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#REDIRECT [[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)]]
| number = ML16356A578
| issue date = 12/16/2016
| title = Fort Calhoun Station Unit 1 - Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR Part 50, Appendix E (Attachments 1, 2, 4, 5 and 6)
| author name = Marik S M
| author affiliation = Omaha Public Power District
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000285
| license number = DPR-040
| contact person =
| case reference number = LIC-16-0109
| package number = ML17005A231
| document type = Calculation, Letter
| page count = 387
}}
 
=Text=
{{#Wiki_filter: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  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.
U.S. Nuclear Regulatory Commission LIC-16-0109 Page 2                        ATTACHMENT 3 CONTAINS INFORMATION BEING WITHHELD FROM PUBLIC DISCLOSURE PER 10 CFR 2.390. UPON SEPARATION, THIS LETTER IS DECONTROLLED 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.
 
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. Attachment 6 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, 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 c: 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 1 0 CFR 2.390 6. List of Regulatory Commitments 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 
 
LIC-16-0109 Attachment 1 Page 1    , 
 
==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
 
==6.0 ENVIRONMENTAL CONSIDERATION==
 
==7.0 REFERENCES==
 
LIC-16-0109 Attachment 1 Page 2 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).
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 Attachment 1 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). In order to allow a reduction in emergency planning requirements commensurate with the hazards associated with FCS's 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 Attachment 1 Page 4    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.
1 10 CFR 50.47(b):  The onsite and, except as provided in paragraph (d) of this section, offsite emergency response plans for nuclear power reactors must meet the following standards: In the Statement of Considerations for the Final Rule for Emergency Planning requirements for Independent Spent Fuel Storage Installations (ISFSIs) and for monitored retrievable storage (MRS) 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 reactor's 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 Attachment 1 Page 5    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 Attachment 1 Page 6    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 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. See the basis for 10 CFR 50.47(b). 3 10 CFR 50.47(b)(2):  On-shift facility licensee responsibilities for emergency response are unambiguously defined, adequate staffing to provide initial facility accident response in key functional areas is maintained at all times, timely augmentation of response capabilities is available and the No exemption is requested.
LIC-16-0109 Attachment 1 Page 7    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 effectively using assistance resources have been made, arrangements to accommodate State and local staff at the licensee's Emergency Operations Facility have been made, and other organizations capable of augmenting the planned response have been identified. Discontinuing offsite emergency planning activities and reducing the scope of onsite emergency planning is acceptable given the significantly reduced offsite consequences when FCS is in the permanently defueled condition. The FCS emergency plan will continue to maintain arrangements for requesting and using 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 action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. FCS will adopt the Permanently Defueled Emergency Action Levels (EALs) consistent with those detailed in Appendix C of Nuclear Energy Institute (NEI) 99-01, "Development of EALs for Non-Passive Reactors," Revision 6 (Reference 7), endorsed by the NRC in a letter dated March 28, 2013 (Reference 8). 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 LIC-16-0109 Attachment 1 Page 8    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 notification, by the licensee, of State and local response organizations and for notification of emergency personnel by all organizations; the content of initial and followup messages to response organizations and the public has been established; and means to provide early notification and clear instruction to the populace within the plume exposure pathway Emergency Planning Zone have been established. Per SECY-00-0145 (Reference 6), after approximately 1 year of spent fuel decay time (and as supported by the SFP analysis), the NRC staff believes an exception to the offsite EPA PAG standard is justified for a zirconium fire scenario considering the low likelihood of this event together with time available to take mitigative or protective actions between the initiating event and before the onset of a postulated fire. 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," (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 Attachment 1 Page 9    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 communications among principal response organizations to emergency personnel and to the public. See the basis for 10 CFR 50.47(b). 8 10 CFR 50.47(b)(7):  Information is made available to the 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. See the basis for 10 CFR 50.47(b). 9 10 CFR 50.47(b)(8):  Adequate emergency facilities and equipment to support the emergency response are provided and maintained. No exemption is requested. 10 10 CFR 50.47(b)(9):  Adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use. See the basis for 10 CFR 50.47(b).
LIC-16-0109 Attachment 1 Page 10    11 10 CFR 50.47(b)(10):  A range of protective actions has been developed for the plume exposure pathway EPZ for emergency workers and the public. In developing this range of actions, consideration has been given to evacuation, sheltering, and, as a supplement to these, the prophylactic use of potassium iodide (KI), as appropriate. Evacuation time estimates have been developed by applicants and licensees. Licensees shall update the evacuation time estimates on a periodic basis. Guidelines for the choice of protective actions during an emergency, consistent with Federal guidance, are developed and in place, and protective actions for the ingestion exposure pathway EPZ appropriate to the locale have been developed. In the unlikely event of a SFP accident, the iodine isotopes which contribute to an offsite dose from an operating reactor accident are not present, so potassium iodide (KI) distribution offsite would no longer serve as an effective or necessary supplemental protective action. Protective actions will be maintained for emergency workers and any offsite emergency responders who would respond to the site. The Commission responded to comments in its Statement of Considerations for the Final Rule for Emergency Planning requirements for ISFSIs and MRS facilities (60 FR 32435), and concluded that, "the offsite consequences of potential accidents at an ISFSI or a MRS would not warrant establishing Emergency 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 Attachment 1 Page 11    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 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. No exemption is requested. 13 10 CFR 50.47(b)(12):  Arrangements are made for medical services for contaminated injured individuals. No exemption is requested. 14 10 CFR 50.47(b)(13):  General plans for recovery and reentry are developed. No exemption is requested. 15 10 CFR 50.47(b)(14):  Periodic exercises are (will be) 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. No exemption is requested. 16 10 CFR 50.47(b)(15):  Radiological emergency response training is provided to those who may be called on to assist in an emergency. No exemption is requested.
