1998 3.350 4/20/1998 S026 3.350 36194 6745 0 .37447 10/5/1996 3.350 10/18/1996 S027 3.350 36245 6745 0.37484 10/5/1996 3.350 10/19/1996 S028 3.350 38922 6201 0 .37491 4/1/1998 3.350 4/19/1998 S029 3.350 38852 6201 0.37435 4/1/1998 3.350 4/19/1998 S030 3.350 38842 6201 0.37487 4/1/1998 3.350 4/20/1998 S031 3.360 38847 6201 0.37397 4/1/1998 3.360 4/19/1998 S032 3.350 35305 6745 0.37441 10/5/1996 3.350 10/20/1996 S033 3.360 38828 6201 0 .37394 4/1/1998 3.360 4/18/1998 S034 3.350 48846 6201 0 .37366 4/1/1998 3.350 4/19/1998 S035 3.350 36839 6745 0.37382 10/5/1996 3.350 10/20/1996 S036 3.360 36653 6745 0.37383 10/5/1996 3.360 10/18/1996 S037 3.350 38889 6201 0.37350 4/1/1998 3.350 4/19/1998 S038 3.350 48575 6201 0.37371 4/1/1998 3.350 4/19/1998 S039 3.350 36580 6745 0.37393 10/5/1996 3.350 10/19/1996 S040 3.350 48653 6201 0.37376 4/1/1998 3.350 4/18/1998 S041 3.350 48849 6201 0.37361 4/1/1998 3.350 4/18/1998 TOOl 4.160 43418 4709 0.37475 5/3/2002 4.160 5/12/2002 T002 4.160 40362 4709 0.37483 5/3/2002 4.160 5/12/2002 T003 4.170 40558 4709 0.37461 5/3/2002 4.170 5/12/2002 T004 4.160 40400 4709 0.37482 5/3/2002 4.160 5/12/2002 TOOS 4.160 40575 4709 0.37457 5/3/2002 4.160 5/12/2002 T006 4.170 42678 4709 0.37481 5/3/2002 4.170 5/12/2002 T007 4.160 43447 4709 0.37510 5/3/2002 4.160 5/12/2002 T008 4.160 42817 4709 0.37496 5/3/2002 4.160 5/12/2002 T009 4.160 41055 4709 0.37436 5/3/2002 4.160 5/12/2002 TOlO 4.150 41057 4709 0.37387 5/3/2002 4.150 5/12/2002 TOll 4.150 40971 4709 0.37426 5/3/2002 4.150 5/12/2002 FC08513, R0 EC 68969 184

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 185 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID T012 4.160 41048 4709 0 .37449 5/3/2002 4.160 5/12/2002 T013 4.160 38480 4709 0 .37412 5/3/2002 4.160 5/14/2002 T014 4.160 33582 4709 0 .37440 5/3/2002 4.160 5/14/2002 T015 4.150 42777 4709 0 .37436 5/3/2002 4.150 5/14/2002 T016 4.160 42780 4709 0 .37441 5/3/2002 4.160 5/14/2002 T017 4.150 38499 4709 0.37407 5/3/2002 4.150 5/14/2002 T018 4.150 38433 4709 0 .37402 5/3/2002 4.150 5/13/2002 T019 4.160 38489 4709 0.37415 5/3/2002 4.160 5/13/2002 T020 4.160 33594 4709 0.37438 5/3/2002 4.160 5/14/2002 T021 4.160 42706 4709 0.37430 5/3/2002 4.160 5/13/2002 T022 4.160 33617 4709 0.37450 5/3/2002 4.160 5/14/2002 T023 4.150 42735 4709 0.37420 5/3/2002 4.150 5/13/2002 T024 4.160 33611 4709 0.37486 5/3/2002 4.160 5/14/2002 T025 3.760 32898 6201 0.37413 4/1/1998 3.760 4/18/1998 T026 3.760 32859 6201 0.37431 4/1/1998 3.760 4/19/1998 T027 3.760 34882 6201 0.37436 4/1/1998 3.760 4/18/1998 T028 3.760 34802 6201 0.37423 4/1/1998 3.760 4/19/1998 T029 3.760 40615 4709 0.37386 5/3/2002 3.760 5/13/2002 T030 3.760 30273 4709 0.37405 5/3/2002 3.760 5/14/2002 T031 3.750 30286 4709 0.37393 5/3/2002 3.750 5/14/2002 T032 3.750 40662 4709 0.37393 5/3/2002 3.750 5/14/2002 T033 3.750 34088 4709 0.37468 5/3/2002 3.750 5/13/2002 T034 3.750 30276 4709 0.37407 5/3/2002 3.750 5/14/2002 T035 3.750 40626 4709 0.37361 5/3/2002 3.750 5/13/2002 T036 3.760 40681 4709 0 .37368 5/3/2002 3.760 5/14/2002 T037 3.750 30252 4709 0 .37431 5/3/2002 3.750 5/14/2002 T038 3.750 34080 4709 0.37507 5/3/2002 3.750 5/13/2002 T039 3.760 34107 4709 0.37490 5/3/2002 3.760 5/14/2002 T040 3.760 34086 4709 0.37501 5/3/2002 3.760 5/14/2002 T041 3.750 40251 4709 0.37460 5/3/2002 3.750 5/13/2002 T042 3.750 37417 6201 0.37434 4/1/1998 3.750 4/19/1998 T043 3.750 40304 4709 0.37485 5/3/2002 3.750 5/14/2002 T044 3.750 37448 6201 0.37481 4/1/1998 3.750 4/18/1998 T045 3.750 40317 4709 0.37454 5/3/2002 3.750 5/14/2002 T046 3.750 40267 4709 0.37480 5/3/2002 3.750 5/13/2002 T047 3.750 37411 6201 0.37473 4/1/1998 3.750 4/19/1998 T048 3.750 37551 6201 0.37442 4/1/1998 3.750 4/18/1998 U001 0.270 10537 5122 0.37500 3/15/2001 0.270 3/28/2001 U002 0.270 10650 5122 0.37600 3/15/2001 0.270 3/29/2001 U003 0.270 10492 5122 0.37500 3/15/2001 0.270 3/30/2001 U004 0.270 10630 5122 0.37600 3/15/2001 0.270 3/30/2001 U005 4.250 46960 5122 0.37431 3/15/2001 4.250 3/29/2001 U006 4.240 46926 5122 0.37436 3/15/2001 4.240 3/29/2001 U007 4.250 46963 5122 0.37455 3/15/2001 4.250 3/28/2001 U008 4.250 48955 5122 0.37514 3/15/2001 4.250 3/30/2001 U009 4.250 46971 5122 0.37414 3/15/2001 4.250 3/30/2001 U010 4.250 49176 5122 0.37543 3/15/2001 4.250 3/29/2001 UOll 4.250 49116 5122 0.37465 3/15/2001 4.250 3/29/2001 U012 4.250 49026 5122 0.37560 3/15/2001 4.250 3/30/2001 U013 4.250 40246 5122 0.37532 3/15/2001 4.250 3/30/2001 U014 4.250 41735 5122 0.37522 3/15/2001 4.250 3/29/2001 FC08513, R0 EC 68969 185

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 186 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD{f ID Assembly ID Assembly ID Assembly ID Assembly ID UOlS 4.250 40279 5122 0 .37588 3/15/2001 4.250 3/29/2001 U016 4.250 41789 5122 0 .37550 3/15/2001 4.250 3/30/2001 U017 4.250 41641 5122 0 .37559 3/15/2001 4.250 3/30/2001 U018 4.250 41632 5122 0.37597 3/15/2001 4.250 3/30/2001 U019 4.250 41646 5122 0.37556 3/15/2001 4.250 3/29/2001 U020 4.250 41644 5122 0 .37590 3/15/2001 4.250 3/29/2001 U021 4.250 43491 5122 0 .37535 3/15/2001 4.250 3/29/2001 U022 4.250 43474 5122 0 .37515 3/15/2001 4.250 3/30/2001 U023 4.250 43485 5122 0 .37521 3/15/2001 4.250 3/29/2001 U024 4.250 43491 5122 0 .37518 3/15/2001 4.250 3/30/2001 U025 4.260 44329 5122 0 .37502 3/15/2001 4.260 3/30/2001 U026 4.250 44339 5122 0.37492 3/15/2001 4.250 3/30/2001 U027 4.250 44328 5122 0.37469 3/15/2001 4.250 3/29/2001 U028 4.260 47457 5122 0.37498 3/15/2001 4.260 3/30/2001 U029 4.260 47446 5122 0.37518 3/15/2001 4.260 3/29/2001 U030 4.260 47401 5122 0.37536 3/15/2001 4.260 3/28/2001 U031 4.250 44201 5122 0.37526 3/15/2001 4.250 3/29/2001 U032 4.260 47301 5122 0.37496 3/15/2001 4.260 3/29/2001 U033 4.250 41484 5122 0.37470 3/15/2001 4.250 3/28/2001 U034 4.260 41644 5122 0.37487 3/15/2001 4.260 3/29/2001 U035 4.260 41628 5122 0.37515 3/15/2001 4.260 3/31/2001 U036 4.250 41469 5122 0.37506 3/15/2001 4.250 3/30/2001 U037 4.250 44968 5122 0 .37449 3/15/2001 4.250 3/30/2001 U038 4.250 44978 5122 0.37471 3/15/2001 4.250 3/29/2001 U039 4.250 45000 5122 0 .37479 3/15/2001 4.250 3/29/2001 U040 4.250 44891 5122 0.37469 3/15/2001 4.250 3/30/2001 U041 4.250 42776 5122 0 .37517 3/15/2001 4.250 3/28/2001 U042 4.250 42863 5122 0.37509 3/15/2001 4.250 3/30/2001 U043 4.250 42874 5122 0.37495 3/15/2001 4.250 3/29/2001 U044 4.250 42780 5122 0.37524 3/15/2001 4.250 3/28/2001 WOOl 4.210 35313 5122 0 .37416 3/15/2001 4.210 3/28/2001 W002 4.210 35242 5122 0 .37364 3/15/2001 4.210 3/28/2001 W003 4.210 35413 5122 0 .37405 3/15/2001 4.210 3/28/2001 W004 4.210 35515 5122 0 .37361 3/15/2001 4.210 3/30/2001 woos 4.220 25485 5122 0 .37451 3/15/2001 4.220 3/29/2001 W006 4.210 33783 5122 0.37439 3/15/2001 4.210 3/28/2001 W007 4.220 25146 5122 0 .37455 3/15/2001 4.220 3/28/2001 W008 4.210 33740 5122 0 .37440 3/15/2001 4.210 3/28/2001 W009 4.220 33982 5122 0.37394 3/15/2001 4.220 3/30/2001 WOlO 4.220 25180 5122 0.37448 3/15/2001 4.220 3/28/2001 W011 4.220 25423 5122 0.37470 3/15/2001 4.220 3/30/2001 W012 4.220 34026 5122 0.37372 3/15/2001 4.220 3/29/2001 W013 3.600 36117 5122 0.37210 3/15/2001 3.600 3/30/2001 W014 3.600 29926 5122 0.37063 3/15/2001 3.600 3/30/2001 WOlS 3.600 36131 5122 0.37164 3/15/2001 3.600 3/28/2001 W016 3.600 29958 5122 0.37090 3/15/2001 3.600 3/28/2001 W017 3.600 36118 5122 0.37182 3/15/2001 3.600 3/30/2001 W018 3.600 30005 5122 0.37103 3/15/2001 3.600 3/29/2001 W019 3.600 35985 5122 0.37216 3/15/2001 3.600 3/28/2001 W020 3.600 30007 5122 0.37126 3/15/2001 3.600 3/30/2001 W021 3.600 37505 5122 0.37192 3/15/2001 3.600 3/29/2001 FC08513, R0 EC 68969 186

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 187 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/f ID Assembly ID Assembly ID Assembly ID Assembly ID W022 3.600 31266 5122 0.37254 3/15/2001 3.600 3/30/2001 W023 3.600 37423 5122 0.37155 3/15/2001 3.600 3/28/2001 W024 3.600 31442 5122 0.37165 3/15/2001 3.600 3/29/2001 W02 5 3.600 37489 5122 0.37154 3/15/2001 3.600 3/29/2001 W026 3.600 36963 5122 0.37168 3/15/2001 3.600 3/29/2001 W027 3.600 37583 5122 0 .37147 3/15/2001 3.600 3/30/2001 W028 3.610 31411 5122 0 .37125 3/15/2001 3.610 3/29/2001 W029 3.600 31459 5122 0 .37154 3/15/2001 3.600 3/30/2001 W030 3.600 37167 5122 0.37125 3/15/2001 3.600 3/30/2001 W031 3.600 37539 5122 0.37188 3/15/2001 3.600 3/28/2001 W032 3.600 37491 5122 0.37127 3/15/2001 3.600 3/30/2001 W033 3.610 37289 5122 0 .37218 3/15/2001 3.610 3/29/2001 W034 3.610 37035 5122 0.37216 3/15/2001 3.610 3/30/2001 W035 3.610 37147 5122 0.37167 3/15/2001 3.610 3/30/2001 W036 3.610 37262 5122 0.37144 3/15/2001 3.610 3/29/2001 W037 3.610 36860 5122 0 .37014 3/15/2001 3.610 3/29/2001 W038 3.610 36913 5122 0.37026 3/15/2001 3.610 3/30/2001 W039 3.610 36777 5122 0.37032 3/15/2001 3.610 3/30/2001 W040 3.610 36885 5122 0.37072 3/15/2001 3.610 3/28/2001 W041 4.240 38275 5122 0.37076 3/15/2001 4.240 3/28/2001 W042 4.230 38351 5122 0.37087 3/15/2001 4.230 3/28/2001 W043 4.240 38247 5122 0.37086 3/15/2001 4.240 3/29/2001 W044 4.240 38415 5122 0.37108 3/15/2001 4.240 3/30/2001 X001 3.510 35800 4212 0.37092 9/12/2003 3.510 9/23/2003 X002 3.510 35762 4212 0.37077 9/12/2003 3.510 9/26/2003 X003 3.510 35709 4212 0.37089 9/12/2003 3.510 9/26/2003 X004 3.510 35743 4212 0.37086 9/12/2003 3.510 9/22/2003 X005 4.170 46151 4212 0.37439 9/12/2003 4.170 9/26/2003 X006 4.170 46136 4212 0.37561 9/12/2003 4.170 9/23/2003 X007 4.170 46203 4212 0.37505 9/12/2003 4.170 9/22/2003 X008 4.170 45565 4212 0.37406 9/12/2003 4.170 9/22/2003 X009 4.170 45561 4212 0.37406 9/12/2003 4.170 9/26/2003 X010 4.170 45440 4212 0.37435 9/12/2003 4.170 9/26/2003 X011 4.170 46286 4212 0.37439 9/12/2003 4.170 9/26/2003 X012 4.170 45529 4212 0.37391 9/12/2003 4.170 9/22/2003 X013 4.190 47767 4212 0.37324 9/12/2003 4.190 9/23/2003 X014 4.190 39355 4212 0.37286 9/12/2003 4.190 9/23/2003 X015 4.190 47533 4212 0.37325 9/12/2003 4.190 9/26/2003 X016 4.190 47666 4212 0.37313 9/12/2003 4.190 9/22/2003 X017 4.190 39317 4212 0.37320 9/12/2003 4.190 9/22/2003 X018 4.190 39272 4212 0.37311 9/12/2003 4.190 9/22/2003 X019 4.190 47604 4212 0.37330 9/12/2003 4.190 9/26/2003 X020 4.190 39383 4212 0.37320 9/12/2003 4.190 9/26/2003 X021 4.190 44473 4212 0.37249 9/12/2003 4.190 9/22/2003 X022 4.190 44526 4212 0.37250 9/12/2003 4.190 9/23/2003 X023 4.190 44249 4212 0.37293 9/12/2003 4.190 9/26/2003 X024 4.190 44369 4212 0.37273 9/12/2003 4.190 9/26/2003 X025 4.190 37726 4212 0.37212 9/12/2003 4.190 9/26/2003 X026 4.190 37677 4212 0.37206 9/12/2003 4.190 9/22/2003 X027 4.190 37659 4212 0.37167 9/12/2003 4.190 9/22/2003 X028 4.190 37746 4212 0.37158 9/12/2003 4.190 9/23/2003 FC08513, R0 EC 68969 187

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 188 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID X029 4.200 39462 4212 0.37163 9/12/2003 4.200 9/26/2003 X030 4.200 44486 4212 0.37131 9/12/2003 4.200 9/23/2003 X031 4.200 40850 4212 0.37140 9/12/2003 4.200 9/26/2003 X032 4.200 40853 4212 0.37135 9/12/2003 4.200 9/26/2003 X033 4.200 44442 4212 0.37122 9/12/2003 4.200 9/22/2003 X034 4.200 40681 4212 0.37144 9/12/2003 4.200 9/23/2003 X035 4.200 40864 4212 0.37125 9/12/2003 4.200 9/23/2003 X036 4.200 44499 4212 0.37100 9/12/2003 4.200 9/26/2003 X037 4.200 39451 4212 0.37173 9/12/2003 4.200 9/23/2003 X038 4.200 39499 4212 0.37161 9/12/2003 4.200 9/26/2003 X039 4.200 39565 4212 0.37179 9/12/2003 4.200 9/23/2003 X040 4.200 44510 4212 0.37133 9/12/2003 4.200 9/25/2003 YOOl 0.340 11666 2530 0.37509 4/19/2008 0.340 5/1/2008 Y002 0.350 11671 2530 0.37520 4/19/2008 0.350 4/30/2008 Y003 0.340 11687 2530 0.37510 4/19/2008 0.340 4/30/2008 Y004 0.340 11658 2530 0.37515 4/19/2008 0.340 5/1/2008 Y005 3.740 39696 1968 0.37159 11/1/2009 3.740 11/11/2009 Y006 3.740 39687 1968 0.37172 11/1/2009 3.740 11/11/2009 Y007 3.740 49259 2530 0.37163 4/19/2008 3.740 4/30/2008 Y008 3.740 33047 4212 0.37143 9/12/2003 3.740 9/23/2003 Y009 3.740 39696 1968 0.37168 11/1/2009 3.740 11/10/2009 YOlO 3.740 49401 1968 0.37159 11/1/2009 3.740 11/11/2009 Y011 3.740 39689 1968 0.37180 11/1/2009 3.740 11/11/2009 Y012 3.740 47084 3118 0.37167 9/9/2006 3.740 9/17/2006 Y013 3.960 44814 3118 0.37273 9/9/2006 3.960 9/17/2006 Y014 3.960 44723 3118 0.37267 9/9/2006 3.960 9/17/2006 Y015 3.960 44668 3118 0.37270 9/9/2006 3.960 9/18/2006 Y016 3.960 44661 3118 0.37260 9/9/2006 3.960 9/18/2006 Y017 3.960 41757 1968 0.37247 11/1/2009 3.960 11/11/2009 Y018 3.960 44681 3677 0.37246 2/26/2005 3.960 3/13/2005 Y019 3.960 43506 1445 0.37240 4/9/2011 3.960 4/23/2011 Y020 3.960 41776 1968 0.37248 11/1/2009 3.960 11/11/2009 Y021 3.960 45045 3677 0.37245 2/26/2005 3.960 3/12/2005 Y022 3.960 43489 1445 0.37263 4/9/2011 3.960 4/23/2011 Y023 3.910 43429 1445 0.37050 4/9/2011 3.910 4/22/2011 Y024 3.910 42418 3118 0.37110 9/9/2006 3.910 9/18/2006 Y025 3.910 42524 3118 0.37118 9/9/2006 3.910 9/18/2006 Y026 3.910 41708 1968 0.37117 11/1/2009 3.910 11/12/2009 Y027 3.920 42484 3118 0.37111 9/9/2006 3.920 9/19/2006 Y028 3.910 42532 3118 0.37125 9/9/2006 3.910 9/19/2006 Y029 3.910 43409 1445 0.37125 4/9/2011 3.910 4/24/2011 Y030 3.910 41687 1968 0.37131 11/1/2009 3.910 11/11/2009 Y031 4.050 55274 1968 0.37431 11/1/2009 4.050 11/11/2009 Y032 4.050 55233 1968 0.37440 11/1/2009 4.050 11/11/2009 Y033 4.060 55230 1968 0.37397 11/1/2009 4.060 11/10/2009 Y034 4.050 55269 1968 0.37408 11/1/2009 4.050 11/11/2009 Y035 4.030 44562 1968 0.37385 11/1/2009 4.030 11/10/2009 Y036 4.030 44544 1968 0.37373 11/1/2009 4.030 11/11/2009 Y037 4.030 44550 1968 0.37375 11/1/2009 4.030 11/11/2009 Y038 4.040 44568 1968 0.37380 11/1/2009 4.040 11/11/2009 Y039 4.020 44386 3677 0.37337 2/26/2005 4.020 3/12/2005 FC08513, R0 EC 68969 188

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 189 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID Y040 4.010 44410 3677 0.37340 2/26/2005 4.010 3/13/2005 Y041 4.030 44669 1968 0.37365 11/1/2009 4.030 11/12/2009 Y042 4.030 44687 1968 0.37361 11/1/2009 4.030 11/12/2009 Y043 4.030 44706 1968 0.37344 11/1/2009 4.030 11/11/2009 Y044 4.020 44687 1968 0.37305 11/1/2009 4.020 11/11/2009 Y045 4.000 42242 3677 0.37256 2/26/2005 4.000 3/12/2005 Y046 4.000 38532 3118 0.37310 9/9/2006 4.000 9/17/2006 Y047 4.000 42173 3677 0.37241 2/26/2005 4.000 3/13/2005 Y048 4.010 38671 3118 0.37316 9/9/2006 4.010 9/19/2006 Y049 4.000 42141 3677 0.37313 2/26/2005 4.000 3/13/2005 YOSO 4.000 38740 3118 0.37328 9/9/2006 4.000 9/18/2006 Y051 4.010 38643 3118 0.37295 9/9/2006 4.010 9/18/2006 Y0 52 4.000 42367 3677 0.37251 2/26/2005 4.000 3/13/2005 Y0 53 3.430 46667 3677 0.37309 2/26/2005 3.430 3/13/2005 Z001 3.440 35649 3677 0.37155 2/26/2005 3.440 3/13/2005 Z002 3.440 35732 3677 0.37150 2/26/2005 3.440 3/12/2005 Z003 3.440 35893 3677 0.37150 2/26/2005 3.440 3/13/2005 Z004 3.440 35840 3677 0.37138 2/26/2005 3.440 3/13/2005 zoos 3.650 37298 3118 0.37141 9/9/2006 3.650 9/18/2006 Z006 3.650 37405 3118 0.37141 9/9/2006 3.650 9/18/2006 Z007 3.650 37387 3118 0.37116 9/9/2006 3.650 9/18/2006 zoos 3.650 37238 3118 0.37136 9/9/2006 3.650 9/17/2006 Z009 3.420 35805 3677 0.37090 2/26/2005 3.420 3/13/2005 ZOlO 3.420 35872 3677 0.37075 2/26/2005 3.420 3/12/2005 Z011 3.430 35908 3677 0.37091 2/26/2005 3.430 3/12/2005 Z012 3.430 36151 3677 0.37092 2/26/2005 3.430 3/12/2005 Z013 3.430 35918 3677 0.37082 2/26/2005 3.430 3/13/2005 Z014 3.430 35957 3677 0.37091 2/26/2005 3.430 3/13/2005 Z015 3.420 36019 3677 0.37074 2/26/2005 3.420 3/13/2005 Z016 3.420 36077 3677 0.37072 2/26/2005 3.420 3/12/2005 Z017 3.680 42323 3118 0.37219 9/9/2006 3.680 9/18/2006 Z018 3.680 42239 3118 0.37229 9/9/2006 3.680 9/18/2006 Z019 3.680 31456 3677 0.37211 2/26/2005 3.680 3/13/2005 Z020 3.680 41856 0 0.37211 3/24/2015 3.680 3/24/2015 Z021 3.690 41851 0 0.37205 3/24/2015 3.690 3/24/2015 Z022 3.680 31448 3677 0.37221 2/26/2005 3.680 3/12/2005 Z023 3.680 42219 3118 0.37213 9/9/2006 3.680 9/17/2006 Z024 3.680 42298 3118 0.37229 9/9/2006 3.680 9/18/2006 Z025 3.630 38797 2530 0.37089 4/19/2008 3.630 5/1/2008 Z026 3.630 37838 3118 0.37074 9/9/2006 3.630 9/18/2006 Z027 3.630 37923 3118 0.37086 9/9/2006 3.630 9/18/2006 Z028 3.630 38791 2530 0.37126 4/19/2008 3.630 5/1/2008 Z029 3.630 35729 3677 0.37114 2/26/2005 3.630 3/13/2005 Z030 3.630 35600 3677 0.37103 2/26/2005 3.630 3/13/2005 Z031 3.630 36268 3677 0.37101 2/26/2005 3.630 3/12/2005 Z032 3.630 36037 3677 0.37083 2/26/2005 3.630 3/13/2005 Z033 3.630 36163 3677 0.37105 2/26/2005 3.630 3/12/2005 Z034 3.630 35888 3677 0.37081 2/26/2005 3.630 3/13/2005 Z035 3.630 35591 3677 0.37084 2/26/2005 3.630 3/12/2005 Z036 3.630 35509 3677 0.37080 2/26/2005 3.630 3/13/2005 Z037 3.630 38817 2530 0.37081 4/19/2008 3.630 4/30/2008 FC08513, R0 EC 68969 189

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 190 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID Z038 3.630 37915 3118 0.37107 9/9/2006 3.630 9/19/2006 Z039 3.630 37852 3118 0.37107 9/9/2006 3.630 9/17/2006 Z040 3.630 38829 2530 0.37114 4/19/2008 3.630 4/30/2008 AA01 3.950 46757 2530 0.37135 4/19/2008 3.950 4/30/2008 AA02 3.950 46798 2530 0.37142 4/19/2008 3.950 4/30/2008 AA03 3.950 41031 2530 0.37155 4/19/2008 3.950 5/1/2008 AA04 3.950 41035 2530 0.37133 4/19/2008 3.950 4/29/2008 AA05 3.950 44560 2530 0.37125 4/19/2008 3.950 4/30/2008 AA06 3.950 39436 1968 0.37126 11/1/2009 3.950 11/11/2009 AA07 3.950 44435 2530 0.37127 4/19/2008 3.950 4/29/2008 AA08 3.950 40317 1968 0.37140 11/1/2009 3.950 11/12/2009 AA09 3.950 40319 1968 0.37133 11/1/2009 3.950 11/11/2009 AA10 3.950 44490 2530 0.37122 4/19/2008 3.950 5/1/2008 AA11 3.950 39439 1968 0.37135 11/1/2009 3.950 11/12/2009 AA12 3.950 44494 2530 0.37124 4/19/2008 3.950 5/1/2008 AA13 3.950 40983 2530 0.37109 4/19/2008 3.950 5/1/2008 AA14 3.950 41047 2530 0.37128 4/19/2008 3.950 4/30/2008 AA15 3.950 46840 2530 0.37140 4/19/2008 3.950 4/30/2008 AA16 3.950 46860 2530 0.37123 4/19/2008 3.950 4/29/2008 AA17 4 .250 51519 2530 0.37206 4/19/2008 4.250 4/30/2008 AA18 4.250 51197 2530 0.37227 4/19/2008 4.250 4/30/2008 AA19 4 .250 51224 2530 0.37232 4/19/2008 4.250 4/30/2008 AA20 4.250 51519 2530 0.37231 4/19/2008 4.250 4/30/2008 AA21 4.180 43668 2530 0.37099 4/19/2008 4.180 5/1/2008 AA22 4.180 46703 1968 0.37080 11/1/2009 4.180 11/12/2009 AA23 4.180 43731 2530 0.37108 4/19/2008 4.180 5/1/2008 AA24 4.180 44256 1968 0.37115 11/1/2009 4.180 11/12/2009 AA25 4.180 44215 1968 0.37084 11/1/2009 4.180 11/10/2009 AA26 4 .180 47706 1968 0.37089 11/1/2009 4 .180 11/10/2009 AA27 4.180 47712 1968 0.37093 11/1/2009 4.180 11/12/2009 AA28 4.190 44216 1968 0.37100 11/1/2009 4.190 11/12/2009 AA29 4.190 44256 1968 0.37098 11/1/2009 4.190 11/11/2009 AA30 4 .190 43754 2530 0.37069 4/19/2008 4 .190 4/30/2008 AA31 4.190 46701 1968 0.37062 11/1/2009 4.190 11/11/2009 AA32 4.190 43693 2530 0.37100 4/19/2008 4.190 4/29/2008 AA33 4.310 44072 1445 0.37307 4/9/2011 4.310 4/23/2011 AA34 4.310 44094 1445 0.37330 4/9/2011 4.310 4/24/2011 AA35 4.310 44093 1445 0.37339 4/9/2011 4 .310 4/23/2011 AA36 4.300 44074 1445 0.37309 4/9/2011 4.300 4/24/2011 AA37 4.280 48361 0 0.37240 3/24/2015 4 .280 3/24/2015 AA38 4.270 35667 3118 0.37292 9/9/2006 4.270 9/18/2006 AA39 4.270 50914 1445 0.37270 4/9/2011 4.270 4/23/2011 AA40 4.280 48362 0 0 .37244 3/24/2015 4.280 3/24/2015 AA41 4.300 46172 0 0.37296 3/24/2015 4.300 3/24/2015 AA42 4.300 45624 0 0.37275 3/24/2015 4.300 3/24/2015 AA43 4.300 45623 0 0.37280 3/24/2015 4.300 3/24/2015 AA44 4.300 46172 0 0.37269 3/24/2015 4.300 3/24/2015 BB01 4.110 49362 1445 0.37353 4/9/2011 4.110 4/23/2011 BB02 4 .110 49356 1445 0.37343 4/9/2011 4.110 4/23/2011 BB03 4.110 49334 1445 0.37363 4/9/2011 4.110 4/23/2011 BB04 4.110 49378 1445 0.37387 4/9/2011 4.110 4/23/2011 FC08513, R0 EC 68969 190

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 191 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID BB05 4.090 45301 0 0 .37342 3/24/2015 4.090 3/24/2015 BB06 4.090 45312 0 0 .37340 3/24/2015 4.090 3/24/2015 BB07 4.090 45757 1968 0 .37302 11/1/2009 4.090 11/11/2009 BB08 4.090 45065 1968 0 .37327 11/1/2009 4.090 11/10/2009 BB09 4 .090 47731 1445 0.37324 4/9/2011 4.090 4/23/2011 BB10 4.090 46380 1445 0.37327 4/9/2011 4.090 4/23/2011 BB11 4.090 46385 1445 0 .37315 4/9/2011 4.090 4/23/2011 BB12 4 .090 47715 1445 0.37326 4/9/2011 4.090 4/23/2011 BB13 4.090 45066 1968 0 .37318 11/1/2009 4.090 11/11/2009 BB14 4 .090 45758 1968 0.37323 11/1/2009 4.090 11/11/2009 BB15 4.090 45314 0 0.37338 3/24/2015 4.090 3/24/2015 BB16 4.090 45300 0 0 .37347 3/24/2015 4.090 3/24/2015 BB17 3.970 40337 1445 0.37078 4/9/2011 3.970 4/22/2011 BB18 3.970 43174 1968 0 .37074 11/1/2009 3.970 11/11/2009 BB19 3.970 43175 1968 0.37069 11/1/2009 3.970 11/12/2009 BB20 3.970 40338 1445 0 .37075 4/9/2011 3.970 4/24/2011 BB21 3.970 43289 1968 0.37070 11/1/2009 3.970 11/12/2009 BB22 3.970 40544 1445 0.37070 4/9/2011 3.970 4/23/2011 BB23 3.970 40546 1445 0.37079 4/9/2011 3.970 4/23/2011 BB24 3.970 43285 1968 0.37076 11/1/2009 3.970 11/10/2009 BB25 3.840 37656 2530 0.37070 4/19/2008 3.840 4/30/2008 BB26 3.840 39743 1968 0.37064 11/1/2009 3.840 11/11/2009 BB27 3.840 39731 1968 0.37057 11/1/2009 3.840 11/10/2009 BB28 3.840 46510 1968 0.37075 11/1/2009 3.840 11/11/2009 BB29 3.840 46530 1968 0.37069 11/1/2009 3.840 11/11/2009 BB30 3.840 41846 0 0.37059 3/24/2015 3.840 3/24/2015 BB31 3.840 38079 2530 0.37058 4/19/2008 3.840 4/30/2008 BB32 3.840 38136 2530 0.37076 4/19/2008 3.840 4/30/2008 BB33 3.840 41858 0 0.37055 3/24/2015 3.840 3/24/2015 BB34 3.840 37748 2530 0.37082 4/19/2008 3.840 4/30/2008 BB35 3.840 37720 2530 0.37068 4/19/2008 3.840 4/30/2008 BB36 3.830 41857 0 0.37103 3/24/2015 3.830 3/24/2015 BB37 3.830 38060 2530 0 .37090 4/19/2008 3.830 4/30/2008 BB38 3.830 37962 2530 0 .37092 4/19/2008 3.830 4/30/2008 BB39 3.840 41847 0 0.37073 3/24/2015 3.840 3/24/2015 BB40 3.830 46532 1968 0.37089 11/1/2009 3.830 11/11/2009 BB41 3.840 46506 1968 0.37100 11/1/2009 3.840 11/10/2009 BB42 3.840 39731 1968 0.37092 11/1/2009 3.840 11/11/2009 BB43 3.840 39744 1968 0.37082 11/1/2009 3.840 11/11/2009 BB44 3.840 37665 2530 0.37073 4/19/2008 3.840 4/29/2008 CC01 4.000 43912 1445 0.38889 4/9/2011 3.980 4/24/2011 CC02 4.000 43979 1445 0.38880 4/9/2011 3.990 4/23/2011 CC03 4.000 41020 0 0.38886 3/24/2015 3.990 3/24/2015 CC04 4.000 40984 0 0.38919 3/24/2015 3.990 3/24/2015 CC05 4 .000 44983 0 0.38898 3/24/2015 3.990 3/24/2015 CC06 4.000 45037 0 0.38921 3/24/2015 3.990 3/24/2015 CC07 4.000 45036 0 0.38866 3/24/2015 3.980 3/24/2015 CC08 4.000 44984 0 0.38881 3/24/2015 3.990 3/24/2015 CC09 4.000 40986 0 0.38919 3/24/2015 3.980 3/24/2015 CC10 4.000 41017 0 0.38933 3/24/2015 3.990 3/24/2015 CC11 4.000 43979 1445 0.38936 4/9/2011 3.990 4/24/2011 FC08513, R0 EC 68969 191

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 192 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/T ID Assembly ID Assembly ID Assembly ID Assembly ID CC12 4.000 43914 1445 0.38925 4/9/2011 3.990 4/22/2011 CC13 4.000 42936 1445 0.38902 4/9/2011 3.970 4/24/2011 CC14 4.000 42941 1445 0.38903 4/9/2011 3.970 4/22/2011 CC15 4.000 42941 1445 0.38893 4/9/2011 3.970 4/24/2011 CC16 4.000 42937 1445 0 .38907 4/9/2011 3.970 4/23/2011 CC17 4.000 39856 1445 0 .38649 4/9/2011 3.700 4/24/2011 CC18 4.000 39882 1445 0.38648 4/9/2011 3.700 4/24/2011 CC19 4.000 42004 1445 0.38652 4/9/2011 3.700 4/23/2011 CC20 4.000 38805 1968 0.38658 11/1/2009 3.700 11/11/2009 CC21 4.000 38807 1968 0.38648 11/1/2009 3.700 11/11/2009 CC22 4.000 41959 1445 0.38639 4/9/2011 3.700 4/24/2011 CC23 4.000 38840 1968 0.38651 11/1/2009 3.700 11/11/2009 CC24 4.000 37079 1445 0.38658 4/9/2011 3.700 4/23/2011 CC25 4.000 38798 1968 0.38668 11/1/2009 3.700 11/11/2009 CC26 4.000 37165 1445 0.38667 4/9/2011 3.700 4/23/2011 CC27 4.000 37181 1445 0.38680 4/9/2011 3.700 4/23/2011 CC28 4.000 38797 1968 0.38636 11/1/2009 3.700 11/10/2009 CC29 4.000 37078 1445 0 .38628 4/9/2011 3.700 4/23/2011 CC30 4.000 38841 1968 0 .38619 11/1/2009 3.700 11/11/2009 CC31 4.000 41959 1445 0 .38620 4/9/2011 3.700 4/22/2011 CC32 4.000 38806 1968 0.38656 11/1/2009 3.700 11/10/2009 CC33 4.000 38806 1968 0 .38639 11/1/2009 3.700 11/11/2009 CC34 4.000 42005 1445 0 .38643 4/9/2011 3.700 4/23/2011 CC35 4.000 39882 1445 0.38627 4/9/2011 3.700 4/23/2011 CC36 4.000 39856 1445 0 .38624 4/9/2011 3.700 4/22/2011 CC37 4.000 43930 1445 0 .38675 4/9/2011 3.880 4/23/2011 CC38 4.000 42849 1445 0 .38673 4/9/2011 3.870 4/24/2011 CC39 4.000 42816 1445 0 .38664 4/9/2011 3.870 4/24/2011 CC40 4.000 43962 1445 0.38571 4/9/2011 3.870 4/24/2011 CC41 4.000 43961 1445 0.38644 4/9/2011 3.870 4/23/2011 CC42 4.000 42815 1445 0.38603 4/9/2011 3.870 4/23/2011 CC43 4.000 42850 1445 0.38604 4/9/2011 3.870 4/23/2011 CC44 4.000 43930 1445 0.38598 4/9/2011 3.870 4/24/2011 DDOl 4.000 48208 0 0.38861 3/24/2015 4.150 3/24/2015 0002 4.000 48253 0 0.38862 3/24/2015 4.150 3/24/2015 0003 4.000 48251 0 0.38847 3/24/2015 4.150 3/24/2015 0004 4.000 48204 0 0.38845 3/24/2015 4.150 3/24/2015 0005 4.000 48800 0 0.38834 3/24/2015 4.150 3/24/2015 0006 4.000 48882 0 0.38844 3/24/2015 4.150 3/24/2015 0007 4.000 48899 0 0.38829 3/24/2015 4.150 3/24/2015 0008 4.000 48733 0 0.38821 3/24/2015 4.150 3/24/2015 0009 4.000 33752 0 0.38837 3/24/2015 4.150 3/24/2015 DOlO 4.000 33730 0 0.38863 3/24/2015 4.150 3/24/2015 DOll 4.000 33734 0 0.38827 3/24/2015 4.150 3/24/2015 0012 4.000 33769 0 0.38797 3/24/2015 4.150 3/24/2015 0013 4.000 40664 0 0.38798 3/24/2015 4.150 3/24/2015 0014 4.000 40565 0 0.38776 3/24/2015 4.150 3/24/2015 0015 4.000 40566 0 0.38856 3/24/2015 4.150 3/24/2015 0016 4.000 40665 0 0.38862 3/24/2015 4.150 3/24/2015 0017 4.000 49074 0 0.38552 3/24/2015 4.030 3/24/2015 0018 4.000 48980 0 0.38504 3/24/2015 4.030 3/24/2015 FC08513, R0 EC 68969 192