LIC-16-0109 Attachment 1 Page 12    17 10 CFR 50.47(b)(16):  Responsibilities for plan development and review and for distribution of emergency plans are established, and planners are properly trained. No exemption is requested. 18 10 CFR 50.47(c)(2):  Generally, the plume exposure pathway EPZ for nuclear power plants shall consist of an area about 10 miles (16 km) in radius and the ingestion pathway EPZ shall consist of an area about 50 miles (80 km) in radius. The exact size and configuration of the EPZs surrounding a particular nuclear power reactor shall be determined in relation to local emergency response needs and capabilities as they are affected by such conditions as demography, topography, land characteristics, access routes, and jurisdictional boundaries. The size of the EPZs also may be determined on a case-by-case basis for gas cooled nuclear reactors and for reactors with an authorized power level less than 250 MW thermal. The plans for the ingestion pathway shall focus on such actions as are appropriate to protect the food ingestion pathway. FCS has developed an analysis 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. The analysis of the potential radiological impact of the postulated accident for FCS in a permanently defueled condition indicates that any releases beyond the site boundary are limited to small fractions of the EPA PAG exposure levels. According to the EPA's "Protective Action Guides and Planning Guidance for Radiological Incidents, Draft for Interim Use and Public Comment," dated March 2013 (PAG Manual), "EPZs are not necessary at those facilities where it is not possible for PAGs to be exceeded off-site."
(Reference 10). Also see the basis for 10 CFR 50.47(b).
LIC-16-0109 Attachment 1 Page 13    19 10 CFR 50 Appendix E 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. No exemption is requested.
LIC-16-0109 Attachment 1 Page 14    20 10 CFR 50 Appendix E IV. Content of Emergency Plans 1. The applicant's emergency plans shall contain, but not necessarily be limited to, information needed to demonstrate compliance with the elements set forth below, , organization for coping with radiological emergencies, assessment actions, activation of emergency organization, notification procedures, emergency facilities and equipment, training, maintaining emergency preparedness, and recovery, and onsite protective actions during hostile action. In addition, the emergency response plans submitted by an applicant for a nuclear power reactor operating license under this part, or for an early site permit (as applicable) or combined license under 10 CFR part 52, shall contain information needed to demonstrate compliance with the standards described in § 50.47(b), and they will be evaluated against those standards. Following docketing of the "Certification of Permanent Removal of Fuel from the Reactor Vessel," in accordance with 10 CFR 50.82(a)(1)(i) and (ii), FCS became a permanently shut down facility with spent fuel stored in the SFP. In the EP Final Rule (76 FR 72596, Nov. 23, 2011) (Reference 11), the NRC defined "hostile action" as, in part, an act directed toward a nuclear power plant or its personnel. This definition is based on the definition of "hostile action" provided in NRC Bulletin 2005-02. NRC Bulletin 2005-02 was not applicable to nuclear power reactors that have permanently ceased operations and have certified that fuel has been removed from the reactor vessel. The NRC excluded non-power reactors (NPRs) from the definition of "hostile action" at that time because an NPR is not a nuclear power plant and a regulatory basis had not been developed to support the inclusion of NPR in that definition. Similarly, a decommissioning power reactor or ISFSI is not a "nuclear reactor" as defined in the NRC's regulations. The following similarities between FCS and NPRs show that the FCS facility should be treated in a similar fashion as an NPR. Similar 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 Attachment 1 Page 15    21 IV.2 This nuclear power reactor license applicant shall also 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. See the basis for 10 CFR 50.47(b)(10). 22 IV.3 Nuclear power reactor licensees shall use NRC approved 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. See the basis for 10 CFR 50.47(b)(10). 23 IV.4 Within 365 days of the later of the date of the availability 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. See the basis for 10 CFR 50.47(b)(10). 24 IV.5 During the years between decennial censuses, nuclear 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 See the basis for 10 CFR 50.47(b)(10).
LIC-16-0109 Attachment 1 Page 16    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 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. See the basis for 10 CFR 50.47(b)(10). 26 IV.7 After an applicant for a combined license under part 52 of this chapter receives its license, the licensee shall conduct at least one review of any changes in the population of its EPZ at least 365 days prior to its scheduled fuel load. The licensee shall estimate EPZ permanent resident population changes No exemption is requested. FCS is not an applicant for a combined license. Therefore, this regulation is not applicable to FCS and an exemption is not necessary.