FC08514 ext enr.xlsx Attachment 0 SpreadSheets FC08513 U-235 Average Enrichment Page 193 of 269 Assembly Initial U- Burn up Assembly 10 235 Enr MWO/T 10 Assembly 10 Assembly 10 Assembly 10 Assembly 10 0019 4.000 48987 0 0 .38508 3/24/2015 4.030 3/24/2015 0020 4.000 49060 0 0.38536 3/24/2015 4.030 3/24/2015 0021 4.000 38016 1445 0.38501 4/9/2011 4.030 4/23/2011 0022 4.000 50296 0 0.38549 3/24/2015 4.030 3/24/2015 0023 4.000 37730 1445 0.38546 4/9/2011 4.030 4/23/2011 0024 4.000 38026 1445 0.38546 4/9/2011 4.030 4/23/2011 0025 4.000 33382 1445 0.38540 4/9/2011 4.030 4/23/2011 0026 4.000 33599 1445 0.38560 4/9/2011 4.030 4/23/2011 0027 4.000 33598 1445 0.38547 4/9/2011 4.030 4/24/2011 0028 4.000 33383 1445 0 .38545 4/9/2011 4.030 4/23/2011 0029 4.000 45812 0 0.38564 3/24/2015 4.030 3/24/2015 0030 4.000 45889 0 0 .38575 3/24/2015 4.030 3/24/2015 0031 4.000 45891 0 0.38563 3/24/2015 4.030 3/24/2015 0032 4.000 45812 0 0.38567 3/24/2015 4.030 3/24/2015 0033 4.000 31421 1445 0.38560 4/9/2011 3.860 4/23/2011 0034 4.000 31364 1445 0.38594 4/9/2011 3.860 4/23/2011 0035 4.000 31364 1445 0.38633 4/9/2011 3.870 4/22/2011 0036 4.000 31420 1445 0.38658 4/9/2011 3.870 4/23/2011 0037 4.000 43157 0 0.38673 3/24/2015 3.870 3/24/2015 0038 4.000 43099 0 0.38544 3/24/2015 3.860 3/24/2015 0039 4.000 43098 0 0.38620 3/24/2015 3.870 3/24/2015 0040 4.000 43156 0 0.38520 3/24/2015 3.850 3/24/2015 0041 4.000 41382 0 0 .38541 3/24/2015 3.850 3/24/2015 0042 4.000 41297 0 0 .38543 3/24/2015 3.850 3/24/2015 0043 4.000 41303 0 0 .38538 3/24/2015 3.850 3/24/2015 0044 4.000 41378 0 0.38542 3/24/2015 3.850 3/24/2015 EEOl 4.000 32665 0 0.38957 3/24/2015 4.150 3/24/2015 EE02 4.000 32575 0 0.38943 3/24/2015 4.150 3/24/2015 EE03 4.000 32579 0 0.38947 3/24/2015 4.150 3/24/2015 EE04 4.000 32666 0 0.38941 3/24/2015 4.150 3/24/2015 EEOS 4.000 27569 0 0.38929 3/24/2015 4.150 3/24/2015 EE06 4.000 27476 0 0 .38918 3/24/2015 4.150 3/24/2015 EE07 4.000 27479 0 0 .38913 3/24/2015 4.150 3/24/2015 EE08 4.000 27565 0 0.38931 3/24/2015 4.150 3/24/2015 EE09 4.000 33427 0 0.38923 3/24/2015 4.150 3/24/2015 EElO 4.000 33426 0 0.38924 3/24/2015 4.150 3/24/2015 EE11 4.000 33420 0 0.38927 3/24/2015 4.150 3/24/2015 EE12 4.000 33432 0 0.38937 3/24/2015 4.150 3/24/2015 EE13 4.000 32913 0 0.38937 3/24/2015 4.150 3/24/2015 EE14 4.000 32862 0 0.38939 3/24/2015 4.150 3/24/2015 EElS 4.000 32867 0 0.38912 3/24/2015 4.150 3/24/2015 EE16 4.000 32907 0 0.38919 3/24/2015 4.150 3/24/2015 EE17 4.000 34430 0 0.38630 3/24/2015 4.030 3/24/2015 EElS 4.000 34376 0 0.38668 3/24/2015 4.030 3/24/2015 EE19 4.000 34373 0 0.38665 3/24/2015 4.030 3/24/2015 EE20 4.000 34434 0 0.38660 3/24/2015 4.030 3/24/2015 EE21 4.000 34508 0 0.38663 3/24/2015 4.030 3/24/2015 EE22 4.000 34478 0 0.38639 3/24/2015 4.030 3/24/2015 EE23 4.000 34475 0 0.38616 3/24/2015 4.030 3/24/2015 EE24 4.000 34507 0 0.38636 3/24/2015 4.030 3/24/2015 EE25 4.000 34091 0 0.38635 3/24/2015 4.030 3/24/2015 FC08513, R0 EC 68969 193

FC08514 ext enr.xlsx Attachment D SpreadSheets FC08513 U-235 Average Enrichment Page 194 of 269 Assembly Initial U- Burn up Assembly ID 235 Enr MWD/f ID Assembly ID Assembly ID Assembly ID Assembly ID EE26 4.000 34048 0 0.38640 3/24/2015 4.030 3/24/2015 EE27 4.000 34100 0 0.38637 3/24/2015 4.030 3/24/2015 EE28 4.000 34076 0 0.38621 3/24/2015 4.030 3/24/2015 EE29 4.000 32461 0 0.38538 3/24/2015 3.860 3/24/2015 EE30 4.000 32033 0 0.38537 3/24/2015 3.860 3/24/2015 EE31 4.000 32039 0 0.38524 3/24/2015 3.860 3/24/2015 EE32 4.000 32462 0 0.38500 3/24/2015 3.860 3/24/2015 EE33 4.000 34858 0 0.38500 3/24/2015 3.860 3/24/2015 EE34 4.000 34827 0 0.38521 3/24/2015 3.860 3/24/2015 EE35 4.000 34832 0 0.38533 3/24/2015 3.860 3/24/2015 EE36 4.000 34854 0 0.38532 3/24/2015 3.860 3/24/2015 FF01 4.000 14936 0 0.38660 3/24/2015 4.280 3/24/2015 FF02 4.000 14877 0 0.38675 3/24/2015 4.280 3/24/2015 FF03 4.000 14879 0 0.38680 3/24/2015 4.280 3/24/2015 FF04 4.000 14936 0 0.38686 3/24/2015 4.280 3/24/2015 FF05 4.000 15143 0 0.38704 3/24/2015 4.280 3/24/2015 FF06 4.000 15154 0 0.38702 3/24/2015 4.280 3/24/2015 FF07 4.000 15157 0 0.38671 3/24/2015 4.280 3/24/2015 FF08 4.000 15140 0 0.38672 3/24/2015 4.280 3/24/2015 FF09 4.000 15880 0 0.38600 3/24/2015 4.250 3/24/2015 FF10 4.000 15879 0 0.38574 3/24/2015 4 .250 3/24/2015 FF11 4.000 15883 0 0.38560 3/24/2015 4 .250 3/24/2015 FF12 4.000 15878 0 0.38588 3/24/2015 4.250 3/24/2015 FF13 4.000 15802 0 0.38578 3/24/2015 4.250 3/24/2015 FF14 4.000 15817 0 0.38584 3/24/2015 4.250 3/24/2015 FF15 4.000 15821 0 0.38525 3/24/2015 4.250 3/24/2015 FF16 4.000 15799 0 0.38598 3/24/2015 4.250 3/24/2015 FF17 4.000 16525 0 0.38433 3/24/2015 4.020 3/24/2015 FF18 4.000 16490 0 0.38430 3/24/2015 4.020 3/24/2015 FF19 4.000 16494 0 0.38420 3/24/2015 4.020 3/24/2015 FF20 4.000 16524 0 0.38444 3/24/2015 4.020 3/24/2015 FF21 4.000 17667 0 0.38404 3/24/2015 4.020 3/24/2015 FF22 4 .000 17769 0 0.38404 3/24/2015 4.020 3/24/2015 FF23 4 .000 17764 0 0.38450 3/24/2015 4.030 3/24/2015 FF24 4.000 17670 0 0.38438 3/24/2015 4.030 3/24/2015 FF25 4.000 16464 0 0.38435 3/24/2015 4.030 3/24/2015 FF26 4.000 16488 0 0.38432 3/24/2015 4.030 3/24/2015 FF27 4.000 16493 0 0.38452 3/24/2015 4.030 3/24/2015 FF28 4.000 16459 0 0.38460 3/24/2015 4.030 3/24/2015 FF29 4.000 16803 0 0 .38390 3/24/2015 4.030 3/24/2015 FF30 4.000 16908 0 0.38432 3/24/2015 4.020 3/24/2015 FF31 4.000 16924 0 0.38401 3/24/2015 4.030 3/24/2015 FF32 4.000 16792 0 0.38359 3/24/2015 4.020 3/24/2015 FF33 4.000 17208 0 0.38304 3/24/2015 3.860 3/24/2015 FF34 4.000 17214 0 0.38298 3/24/2015 3.860 3/24/2015 FF35 4 .000 17220 0 0.38281 3/24/2015 3.860 3/24/2015 FF36 4 .000 17210 0 0.38288 3/24/2015 3.860 3/24/2015 FF37 4.000 17188 0 0.38307 3/24/2015 3.860 3/24/2015 FF38 4.000 17196 0 0.38311 3/24/2015 3.860 3/24/2015 FF39 4.000 17203 0 0.38319 3/24/2015 3.860 3/24/2015 FF40 4.000 17178 0 0.38300 3/24/2015 3.860 3/24/2015 FC08513, R0 EC 68969 194

FC08514 ext enr.xlsx Attachment 0 SpreadSheets FC08513 U-235 Average Enrichment Page 195 of 269 Assembly Initial U- Burn up Assembly 10 235 Enr MWO/T 10 Assembly 10 Assembly 10 Assembly 10 Assembly 10 GG01 3.760 0 20 0.38680 3/4/2015 3.630 3/4/2015 GG02 3.760 0 20 0.38759 3/4/2015 3.630 3/4/2015 GG03 3.760 0 20 0.38747 3/4/2015 3.630 3/4/2015 GG04 3.760 0 20 0.38691 3/4/2015 3.630 3/4/2015 GG05 3.760 0 20 0.38784 3/4/2015 3.630 3/4/2015 GG06 3.760 0 27 0.38742 2/25/2015 3.630 2/25/2015 GG07 3.760 0 20 0.38750 3/4/2015 3.630 3/4/2015 GG08 3.760 0 20 0.38706 3/4/2015 3.630 3/4/2015 GG09 3.730 0 27 0.38639 2/25/2015 3.610 2/25/2015 GG10 3.730 0 27 0.38632 2/25/2015 3.610 2/25/2015 GG11 3.730 0 27 0.38683 2/25/2015 3.610 2/25/2015 GG12 3.730 0 27 0.38621 2/25/2015 3.610 2/25/2015 GG13 3.730 0 20 0.38612 3/4/2015 3.610 3/4/2015 GG14 3.730 0 20 0.38724 3/4/2015 3.600 3/4/2015 GG15 3.730 0 20 0.38709 3/4/2015 3.600 3/4/2015 GG16 3.730 0 34 0.38677 2/18/2015 3.610 2/18/2015 GG17 3.350 0 34 0.38445 2/18/2015 3.270 2/18/2015 GG18 3.350 0 34 0.38449 2/18/2015 3.270 2/18/2015 GG19 3.350 0 34 0.38459 2/18/2015 3.270 2/18/2015 GG20 3.350 0 20 0.38460 3/4/2015 3.270 3/4/2015 GG21 3.350 0 20 0.38454 3/4/2015 3.270 3/4/2015 GG22 3.350 0 27 0.38440 2/25/2015 3.270 2/25/2015 GG23 3.350 0 27 0.38438 2/25/2015 3.270 2/25/2015 GG24 3.350 0 27 0.38449 2/25/2015 3.270 2/25/2015 GG25 3.350 0 27 0.38464 2/25/2015 3.270 2/25/2015 GG26 3.350 0 27 0.38463 2/25/2015 3.270 2/25/2015 GG27 3.350 0 27 0.38471 2/25/2015 3.270 2/25/2015 GG28 3.350 0 27 0.38440 2/25/2015 3.270 2/25/2015 GG29 3.350 0 13 0.38460 3/11/2015 3.270 3/11/2015 GG30 3.350 0 13 0.38457 3/11/2015 3.270 3/11/2015 GG31 3.350 0 13 0.38448 3/11/2015 3.270 3/11/2015 GG32 3.350 0 13 0.38451 3/11/2015 3.270 3/11/2015 GG33 3.350 0 13 0.38437 3/11/2015 3.270 3/11/2015 GG34 3.350 0 13 0.38465 3/11/2015 3.270 3/11/2015 GG35 3.350 0 13 0.38439 3/11/2015 3.270 3/11/2015 GG36 3.350 0 13 0.38471 3/11/2015 3.270 3/11/2015 GG37 3.350 0 13 0.38460 3/11/2015 3.270 3/11/2015 GG38 3.350 0 13 0.38415 3/11/2015 3.270 3/11/2015 GG39 3.350 0 13 0.38468 3/11/2015 3.270 3/11/2015 GG40 3.350 0 13 0.38446 3/11/2015 3.270 3/11/2015 FC08513, R0 EC 68969 195

tallies_gamm20161017.xlsx Attachment D SpreadSheets FC08513 Gamma Tally Results FCS 14MOg FC08513, R0 EAB Dose CR EAB Dose ~~g ~CW9cQe~ (14mog) Dose(14mog) Error (14mog) Error (14mog) nps mRem/hr EAB error EAB vov EAB slope EAB fom mRem/hr CR error CR vov CR slope CR fom mRem/hr mRem/hr 4.0265318E+08 6.591E-04 0.5942 0.9264 1.7 9.5E-03 9.5316E-05 0.4717 0.3369 1.6 1.5E-02 8.0530637E+08 4.872E-04 0.4395 0.6738 1.8 8.6E-03 2.7896E-04 0.5776 0.5716 1.6 5.0E-03 1.2079596E+09 4.224E-04 0.3470 0.6077 1.7 9.2E-03 1.9971E-04 0.5385 0.5692 1.5 3.8E-03 1.6106127E+09 3.964E-04 0.2858 0.5411 1.7 l.OE-02 1.8288E-04 0.4575 0.4943 1.5 4.0E-03 2.0132659E+09 1.473E-03 0.7517 0.9861 1.7 1.2E-03 1.5396E-04 0.4354 0.4912 1.6 3.6E-03 2.4159191E+09 1.299E-03 0.7111 0.9824 1.8 1.1E-03 1.3851E-04 0.4058 0.4795 1.6 3.5E-03 2.8185723E+09 1.175E-03 0.6738 0.9805 1.7 l.lE-03 1.2464E-04 0.3871 0.4764 1.6 3.3E-03 3.2212255E+09 1.154E-03 0.6040 0.9587 1.7 1.2E-03 1.2634E-04 0.3440 0.4271 1.5 3.7E-03 3.6238787E+09 1.090E-03 0.5698 0.9478 1.8 1.2E-03 1.1738E-04 0.3296 0.4244 1.5 3.6E-03 4.0265318E+09 1.204E-03 0.4944 0.7516 1.8 1.4E-03 1.3338E-04 0.3214 0.3000 1.5 3.4E-03 4.4291850E+09 1.109E-03 0.4880 0.7516 1.8 1.4E-03 1.2901E-04 0.3046 0.2907 1.5 3.5E-03 4.8318382E+09 1.065E-03 0.4662 0.7490 1.8 1.4E-03 1.2521E-04 0.2913 0.2769 1.6 3.5E-03 5.0000000E+09 1.034E-03 0.4639 0.7490 1.8 1.3E-03 1.2267E-04 0.2875 0.2764 1.6 3.5E-03 5.6371446E+09 1.164E-03 0.3876 0.6028 1.7 1.7E-03 1.1937E-04 0.2660 0.2608 1.6 3.6E-03 6.4424509E+09 1.371E-03 0.3440 0.3791 1.7 1.9E-03 1.1075E-04 0.2520 0.2561 1.6 3.5E-03 EC 68969 7.2477573E+09 1.260E-03 0.3329 0.3785 1.7 1.8E-03 1.1823E-04 0.2324 0.1975 1.6 3.7E-03 8.0530637E+09 1.161E-03 0.3251 0.3782 1.7 1.7E-03 1.1826E-04 0.2158 0.1769 1.6 3.9E-03 8.8583700E+09 1.158E-03 0.3029 0.3482 1.7 1.8E-03 1.1429E-04 0.2038 0.1741 1.6 4.0E-03 9.6636764E+09 1.090E-03 0.2951 0.3478 1.7 1.7E-03 1.0969E-04 0.1974 0.1655 1.6 3.9E-03 l.OOOOOOOE+10 1.059E-03 0.2935 0.3478 1.7 1.7E-03 1.0735E-04 0.1950 0.1652 1.6 3.8E-03 1.0416E-03 1.3630E-04 1.64E-03 1.88E-04 196

tallies_gamm20161017.xlsx Attachment D SpreadSheets FC08513 Gamma Tally Results FCS 18MOg FC08513, R0 EAB Dose CR EAB Dose P1u~£ e 2rr6o~11>?~ {18mog) Dose(18mog) Error {18mog) Error {18mog) nps mRem/hr EAB error EAB vov EAB slope EAB fom mRem/hr CR error CR vov CR slope CR fom mRem/hr mRem/hr 4.02653E+08 5.4485E-04 0.5941 0.9262 1.7 6.7E-04 7.8906E-05 0.4716 0.3374 1.6 l.lE-03 8.05306E+08 4.0254E-04 0.4395 0.6738 1.8 6.0E-04 2.3080E-04 0.5777 0.5722 1.6 3.SE-04 1.20796E+09 3.4907E-04 0.3470 0.6077 1.7 6.4E-04 1.6521E-04 0.5386 0.5698 1.5 2.7E-04 1.61061E+09 3.2760E-04 0.2857 0.5411 1.7 7.1E-04 1.5128E-04 0.4576 0.4949 1.5 2.8E-04 2.01327E+09 1.2176E-03 0.7516 0.9861 1.7 8.2E-05 1.2733E-04 0.4356 0.4917 1.5 2.SE-04 2.41592E+09 1.0736E-03 0.7110 0.9824 1.8 7.7E-05 1.1457E-04 0.4059 0.4800 1.6 2.4E-04 2.81857E+09 9.7174E-04 0.6737 0.9805 1.7 7.4E-05 1.0312E-04 0.3872 0.4769 1.6 2.2E-04 3.22123E+09 9.5482E-04 0.6033 0.9588 1.7 8.0E-05 1.0345E-04 0.3475 0.4283 1.5 2.4E-04 3.62388E+09 9.0215E-04 0.5692 0.9478 1.8 8.0E-05 9.6153E-05 0.3328 0.4256 1.5 2.4E-04 4.02653E+09 9.9591E-04 0.4940 0.7519 1.8 9.6E-05 1.0938E-04 0.3237 0.3003 1.5 2.2E-04 4.42919E+09 9.1718E-04 0.4876 0.7519 1.8 9.0E-05 1.0581E-04 0.3066 0.2910 1.5 2.3E-04 4.83184E+09 8.8082E-04 0.4658 0.7493 1.8 9.0E-05 1.0275E-04 0.2932 0.2771 1.6 2.3E-04 5.00000E+09 8.5547E-04 0.4635 0.7493 1.8 8.8E-05 1.0068E-04 0.2893 0.2766 1.6 2.3E-04 5.63714E+09 9.6275E-04 0.3873 0.6029 1.7 l.lE-04 9.8168E-05 0.2674 0.2602 1.6 2.3E-04 6.44245E+09 1.1354E-03 0.3433 0.3792 1.6 1.2E-04 9.1104E-05 0.2532 0.2555 1.6 2.3E-04 EC 68969 7.24776E+09 1.0420E-03 0.3326 0.3785 1.7 1.2E-04 9.7243E-05 0.2335 0.1971 1.6 2.4E-04 8.05306E+09 9.6048E-04 0.3248 0.3783 1.7 l.lE-04 9.7341E-05 0.2168 0.1763 1.6 2.SE-04 8.85837E+09 9.5782E-04 0.3027 0.3484 1.7 1.2E-04 9.4124E-05 0.2046 0.1736 1.6 2.6E-04 9.66368E+09 9.0125E-04 0.2949 0.3480 1.7 l.lE-04 9.0353E-05 0.1981 0.1650 1.6 2.SE-04 1.00000E+10 8.7565E-04 0.2934 0.3480 1.7 l.lE-04 8.8403E-05 0.1958 0.1647 1.6 2.SE-04 8.614E-04 1.1231E-04 1.36E-03 1.55E-04 Desired error Desired vov Desired nps (0.05) {0.10) slope (3 .00) O.OOOOOE+OO 0.05 0.10 3.00 l.OOOOOE+10 0.05 0.10 3.00 197

FC08513, R0 EAB & CR Gamma Dose Rate (mRem/hr) Page 198 of 269 3.0000E-04 1.600E-03 2.SOOOE-04 1.3SOE-03 2.0000E-04 l.lOOE-03

             ..r::.

E l.SOOOE-04 8.SOOE-04 EC 68969 Ql a: E l.OOOOE-04 6.000E-04 S.OOOOE-05 3.SOOE-04 O.OOOOE+OO l .OOOE-04 O.OOE+OO l.OOE+09 2.00E+09 3.00E+09 4.00E+09 S.OOE+09 6.00E+09 7.00E+09 8.00E+09 9.00E+09 l.OOE+lO Total Particles

                             ~ CR    Dose(14mog) mRem/hr         CR Dose(18mog) mRem/hr        ~ EAB        Dose (14mog) mRem/hr     ~   EAB Dose (18mog) mRem/hr 198

FC08513, R0 Relative Error Page 199 of 269 0.70 0.90 0.80 0.60 0.70 0.50 I 0.60 0 6 0.40

             ~                                                                                                                                                           0.50 c

0 EC 68969 u LL 0.40

              .... 0.30 0

w 0.30 0.20 0.20 0.10 0.10 0.00 0.00 O.OE+OO l.OE+09 2.0E+09 3.0E+09 4.0E+09 5.0E+09 6.0E+09 7.0E+09 8.0E+09 9.0E+09 l.OE+10 Total Particles

                                              -+- CRerror             CR error   - - Desired error (0.05)   .,._EAB error             EAB error 199

FC08513, R0 Variance of the Variance Page 200 of 269 1.10 1.00 0.90 0.80 0.70

             ~ 0.60 ci v

EC 68969 0 0.50 0.40 0.30 0.20 0.10 0.00 O.OOE+OO 1.00E+09 2.00E+09 3.00E+09 4.00E+09 5.00E+09 6.00E+09 7.00E+09 8.00E+09 9.00E+09 l.OOE+10 Total Particles

                                              ~ EABvov    ~ CRvov    ~ EABvov                (R VOV   -    Desired         (0.10) 200 VOV

FC08513, R0 PDF Slope Page 201 of 269 3.20 3.00 2.80 2.60 0 q m 2.40 v QJ a. EC 68969 0 Vi u.. 2.20 0 0.. 2.00 1.80 1.60 1.40 O.OOE+OO 1.00E+09 2.00E+09 3.00E+09 4.00E+09 S.OOE+09 6.00E+09 7.00E+09 8.00E+09 9.00E+09 1.00E+10 Total Particles

                                           .,._ EAB slope     .,._ CR slope   _._ EAB slope            CR slope    - - Desired slope (3.00) 201

FC08513, R0 Figure of Merit Page 202 of 269 1.6000E-02 1.4000E-02 1.2000E-02 l .OOOOE-02

             ~    8.0000E-03 EC 68969 LL 6.0000E-03 4.0000E-03 2.0000E-03 O.OOOOE+OO O.OOE+OO   l.OOE+09   2.00E+09      3.00E+09    4.00E+09       S.OOE+09       6.00E+09   7.00E+09   8.00E+09   9.00E+09     l.OOE+lO Total Particles

_._ EAB fom _._ CR fom _._ EAB fom CR fom 202

tallies_neut20131017.xlsx Attachment D SpreadSheets FC08513 Neutron Tally Results Summary of Gamma and Neutron Results Page 203 of 269 14 Month EAB 14 Month CR 18 Month EAB 18 Month CR Neutron Gamma 1....-------' Total FC08513, R0 EC 68969 203

tallies_neut20131017.xlsx Attachment D SpreadSheets FC08513 Neutron Tally Results FCS 14M0n FC08513, R0 EAB Dose CR EAB DcHilQil ~~die lfii nps (14mon) EAB error EAB VOV EAB slope EAB fom Dose(14mon) CR error CR vov CR slope CR fom Plus Error Error (14mon) 3.1457280E+06 5.5506E-06 1.782E-01 2.265E-01 1.80 5.1E-02 1.946E-03 3.880E-01 2.959E-01 1.20 l.lE-02 6.2914560E+06 5.5102E-06 1.345E-01 2.032E-01 2.00 5.1E-02 1.879E-03 2.889E-01 1.686E-01 1.30 l.lE-02 6.2914560E+06 5.5102E-06 1.345E-01 2.032E-01 2.00 5.1E-02 1.879E-03 2.889E-01 1.686E-01 1.30 l.lE-02 9.4371840E+06 5.3999E-06 1.133E-01 1.195E-01 2.20 4.8E-02 1.760E-03 2.381E-01 1.173E-01 1.30 l.lE-02 1.2582912E+07 6.0547E-06 1.287E-01 1.876E-01 2.10 2.8E-02 1.554E-03 2.275E-01 1.117E-01 1.30 8.8E-03 1.5728640E+07 1.1182E-05 2.539E-01 4.204E-01 1.90 5.7E-03 1.834E-03 2.185E-01 1.641E-01 1.30 7.7E-03 1.8874368E+07 1.0251E-05 2.316E-01 4.149E-01 2.00 5.7E-03 1.733E-03 2.030E-01 1.407E-01 1.40 7.4E-03 2.2020096E+07 1.1553E-05 2.057E-01 2.612E-01 1.90 6.2E-03 1.623E-03 1.889E-01 1.321E-01 1.50 7.3E-03 2.5165824E+07 1.1146E-05 1.876E-01 2.556E-01 2.00 6.5E-03 1.555E-03 1.776E-01 1.194E-01 1.60 7.2E-03 2.8311552E+07 1.0529E-05 1.771E-01 2.523E-01 2.00 6.4E-03 1.506E-03 1.679E-01 1.080E-01 1.60 7.2E-03 3.1457280E+07 1.0482E-05 1.629E-01 2.361E-01 2.00 6.7E-03 1.432E-03 1.610E-01 1.031E-01 1.70 6.9E-03 3.4603008E+07 1.0182E-05 1.531E-01 2.316E-01 2.10 6.8E-03 1.378E-03 1.534E-01 9.940E-02 1.80 6.8E-03 3.5000000E+07 1.0176E-05 1.517E-01 2.306E-01 2.10 6.9E-03 1.364E-03 1.532E-01 9.940E-02 1.80 6.7E-03 3.77 48736E+07 1.0632E-05 1.410E-01 1.952E-01 2.10 5.1E-03 1.348E-03 1.467E-01 9.250E-02 1.80 4.8E-03 4.4040192E+07 1.0550E-05 1.259E-01 1.720E-01 2.10 3.6E-03 1.435E-03 1.362E-01 7.620E-02 1.90 3.1E-03 5.0331648E+07 1.0192E-05 1.162E-01 1.599E-01 2.10 2.9E-03 1.425E-03 1.302E-01 6.970E-02 2.00 2.3E-03 5.6623104E+07 1.0035E-05 1.093E-01 1.388E-01 2.10 2.5E-03 1.406E-03 1.216E-01 6.230E-02 2.20 2.0E-03 EC 68969 6.2914560E+07 1.0128E-05 1.010E-01 1.219E-01 2.10 2.4E-03 1.366E-03 1.155E-01 5.700E-02 2.30 1.8E-03 6.9206016E+07 1.0046E-05 9.550E-02 1.099E-01 2.10 2.2E-03 1.547E-03 1.415E-01 2.487E-01 2.30 1.0E-03 7.5497472E+07 1.1073E-05 1.085E-01 1.862E-01 2.00 1.5E-03 1.519E-03 1.335E-01 2.390E-01 2.30 9.9E-04 8.1788928E+07 1.1576E-05 1.098E-01 1.423E-01 2.00 1.3E-03 1.482E-03 1.278E-01 2.281E-01 2.50 9.5E-04 8.8080384E+07 1.1358E-05 1.043E-01 1.397E-01 2.10 1.3E-03 1.456E-03 1.221E-01 2.187E-01 2.60 9.3E-04 9.4371840E+07 1.1220E-05 9.920E-02 1.362E-01 2.30 1.3E-03 1.384E-03 1.200E-01 2.179E-01 2. 70 8.7E-04 l.OOOOOOOE+08 1.0897E-05 9.660E-02 1.354E-01 2.30 1.2E-03 1.4437E-03 1.161E-01 1.751E-01 2.80 8.5E-04 7.2500000E+08 1.5595E-05 1.1580E-01 1.8085E-03 9.0700E-02 1.91E-05 I 2.13E-03 204

tallies_neut20131017.xlsx Attachment D SpreadSheets FC08513 Neutron Tally Results FCS 18M0n tally 15 tally 25 FC08513, R0 EAB Dose CR EAB D~i19t OOodiU~ nps (18m on) EAB error EAB vov EAB slope EAB fom Dose(18mon) CR error CRvov CR slope CR fom Plus Error Error (18mon) 5.0331648E+07 1.0360E-05 1.3880E-01 1.7770E-01 2.10 8.4E-04 1.5929E-03 1.8910E-01 2.8310E-01 2.00 4.5E-04 1.0066330E+08 1.1342E-05 1.1260E-01 2.2500E-01 2.10 6.3E-04 1.4876E-03 1.2080E-01 1.4640E-01 2.70 5.5E-04 1.5099494E+08 1.3747E-05 1.0380E-01 1.0650E-Ol 2.10 5.0E-04 1.6460E-03 9.5500E-02 7.8100E-02 2.60 5.9E-04 2.0132659E+08 1.3273E-05 8.7100E-02 8.7200E-02 2.30 5.3E-04 1.6499E-03 7.9400E-02 5.4600E-02 2.80 6.4E-04 2.5165824E+08 1.3310E-05 7.6100E-02 6.7500E-02 2.70 5.6E-04 1.7010E-03 7.0900E-02 4.2100E-02 3.20 6.4E-04 3.0198989E+08 1.3783E-05 6.9300E-02 5.6900E-02 2.90 5.6E-04 1.7342E-03 6.4000E-02 3.2600E-02 3.50 6.6E-04 3.5232154E+08 1.4124E-05 6.4600E-02 4.3800E-02 2.90 5.5E-04 1.7089E-03 5.8100E-02 2.7800E-02 3.90 6.8E-04 4.0265318E+08 1.7055E-05 1.8880E-01 8.6880E-Ol 2.60 5.6E-05 1.9534E-03 1.4180E-01 7.9060E-01 3.20 1.0E-04 4.5298483E+08 1.7312E-05 1.6790E-01 8.1860E-Ol 2.50 6.3E-05 1.9429E-03 1.2810E-01 7.5680E-01 3.30 l.lE-04 5.0331648E+08 1.7236E-05 1.5290E-01 7.9330E-01 2.50 6.9E-05 1.9028E-03 1.1840E-01 7.3870E-01 3.10 1.1E-04 5.5364813E+08 1.7087E-05 1.4100E-01 7.7640E-01 2.50 7.4E-05 1.8580E-03 1.1060E-01 7.2810E-Ol 3.00 1.2E-04 6.0397978E+08 1.6951E-05 1.3130E-01 7.5200E-Ol 2.60 7.8E-05 1.8157E-03 1.0410E-01 7.1870E-01 3.20 1.2E-04 6.5431142E+08 1.6618E-05 1.2410E-01 7.4120E-01 2.70 8.0E-05 1.7817E-03 9.8300E-02 7.0950E-01 3.10 1.3E-04 7.0464307E+08 1.6364E-05 1.1770E-01 7.2460E-01 2.80 8.3E-05 1.7739E-03 9.3400E-02 6.5960E-01 3.00 1.3E-04 7 .2500000E+08 1.6178E-05 1.1580E-01 7.2350E-01 2.80 8.3E-05 1.7783E-03 9.0700E-02 6.5310E-01 3.00 1.4E-04 1.4983E-05 1.7551E-03 1.84E-05 I 2.07E-03 EC 68969 Desired error Desired vov Desired slope nps (0.05) (0.10) (3.00) O.OOOOOOOE+OO 0.05 0.10 3.00 1.0000000E+09 0.05 0.10 3.00 Ratio of 14 mo to 18 mo CR EAB Dose Dose(14mon nps (14mon) ) 1.00000E+08 1.0897E-05 1.0408E+OO 1.444E-03 1.0304E+OO pseudo 7.25E+08 1.559SE-OS l.BOSSE-03 CR EAB Dose Dose(18mon nps (18mon) ) 1.00663E+08 1.1342E-05 1.4876E-03 7.25000E+08 1.4983E-05 1.7551E-03 205

FC08513, R0 EAB & CR Neutron Dose Rate (mRem/hr) Page 206 of 269 2.500E-03 2.500E-05 2.250E-03 2.250E-05 2.000E-03 2.000E-05 1.750E-03 1.750E-05

             ..c.

E 1.500E-03 1.500E-05 EC 68969 Q) 0::: E 1.250E-03 1.250E-05 l .OOOE-03 l.OOOE-05 7.500E-04 7.500E-06 5.000E-04 5.000E-06 O.OOE+OO l .OOE+08 2.00E+08 3.00E+08 4.00E+08 5.00E+08 6.00E+08 7.00E+08 8.00E+08 Total Particles

                                              ~ CR   Dose(14mon)     CR Dose(18mon)     ~ EAB       Dose (14mon)    - . - EAB Dose (18mon) 206

FC08513, R0 Relative Error Page 207 of 269 0.45 0.40 0.35 0.30 Lil 0 c:i

!... 0.25 c::

0 EC 68969

             *~

u ro u.. 0.20 0 LU 0.15 0.10 0.05 0.00 O.OOE+OO 1.00E+08 2.00E+08 3.00E+08 4.00E+08 5.00E+08 6.00E+08 7.00E+08 8.00E+08 Total Particles 207

                                             -   EAB error   -   CR error     -     EAB error          CR error   - - Desired error (0.05)

FC08513, R0 Variance of the Variance Page 208 of 269 1.00 0.90 0.80 0.70 I 0.60 0

             .-I ci v 0.50 EC 68969 0

0.40 0.30 0.20 0.10 0.00 O.OE+OO 1.0E+08 2.0E+08 3.0E+08 4.0E+08 S.OE+08 6.0E+08 7.0E+08 8.0E+08 Total Particles

                                          ~ EABvov   ~ CRvov      ~ EABvov              CR vov    - - Desired vov (0.10) 208

FC08513, R0 PDF Slope Page 209 of 269 4.20 3.80 3.40 3.00 0 q m v

             ~ 2.60 EC 68969 0

Vi u.. 0 a.. 2.20 1.40 1.00 O.OOE+OO l .OOE+08 2.00E+08 3.00E+08 4.00E+08 5.00E+08 6.00E+08 7.00E+08 8.00E+08 Total Particles

                                            ~ EAB   slope   ~ CRslope   ~ EAB   slope          CR slope   - - Desired slope (3.00) 209

FC08513, R0 Figure of Merit Page 210 of 269 2.00E-02 l.SOE-02

             ~    l .OOE-02 EC 68969 LL S.OOE-03 O.OOE+OO O.OOE+OO   l.OOE+08   2.00E+08          3.00E+08         4.00E+08         S.OOE+08     6.00E+08   7.00E+08     8.00E+08 Total Particles
                                                  ~ EAB   fom     ~ CR     fom   ~ EAB      fom       CR fom 210

tallies_neut20161017.xlsx Attachment D SpreadSheets FC08513 Neutron Tally Results Summary of Gamma and Neutron Results Page 211 of 269 14 Month EAB 14 Month CR 18 Month EAB 18 Month CR Neutron Gamma L . . . - - - - - - . . . . 1 Total FC08513, R0 EC 68969 211

tallies_neut20161017 .xlsx Attachment D SpreadSheets FC08513 Neutron Tally Results FCS 14MOn FC08513, R0 EAB Dose CR EAB DcH8gt ~ 0~~ i~ nps (14mon) EAB error EAB vov EAB slope EAB fom Dose(14mon) CR error CR vov CR slope CR fom Plus Error Error (14mon) 3.1457280E+06 5.5506E-06 1.782E-01 2.265E-01 1.80 5.1E-02 1.946E-03 3.880E-01 2.959E-01 1.20 l.lE-02 6.2914560E+06 5.5102E-06 1.345E-01 2.032E-01 2.00 5.1E-02 1.879E-03 2.889E-01 1.686E-01 1.30 l.lE-02 6.2914560E+06 5.5102E-06 1.345E-01 2.032E-01 2.00 5.1E-02 1.879E-03 2.889E-01 1.686E-01 1.30 l.lE-02 9.4371840E+06 5.3999E-06 1.133E-01 1.195E-01 2.20 4.8E-02 1.760E-03 2.381E-01 1.173E-01 1.30 l.lE-02 1.2582912E+07 6.0547E-06 1.287E-01 1.876E-01 2.10 2.8E-02 1.554E-03 2.275E-01 1.117E-01 1.30 8.8E-03 1.5728640E+07 1.1182E-05 2.539E-01 4.204E-01 1.90 5.7E-03 1.834E-03 2.185E-01 1.641E-01 1.30 7.7E-03 1.8874368E+07 1.0251E-05 2.316E-01 4.149E-01 2.00 5.7E-03 1.733E-03 2.030E-01 1.407E-01 1.40 7.4E-03 2.2020096E+07 1.1553E-05 2.057E-01 2.612E-01 1.90 6.2E-03 1.623E-03 1.889E-01 1.321E-01 1.50 7.3E-03 2.5165824E+07 1.1146E-05 1.876E-01 2.556E-01 2.00 6.SE-03 1.555E-03 1.776E-01 1.194E-01 1.60 7.2E-03 2.8311552E+07 1.0529E-05 1.771E-01 2.523E-01 2.00 6.4E-03 1.506E-03 1.679E-01 1.080E-01 1.60 7.2E-03 3.1457280E+07 1.0482E-05 1.629E-01 2.361E-01 2.00 6.7E-03 1.432E-03 1.610E-01 1.031E-01 1.70 6.9E-03 3.4603008E+07 1.0182E-05 1.531E-01 2.316E-01 2.10 6.8E-03 1.378E-03 1.534E-01 9.940E-02 1.80 6.8E-03 3.5000000E+07 1.0176E-05 1.517E-01 2.306E-01 2.10 6.9E-03 1.364E-03 1.532E-01 9.940E-02 1.80 6.7E-03 3.7748736E+07 1.0632E-05 1.410E-01 1.952E-01 2.10 5.1E-03 1.348E-03 1.467E-01 9.250E-02 1.80 4.8E-03 4.4040192E+07 1.0550E-05 1.259E-01 1.720E-01 2.10 3.6E-03 1.435E-03 1.362E-01 7.620E-02 1.90 3.1E-03 5.0331648E+07 1.0192E-05 1.162E-01 1.599E-01 2.10 2.9E-03 1.425E-03 1.302E-01 6.970E-02 2.00 2.3E-03 5.6623104E+07 1.0035E-05 1.093E-01 1.388E-01 2.10 2.5E-03 1.406E-03 1.216E-01 6.230E-02 2.20 2.0E-03 EC 68969 6.2914560E+07 1.0128E-05 1.010E-01 1.219E-01 2.10 2.4E-03 1.366E-03 1.155E-01 5.700E-02 2.30 1.8E-03 6.9206016E+07 1.0046E-05 9.550E-02 1.099E-01 2.10 2.2E-03 1.547E-03 1.415E-01 2.487E-01 2.30 1.0E-03 7.5497472E+07 1.1073E-05 1.085E-01 1.862E-01 2.00 1.5E-03 1.519E-03 1.335E-01 2.390E-01 2.30 9.9E-04 8.1788928E+07 1.1576E-05 1.098E-01 1.423E-01 2.00 1.3E-03 1.482E-03 1.278E-01 2.281E-01 2.50 9.5E-04 8.8080384E+07 1.1358E-05 1.043E-01 1.397E-01 2.10 1.3E-03 1.456E-03 1.221E-01 2.187E-01 2.60 9.3E-04 9.4371840E+07 1.1220E-05 9.920E-02 1.362E-01 2.30 1.3E-03 1.384E-03 1.200E-01 2.179E-01 2.70 8.7E-04 1.0000000E+08 1.0897E-05 9.660E-02 1.354E-01 2.30 1.2E-03 1.4437E-03 1.161E-01 1.751E-01 2.80 8.5E-04 7.2500000E+08 1.5595E-05 1.1580E-01 1.8085E-03 9.0700E-02 1.91E-05 I 2.13E-03 212