LIC-16-0109 Attachment 1 Page 17    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 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: No exemption is requested. 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 Attachment 1 Page 18    29 A.2. A description of the onsite emergency response 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. No exemption is requested. 30 A.3. A description, by position and function to be performed, of the licensee's headquarters personnel who will be sent to the plant site to augment the onsite emergency organization. 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 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 persons within the licensee organization who will be responsible for making offsite dose projections and a description of how these projections will be made and the results transmitted to State and local authorities, NRC, and other appropriate governmental entities. FCS has developed an analysis 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. FCS will maintain the capability to determine if a radiological release 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 Attachment 1 Page 19    are responsible for deciding what, if any, protective actions should be taken. 32 A.5. Identification, by position and function to be performed, of other employees of the licensee with special qualifications for coping with emergency conditions that may arise. Other persons with special qualifications, such as consultants, who are not employees of the licensee and who may be called upon for assistance for emergencies shall also be identified. The special qualifications of these persons shall be described. As indicated by the FCS adiabatic heatup analysis, the time available to initiate compensatory actions in the event of a loss of SFP cooling or inventory precludes the need to identify and describe the special qualifications of these individuals in the emergency plan. The number of staff at FCS when it is in the permanently defueled state will be small but will be commensurate with the need to 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 support of the licensee's emergency organization. No exemption is requested. 34 A.7. By June 23, 2014, identification of, and a description of the assistance expected from, appropriate State, local, and Federal agencies with responsibilities for coping with emergencies, including hostile action at the site. For purposes of this appendix, "hostile action" is defined as an act directed toward a nuclear power plant or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. A decommissioning power reactor has a low likelihood of a credible accident resulting in radiological releases requiring offsite protective measures. For this reason and those described in the basis for Section IV.1 of 10 CFR Part 50, Appendix E, a decommissioning power reactor is not a facility that falls within the definition of "hostile action." Similarly, for security, risk insights can be used to determine which targets are important to protect against sabotage. A level of security commensurate with the consequences of a sabotage event is required and is evaluated on a site-specific basis. The severity of the consequences declines as fuel ages, and over time, the underlying LIC-16-0109 Attachment 1 Page 20    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 for planning for, ordering, and controlling appropriate protective actions, including evacuations when necessary. Offsite emergency measures are limited to support provided by local police, fire departments, and ambulance and hospital services as 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 licensees, a detailed analysis demonstrating that on shift personnel assigned emergency plan implementation functions Responsibilities of the on-shift and emergency response personnel will be detailed in the Permanently Defueled Emergency Plan and implementing procedures and will be regularly tested through drills LIC-16-0109 Attachment 1 Page 21    are not assigned responsibilities that would prevent the timely performance of their assigned functions as specified in the emergency plan. and exercises, audited, and inspected by FCS and the NRC. The duties of the on-shift personnel at a decommissioning reactor facility 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 B.1. The means to be used for determining the magnitude of, and for continually assessing the impact of, the release of radioactive materials shall be described, including emergency FCS will develop EALs consistent with the Permanently Defueled EALs detailed in Appendix C of NEI 99-01, Revision 6 (Reference 7). FCS proposes to continue to review EALs with the State of Nebraska and Washington County on an annual basis. However, based upon the reduced scope of EALs for the permanently LIC-16-0109 Attachment 1 Page 22    action levels that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other Federal agencies, and the emergency action levels that are to be used for 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. defueled facility, the scope of the annual review of EALs is expected to be limited (i.e., informal mailings, etc.). Also see the basis for Section IV.1 for the justification from the requirements in Appendix E related to "hostile action." 38 B.2. A licensee desiring to change its entire emergency action 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. No exemption is requested. 39 C. Activation of Emergency Organization C.1. The entire spectrum of emergency conditions that involve the alerting or activating of progressively larger segments of the total emergency organization shall be described. The The Permanently Defueled EALs, developed consistent with Appendix C of NEI 99-01, Revision 6 (Reference 7), will be adopted, as previously described. This scheme eliminates the Site Area Emergency and General Emergency event classifications.
LIC-16-0109 Attachment 1 Page 23    communication steps to be taken to alert or activate emergency personnel under each class of emergency shall be described. Emergency action levels (based not only on onsite and offsite radiation monitoring information but also on readings from a number of sensors that indicate a potential emergency, such as the pressure in containment and the response of the Emergency Core Cooling System) for notification of offsite agencies shall be described. The existence, but not the details, of a message authentication scheme shall be noted for such agencies. The emergency classes defined shall include: (1) Notification of unusual events, (2) alert, (3) site area emergency, and (4) general emergency. These classes are further discussed in NUREG-0654/FEMA-REP-1. Additionally, the need to base EALs on containment parameters is no longer appropriate. The EAL scheme presented in NEI 99-01, Revision 6 was endorsed by the NRC in a letter dated March 28, 2013 (ML 12346A463). No offsite protective actions are anticipated to be necessary, so classification above the Alert level is no longer required. In the event of an accident at a defueled facility that meets the conditions for relaxation of emergency planning requirements, there will be available time for event mitigation, and if necessary, implementation of offsite protective actions using a comprehensive approach to emergency planning. 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. Containment parameters will not provide an indication of the 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 Attachment 1 Page 24    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 establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and shall promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. Licensees shall not construe these criteria as a grace period to attempt to restore plant conditions to avoid declaring an emergency action due to an emergency action level that has been exceeded. Licensees shall not construe these criteria as preventing implementation of response actions deemed by the licensee to be necessary to protect public health and safety provided that any delay in declaration does not deny the State and local authorities the opportunity to implement measures necessary to protect the public health and safety. In the Proposed Rule (74 FR 23254) (Reference 13) to amend certain emergency planning requirements for 10 CFR Part 50, the NRC asked for public comment on whether the NRC should add requirements for non-power reactor licensees to assess, classify, and declare an emergency condition within 15 minutes and promptly declare an emergency condition. The NRC received several comments on these issues. The NRC believed there may be a need for the NRC to be aware of security-related events early on so that an assessment of the likelihood that the event is part of a larger coordinated attack can be made. However, the NRC determined that further analysis and stakeholder interactions are needed prior to changing the requirements for non-power reactor licensees. Therefore, the NRC did not include requirements in the 2011 EP Final Rule (Reference 11) for non-power reactor licensees to assess, classify, and declare an emergency condition within 15 minutes and promptly declare an emergency condition.