tallies_neut20161017.xlsx Attachment D SpreadSheets FC08513 Neutron Tally Results FCS 18M0n tally 15 tally 25 FC08513, R0 EAB Dose CR EAB D<H8gt 1Z11 OdiU~ nps (18m on) EAB error EAB VOV EAB slope EAB fom Dose(18mon) CR error CR vov CR slope CR fom Plus Error 1 Error (18mon) 5.0331648E+07 1.0360E-05 1.3880E-01 1.7770E-01 2.10 8.4E-04 1.5929E-03 1.8910E-01 2.8310E-01 2.00 4.5E-04 1.0066330E+08 1.1342E-05 1.1260E-01 2.2500E-01 2.10 6.3E-04 1.4876E-03 1.2080E-01 1.4640E-01 2.70 5.5E-04 1.5099494E+08 1.3747E-05 1.0380E-01 1.0650E-01 2.10 5.0E-04 1.6460E-03 9.5500E-02 7.8100E-02 2.60 5.9E-04 2.0132659E+08 1.3273E-05 8.7100E-02 8.7200E-02 2.30 5.3E-04 1.6499E-03 7.9400E-02 5.4600E-02 2.80 6.4E-04 2.5165824E+08 1.3310E-05 7.6100E-02 6.7500E-02 2.70 5.6E-04 1.7010E-03 7.0900E-02 4.2100E-02 3.20 6.4E-04 3.0198989E+08 1.3783E-05 6.9300E-02 5.6900E-02 2.90 5.6E-04 1.7342E-03 6.4000E-02 3.2600E-02 3.50 6.6E-04 3.5232154E+08 1.4124E-05 6.4600E-02 4.3800E-02 2.90 5.5E-04 1.7089E-03 5.8100E-02 2.7800E-02 3.90 6.8E-04 4.0265318E+08 1.7055E-05 1.8880E-01 8.6880E-01 2.60 5.6E-05 1.9534E-03 1.4180E-01 7.9060E-01 3.20 1.0E-04 4.5298483E+08 1.7312E-05 1.6790E-01 8.1860E-01 2.50 6.3E-05 1.9429E-03 1.2810E-01 7.5680E-01 3.30 l.lE-04 5.0331648E+08 1.7236E-05 1.5290E-01 7.9330E-01 2.50 6.9E-05 1.9028E-03 1.1840E-01 7.3870E-01 3.10 1.1E-04 5.5364813E+08 1.7087E-05 1.4100E-01 7.7640E-01 2.50 7.4E-05 1.8580E-03 1.1060E-01 7.2810E-01 3.00 1.2E-04 6.0397978E+08 1.6951E-05 1.3130E-01 7.5200E-01 2.60 7.8E-05 1.8157E-03 1.0410E-01 7.1870E-01 3.20 1.2E-04 6.5431142E+08 1.6618E-05 1.2410E-01 7.4120E-01 2.70 B.OE-05 1.7817E-03 9.8300E-02 7.0950E-01 3.10 1.3E-04 7.0464307E+08 1.6364E-05 1.1770E-01 7.2460E-01 2.80 8.3E-05 1.7739E-03 9.3400E-02 6.5960E-01 3.00 1.3E-04 7 .2500000E+08 1.6178E-05 1.1580E-01 7.2350E-01 2.80 8.3E-05 1.7783E-03 9.0700E-02 6.5310E-01 3.00 1.4E-04 1.4983E-05 1.7551E-03 1.84E-05 I 2.07E-03 EC 68969 Desired error Desired vov Desired slope nps (0.05) (0.10) (3.00) O.OOOOOOOE +00 0.05 0.10 3.00 l.OOOOOOOE+09 0.05 0.10 3.00 Ratio of 14 mo to 18 mo CR EAB Dose Dose(14mon nps (14mon) ) l.OOOOOE+08 1.0897E-05 1.0408E+OO 1.444E-03 1.0304E+OO pseudo 7.25E+08 1.559SE-05 1.8085E-03 CR EAB Dose Dose(18mon nps (18mon) ) 1.00663E+08 1.1342E-05 1.4876E-03 7.25000E+08 1.4983E-05 1.7551E-03 213

FC08513, R0 EAB & CR Neutron Dose Rate (mRem/hr) Page 214 of 269 2.500E-03 2.500E-05 2.250E-03 2.250E-05 2.000E-03 2.000E-05 I 1.750E-03 1.750E-05

             ..c.

E 1.500E-03 1.500E-05 EC 68969

              <IJ a::

E 1.250E-03 1.250E-05 l.OOOE-03 l .OOOE-05 7.500E-04 7.500E-06 5.000E 5.000E-06 O.OOE+OO l.OOE+OS 2.00E+08 3.00E+08 4.00E+08 5.00E+08 6.00E+08 7.00E+08 8.00E+08 Total Particles

                                            -+- CR Dose(14mon)      CR Dose(18mon)     -+- EAB Dose (14mon)    -+- EAB Dose (18m on) 214

FC08513, R0 Relative Error Page 215 of 269 0.45 0.40 0.35 0.30

             ;n 0

0

!... 0.25 c:

0 EC 68969 u ro u.. 0.20 0 w 0.15 0.10 0.05 0.00 O.OOE+OO 1.00E+08 2.00E+08 3.00E+08 4 .00E+08 5.00E+08 6.00E+08 7.00E+08 8.00E+08 Total Particles 215

                                            -+- EAB error   -+- CR error      -+- EAB error           CR error   - - Desired error (0.05)

FC08513, R0 Variance of the Variance Page 216 of 269 1.00 0.90 0.80 0.70 0.60 0

             ~

c:i v 0.50 EC 68969 0 0.40 0.30 0.20 0.10 O.OE+OO l.OE+08 2.0E+08 3.0E+08 4.0E+08 S.OE+08 6.0E+08 7.0E+08 8.0E+08 Total Particles

                                         ......_. EAB vov   ......_. CR vov   ......_. EAB vov          CR vov    - - Desired vov (0.10) 216

FC08513, R0 PDF Slope Page 217 of 269 4.20 3.80 3.40 3.00 v

              ~ 2.60 EC 68969 0

Vi Ll.. 0 a.. 2.20 1.80 1.40 1.00 O.OOE+OO l.OOE+08 2.00E+08 3.00E+08 4.00E+08 S.OOE+08 6.00E+08 7.00E+08 8.00E+08 Total Particles 217 - EAB slope - CR slope - EAB slope CR slope - - Desired slope (3.00)

FC08513, R0 Figure of Merit Page 218 of 269 2.00E-02 l .SOE-02

             ~     l.OOE-02 EC 68969 u..

S.OOE-03 O.OOE+OO O.OOE+OO l.OOE+08 2.00E+08 3.00E+08 4.00E+08 5.00E+08 6.00E+08 7.00E+08 8.00E+08 Total Particles

                                                   -+- EAB fom   -+- CR fom        EAB fom       CR fom 218

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 219 of 269 Attachment E: Additional References Attachment E1 Reference 5: Pages 107,108,240,241 , 281 , 287-290,320,321,350 & 351. Attachment E2 Reference 10, All Pages. Attachment E3 Reference 18, Page 941 . Attachment E4 DOW CHEMICAL CO TRYMER 9501-4. Attachment E5 Hall1 05a.txt, (input file), Hall1 05ao.txt (output summary), Hall1 05a.sum5 (MCNP5 test case summary) Hall1 05a.sum6 (MCNP6.1 test case summary) FC08513, R0 EC 68969 219

Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Page 220 )f 269 Mg 0.032649 AI 0.010830 Si 0.034479 K 0.001138 Ca 0.321287 Fe 0.007784 matname Concrete, Oak Ridge (ORNL) density 2.300000 Comments and References Data from Petrie et al. (2000). Weight fractions are adjusted so they sum to unity. Also listed as ORNL concrete in Table 1 of Carter (1978), with reference to Maerker and Muckenthaler (1966). 95 Concrete, Ordinary (NBS 03) Formula= - Molecular weight (g/mole) = - Density (g/cm3) = 2.350000 Total atom density (atoms/b-ern)= 7.950E-02 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data were calculated from the input weight fractions . Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Dens it~ H 1001 1000 0.008485 0.149867 0.011914 c 6000 6000 0.050064 0.074204 0.005899 0 8016 8000 0.473483 0.526832 0.041881 Mg 12000 12000 0.024183 0.017713 0.001408 AI 13027 13000 0.036063 0.023794 0.001892 Si 14000 14000 0.145100 0.091972 0.007311 s 16000 16000 0.002970 0.001649 0.000131 K 19000 19000 0.001697 0.000773 0.000061 Ca 20000 20000 0.246924 0.109680 0.008719 Fe 26000 26000 0.011031 0.003516 0.000280 Total 1.000000 1.000000 0.079496 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 1001 -0.008485 1001 0.149867 1001 0.011914 6000 -0.050064 6000 0.074204 6000 0.005899 8016 -0.473483 8016 0.526832 8016 0.041881 12000 -0.024183 12000 0.017713 12000 0.001408 13027 -0.036063 13027 0.023794 13027 0.001892 14000 -0.145100 14000 0.091972 14000 0.007311 16000 -0.002970 16000 0.001649 16000 0.000131 19000 -0.001697 19000 0.000773 19000 0.000061 20000 -0.246924 20000 0.109680 20000 0.008719 26000 -0.011031 26000 0.003516 26000 0.000280 Page 107 of357 FC08513, R0 EC 68969 220

Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Pa~e 221 pf 269 Photons 1000 -0.008485 1000 0.149867 1000 0.01 914 6000 -0.050064 6000 0.074204 6000 0.005899 8000 -0.473483 8000 0.526832 8000 0.041881 12000 -0.024183 12000 0.017713 12000 0.001408 13000 -0.036063 13000 0.023794 13000 0.001892 14000 -0.145100 14000 0.091972 14000 0.007311 16000 -0.002970 16000 0.001649 16000 0.000131 19000 -0.001697 19000 0.000773 19000 0.000061 20000 -0.246924 20000 0.109680 20000 0.008719 26000 -0.011031 26000 0.003516 26000 0.000280 CEPXS Form : material H 0.008485 c 0.050064 0 0.473483 Mg 0.024183 AI 0.036063 Si 0.145100 s 0.002970 K 0.001697 Ca 0.246924 Fe 0.011031 matname Concrete, Ordinary (NBS 03) density 2.350000 Comments and References Density= 2.35 g/cm3, and weight fractions calculated from partial densities (g/cm3) listed for each element in Table 8.8 of Shultis and Faw (1996), and extracted from ANSI/ANS-6.4-1985. 96 Concrete, Ordinary (NBS 04) Formula= Molecular weight (g/mole) = Density (g/cm3) = 2.350000 Total atom density (atoms/b-ern)= 7.533E-02 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data were calculated from the input weight fractions. Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Dens it~ H 1001 1000 0.005558 0.103586 0.007804 0 8016 8000 0.498076 0.584810 0.044057 Na 11023 11000 0.017101 0.013974 0.001053 Mg 12000 12000 0.002565 0.001983 0.000149 AI 13027 13000 0.045746 0.031850 0.002399 Si 14000 14000 0.315092 0.210755 0.015877 s 16000 16000 0.001283 0.000751 0.000057 K 19000 19000 0.019239 0.009244 0.000696 Ca 20000 20000 0.082941 0.038877 0.002929 Fe 26000 26000 0.012398 0.004171 0.000314 Page 108 of357 FC08513, R0 EC 68969 221

Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Page 222 pf 269 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 1001 -0.143716 1001 0.666662 1001 0.079855 6000 -0.856284 6000 0.333338 6000 0.039929 Photons 1000 -0.143716 1000 0.666662 1000 0.079855 6000 -0.856284 6000 0.333338 6000 0.039929 CEPXS Form : material H 0.143716 c 0.856284 matname Polyethylene, Non-borated density 0.930000 Comments and References Density= 0.93 g/cm3 and weight fractions from http://physics.nist.gov/PhysRefData/XrayMassCoef/tab2.html (NIST 1996). High density polyethylene (HOPE) is 0.944 to 0.965 g/cm3 (http://www.bpf.co.uk/Piastipedia/Polymers/HDPE.aspx). Low density polyethylene (LDPE) is 0.917 to 0.930 g/cm3 (http://www.bpf.co.uk/Piastipedia/Polymers/LDPE.aspx). Automation Creations (2010) at http://www.matweb.com/search/QuickText.aspx has molded HOPE= 0.918-1 .05g/cm3 and MOPE= 0.926-0.95. Density= 0.94 g/cm3 at http://physics.nist.gov/cgi-bin/Star/compos.pl?matno=221 (NIST 1998). Density= 0.92 g/cm3 on pg 138 of Brewer (2009). The range of density values is discussed further at http ://en.wikiped ia. org/wiki/Polyethylene. 249 Polyisocyanurate (PIR) Formula= C15H10N202 Molecular weight (g/mole) = 250.2521 Density (g/cm3) = 0.048200 Total atom density (atoms/b-ern) = 3.364E-03 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data was calculated from the input formula. Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Dens it~ H 1001 1000 0.040277 0.344828 0.001160 c 6000 6000 0.719916 0.517241 0.001740 N 7014 7000 0.111941 0.068966 0.000232 0 8016 8000 0.127866 0.068966 0.000232 Total 1.000000 1.000000 0.003364 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 1001 -0.040277 1001 0.344828 1001 0.001160 6000 -0.719916 6000 0.517241 6000 0.001740 7014 -0.111941 7014 0.068966 7014 0.000232 8016 -0.127866 8016 0.068966 8016 0.000232 Photons 1000 -0.040277 1000 0.344828 1000 0.001160 6000 -0.719916 6000 0.517241 6000 0.001740 Page 240 of 357 FC08513, R0 EC 68969 222

Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. I Page 223 )f 269 7000 -0.111941 7000 0.068966 7000 O.OOU232 8000 -0.127866 8000 0.068966 8000 0.000232 CEPXS Form : material H 0.040277 c 0.719916 N 0.111941 0 0.127866 matname Polyisocyanurate (PIR) density 0.048200 Comments and References Called PIR, polyiso, ISO, or isocyanurate (http://en.wikipedia.org/wiki/Polyisocyanurate). Formula from http://webbook. nist.gov/cgi/cbook.cgi?Name=Polyisocyanurate&U nits=SI&Units=S l&cTG=1 &cTC= 1&c TP= 1&cTR=1 &cPI=1 . Density range = 0.0264 to 0.096 g/cm3 at http://www.fpcfoam.com/polyiso-tech.html. Density= 0.0264, 0.0288, 0.048, 0.064, and 0.096 g/cm3 at http://www.fpcfoam .com/polyiso-tech.html. Density range= 0.033 to 0.32 g/cm3 at www.kingspantarec.com/en/pdf/tarecpir_datasheet.pdf. Density= 0.0482 g/cm3 for nominal 3.0 lb/ft3 density on ISC-C1 datasheet available from http://www.dyplastproducts.com/ISOC1_polyisocyanurate_insulation.htm . Nominal densities are available at 2.0, 2.5, 3, 4, 6, and 10 lb/ft3. 250 Polypropylene (PP) Formula= C3H6 Molecular weight (g/mole) = 42.07974 Density (g/cm3) = 0.900000 Total atom density (atoms/b-cm) = 1.159E-01 The above density is estimated to be accurate to 2 significant digits. Uncertainties are not addressed. The following data were calculated from the input weight fractions. Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Densit~ H 1001 1000 0.143711 0.666653 0.077277 c 6000 6000 0.856289 0.333347 0.038641 Total 1.000000 1.000000 0.115917 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 1001 -0.143711 1001 0.666653 1001 0.077277 6000 -0.856289 6000 0.333347 6000 0.038641 Photons 1000 -0.143711 1000 0.666653 1000 0.077277 6000 -0.856289 6000 0.333347 6000 0.038641 CEPXS Form : material H 0.143711 c 0.856289 matname Polypropylene (PP) density 0.900000 Page 241 of357 FC08513, R0 EC 68969 223

Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Page 224 pf 269 p 0.000228 s 0.000149 Cr 0.188100 Mn 0.009900 Fe 0.694713 Ni 0.091575 matname Steel, Boron Stainless density 7.870000 Comments and References 1.0 wt% boron in the 304 stainless steel specified below. Density from pg II .F.1-2 of Carteret al. (1968). 294 Steel, Carbon Formula= - Molecular weight (g/mole) = - Density (g/cm3) = 7.820000 Total atom density (atoms/b-ern)= 8.587E-02 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data were calculated from the input weight fractions. Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Density c 6000 6000 0.005000 0.022831 0.001960 Fe 26000 26000 0.995000 0.977169 0.083907 Total 1.000000 1.000000 0.085867 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 6000 -0.005000 6000 0.022831 6000 0.001960 26000 -0.995000 26000 0.977169 26000 0.083907 Photons 6000 -0.005000 6000 0.022831 6000 0.001960 26000 -0.995000 26000 0.977169 26000 0.083907 CEPXS Form : material c 0.005000 Fe 0.995000 matname Steel, Carbon density 7.820000 Comments and References See Brewer (2009). Page 281 of357 FC08513, R0 EC 68969 224

Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Page 225 of 269 299 Steel, Stainless 304L Formula= - Molecular weight (g/mole) = - Density (g/cm3) = 8.000000 Total atom density (atoms/b-ern)= 8.758E-02 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data were calculated from the input weight fractions. Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Dens it~ c 6000 6000 0.000150 0.000687 0.000060 Si 14000 14000 0.005000 0.009793 0.000858 p 15031 15000 0.000230 0.000408 0.000036 s 16000 16000 0.000150 0.000257 0.000023 Cr 24000 24000 0.190000 0.201015 0.017605 Mn 25055 25000 0.010000 0.010013 0.000877 Fe 26000 26000 0.694480 0.684101 0.059912 Ni 28000 28000 0.100000 0.093725 0.008208 Total 1.000010 1.000000 0.087578 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 6000 -0.000150 6000 0.000687 6000 0.000060 14000 -0.005000 14000 0.009793 14000 0.000858 15031 -0.000230 15031 0.000408 15031 0.000036 16000 -0.000150 16000 0.000257 16000 0.000023 24000 -0.190000 24000 0.201015 24000 0.017605 25055 -0.010000 25055 0.010013 25055 0.000877 26000 -0.694480 26000 0.684101 26000 0.059912 28000 -0.100000 28000 0.093725 28000 0.008208 Photons 6000 -0.000150 6000 0.000687 6000 0.000060 14000 -0.005000 14000 0.009793 14000 0.000858 15000 -0.000230 15000 0.000408 15000 0.000036 16000 -0.000150 16000 0.000257 16000 0.000023 24000 -0.190000 24000 0.201015 24000 0.017605 25000 -0.010000 25000 0.010013 25000 0.000877 26000 -0.694480 26000 0.684101 26000 0.059912 28000 -0.100000 28000 0.093725 28000 0.008208 CEPXS Form : material c 0.000150 Si 0.005000 p 0.000230 s 0.000150 Cr 0.190000 Mn 0.010000 Fe 0.694480 Ni 0.100000 Page 287 of 357 FC08513, R0 EC 68969 225

Attachment E 1 FC08513 PIET-43741-TM-963 P~-15870Rev. 1 Page 226 )f 269 matname Steel, Stainless 304L density 8.000000 Comments and References Density= 8.00 g/cm3 and weight fractions from http://www. matweb.com/search/DataSheet.aspx?MatGUID=e2147b8f727343b0b0d51 efe02a6127e (Automation Creations 201 0). Weight fractions for Cr and Ni set at the average of the allowed range. Weight fractions for C, Si, P, S, and Mn assumed to be 50% of their upper limits. Weight fraction of Fe set so the total sums to unity. 300 Steel, Stainless 316 Formula= - Molecular weight (g/mole) = - Density (g/cm3) = 8.000000 Total atom density (atoms/b-ern)= 8.655E-02 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data were calculated from the input weight fractions. Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Densit:t c 6000 6000 0.000410 0.001900 0.000164 Si 14000 14000 0.005070 0.010048 0.000870 p 15031 15000 0.000230 0.000413 0.000036 s 16000 16000 0.000150 0.000260 0.000023 Cr 24000 24000 0.170000 0.181986 0.015751 Mn 25055 25000 0.010140 0.010274 0.000889 Fe 26000 26000 0.669000 0.666811 0.057714 Ni 28000 28000 0.120000 0.113803 0.009850 Mo 42000 42000 0.025000 0.014504 0.001255 Total 1.000000 1.000000 0.086553 MCNP Form I Weight Fractions 1 Atom Fractions I Atom Densities Neutrons 6000 -0.000410 6000 0.001900 6000 0.000164 14000 -0.005070 14000 0.010048 14000 0.000870 15031 -0.000230 15031 0.000413 15031 0.000036 16000 -0.000150 16000 0.000260 16000 0.000023 24000 -0.170000 24000 0.181986 24000 0.015751 25055 -0.010140 25055 0.010274 25055 0.000889 26000 -0.669000 26000 0.666811 26000 0.057714 28000 -0.120000 28000 0.113803 28000 0.009850 42000 -0.025000 42000 0.014504 42000 0.001255 Photons 6000 -0.000410 6000 0.001900 6000 0.000164 14000 -0.005070 14000 0.010048 14000 0.000870 15000 -0.000230 15000 0.000413 15000 0.000036 16000 -0.000150 16000 0.000260 16000 0.000023 24000 -0.170000 24000 0.181986 24000 0.015751 25000 -0.010140 25000 0.010274 25000 0.000889 26000 -0.669000 26000 0.666811 26000 0.057714 Page 288 of 357 FC08513, R0 EC 68969 226

Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Pa_ge 227 )f 269 28000 -0.120000 28000 0.113803 28000 0.009850 42000 -0.025000 42000 0.014504 42000 0.001255 CEPXS Form : material c 0.000410 Si 0.005070 p 0.000230 s 0.000150 Cr 0.170000 Mn 0.010140 Fe 0.669000 Ni 0.120000 Mo 0.025000 matname Steel, Stainless 316 density 8.000000 Comments and References Density= 8.00 g/cm3 and weight fractions from http://www.matweb.com/search/DataSheet.aspx?MatGUI D=50f320bd 1daf4fa 7965448c30d3114ad&ckck= 1 (Automation Creations 2010). Density= 8.03 g/cm3 and same weight fractions at http://www.espi-metals.com/technicaldata.htm . Same weight fractions at http://www.engineersedge.com/stainless_steel.htm . Similar to Petrie et al. (2000). Density= 8.027 g/cm3 at http://www.upmet.com/304-physical.shtml. Weight fractions for Cr, Fe, Ni, and Mo set at the average of the allowed range. Weight fractions for C, Si, P, S, and Mn set at 50.7% of their upper limits to allow the total to sum to unity. 301 Steel, Stainless 316L Formula = Molecular weight (g/mole) = Density (g/cm3) = 8.000000 Total atom density (atoms/b-ern)= 8.698E-02 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data were calculated from the input weight fractions. Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Dens it~ c 6000 6000 0.000300 0.001384 0.000120 Si 14000 14000 0.010000 0.019722 0.001715 p 15031 15000 0.000450 0.000805 0.000070 s 16000 16000 0.000300 0.000518 0.000045 Cr 24000 24000 0.170000 0.181098 0.015751 Mn 25055 25000 0.020000 0.020165 0.001754 Fe 26000 26000 0.653950 0.648628 0.056416 Ni 28000 28000 0.120000 0.113247 0.009850 Mo 42000 42000 0.025000 0.014434 0.001255 Total 1.000000 1.000000 0.086977 Page 289 of 357 FC08513, R0 EC 68969 227

Attachment E1 FC08513 PIET-43 741-TM-963 PNNL-15870 Rev. I D<:>no??R f269 MCNP Form I Weight Fractions I Atom Fractions I Atom Densitiesv Neutrons 6000 -0.000300 6000 0.001384 6000 0.000120 14000 -0.010000 14000 0.019722 14000 0.001715 15031 -0.000450 15031 0.000805 15031 0.000070 16000 -0.000300 16000 0.000518 16000 0.000045 24000 -0.170000 24000 0.181098 24000 0.015751 25055 -0.020000 25055 0.020165 25055 0.001754 26000 -0.653950 26000 0.648628 26000 0.056416 28000 -0.120000 28000 0.113247 28000 0.009850 42000 -0.025000 42000 0.014434 42000 0.001255 Photons 6000 -0.000300 6000 0.001384 6000 0.000120 14000 -0.010000 14000 0.019722 14000 0.001715 15000 -0.000450 15000 0.000805 15000 0.000070 16000 -0.000300 16000 0.000518 16000 0.000045 24000 -0.170000 24000 0.181098 24000 0.015751 25000 -0.020000 25000 0.020165 25000 0.001754 26000 -0.653950 26000 0.648628 26000 0.056416 28000 -0.120000 28000 0.113247 28000 0.009850 42000 -0.025000 42000 0.014434 42000 0.001255 CEPXS Form: material c 0.000300 Si 0.010000 p 0.000450 s 0.000300 Cr 0.170000 Mn 0.020000 Fe 0.653950 Ni 0.120000 Mo 0.025000 matname Steel, Stainless 316L density 8.000000 Comments and References Density= 8.00 g/cm3 and weight fractions from http://www.matweb.com/search/DataSheet.aspx?MatGUID=530144e2752b47709a58ca8fe0849969 (Automation Creations 201 0). Fe calculated so the elements sum to unity. Weight fractions for all elements set at specified value, except weight fraction for Fe increased by 0.00395 so weight fractions sum to unity. 302 Steel, Stainless 321 Formula = Molecular weight (g/mole) = Density (g/cm3) = 8.000000 Total atom density (atoms/b-ern)= 8.816E-02 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data were calculated from the input weight fractions. Page 290 of 357 FC08513, R0 EC 68969 228

Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Page 22S of269 U-236 92236 92000 0.000125 0.000046 0.000004 U-238 92238 92000 0.880691 0.323072 0.025131 Total 1.000000 1.000000 0.077788 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 6000 -0.091692 6000 0.666667 6000 0.051859 92234 -0.000243 92234 0.000090 92234 0.000007 92235 -0.027249 92235 0.010124 92235 0.000788 92236 -0.000125 92236 0.000046 92236 0.000004 92238 -0.880691 92238 0.323072 92238 0.025131 Photons 6000 -0.091692 6000 0.666667 6000 0.051859 92000 -0.000243 92000 0.000090 92000 0.000007 92000 -0.027249 92000 0.010124 92000 0.000788 92000 -0.000125 92000 0.000046 92000 0.000004 92000 -0.880691 92000 0.323072 92000 0.025131 CEPXS Form: material c 0.091692 U-234 0.000243 U-235 0.027249 U-236 0.000125 U-238 0.880691 matname Uranium Dicarbide density 11.280000 Comments and References Density from http://physics.nist.gov/cgi-bin/Star/compos.pl?matno=270 (NIST 1998). Formula from pgs 4 - 97 of Lide (2008). Uranium isotopics assumed for LEU: Wt% U234/235/236/238 = 0.0267/3.0/0.0138/96.9595. 334 Uranium Dioxide Formula= U02 Molecular weight (g/mole) = 269.9568909 Density (g/cm3) = 10.960000 Total atom density (atoms/b-ern)= 7.335E-02 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data was calculated from the input formula. Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Dens it~ 0 8016 8000 0.118533 0.666667 0.048899 U-234 92234 92000 0.000235 0.000090 0.000007 U-235 92235 92000 0.026444 0.010124 0.000743 U-236 92236 92000 0.000122 0.000046 0.000003 U-238 92238 92000 0.854666 0.323072 0.023697 Total 1.000000 1.000000 0.073348 Page 320 of 357 FC08513, R0 EC 68969 229

Attachment E1 FC08513 PIET -43741-TM-963 PNNL-15870 Rev. 1 Page 23C of269 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 8016 -0.118533 8016 0.666667 8016 0.048899 92234 -0.000235 92234 0.000090 92234 0.000007 92235 -0.026444 92235 0.010124 92235 0.000743 92236 -0.000122 92236 0.000046 92236 0.000003 92238 -0.854666 92238 0.323072 92238 0.023697 Photons 8000 -0.118533 8000 0.666667 8000 0.048899 92000 -0.000235 92000 0.000090 92000 0.000007 92000 -0.026444 92000 0.010124 92000 0.000743 92000 -0.000122 92000 0.000046 92000 0.000003 92000 -0.854666 92000 0.323072 92000 0.023697 CEPXS Form : material 0 0.118533 U-234 0.000235 U-235 0.026444 U-236 0.000122 U-238 0.854666 matname Uranium Dioxide density 10.960000 Comments and References Density from http://physics.nist.gov/cgi-bin/Star/compos.pl?matno=272 (NIST 1998). Also called uranium dioxide. Paxton and Pruvost (1986) appears to have weight fractions appropriate for U03 instead of U02. Density and formula also from pg M8.2.4 of Petrie et al. (2000). Uranium isotopics assumed for LEU: Wt% U234/235/236/238 = 0.0267/3.0/0.0138/96.9595. 335 Uranium Hexafluoride Formula= UF6 Molecular weight (g/mole) = 351.9485101 Density (g/cm3) = 4.680000 Total atom density (atomslb-cm) = 5.606E-02 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data was calculated from the input formula. Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Dens it~ F 9019 9000 0.323884 0.857143 0.048047 U-234 92234 92000 0.000181 0.000039 0.000002 U-235 92235 92000 0.020283 0.004339 0.000243 U-236 92236 92000 0.000093 0.000020 0.000001 U-238 92238 92000 0.655559 0.138460 0.007761 Total 1.000000 1.000000 0.056055 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 9019 -0.323884 9019 0.857143 9019 0.048047 92234 -0.000181 92234 0.000039 92234 0.000002 Page 321 of 357 FC08513, R0 EC 68969 230

Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Page 231 of 269 Total 1.000000 1.000000 0.043481 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 8016 -0.001197 8016 0.006796 8016 0.000296 24000 -0.000997 24000 0.001743 24000 0.000076 26000 -0.000997 26000 0.001623 26000 0.000071 28000 -0.000499 28000 0.000772 28000 0.000034 40000 -0.982348 40000 0.978381 40000 0.042541 50000 -0.013962 50000 0.010686 50000 0.000465 Photons 8000 -0.001197 8000 0.006796 8000 0.000296 24000 -0.000997 24000 0.001743 24000 0.000076 26000 -0.000997 26000 0.001623 26000 0.000071 28000 -0.000499 28000 0.000772 28000 0.000034 40000 -0.982348 40000 0.978381 40000 0.042541 50000 -0.013962 50000 0.010686 50000 0.000465 CEPXS Form : material 0 0.001197 Cr 0.000997 Fe 0.000997 Ni 0.000499 Zr 0.982348 Sn 0.013962 matname Zircaloy-2 density 6.560000 Comments and References See http://www. matweb.com/search/DataSheet. aspx?MatGU ID=eb 1dad5ce 1ad4a 1f9e92f86d5b44 740d&ckc k=1 (Automation Creations 2010) and pg 201 of Paxton and Pruvost (1986), revision issued July 1987. Weight fractions normalized to 1.0. 369 Zircaloy-4 Formula = Molecular weight (g/mole) = Density (g/cm3) = 6.560000 Total atom density (atoms/b-cm) = 4.350E-02 The above density is estimated to be accurate to 3 significant digits. Uncertainties are not addressed. The following data were calculated from the input weight fractions . Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Densitll 0 8016 8000 0.001196 0.006790 0.000295 Cr 24000 24000 0.000997 0.001741 0.000076 Fe 26000 26000 0.001994 0.003242 0.000141 Zr 40000 40000 0.981858 0.977549 0.042520 Sn 50000 50000 0.013955 0.010677 0.000464 Total 1.000000 1.000000 0.043497 Page 350 of 357 FC08513, R0 EC 68969 231

Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. I Page 232 of269 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 8016 -0.001196 8016 0.006790 8016 0.000295 24000 -0.000997 24000 0.001741 24000 0.000076 26000 -0.001994 26000 0.003242 26000 0.000141 40000 -0.981858 40000 0.977549 40000 0.042520 50000 -0.013955 50000 0.010677 50000 0.000464 Photons 8000 -0.001196 8000 0.006790 8000 0.000295 24000 -0.000997 24000 0.001741 24000 0.000076 26000 -0.001994 26000 0.003242 26000 0.000141 40000 -0.981858 40000 0.977549 40000 0.042520 50000 -0.013955 50000 0.010677 50000 0.000464 CEPXS Form : material 0 0.001196 Cr 0.000997 Fe 0.001994 Zr 0.981858 Sn 0.013955 matname Zircaloy-4 density 6.560000 Comments and References See http://www. matweb.com/search/DataSheet.aspx?MatGUID=e36a9590eb5945de94d89a35097b ?faa (Automation Creations 201 0). Weight fractions normalized to 1.0. 370 Zirconium Formula= Zr Molecular weight (g/mole) = 91.224 Density (g/cm3) = 6.506000 Total atom density (atoms/b-ern)= 4.295E-02 The above density is estimated to be accurate to 4 significant digits. Uncertainties are not addressed. The following data was calculated from the input formula. Weight Atom Atom Element Neutron ZA Photon ZA Fraction Fraction Densit~ Zr 40000 40000 1.000000 1.000000 0.042949 Total 1.000000 1.000000 0.042949 MCNP Form I Weight Fractions I Atom Fractions I Atom Densities Neutrons 40000 -1 .000000 40000 1.000000 40000 0.042949 Photons 40000 -1 .000000 40000 1.000000 40000 0.042949 CEPXS Form: material Zr 1.000000 matname Zirconium Page 351 of 357 FC08513, R0 EC 68969 232

FC08513, R0 Neutron reactions

• Attachment E3 Effective FC08513 energy Barns per decay,' Page 248 of 269 Mass Abundance Decay MeV 2200 m/s Reso-Element no., (a/o) or cross nance (Symbol)

Mass, amu half-life Type Percent Total 'Y Prod. section integral At. no., Z A Neutron 1.00866520 12m 100 (0.31) 0 Hydrogen 1.00797 (H) 1 1.00782519 99 .985 a/o 'Y 0.332 2 2.01410222 0.015 a/o 'Y 0.00053 3 3.01604971 12.3 y 100 0.0057 0 Helium 4.0026 (He) 3 3.01602973 0.000137 a/o p 5327

                                                                                                                         'Y                           2390
                   .2 4      4.00260312    99 .999863 a/o                                                            0 EC 68969 6      6.0188927     0.8 8                         100      1.58       0 Lithium                  6.939                                                                       a           70.7 (Li)              6      6.0151247     7.5632 a/o                                                    a           940 7      7.0160039     92.4368 afo                                                   'Y          0.037 3

8 8.0224871 0.84 8 -,2a 100 8.49 0

                                                                                                                         'Y          0.0092            0.004 Beryllium          9      9.0121855    100 a/o (Be)              10     10.0135344    2.5E6y                        100      (0.22)      0          'Y          <0 .00 1 4

a 759 341 Boron 10.811 a '3837 1722 (B) 10 10.0129388 19.61 a/o 11 11.0093053 80.39 a/o 'Y 0.0055 5 12 12.0143537 0.0203 s 100 6.381 0.058

                                                                                                                          'Y         O.OQ34            0.0015 Carbon                   12.01115 (C)               11     11.0114317    20 .3 m             +         100       (1.40)        1.022 0.0034           . 0.0015 6                 12     12.0000000    98 .893 a/o                                                    'Y
             't                                                                                                                      0.0009             0;0013 j

13 13.003354 1.107 a/o 'Y 14.00324197 5730 y 100 0.045 0 0 14 248

1Q/1412016 Attachment E4 www.hazard.com/msds/12/cdrv'cdnsx.html FCOSS 13 DOW CHEMICAL CO-- TRYMER 9501-4 RIGID FOAM INSULATION, 02837 -- 9320-00N056603

    =====================             Product Identification   =====================           Page 249 of 269 Product ID :TRYMER 9501-4 RIGID FOAM INSULATION, 02837 MSDS Date:03/18/1992 FSC:9320 NIIN:00N056603 MSDS Number: CDNSX
    === Responsible Party Company Name:DOW CHEMICAL CO Address:2020 DOW CNTR City:MIDLAND State:MI ZIP:48674 Country:US Info Phone Num:517-636-1000 Emergency Phone Num:517-636-4400 CAGE:0BG07
   === Contractor Identification Company Name:DOW CHEMICAL CO Address:100 LARKIN CENTER Box:City:MIDLAND State :MI ZIP:48674 Country:US Phone:800-258-2436 CAGE:96717 Company Name:DOW CHEMICAL CO THE Address:1801 DOW CTR City:MIDLAND State:MI ZIP:48674-1801 Country:US Phone:517-636-4400 I 800-258-2436 CAGE:0BG07
   =============         Composition/Information on Ingredients           =============

Ingred Name:HNDLG & STOR:INADEQ VENTILATION, IMPROPER DISPOSAL AND/OR MISAPPLICATION. STORE POLYURETHANE & POLYISOCYANURATE(ING 21 RTECS #:9999999ZZ Ingred Name:ING 20:FOAM BUNS AND SHEETS WITH ADEQUATE AISLEWAYS TO PERMIT ACCESS TO ALL AREAS. FOR MORE INFO CONTACT NEHC . RTECS #:9999999ZZ Ingred Name:POLYMERIZED POLYURETHANE MODIFIED POLYISOCYANURATE RIGID CELLULAR PLASTIC Fraction by Wt : 85 - 97% OSHA PEL:N/K ACGIH TLV:N/K Ingred Name :TRICHLOROFLUOROMETHANE (CFC-11) (SARA 313) (CERCLA) CAS:75-69-4 RTECS #:PB6125000 Fraction by Wt: 3-15% OSHA PEL:1000 PPM ACGIH TLV:C, 1000 PPM EPA Rpt Qty:5000 LBS DOT Rpt Qty:5000 LBS Ozone Depleting Chemical:1 Ingred Name:SUPDAT:SUCH FIRES CAN BURN RAPIDLY & PRDCE INTENSE HEAT, hl.tp:/lwww.hazard.com/msds/12/cdrv'cdnsx.html 1/4 FC08513, R0 EC 68969 249

1011412016 www.hazard.com/msds/12/cdr"Vcdnsx.html Attachment E4 FC08513 DENSE SMOKE & IRRITATING/TOX GASES. RIGID POLYURETHANE (lNG 4) RTECS #:9999999ZZ Ingred Name:ING 3:FOAMS AUTOIGNITE @ 650-800F(343-427C) & RIGID Page 250 of 269 POLYISOCYANURATE FOAMS@ 900-1000F(482-538C). C0*2, CO, (lNG 5) RTECS #:9999999ZZ Ingred Name:ING 4:POSS TRACES OF HYDROGEN CYANIDE, HALOGEN ACIDS, & NITROGEN OXIDES EVOLVED UNDER FIRE CNDTNS. PROBABILITY (lNG 6) RTECS #:9999999ZZ Ingred Name:ING 5:0F DUST EXPLO FROM POLYURETHANE/POLYISOCYANURATE DUST IS VERY LOW, HOWEVER, DO NOT SMOKE/USE NAKED LIGHTS, (lNG 7) RTECS #:9999999ZZ Ingred Name:ING 6:0PEN FLAMES, SPACE HEATERS/OTHER IGNIT SOURCES NEAR RIGID FOAM FABRICATING OPERATIONS/NEAR STORED BUNS OR (lNG 8) RTECS #:9999999ZZ Ingred Name:ING 7:SHEETS. INSTALL FOAM ONLY AFTER ALL WELDING, CUTTING/OTHER HOT WORK HAS BEEN COMPLETED. IF HOT WORK MUST (lNG 9) RTECS #:9999999ZZ Ingred Name:ING 8:BE DONE AFTER FOAM HAS BEEN INSTALLED, HOT WORK TRADE MUST BE WARNED:REMOVE FOAM FROM IMMED WORK AREA TO (lNG 10) RTECS #:9999999ZZ Ingred Name:ING 9:SUFFICIENT DIST THAT HEAT TRANSMITTED FROM TORCH/THRU METAL WILL NOT IGNITE FOAM. REMOVE ALL COMBUST MATL (lNG 11) RTECS #:9999999ZZ Ingred Name:ING 10:FROM VICIN OF & IMMED BELOW WORK AREA. POST FIRE WATCHER EQUIPPED W/FIRE EXTING DURING & FOR 30 MINS (lNG 12) RTECS #:9999999ZZ Ingred Name:ING 11 :AFTER HOT OPERATIONS. STOP WORK IMMED IF FOAM BEGINS TO SMOKE & REMOVE MORE FOAM FROM WORK AREA. WHEN (lNG 13) RTECS #:9999999ZZ Ingred Name:ING 12:HOT-WIRE CUTTING RIGID POLYURETHANE/POLYISOCYANURATE FOAM, KEEP FIRE EXTING NEARBY. WORK SHOULD BE (lNG 14) RTECS #:9999999ZZ Ingred Name:ING 13:CARRIED OUT IN WELL VENTILATED AREA - DO NOT BREATHE FUMES. RTECS #:9999999ZZ Ingred Name:HAZ DECOMP PROD:OXIDES UNDER FIRE CONDITIONS. RTECS #:9999999ZZ Ingred Name:EFTS OF OVEREXP:RELEASED FROM MATL, ESPECIALLY WHEN IT IS CUT. IN MAN, EXPOS TO CONCS >2500 PPM CAN CAUSE CNS, (lNG 17) RTECS # :9999999ZZ Ingred Name : ING 16:ANESTH/NARC EFTS; @LEVELS >5000 PPM, IT MAY INCR SENSITIVITY TO EPINEPHRINE & INCR IRREGULAR HEARTBEATS.(ING 18) RTECS #:9999999ZZ Ingred Name:ING 17 : CONCS OF BLOWING AGENT ANTIC INCIDENTAL TO PROPER INDUST HNDLG ARE EXPECTED TO BE WELL BELOW EXPOS (lNG 19) RTECS #:9999999ZZ Ingred Name:ING 18:GUIDELINES. RPTD EXCESS EXPOS TO DUSTS MAY CAUSE RESP IRRIT & POSSIBLY OTHER RESP EFTS. RTECS #:9999999ZZ http://www.hazard.com/msds/12/cdrVcdnsx.html 214 FC08513, R0 EC 68969 250

10114/2016 www.hazard.com/msdslf2/cdrv'cdnsx.html Attachment E4 FC08513

    =====================             Hazards Identification    =====================

LDSe LCSe Mixture : NONE SPECIFIED BY MANUFACTURER. Page 251 of 269 Routes of Entry: Inhalation :YES Skin:YES Ingestion:YES Reports of Carcinogenicity:NTP:NO IARC:NO OSHA : NO Health Hazards Acute and Chronic:CHLOROFLUOROCARB (CFC) MATLS HAVE PRDCED SENSIT OF MYOCARDIUM TO EPINEPHRINE IN LAB ANIMALS & COULD HAVE SIMILAR EFT IN HUMANS. ADRENOMIMETICS (I.E., EPINEPHRINE) MAY BE CONTRA-INDICATED EXCEPT FOR LI FE-SUSTAINING USES IN HUMANS ACUTELY/CHRONICALLY EXPOS TO CFCS. ACUTE:EYE:SOLID/DUST MAY CAUSE IRRIT OR(EFTS OF OVEREXP) Explanation of Carcinogenicity : NOT RELEVANT. Effects of Overexposure:HLTH HAZ:CORNEAL INJURY DUE TO MECH ACTION. SKIN:MECH INJURY ONLY. INGEST :MAY CAUSE CHOKING. INHAL:DUST MAY CAUSE IRRIT TO UPPER RESP TRACT. VAPS/FUMES GENERATED IN THERM OPERATIONS SUCH AS HOT-WIRE C UTTING MAY CAUSE IRRIT UNLESS AREA IS ADEQ VENTILATED. SM AMTS OF BLOWING AGENT TRICHLOROFLUOROMETHANE ARE (lNG 16) Medical Cond Aggravated by Exposure:NONE SPECIFIED BY MANUFACTURER.