LIC-16-0109 Attachment 1 Page 25    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 Attachment 1 Page 26    for discussion on the similarity between a permanently defueled reactor and a non-power reactor. 41 D. Notification Procedures 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. See the basis for 10 CFR 50.47(b) and 10 CFR 50.47(b)(10). 42 D.2. Provisions shall be described for yearly dissemination to 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. See the basis for Section IV.D.1. 43 D.3. A licensee shall have the capability to notify responsible State and local governmental agencies within 15 minutes after declaring an emergency. The licensee shall demonstrate that While the capability needs to exist for the notification of offsite government agencies within a specified time period, previous exemptions have allowed for extending the State and local LIC-16-0109 Attachment 1 Page 27    the appropriate governmental authorities have the capability to make a public alerting and notification decision promptly on being informed by the licensee of an emergency condition. Prior to initial operation greater than 5 percent of rated thermal power of the first reactor at a site, each nuclear power reactor licensee shall demonstrate that administrative and physical means have been established for alerting and providing prompt instructions to the public within the plume exposure pathway EPZ. The design objective of the prompt public alert and notification system shall be to have the capability to essentially complete the initial alerting and initiate notification of the public within the plume exposure pathway EPZ within about 15 minutes. The use of this alerting and notification capability will range from immediate alerting and notification of the public (within 15 minutes of the time that State and local officials are notified that a situation exists requiring urgent action) to the more likely events where there is substantial time available for the appropriate governmental authorities to make a judgment whether or not to activate the public alert and notification system. The alerting and notification capability shall additionally include administrative and physical means for a backup method of public alerting and notification capable of being used in the event the primary method of alerting and notification is unavailable during an emergency to alert or notify all or portions of the plume exposure pathway EPZ population. The backup method shall have the capability to alert and notify the public within the plume exposure pathway EPZ, but does not need to meet the 15-minute design government agencies' notification time up to 60 minutes based on the site-specific justification provided. FCS proposes to complete emergency notification to the State of Nebraska within 60 minutes after an emergency declaration or a change in classification. This timeframe is consistent with the 10 CFR 50.72(a)(3) notification to the NRC and is appropriate because in the permanently defueled condition, the rapidly developing scenarios associated with events initiated during reactor power operation are no longer credible and there is no need for State or local response organizations to implement any protective actions. 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 60-minute notification to the State of Nebraska. Emergency management officials with both states have agreed that the proposed notification to Nebraska within 60 minutes is appropriate. FCS analyses demonstrate that 530 days (1 year, 165 days) after permanent cessation of power operations, no remaining postulated accidents at FCS will result in radiological releases requiring offsite 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 Attachment 1 Page 28    objective for the primary prompt public alert and notification system. When there is a decision to activate the alert and notification system, the appropriate governmental authorities 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. emergency management plan. Therefore, there is no need to maintain an Alert and Notification System. Also see the basis for 10 CFR 50.47(b) and 10 CFR 50.47(b)(10). 44 D.4. If FEMA has approved a nuclear power reactor site's alert and notification design report, including the backup alert and 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. See the basis for Section IV.D.3 regarding the alert and notification system requirements. 45 E. Emergency Facilities and Equipment Adequate provisions shall be made and described for emergency facilities and equipment, including: No exemption is requested.
LIC-16-0109 Attachment 1 Page 29    E.1. Equipment at the site for personnel monitoring; 46 E.2. Equipment for determining the magnitude of and for continuously assessing the impact of the release of radioactive materials to the environment; No exemption is requested. 47 E.3. Facilities and supplies at the site for decontamination of onsite individuals; No exemption is requested. 48 E.4. Facilities and medical supplies at the site for appropriate emergency first aid treatment; No exemption is requested. 49 E.5. Arrangements for medical service providers qualified to handle radiological emergencies onsite; No exemption is requested. 50 E.6. Arrangements for transportation of contaminated injured individuals from the site to specifically identified treatment facilities outside the site boundary; No exemption is requested. 51 E.7. Arrangements for treatment of individuals injured in support of licensed activities on the site at treatment facilities outside the site boundary; No exemption is requested. 52 E.8.a(i) A licensee onsite technical support center and an emergency operations facility from which effective direction can be given and effective control can be exercised during an emergency; FCS analyses demonstrate that 530 days (1 year, 165 days) after permanent cessation of power operations, no remaining postulated accidents at FCS will result in radiological releases requiring offsite 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 Attachment 1 Page 30    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 operational support center; NUREG-0696, "Functional Criteria for Emergency Response 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 Attachment 1 Page 31    54 E.8.b. For a nuclear power reactor licensee's emergency operations facility required by paragraph 8.a of this section, either a facility located between 10 miles and 25 miles of the nuclear power reactor site(s), or a primary facility located less than 10 miles from the nuclear power reactor site(s) and a backup facility located between 10 miles and 25 miles of the nuclear power reactor site(s). An emergency operations 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: In accordance with paragraph 8.e. the requirements of paragraph 8.b.(1) - (5) do not apply to the FCS EOF because it was an approved facility prior to December 23, 2011. However, the exemption is requested to clearly reflect that the requirement no longer applies to FCS in a permanently shut down and defueled condition. See also basis for 10 CFR 50.47(b)(3). 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 Attachment 1 Page 32    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 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; See the basis for 10 CFR 50.47(b)(3). 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 Attachment 1 Page 33    63 E.8.d. For nuclear power reactor licensees, an alternative 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. See the basis for Section IV.1 regarding hostile action. 64 E.8.e. A licensee shall not be subject to the requirements of paragraph 8.b of this section for an existing emergency operations facility approved as of December 23, 2011; See the basis for 10 CFR 50.47(b)(3) and Appendix E, Section IV.E.8.b. 65 E.9. At least one onsite and one offsite communications system; each system shall have a backup power source. All communication plans shall have arrangements for emergencies, including titles and alternates for those in charge at both ends of the communication links and the See the basis for 10 CFR 50.47(b) and (b)(10). FCS will maintain communications with the State of Nebraska, Washington County, and the NRC. The onsite response facilities will be combined into a single facility, as described in IV.E.8.a(ii).
LIC-16-0109 Attachment 1 Page 34    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 response organizations. Such communications systems shall be tested annually. No exemption is requested. 67 E.9.c. Provision for communications among the nuclear power reactor control room, the onsite technical support center, and the emergency operations facility; and among the nuclear facility, the principal State and local emergency operations centers, and the field assessment teams. Such communications systems shall be tested annually. FCS analyses demonstrate that 530 days (1 year, 165 days) after permanent cessation of power operations, no remaining postulated accidents at FCS will result in radiological releases requiring offsite 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 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 Attachment 1 Page 35    68 E.9.d. Provisions for communications by the licensee with NRC Headquarters and the appropriate NRC Regional Office Operations Center from the nuclear power reactor control room, the onsite technical support center, and the emergency operations facility. Such communications shall be tested monthly. The functions of the Control Room, EOF, TSC, and OSC may be combined into one or more locations due to the smaller facility staff and the greatly reduced interaction required with State and local emergency response facilities. An onsite facility will continue to be maintained, from which effective command and control will be 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 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: No exemption is requested. 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 Attachment 1 Page 36    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 shall be made available to local services personnel; e.g., local emergency services/Civil Defense, local law enforcement personnel, local news media persons. Because there will no longer be any expected actions that must be taken by the public during an emergency, it is no longer necessary to pre-plan the dissemination of this information to the public or to provide radiological orientation training to local news media persons.
LIC-16-0109 Attachment 1 Page 37    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 emergency preparedness exercises as follows: Exercises shall test the adequacy of timing and content of implementing procedures and methods, test emergency equipment and communications networks, test the public alert and notification system, and ensure that emergency organization personnel are familiar with their duties.3 FCS analyses demonstrate that 530 days (1 year, 165 days) after permanent cessation of power operations, no remaining postulated accidents at FCS will result in radiological releases requiring offsite 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 emergency management plan. Therefore, the public alert and 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 licensee, State, and local emergency plans as is reasonably achievable without mandatory public participation shall be conducted for each site at which a power reactor is located. Nuclear power reactor licensees shall submit exercise scenarios under § 50.4 at least 60 days before use in a full participation exercise required by this paragraph 2.a. FCS will continue to invite the State of Nebraska and Washington County to participate in the periodic drills and exercises conducted to assess their ability to perform responsibilities related to an emergency at FCS, to the extent defined by the FCS emergency plan. Because the need for offsite emergency planning is relaxed due to the low probability of the postulated accident or other credible events that would be expected to result in an offsite radioactive release that would exceed the EPA PAGs and the available time for event mitigation, no offsite emergency plans will be in place to test. The intent of submitting exercise scenarios at power reactors is to verify that licensees utilize different scenarios in order to prevent the preconditioning of responders at power reactors. For defueled sites, there are limited events that could occur and the previously routine 82 F.2.a.(i) For an operating license issued under this part, this exercise must be conducted within two years before the issuance of the first operating license for full power (one authorizing operation above 5 percent of rated power) of the first reactor and shall include participation by each State and local government within the plume exposure pathway EPZ LIC-16-0109 Attachment 1 Page 38    and each state within the ingestion exposure pathway EPZ. If the full participation exercise is conducted more than 1 year prior to issuance of an operating licensee for full power, an exercise which tests the licensee's onsite emergency plans 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. progression to General Emergency in power reactor site scenarios is not applicable to a decommissioning site. OPPD considers FCS to be exempt from F.2.a.(i) - (iii) because FCS will be exempt from the umbrella provision of Section IV.F.2.a. 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 Attachment 1 Page 39    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 exercise of its onsite emergency plan every 2 years. Nuclear power reactor licensees shall submit exercise scenarios under § 50.4 at least 60 days before use in an exercise required by this paragraph 2.b. The exercise may be included in the full participation biennial exercise required by paragraph 2.c. of this section. In addition, the licensee shall take actions necessary to ensure that adequate emergency response capabilities are maintained during the interval between biennial exercises by conducting drills, including at least one drill involving a combination of some of the principal functional areas of the licensee's onsite emergency response capabilities. The principal functional areas of emergency response include activities such as management and coordination of emergency response, accident assessment, event classification, notification of offsite authorities, assessment of the onsite and offsite impact of radiological releases, protective action recommendation development, See the basis for Section IV.F.2.a. The low probability of the postulated accident or other credible events that would result in an offsite radioactive release that would exceed the EPA PAGs and the available time for event mitigation at FCS during decommissioning render the TSC, OSC, and EOF unnecessary. The principal functions required by regulation can be performed at a single onsite location that does not meet the requirements of the TSC, OSC, or EOF. The onsite response facilities at FCS will be combined into a single facility. FCS will continue to conduct biennial exercises and will invite the State of Nebraska and local support organizations (firefighting, law enforcement, and ambulance/medical services) to participate in periodic drills and exercises to assess their ability to perform responsibilities related to an emergency at FCS, to the extent defined by the FCS emergency plan. The intent of submitting exercise scenarios for use by power reactor licensees is to check that licensees utilize different scenarios in LIC-16-0109 Attachment 1 Page 40    protective action decision making, plant system repair and mitigative action implementation. During these drills, activation of all of the licensee's emergency response facilities (Technical Support Center (TSC), Operations Support Center (OSC), and the Emergency Operations Facility (EOF)) would 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. order to prevent the preconditioning of responders at power reactors. In FCS's permanently shut down and defueled condition, there are limited events that could occur and the previously routine progression to General Emergency in scenarios is not applicable to a decommissioning site. 86 F.2.c. Offsite plans for each site shall be exercised biennially 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: See the basis for Section IV.F.2.a. 87 F.2.c.(1) Conduct an exercise biennially of its onsite emergency plan; LIC-16-0109 Attachment 1 Page 41    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 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 See the basis for Section IV.2.
LIC-16-0109 Attachment 1 Page 42    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 located within the plume exposure pathway EPZ to participate in the licensee's drills when requested by such State or local government. See the basis for Section IV.2. 94 F.2.f. Remedial exercises will be required if the emergency plan is not satisfactorily tested during the biennial exercise, such that NRC, in consultation with FEMA, cannot (1) find reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency or (2) determine that the Emergency Response Organization (ERO) has maintained key skills specific to emergency response. The extent of State and local participation in remedial exercises must be sufficient to show that appropriate corrective measures have been taken regarding the elements of the plan not properly tested in the previous exercises. FEMA is responsible for evaluating the adequacy of an offsite response exercise. No action is expected from State or local government organizations in response to an event at a decommissioning site other than receiving notification of the emergency and firefighting, law enforcement, and ambulance/medical response services. Letters of Agreement will continue to be in place for those services. Offsite response organizations will continue to implement actions to protect the health and safety of the public as they would at any other industrial site. 95 F.2.g. All exercises, drills, and training that provide 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. No exemption is requested.