    =======================              First Aid Measures  =======================

First Aid:INGEST:CALL MD IMMEDIATELY . EYES:IRRIGATE IMMED W/WATER FOR

         @LST 15 MINS. MECH EFTS ONLY. SKIN:WASH OFF IN FLOWING WATER/SHOWER . INHAL:REMOVE TO FRESH AIR IF EFFECTS OCCUR. CONSULT PHYSICIAN.
   =====================              Fire Fighting Measures    =====================

Extinguishing Media : IF STORED/IN-PLACE POLYURETHANE/POLY-ISOCYANURATE FOAM SHOULD IGNITE, EXTING FIRE IMMED BY DRENCHING W/WATER (SUPDAT) Fire Fighting Procedures:NIOSH APPRVD PRESS DEMAND SCBA & FULL PROT EQUIP. PROTECT ALL INDOOR BUN & SHEET STOR AREAS W/FUSIBLE SPRINKLERS. MAINTAIN MIN CLEARANCE OF 6 FT (SUPDAT) Unusual Fire/Explosion Hazard:THERM DECOMP PRODS MAY INCL FLUORIDES, CHLORIDES, & PHOSGENE . RIGID POLYURETHANE & POLYISOCYANURATE FOAMS, IN COMMON W/OTHER ORG MATLS SUCH AS (SUPDAT)

   ==================            Accidental Release Measures       ==================

Spill Release Procedures:NOT APPLICABLE. Neutralizing Agent : NONE SPECIFIED BY MANUFACTURER.

   ======================              Handling and Storage   ======================

Handling and Storage Precautions:POTNTL RISKS ASSOC W/RIGID POLYURETHANE & POLYISOCYANURATE FOAMS ARISE FROM DUST, FIRE & TOX THERM DECOMP PRODS & MAY RSLT FROM IMPROPER STOR,(ING 2e) Other Precautions:NO SMOKING IN AREA OF USE . DO NOT USE IN GENERAL VICINITY OF ARC WELDING, OPEN FLAMES OR HOT SURFACES. HEAT AND/OR UV RADIATION MAY CAUSE THE FORMATION OF CHLORIDES, FLUORIDES OR PHOSGENE .

   =============        Exposure Controls/Personal Protection            =============

Respiratory Protection:ATM LEVELS SHOULD BE MAINTAINED BELOW EXPOS GUIDELINE. WHEN RESP PROT IS REQD FOR CERTAIN OPERATIONS, USE NIOSH APPRVD AIR-PURIFYING RESP. IN DUSTY ATM, USE NIOSH APPRVD DUST RESPIRATOR . Ventilation:CONTROL AIRBORNE CONCS BELOW EXPOS GUIDELINE. USE LOC EXHAUST VENT FOR DUSTY OPERATIONS/WHEN HOT-WIRE CUTTING. Protective Gloves:IMPERVIOUS GLOVES . Eye Protection:ANSI APPROVED SAFETY GLASSES . Other Protective Equipment :ANSI APPRVD EYE WASH & DELUGE SHOWER . t"dtp://www.hazard.com/msdslf2/cdrv'cdnsx.html 314 FC08513, R0 EC 68969 251

10/1412016 www.hazard.com/msds/12/cdnlcdnsx.html FC08513 Attachment E4 Work Hygienic Practices:NONE SPECIFIED BY MANUFACTURER. Supplemental Safety and Health EXTING MEDIA:SPRAY FROM FIRE HOSE. FOR SM FIRES, USE WATER SPRAY, FOAM, C0*2/DRY CHEM EXITINGUISHERS. FIRE FIGHT PROC:BETWEEN TOPS OF FOAM Page 252 of 269 STACKS & SPRINKLER HEADS. EXPLO HAZ:PAPER, WOOD, COTTON & RUB B, CAN PRESENT UNREASONABLE FIRE RISKS IN CERTAIN MISAPPLIC WHEN EXPOS TO IGNIT SOURCES IN AIR. ONCE IGNITED, (ING 3)

   ==================          Physical/Chemical Properties       ==================

Evaporation Rate &

Reference:

NOT KNOWN Appearance and Odor:RIGID CELLULAR PLASTIC, NO ODOR.

   =================          Stability and Reactivity Data       =================

Stability Indicator/Materia ls to Avoid:YES NONE KNOWN. Stability Condition to Avoid:NONE SPECIFIED BY MANUFACTURER. Hazardous Decomposition Products:CHLORIDES, FLUORIDES & PHOSGENE C0*2, CO, POSS TRACES OF HYDROGEN CYANIDE, HALOGEN ACIDS & NITROGEN (ING 15)

   ====================            Disposal Considerations   ====================

Waste Disposal Methods:INCINERATE OR BURY IN AN APPROVED LANDFILL ACCORDING TO LOCAL, STATE, AND FEDERAL REGULATIONS. Disclaimer (provided with this information by the compiling agencies): This information is formulated for use by elements of the Department of Defense.. The United States of America in no manner whatsoever, expressly or implied, warrants this information to be accurate and disclaims all liability for its use. Any person utilizing this document should seek competent professional advice to verify and assume responsibility for the suitability of this information to their particular situation. tttp://www.hazard.com/msds/12/cdnlcdnsx.html 414 FC08513, R0 EC 68969 252

Attachment E BENCHMARK Case Results FC08513 FC08513, R0 Page 253 of 269 Thread Name & Version = MCNP5, 1.68 I I I <= 1-1

              +---------------------------------------------------------------------+

Copyright 2818. Los Alamos National Security, LLC. All rights reserved. This material was produced under U.S. Government contract DE-AC52-86NA25396 for Los Alamos National Laboratory, which is operated by Los Alamos National Security, LLC, for the U.S. Department of Los Energy. The Government is granted for itself and others acting on its behalf a paid-up, nonexclusive, irrevocable worldwide license in this material to reproduce, prepare derivative works, and perform publicly and display publicly. Beginning five (5) years after 2818, subject to additional five-year worldwide renewals, the Government is granted for itself and others acting on its behalf a paid-up, nonexclusive, irrevocable worldwide license in this material to reproduce, prepare derivative works, distribute copies to the public, perform publicly and display publicly, and to permit others to do so. NEITHER THE UNITED STATES NOR THE UNITED EC 68969 STATES DEPARTMENT OF ENERGY, NOR LOS ALAMOS NATIONAL SECURITY, LLC, NOR ANY OF THEIR EMPLOYEES, MAKES ANY WARRANTY, EXPRESS OR IMPLIED, OR ASSUMES ANY LEGAL LIABILITY OR RESPONSIBILITY FOR THE ACCURACY, COMPLETENESS, OR USEFULNESS OF ANY INFORMATION, APPARATUS, PRODUCT, OR PROCESS DISCLOSED, OR REPRESENTS THAT ITS USE WOULD NOT INFRINGE PRIVATELY OWNED RIGHTS.

               +----------------------------- ----------------------------------------+

1mcnp version 5 ld=89282818 87/14/11 13:11:48

              *************************************************************************                       probid    87/14/11 13:11:48 n=hall185a tasks 4 1tally fluctuation charts tally         4                            tally         34                           tally        44 nps      mean       error     vov slope    fom      mean       error     vov slope    fom      mean       error     vov slope    fom 8192888   2.3587E-14   8.4688   8.7587 2.8 1.2E-81   1.9187E-81   8.5698   8.9589 1.9 8.3E-82   3.2888E-82   8.1824   8.2469 2.6 8.1E-81 16384888   2.2313E-14   8.2972   8.4899 1.9 1.4E-81   1.6214E-81   8.3552   8.7723 1.9 9.7E-82   3.5885E-82   8.1231   8.1211 2.1 8.1E-81 24576888   3.7911E-14   8.2323   8.1475 1.7 1.3E-81   1.7643E-81   8.2368   8.5616 1.8 1.2E-81   3.6983E-82   8.1282   8.2169 2.3 4.7E-81 32768888   4.6842E-14   8.2215   8.1219 1.6 1.8E-81   2.2123E-81   8.2497   8.4843 1.8 7.9E-82   3.5676E-82   8.8979   8.1818 2.2 5.2E-81 48968888   3.8782E-14   8.2111   8.1213 1.7 8.3E-82   1.9141E-81   8.2312   8.4821 1.9 6.9E-82   3.3821E-82   8.8865   8.1515 2.3 4.9E-81 49152888   3.5725E-14   8.1954   8.1113 1.6 7.9E-82   1.9475E-81   8.2898   8.2948 2.8 6.9E-82   5.8388E-82   8.3341   8.9588 2.1 2.7E-82 Page 1 253

Attachment E BENCHMARK Case Results FC08513 FC08513, R0 57344eee 3 . 4564E-14 e.1816 e.e978 1.6 7.7E-e2 2.e123E-e1 e.1971 e.2211 1.9 6.5E-92 4.9239E-e2 e.2949 e.9211 2 . 1 2 . 9 E-e~age 254 of 269 65536eee 5.2265E-14 9.2936 9.5667 1.6 2.5E-e2 2.2868E-91 9.1758 9.1552 1.8 7 .eE-e2 6.4263E-92 9.2965 9.4699 1.9 2.5E-92 73728eee 4.9845E-14 9.2753 9.5535 1.7 2.5E-e2 2.3626E-e1 e.1732 e.1441 1.8 6.4E-e2 8.4795E-e2 e.3462 e.4963 2.e 1.6E-e2 8192eeee 5.9239E-14 9.2599 9.5198 1.6 2.7E-e2 2.473eE-e1 e.1622 e.1132 1.8 6.4E-e2 8.7632E-92 9.3134 9.4393 2.e 1.7E-e2 9e112eee 4.9269E-14 9.2352 9.4893 1.6 2.7E-e2 2.4227E-e1 e.1529 e.1e55 1.8 6.5E-e2 8.2632E-e2 e.3922 e.43ee 2.e 1.7E-e2 983e4eee 4.9743E-14 9.2195 9.4396 1.6 2.8E-e2 2.5897E-e1 e.15e5 e.e914 1.7 6.1E-e2 7.9472E-e2 e.2882 e.4289 2.e 1. 7E-e2 1eeeeeeee 4.9329E-14 9.2177 9.4392 1.6 2.8E-e2 2.5684E-e1 e.1488 e.e911 1.8 6.1E-e2 7.8821E-e2 e.2857 e.4288 2.1 1.6E-e2 1mesh-based weight window generator print table 199 normalization: 9-group: 3.4962E-99 1-group: 3.4962E-e9 327691 particles escaped wwg mesh. dump no. 2 on file hall195ar nps 1eeeeeeee call = 883937219 ctm = 743.93 nrn = 11881778178 tally data written to file hall195am EC 68969 23 warning messages so far. run terminated when 1eeeeeeee particle histories were done. computer time = 743.11 minutes Page 2 254

Attachment E BENCHMARK Case Results FC08513 FC08513, R0 Page 255 of 269 Thread Name & Version = MCNP5, 1.69 I I I I_) I

              +------------------------------------------------------ ---------------+

Copyright 2919. Los Alamos National Security, LLC. All rights reserved. This material was produced under U.S . Government contract DE-AC52-96NA25396 for Los Alamos National Laboratory, which is operated by Los Alamos National Security, LLC, for the U.S. Department of Los Energy. The Government is granted for itself and others acting on its behalf a paid-up, nonexclusive, irrevocable worldwide license in this material to reproduce, prepare derivative works, and perform publicly and display publicly. Beginning five (5) years after 2919, subject to additional five-year worldwide renewals, the Government is granted for itself and others acting on its behalf a paid-up, nonexclusive, irrevocable worldwide license in this material to reproduce, prepare derivative works, distribute copies to the public, perform publicly and display publicly, and to permit others to do so. NEITHER THE UNITED STATES NOR THE UNITED EC 68969 STATES DEPARTMENT OF ENERGY, NOR LOS ALAMOS NATIONAL SECURITY, LLC, NOR ANY OF THEIR EMPLOYEES, MAKES ANY WARRANTY, EXPRESS OR IMPLIED, OR ASSUMES ANY LEGAL LIABILITY OR RESPONSIBILITY FOR THE ACCURACY, COMPLETENESS, OR USEFULNESS OF ANY INFORMATION, APPARATUS, PRODUCT, OR PROCESS DISCLOSED, OR REPRESENTS THAT ITS USE WOULD NOT INFRINGE PRIVATELY OWNED RIGHTS.

              +--------------------------------------- - ----------------- - -----------+

1mcnp version 5 ld=99292919 19/98/16 97:29 : 15

              *************************************************************************                                         probid      19/98/16 97:29 : 15 i=hall195a.txt o=hall195a.out5 1tally fluctuation charts tally             4                               tally             34                                tally           44 nps        mean       error         vov slope     fom        mean       error         vov slope       fom        mean       error       vov slope    fom 8192999   2.3455E-14     9 . 4683   9.7643 2.9 4.8E-91      1.9219E-91     9.5696     9.9512 1.9 3.2E-91        3.2856E-92     9.1821   9.2456 2.5 3.1E+99 16384999   2.2389E-14     9.2978     9.4967 1.9 6.2E-91      1.6265E-91     9.3551     9 . 7719 1 . 9 4 .4E-91   3.5929E-92     9.1226   9.1194 2.1 3. 7E+99 24576999   3.6196E-14     9.2313     9.1632 1.7 7.1E-91      1.7533E-91     9.2377     9.5648 1.8 6.8E-91        3.6927E-92     9.1299   9.2185 2.3 2.6E+99 32768999   4.5999E-14     9.2243     9 . 1397 1.7 5.7E-91    2.2969E-91     9 . 2591   9.4915 1 . 8 4 . 6E-91    3.5644E-92     9.9977   9.1821 2.2 3.9E+99 49969999   3.7943E-14     9.2136     9.1399 1.7 5.9E-91      1.9999E-91     9.2315     9.3993 1.9 4.2E-91        3.3788E-92     9.9864   9.1525 2.3 3.9E+99 49152999   3.5982E-14     9.1975     9.1191 1.6 4.8E-91      1 . 9443E-91   9.2999     9 . 2926 2.9 4 . 3E-91    5.9279E-92     9.3343   9.9519 2.2 1. 7E-91 57344999   3.4994E-14     9.1833     9.1941 1.6 4.8E-91      2.9996E-91     9.1972     9.2197 1.9 4.1E-91        4.9246E-92     9.2948   9.9212 2.1 1.9E-91 255                   65536999   4 . 6736E-14   9.3149     9.7383 1.7 1.4E-91      2.2914E-91     9 . 1794   9.1799 1.8 4.4E-91        6.5887E-92     9.3974   9.4893 1.9 1. 5E-91 73728999   4.4898E-14     9.2939     9.7197 1.7 1.5E-91      2.2859E-91     9.1767     9.1539 1.8 4.1E-91        8 . 6181E-92   9.3478   9.4717 1.9 1.9E-91 81929999   4.5938E-14     9.2637     9 . 6599 1.6 1. 7E-91   2.4977E-91     9.1648     9.1196 1 . 8 4.2E-91      8.8951E-92     9.3149   9 . 4119 1.9 1.2E-91 Page 1

Attachment E BENCHMARK Case Results FC08513 FC08513, R0 90112000 4.5333E-14 0.2458 0.6301 1.6 1.7E-01 2.3628E-01 0.1552 0.1124 1.8 4.4E-01 8.3836E-02 0 . 3038 0.4197 2 .e1 . 1 E~'ige 256 of 269 983e4eee 4.6108E-14 0.2283 9.5609 1.6 1 . 9E-01 2.5256E-e1 9.1526 9.9951 1.7 4.1E-91 8 . e598E-92 e.2898 e.4097 2.0 1.2E-01 1eeeeeeee 4.5757E-14 0.2262 9.5604 1.6 1.9E-01 2.5143E-01 9.1598 0.0948 1.7 4.2E-01 7 . 9930E - 02 0.2873 0.4096 2.1 1.2E-01 1mesh-based weight window generator print table 199 normalization: a-group: 3.2430E-09 1-group: 3.2439E-99 327605 particles escaped wwg mesh. dump no. 2 on file runtpf nps = 10eeeeeee coll = 883922966 ctm = 105.19 nrn = 11889967778 tally data written to file mctam 23 warning messages so far. run terminated when 1eeeeeeee particle histories were done. EC 68969 computer time = 105.29 minutes mcnp version 5 09292010 10/08/16 09 :07 : 27 probid 10/08/16 07:20:15 256 Page 2

Attachment E BENCHMARK Case Results FC08513 FC08513, R0 Page 257 of 269 Code Name & Version MCNP6, 1.0 _I _I _I _I _I _I _I _!_!_I _I _1_1 _I _I _I _I _I _!_! _I _I _I _I _I _! _! _I _I _I _I _I_!_! _I_!_! _I _I _I _I _!_! _I _! _I _I _I _!_!_I _I _I _I _1_1

              +------------------------------------------------------ ---------------+

Copyright 2008. Los Alamos National Security, LLC. All rights reserved. This material was produced under U.S. Government contract DE-AC52-06NA25396 for Los Alamos National Laboratory, which is operated by Los Alamos National Security, LLC, for the U.S. Department of Energy. The Government is granted for itself and others acting on its behalf a paid-up, nonexclusive, irrevocable worldwide license in this material to reproduce, prepare derivative works, and perform publicly and display publicly. Beginning five (5) years after 2008, subject to additional five-year worldwide renewals, the Government is granted for itself and others acting on its behalf a paid-up, nonexclusive, irrevocable worldwide license EC 68969 in this material to reproduce, prepare derivative works, distribute copies to the public, perform publicly and display publicly, and to permit others to do so. NEITHER THE UNITED STATES NOR THE UNITED STATES DEPARTMENT OF ENERGY, NOR LOS ALAMOS NATIONAL SECURITY, LLC, NOR ANY OF THEIR EMPLOYEES, MAKES ANY WARRANTY, EXPRESS OR IMPLIED, OR ASSUMES ANY LEGAL LIABILITY OR RESPONSIBILITY FOR THE ACCURACY, COMPLETENESS, OR USEFULNESS OF ANY INFORMATION, APPARATUS, PRODUCT, OR PROCESS DISCLOSED, OR REPRESENTS THAT ITS USE WOULD NOT INFRINGE PRIVATELY OWNED RIGHTS.