LIC-16-0109 Attachment 1 Page 43    96 F.2.h. The participation of State and local governments in an 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. No exemption is requested. 97 F.2.i. Licensees shall use drill and exercise scenarios that provide reasonable assurance that anticipatory responses will not result from preconditioning of participants. Such scenarios for nuclear power reactor licensees must include a wide spectrum of radiological releases and events, including hostile action. Exercise and drill scenarios as appropriate must emphasize coordination among onsite and offsite response organizations. At FCS, there will be limited events that could occur that could result in radiological releases that exceed the EPA PAGs and the previously routine progression to General Emergency in power reactor site scenarios will not be applicable. Therefore, FCS does not expect to demonstrate response to a wide spectrum of events. 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 basis for Section IV.1 regarding hostile action. 98 F.2.j. The exercises conducted under paragraph 2 of this section by nuclear power reactor licensees must provide the opportunity for the ERO to demonstrate proficiency in the key skills necessary to implement the principal functional areas of emergency response identified in paragraph 2.b of this section. Each exercise must provide the opportunity for the ERO to demonstrate key skills specific to emergency response duties in the control room, TSC, OSC, EOF, and joint information center. Additionally, in each eight calendar See the basis for Section IV.F.2. Periodic drills and exercises will be completed to demonstrate ERO proficiency in key skills necessary to implement the principal functional areas of emergency response as applicable for the permanently defueled plant status. Critiques will follow each drill or exercise activity. FCS will continue to invite the State of Nebraska and local support organizations to participate in the periodic drills and exercises to assess their ability to perform responsibilities LIC-16-0109 Attachment 1 Page 44    year exercise cycle, nuclear power reactor licensees shall vary the content of scenarios during exercises conducted 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. related to an emergency at FCS to the extent defined by the FCS emergency plan. 99 G. Maintaining Emergency Preparedness 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. No exemption is requested.
LIC-16-0109 Attachment 1 Page 45    100 H. Recovery 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. No exemption is requested. 101 I. Onsite Protective Actions During Hostile Action 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 licensee's emergency plan. See the basis for Section IV.1. 102 10 CFR 50 Appendix E  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 applicant's or licensee's detailed implementing procedures for its emergency plan shall be submitted to the Commission as specified in § 50.4. No exemption is requested.
LIC-16-0109 Attachment 1 Page 46    103 10 CFR 50 Appendix E  VI. Emergency Response Data System 1. The Emergency Response Data System (ERDS) is a direct near real-time electronic data link between the licensee's onsite computer system and the NRC Operations Center that provides for the automated transmission of a limited data set 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.The regulation that identifies the requirement to maintain the Emergency Response Data System (ERDS) is not applicable to nuclear power facilities that are permanently shut down. When FCS is permanently defueled, this system will no longer be necessary to transmit safety system parameter data. No exemption is requested because this change in the ERDS data requirement is identified in 10 CFR Part 50, Appendix E, Section VI.2.
LIC-16-0109 Attachment 1 Page 47    104 10 CFR 50 Appendix E  Footnotes 3, 4, 5, and 6 are proposed for exemption. OPPD considers FCS to be exempt from Footnotes 3, 4, 5, and 6 because FCS will be exempt from the umbrella provisions of Section F.2.
LIC-16-0109 Attachment 1 Page 48 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 Attachment 1 Page 49    1 Applicable design DBAs (i.e., fuel handling accident in the spent fuel storage facility, waste gas system release, and cask handling accident if the cask handling system is not licensed as single-failure-proof) (Indicates that any radiological release would not exceed the limits of EPA PAGs at EAB); The postulated accident that will remain applicable to FCS and could contribute to dose upon implementation of the requested exemptions is the fuel handling accident (FHA) in the Auxiliary Building, where the SFP is located. The results of the analysis indicate that the dose at the EAB would not exceed the EPA PAGs within 10 days after permanent cessation of power operations. This analysis is described in Section 4.2. 2 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); FCS performed an analysis that conservatively evaluates the length of time (number of hours) it takes for uncovered spent fuel assemblies in the SFP to reach the temperature at which the zirconium cladding would fail. Based on the limiting fuel assembly for decay heat and adiabatic heat up analysis, at 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 spent fuel cooling is lost. This analysis is described in Section 4.3 and is included in Attachment 2. 3 Loss of SFP water inventory resulting in radiation exposure at the EAB and control room; (Indicates that any release is less than EPA PAGs at EAB); and FCS performed an analysis to determine the offsite radiological impact of a complete loss of SFP water. It was determined that the gamma radiation dose rate at the EAB would be limited to small fractions of the EPA PAG exposure levels. This analysis is described in Section 4.4 and is included in Attachment 3. 4 Considering the site-specific seismic hazard, either an evaluation demonstrating a high confidence of a low-probability (less than 1 x 10-5 per year) of seismic failure of the spent fuel storage pool structure or an analysis demonstrating the fuel has decayed sufficiently that natural air flow in a completely drained pool would maintain peak cladding temperature below 565 degrees Celsius (the point of incipient cladding damage) (Indicates that any release is less than EPA PAGs at EAB). 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 Near-Term Task Force (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 high confidence of a low-probability of seismic failure (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 LIC-16-0109 Attachment 1 Page 50    integrity evaluation, to meet the SFP Evaluation Guidance (Item 9 in Enclosure 1 of the NRC's 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 described in NUREG-1738 are 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. IDCs and SDAs are addressed in Section 4.5 and Tables 4 and
: 5. 6 Verify that the licensee presents a determination that there is sufficient resources and adequately trained personnel available on-shift to initiate mitigative actions within the 10-hour minimum time period that will prevent an offsite radiological release that exceeds the EPA PAGs at the EAB. The onsite restoration plans for repair of the SFP cooling system and to provide makeup water to the SFP are incorporated into FCS procedures. There are multiple ways to initiate mitigative actions and add makeup water to the SFP within the 10-hour minimum time period with or without entry to the SFP floor. Refer to SDA 2 in Table 5. 7 Verify that mitigation strategies are consistent with that required by the Permanently Defueled Technical Specifications or by retained license conditions. 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 were developed in response to 10 CFR 50.54(hh)(2) and 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. Refer to SDA 4 in Table 5.