              +------------------------------------------------------ ---------------+

1mcnp version 6 ld=05l08l13 10108116 00:02:59

              *************************************************************************                                     probid    10108116 00:02:59 i=hall105a.txt o=hall105a.out 1tally fluctuation charts tally          4                                tally          34                             tally         44 nps         mean          error     vov slope    fom         mean         error      vov slope      fom      mean       error     vov slope    fom 8192000      2.3455E-14      0.4683   0.7643 2.0 3.6E-01      1.9210E-01     0.5696    0.9512 1.9 2 . 4E-01   3.2856E-02   0.1821   0.2456 2.5 2.4E+00 16384000      2.2389E-14      0.2978   0.4067 1.9 4.7E-01      1.6265E-01     0.3551    0.7710 1.9 3.3E-01     3.5029E-02   0.1226   0.1194 2.1 2.8E+00 24576000      3.6196E-14      0.2313   0.1632 1.7 5.0E-01      1.7533E-01     0.2377    0.5648 1.8 4. 7E-01    3.6927E-02   0.1200   0.2185 2.3 1.9E+00 32768000      4.5098E-14      0.2243   0.1307 1.7 3.8E-01      2.2068E-01     0.2501    0.4015 1.8 3.1E-01     3.5644E-02   0.0977   0.1821 2.2 2.0E+00 257                40960000      3.7943E-14      0.2136   0.1300 1.7 3.3E-01      1.9099E-01     0.2315    0.3993 1.9 2.8E-01     3.3788E-02   0.0864   0.1525 2.3 2.0E+00 49152000      3.5081E-14      0.1975   0.1191 1.6 3.2E-01      1.9443E-01     0.2099    0.2926 2.0 2.8E-01     5.0270E-02   0.3343   0.9510 2.2 1.1E-01 Page 1

Attachment E BENCHMARK Case Results FC08513 FC08513, R0 Page 258 of 269 57344999 3.4994E-14 9.1834 9.1941 1.6 3.2E-91 2.9996E-91 9.1972 9.2197 1.9 2.7E-91 4.9246E-92 9.2948 9.9212 2.1 1.2E-91 65536999 4.6736E-14 9.3149 9.7383 1.7 9.3E-92 2.2914E-91 9.1794 9.1799 1.8 2.9E-91 6.5887E-92 9.3974 9.4893 1.9 9.8E-92 73728999 4.4898E-14 9.2939 9.7197 1.7 9.5E-92 2.2859E-91 9.1767 9.1539 1.8 2.6E-91 8.6181E-92 9.3478 9.4717 1.9 6.7E-92 81929999 4.5938E-14 9.2637 9.6599 1.6 1.9E-91 2.4977E-91 9.1648 9 . 1196 1.8 2.7E-91 8.8951E-92 9.3149 9.4119 1.9 7.4E-92 99112999 4.5333E-14 9.2458 9.6391 1.6 1.1E-91 2 . 3628E-91 9.1552 9.1124 1.8 2.8E-91 8.3836E-92 9.3938 9.4197 2.9 7.2E-92 98394999 4.6198E-14 9.2283 9.5699 1.6 1.2E-91 2.5256E-91 9.1526 9.9951 1.7 2.6E-91 8.9598E-92 9.2898 9 . 4997 2.9 7.3E-92 199999999 4.5757E-14 9.2262 9.5694 1.6 1. 2E-91 2.5143E-91 9 .1598 9.9948 1.7 2.7E-91 7.9939E-92 9.2873 9.4996 2.1 7.4E-92 1mesh-based weight window generator print table 199 normalization: 1-group : 3.2439E-99 1-group: 3.2439E-99 327629 particles escaped wwg mesh. dump no. 2 on file runtpe nps = 199999999 coll = 883985942 ctm = 164.77 nrn = 11889994117 tally data written to file metal EC 68969 29 warning messages so far. run terminated when 199999999 particle histories were done. computer time= 164.78 minutes mcnp version 6 95/98/13 19/98/16 92:49:56 pro bid 19/98/16 99:92:59 258 Page 2

Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 259 of 269 Attachment F: File Listings Additional References MCNP Directory Top Directory Level. Directory Level 14mog. Directory Level14mon. Directory Level18mog . Directory Level 18mon. Directory Level Benchmark. FC08513, R0 EC 68969 259

Attachment F FC08513 FC08513, R0 Page 260 of 269 D X

             ~             v Home 1'

Sha re View This PC > Local Disk (C:) > Users > Jo e > MCNP_RUNS > SFP > RU N v b Sea rch RUN p I Program Files .... Name "' Date mo.dified Type Size Program Files (x86) --- 14M0gamma 10/ 17/2016 9:S2 AM Fil e folder Recovery 14MOneutron 10/ 17/ 20161 :02 PM File fo lder

• SWSetup 18MOgamma 10/ 17/ 2016 1:06PM Fi le folder Users 18MOneutron 10/ 17/ 2016 9:S4 AM Fi le folder I Default.migrated Benchmark 10/ 17/ 2016 1:38PM Fi le folder I

EC 68969 DefaultAppPool 3D Objects (£] Contacts

• Desktop

                           ~ Documents
                       . . Downloads Favorites I    Google Drive 5 items 260

Attachment F FC08513 FC08513, R0 Page 26 1 of 269

                            -;- I 14MOgamma                                                                                                                                0      X Ho me!

y 1' Share! Vie!W Local Disk (C:) ) Users ) Joe ) MCNP_RU NS ) SFP ) RU N ) 14MOgamma Search 14MOgamma p god iva Name Date modified Type Size RNP ]l FCS_14MOg_EABCR.inp 10/ 4/ 2016 8:36AM INP File 17 KB SFP J] FCS_14MOg_EABCR.out 10/4/ 2016 9:04AM OUT File 118 KB

• metal ex 0 FCS_14MOg_EABCR.tap 10/4/ 2016 9:03AM TAP File 42,361 KB
• RUN .J] FCS_14MOg_EABCRca.out 10/ 17/ 2016 9:22AM OUT File 3 KB 14MOgamma FCS_14MOg_EABCRcb.out 10/ 17/2016 9:22AM OUT File 59 KB EC 68969

                                                                    ]l FCS_14MOg_EABCRcc.out               10/ 17/ 2016 9:22AM    OUT File                     10 KB

• 14M0neutron FCS_14MOg_EABCRcd.out 10/ 17/ 2016 9:22 AM OUT File 10 KB 18MOgamma Jj FCS_14MOg_EABCRce.out 10/ 17/ 2016 9:22AM OUT File 10 KB 18M0neutron .]) FCS_14MOg_EABCRd.out 10/17/2016 9:22AM OUT File 10 KB
• save FCS_14MOg_EABCRcg.out 10/ 17/2016 9:22AM OUT File 10 KB
• case runs old FCS_14MOg_EABCRch.out 10n12016 9:20 PM OUT File 121 KB
• Cases for reference ]I FCS_14MOg_EABCRci.out 10/ 10/ 2016 7:33AM OUT File 126 KB FCS_imp ~ FCS_14MOg_EABCRci.txt 10/10/ 2016 8:41 AM Text Document 3 KB
• FCS_WWbase 0 FCS_14MOgEABCR.tap 10/9/2016 12:24 AM TAP File 294,147 KB OldM3 SG_Cases 14 items 261

Attachment F FC08513 FC08513, R0 Page 262 of 269

• I 14M Oneutron 0 X Home Sha re View 'v *

                               .,. 1' * ,        Joe , MC NP_RUNS , SFP , RUN > 14M0neutron                                               Search 1 MOn eutron                       jJ
                 ...... Downloads                                  Name                                          Date modified          Type                Size Favorites                             Jj FCS_14MOn_EABCR.inp                       10/ 13/ 2016 5:21 PM   IN P File                   17 KB

• Google Drive ,Jl FCS_14M0n_EABCR.out 10/ 3/ 2016 5:13PM OUT File 173 KB f4 Links 0 FCS_ 14M0n_EABCR.tap 10/ 14/201610:48 AM TAP File 640,706 KB MCNP_RU NS Jj FCS_14MOn_EABCRca.out 10/ 6/ 2016 10:54 PM OUT Fil e 77 KB
• Benchmark Jj FCS_14MOn_EABCRcb.out 10/ 8/ 20164:2.9 AM OUT Fil e 116 KB
• examples_files J FCS_14MOn_EABCRcc.out 10/17/ 2016 2:21PM OUT File 3 KB

                                                                    ] FCS_14MOn_EABCRcd.out                      10/ 17/ 2016 2:22PM    OUT File                    85 KB EC 68969 god iva Jl FCS_14Mon_EABCRce.out                     10/17/ 2016 2: 19PM    OUT File                   118 KB RNP                                  liJ FCS_14MOn_EABCRci:-txt-                  10/ 17/ 2016 2:28PM    Text Document                 3 KB 0   FCS_Cont                                 10/ 17/2016 1:05AM     Fil e                         1 KB FCS_SFP_H20

• metal ex RUN I 14MOgamma 14MOneutron 18MOgamma 18M0neutron Benchmark
• save SCE Files

                          .       test 262                11          Music 10 items           1 item selected 117 KB                                                                                                                       m--     r,;]
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Attachment F FC08513 FC08513, R0 Page 263 of 269 I

• Home v ..,..

I 18MOga mma Share View Local Disk (C:) , Users , Joe , MCNP _RUNS , SFP , RUN , 18MOgamma 5earch lSMOgamma D v. p X I godiva Name Date modified Type Size RNP J FCS_18MOg_EABCR.inp 10/4/ 2016 8:36AM INP File 17 KB I SFP .) FCS_18MOg_EABCR.out 10/7/ 2016 9:18 PM OUT File 192 KB I metal ex D FCS_18MOg_EABCR.tap 10/ 11/ 2016 5:04 PM TAP File 440,819 KB I RUN ~ FCS_18MOg_EABCR.txt 10/ 10/ 2016 8:50AM Text Document 2 KB I 14MOgamma ]l FCS_18MOg_EABCRca.out 10/ 11/ 20167:51 PM OUT File 134 KB EC 68969

                                                                           ~ FCS_18MOg_EABCRca.txt                10/ 11/ 2016 5:33 PM   Text Document                 2 KB I         14MOneutron D FCS_Cont                             10/ 10/ 2016 3:35 PM   File                          1 KB 18MOgamma
                                                                           .]l FCS_Cont..dat                      10/ 11 / 2016 7:51PM   DAT File                      1 KB I         18MOneutron case runs old Cases for reference FCS_imp FCS_WWbase I         OldM3 I         SG_Cases rrr r :1.. ...

8 items 263

Attachment F FC08513 FC08513, R0 Page 264 of 269

• I 18MOneutron D X Home Share View

                                        << Lo<:!!l Disl>   ~C. ::   ~   Users > Joe > MCNP_RUNS , SFP , RU N , 18MOneut ron                                  Sea rch 18MO neut ron                    p
                                                                       "'   Name                                   v   Date modified            Type                     Size 0   FCS_18MOn_EAB.tap                      10/ 10/ 2016 3:09 PM     TAP Fil e                3,71 8, 11 2 KB J   FCS_18M0n_EABCR.inp                    10/4/ 2016 8:37AM        IN P File                        17 KB metal ex                                    Jj FCS_18MOn_EABCR.out                     10/ 10/ 2016 3:39 PM     OUT File                        501 KB

• RUN ] FCS_18MOn_EABCRca.out 10/ 10/ 2016 3:38 PM OUT File 3 KB 14MOgamma J FCS_18MOn_EABCRcb.out 10/ 10/ 2016 3:38 PM OUT File 3 KB
• 14MOneutron Jl FCS_18MOn_EABCRcc.out 10/ 10/ 2016 3:39 PM OUT File 116 KB

                                                                            ~ FCS_18MOn_EABCRcc.txt                    10/ 10/ 2016 3:28 PM     Text Doc ument                      2 KB EC 68969 18MOgamma 0   FCS_Cont                               10/ 10/ 2016 12:22 ... File                                1 KB 18MOneutron                               ~Jl FCS_Cont.dat                           10/ 10/ 2016 3:38 PM     OAT Fil e                           1 KB

• save case runs old Ca.ses for reference FCS_imp FCS_WWba.se OldM3 SG_Cases e-r r- r-:1**

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Attachment F FC085 13 FC08513, R0 Page 265 of 269

• I Eenchmark D X Home Share View

                                                      <<   Local Disk (C:) > Users ) Joe ) MCNP_RUNS > SFP > RUN > Benchmark                        Searc h Ben chmark                 p Output                            1<.

Name Date modified Type Size

• examples_files ] ha1110Sa.out5 OUTS File 10/ 8/ 2016 9:07AM 202 KB gociva )) haii10Sa.out6 10/ 8/ 2016 2:49AM OUT6 File 218 KB

                    >                 RNP                                         ~ haii1 0Sa.txt                    7/ 8/ 2015 5:07 AM    Text Document                 13 KB
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• SFP E) haii1 0Sae.txt 7/ 8/ 2015 5:07 AM Text Doc ument 126 KB
• metal ex ~ haii10Sao.txt 7/ 8/ 2015 5:07 AM Text Document 203 KB 0 metal EC 68969 10/ 8/ 2016 2:49 AM File 7 KB v RUN 14MOgamma 0 mctam 10/8/ 2016 9:07 AM File 6 KB 0 runtpe 10/ 8/ 2016 2:49AM File 1,373 KB 14MOneutron 0 runtpf 10/ 8/ 2016 9:07AM Fil e 1,389 KB 18MOgamma 0 wwout 10/ 8/ 2016 2:49AM File 126 KB 18MOneutron D wwouu 10/ 8/ 2016 9:07AM Fil e 126 KB Benchmark

                        ..,               save

• case runs old Ca.ses for reference FCS_imp y

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Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 266 of 269 Attachment G: Reviewer Alternate Calculation FC08513, R0 EC 68969 266

Attachment G Alternate Calculation FC08513 Joe Willett Page 267 of 269 From: Steve Gebers, CHP <gebers@centurylink.net > Sent: Wednesday, October 12, 2016 12:58 PM To: 'Jan Bostelman Wildblue'; 'Joseph E Willett'

Subject:

18 Month EAB exposure rate Skyshine.doc Attachments: 18 Month EAB exposure rate Skyshine.doc.docx I attached a MicroSkyshine estimate for the site boundary (900-meter) using the geometry Joe had developed. The results are 4E-05-mR/hr. MicroSkyshine will only except energies between 0.1 to 10-Mev, so below 0.11 didn't use, and flux for 11-MeV was entered as 10-MeV. I might make a run adding all lower energies into 0.1 and see what the difference is between the two runs . Steve

• Virus-free. www.avast.com FC08513, R0 EC 68969 267

Attachment G Alternate Calculation FC08513 Page 268 of 269 C..lnlonullon Geometry Rectangular Volume Source Behind a Wall Name Rectangular Volume Source Behind a Wall File C:\Users\Steve\CioudStation\Quantum\BE Wofk\Decommission FCS\mcnp shine\SFP Model for SB Dose.sky Time Wednesday, October 12, 201612:49:03 PM Energy~ DaHRMullll Group II Energy (MeV) I Acdvly (PhofoMIMCt Radii I PtiCIIIon [ mR/hr 1 0.100000 2.1200E+17 4 .8576E-49 4 .2464E-25 2 0 .150000 2.5800E+17 1.1213E-50 1.1929E-26 3 0 .300000 1.9500E+17 4.3329E-35 3.4640E-11 4 0.450000 5.5900E+16 2 .2753E-32 5.2447E-09 5 0.700000 2.0400E+18 1.0500E-30 0.000009 6 1.000000 3.5100E+17 8.0622E-30 0.000012 7 1.500000 6.2200E+16 4 .8060E-29 0.000012 8 2.000000 4 .7000E+15 1.2203E-28 0.000002 9 2.500000 4 .7500E+15 2.1335E-28 0 .000004 10 3.000000 9.3000E+13 3.2346E-28 1.2404E-07 11 4.000000 8.3300E+12 5.2024E-28 1.7870E-08 12 6 .000000 1.0400E+10 8.2986E-28 3.5588E-11 13 8 .000000 1.2000E+09 9.7147E-28 4.8070E-12 14 10.000000 1.3800E+08 1.0380E-27 5.9067E- 13 Totals: 3.9700E+01 3.1 837E+18 4 .0756E-27 0.000040 1r ca.om.Q ........... (Unlla:....., X Jl v.lue 2.9528E+03 Jl fl y If Vlllue 5.2400E+01 II z II Vllue O.OOOOE+OO Length 9.2025E+01 Vllidth 3.5483E+01 Height -4.2000E+01 Diameter 10.000000 Radius One 0 .041667 Radius Two 0.041667 Thickness One 1.600000 Thickness Two 4 .9800E+01

i .l llllildll . . . . . .

           ........        II          Amllltnt AJr II           Clwwlllb                        a.-SIIIeld             [I      Sounle'#a-AJr                      0 .001200                                                           0.012000 Water Concrete                                                       2.350000 Iron Lead Zirconium Urania                                                                                                                       2.568020
                                                                                                                               ~

Buildup...., llld AtiiMUIIOft Cotllal.... 0.100000 Blllldup...., Alit 1.000000 Clwwlllb 2.0411E+01 0 a.-r8MIIII 2.806908 II --~ ~ 6.8137E+02 7.0459E+02 FC08513, R0 EC 68969 268

Attachment G Alternate Calculation FC08513 Page 269. of 269 0.150000 1.000000 1.6423E+01 2.469933 9.0014E+02 9.1904E+02 0.300000 1.000000 1.2400E+01 1.945346 1.8406E+02 1.9641E+02 0.450000 1.000000 1.0502E+01 1.657040 8.6263E+01 9.6422E+01 0.700000 1.000000 6.651117 1.368761 4.5364E+01 5.5404E+01 1.000000 1.000000 7.312932 1.158262 3.0166E+01 3.6657E+01 1.500000 1.000000 5.956011 0.942601 2.1678E+01 2.65nE+01 2.000000 1.000000 5.136587 0.810014 1.6667E+01 2.4834E+01 2.500000 1.000000 4.590476 0.718996 1.7827E+01 2.3136E+01 3.000000 1.000000 4.187659 0.652274 1.7005E+01 2.1945E+01 4.000000 1.000000 3.654747 0.560835 1.6566E+01 2.0602E+01 6.000000 1.000000 3.089745 0.459561 1.6993E+01 2.0542E+01 8.000000 1.000000 2.607818 0.405260 1.7662E+01 2.1095E+01 10.000000 1.000000 2.648517 0.372494 1.6667E+01 2.1908E+01 llllllgetllaw..._..... Numerical Quadrature Thirty Two Length Segments (M): 20

                 \Mdth Segments (N):                                         20 Vertical Segments (C):                                       20 FC08513, R0                                          EC 68969                                     269

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OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION DOCKET NUMBER 50-285 / LICENSE NUMBER CPR-40 ATTACHMENT 6 LIST OF REGULATORY COMMITMENTS

LIC-16-0109 Page 1 Regulatory Commitments This table identifies actions discussed in this letter for which OPPD commits to perform. Any other actions discussed in this submittal are described for the NRCs information and are not commitments. TYPE (Check one) SCHEDULED ONE-TIME CONTINUING COMPLETION DATE COMMITMENT ACTION COMPLIANCE (If Required) Revise the USAR to include a description X Complete in accordance of how the FCS Spent Fuel Pool design with next scheduled and operational characteristics meets or USAR update following compares with the NUREG-1738 Industry exemption approval. Decommissioning Commitments (IDCs) and Staff Decommissioning Assumptions (SDAs).}}