LIC-16-0109 Attachment 1 Page 51      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). 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 Attachment 1 Page 52      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.
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 Attachment 1 Page 53    NUREG-1738 contains the results of the NRC staff's 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 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 Attachment 1 Page 54 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 Attachment 1 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 NRC's 50.54(f) letter), thus supporting SDA No. 6 of NUREG-1738.
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 Attachment 1 Page 56    1 Cask drop analyses will be performed or single failure-proof cranes will be in use for handling of heavy loads (i.e.,
phase II of NUREG-0612 will be implemented). The FCS Auxiliary Building crane with its main hook is used to move the spent fuel casks into the SFP. The capacity of the crane lifting system using the main hook is 106 tons to accommodate the ISFSI spent fuel casks. The design of this lifting system is single failure-proof. Therefore, the likelihood of dropping the spent fuel casks into the SFP or in the Auxiliary Building is extremely 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 FCS's 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 Attachment 1 Page 57    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 will be in place to ensure that onsite and offsite resources can be brought to bear during an event. FCS procedures are in place to ensure onsite and offsite resources can be brought to bear during an event, including: AOP-01, Acts of Nature, AOP-06, Fire Emergency, AOP-37, Security Event, AOP-31, 161kV Grid Malfunctions, AOP-38, Blair Water Main Trouble, and OCAG-1/2/3/4, 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 Attachment 1 Page 58    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 communication between onsite and offsite organizations during severe weather and seismic events. FCS maintains the following procedures to provide guidance for establishing and maintaining communications between offsite agencies and the onsite ERO during severe 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 developed which will include access to portable pumps and emergency power to supplement onsite resources. The plan would principally identify organizations or suppliers where offsite resources could be obtained in a timely manner. The following procedures provide guidance for access to offsite resources including a listing of providers, contact information and resources, as well as mitigation strategies for SFP damage and water supply: 1. OP-AA-201-010-1001, B.5.b Mitigating Strategies Equipment Expectations 2. OCAG-1, Operational Contingency Action Guideline ***10 CFR 2.390***** 3. OCAG-2, Operational Contingency Action Guideline for Dam Failure Predicted Flood Greater than 1014 Feet MSL ***10 CFR 2.390***
LIC-16-0109 Attachment 1 Page 59    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 readouts and alarms in the control room (or where personnel are stationed) for SFP temperature, water level, and area radiation levels. The FCS design is consistent with this IDC, which includes multiple readouts and alarms for SFP temperature, water level, and area radiation monitors. Additionally, FCS maintains procedure AOP-36, Loss of Spent Fuel Pool Cooling, to guide the response in the event that cooling or water level is lost in the SFP. The Control Room and local operating panel (AI-100) are also 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 leading to fuel uncovery in the event of seal failure shall be self-limiting to leakage or otherwise engineered so that drainage cannot occur. The design of the SFP and its cooling system and connections to the pool are such that the SFP cannot be drained below the level of the top of the stored fuel when in its storage rack. Additionally, drainage grooves are provided behind the stainless steel liner to permit detection of any liner leakage. On a quarterly basis, the liner leakage is evaluated per preventative 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 Attachment 1 Page 60    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 7'4" wide at the top and tapers to 4'6" 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 reduce the likelihood of rapid draindown events will include (1) prohibitions on the use of pumps that lack adequate siphon protection or (2) controls for pump suction and discharge points. The functionality of anti-siphon devices will be periodically verified. 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). The SFP is designed to reduce the likelihood of rapid drain down events. Specifically, the SFP cooling lower suction, which is still well above the top of stored fuel, has a manual valve that is 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 Attachment 1 Page 61    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 place to provide repair of the SFP cooling systems or to provide access for makeup water to the SFP. The plan will provide for remote alignment of the makeup source to the SFP without requiring entry to the refuel floor. FCS practices align with this IDC. The onsite restoration plan for repair of the SFP cooling system and for makeup water to the SFP are incorporated into procedures AOP-36, Loss of Spent Fuel Pool Cooling, and OCAG-1, Operational Contingency Action Guideline ***10 CFR 2.390**** (also OCAGs-2/3/4) AOP-36 establishes multiple makeup sources from onsite that include: 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 SFP operations that have the potential to rapidly decrease SFP inventory.
These administrative controls may require additional operations or management review, management physical presence for designated operations or administrative limitations such as restrictions on heavy load movements. 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). 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 LIC-16-0109 Attachment 1 Page 62    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 FCS's 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 Attachment 1 Page 63    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 pool makeup system components will be performed and administrative controls for equipment out of service will be implemented to provide added assurance that the components would be available, if needed. FCS practices align with this IDC. SFP makeup strategies are outlined in AOP-36 and OCAG documents as previously outlined. One of the primary makeup strategies is from the Fire Protection System. The Fire Protection System has redundant pumping capability and power supplies to ensure alternate SFP makeup functions. The system is supplied by redundant pumps, one diesel driven and one electric motor driven, each design rated for 2000 gpm at 125 psig discharge pressure. Both pumps take suction from the plant intake cooling water structure 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 Attachment 1 Page 64    additional clarifying information when crediting or considering available systems and equipment to fulfill SFP only shutdown conditions.
LIC-16-0109 Attachment 1 Page 65    1 Licensee's SFP cooling design will be at least as capable as that assumed in the risk assessment, including instrumentation. Licensees will have at least one motor-driven and one diesel-driven fire pump capable of delivering inventory to the SFP. The FCS design is consistent with this SDA. The SFP cooling system, SFP racks, and Auxiliary Building which houses the SFP were designed and constructed to Seismic Class I standards. The SFP cooling system heat exchangers are ultimately cooled by the Missouri River (ultimate heat sink) using redundant pumps that are supplied by redundant power sources. The pumps are normally powered from offsite power, but can be supplied from an alternate reliable power source. The makeup system is adequate to provide water at the required capacity. 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 performed at least once per shift by the operators. Procedures will be developed for and employed by the operators to provide guidance on the capability and availability of onsite and offsite inventory makeup sources and time available to initiate these sources for various loss of cooling or inventory events. FCS operators perform a walk down of the SFP systems once per shift as driven by operator rounds and by surveillance testing procedure. The onsite restoration plan for repair of the SFP cooling system and for makeup water to the SFP are incorporated into procedures AOP-36, Loss of Spent Fuel Pool Cooling, and OCAG-1, Operational Contingency Action Guideline ***10CFR2.390****. (also OCAG-2/3/4) AOP-36 establishes multiple makeup sources from onsite that include: 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.
LIC-16-0109 Attachment 1 Page 66    There are multiple ways to add makeup water to the SFP with or without entry to the pool floor. 3 Control room instrumentation that monitors SFP temperature and water level will directly measure the parameters involved. Level instrumentation will provide alarms at levels associated with calling in offsite resources and with declaring a general emergency. The FCS design is consistent with this SDC, which include multiple readouts and alarms for SFP temperature, water level, and area radiation monitors. Additionally, FCS maintains procedure AOP-36, Loss of Spent Fuel Pool Cooling, to guide the response in the event that cooling or level is lost in the SFP. The Control Room is also 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 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 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 drain paths in the SFP that could lower the pool level (by draining, suction, or pumping) more than 15 feet below the normal pool operating level and that licensee must initiate recovery using offsite sources. The normal pool operating level is 1035'-6". The SFP cooling system is normally supplied through an 8" pipe at 1034'-0". The SFP cooling system discharges through an 8" pipe. The discharge pipe enters the SFP at 1034'-0" and ends at 1031'-7" (a 90 degree elbow to direct flow downward). This discharge line is provided with a 1/2 inch hole to act as a vacuum breaker. The discharge line is also the normal borated makeup path for the SFP. The SFP cooling system has an additional lower suction point at an elevation of 1011'-4", which 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 Attachment 1 Page 67    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 7'4" wide at the top and tapers to 4'6" 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 be performed for facilities with nonsingle failure-proof systems. The analyses and any mitigative actions necessary to preclude catastrophic damage to the SFP that would lead to a rapid pool draining would be sufficient to demonstrate that there is high confidence in the facilities ability to withstand a heavy load drop. The FCS design is in alignment with this description. 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 FCS's 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 LIC-16-0109 Attachment 1 Page 68    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 successfully complete the seismic checklist provided in Appendix 2B to this study [NUREG-1738]. If the checklist cannot be successfully completed, the decommissioning plant will perform a plant specific seismic risk assessment of the SFP and demonstrate that SFP seismically 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 LIC-16-0109 Attachment 1 Page 69    induced structural failure and rapid loss of inventory is less than the generic bounding estimates provided in this study (<1 x10-5 per year including non-seismic events). 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 NRC's 50.54(f) letter), thus supporting SDA No. 6 of NUREG-1738. 7 Licensees will maintain a program to provide surveillance and monitoring of Boraflex in high-density spent fuel racks until such time as spent fuel is no longer stored in these high-density racks. The FCS spent fuel racks utilize Boral rather than Boraflex as a neutron absorbing material. There are two regions or types of racks in the SFP that both contain Boral as described in USAR 9.5. The Boral is attached as panels between each storage cell. The panels are protected with a stainless steel sheath. The technical specifications describe the surveillance test requirements, which requires poison 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 Attachment 1 Page 70    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.
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.
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 Attachment 1 Page 71 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 FCS's 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. 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.
 
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 Attachment 1 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).
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 Company's 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.
 
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 Attachment 1 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. 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.
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.
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:
LIC-16-0109 Attachment 1 Page 74    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 Attachment 1 Page 75    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. 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. No construction activities are associated with the proposed exemptions.
See the no significant hazards considerations discussion in Item (i)(1) above.
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.
 
LIC-16-0109 Attachment 1 Page 76    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 Attachment 1 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) 
 
Design Analysis Cover Sheet Page 1 CC-AA-309-1 001 Revision 8 Design Analysis I Last Page No.* p. 2 of Attachment 3 Analysis No.:' FC08104 Revision: 2 0 Minor 0 Title:' Maximum Cladding Temperature Analysis for Adiabatic Heat-Up of Spent Fuel Assembly Revision: 5 Station(s): 7 Unit No.:
* Discipline:
* Fort Calhoun Component(s):" Descrip. Code/Keyword: '0 Safety/QA Class: " System Code: '2 Structure: " Document No.: TOOl 16-048 NFA, Rev. 0 TODI16-059 NFA, Rev. 0 EA14-006, Rev. 0 EA14-012, Rev. 0 1 ME CALC Non-CQE AC-SFP Auxiliary Building CONTROLLED DOCUMENT REFERENCES '5 From/To Document No.: From USAR &sect;3.1, Rev. 10 From USAR Fig. 3.1-2 From From From/To From From Is this Design Analysis Safeguards Information?'" Yes D No Does this Design Analysis contain Unverified Assumptions? '7 Yes D This Design Analysis SUPERCEDES: '" N/A If yes, see SY -AA-1 01-1 06 If yes, A TIIAR#: ------l 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 Testing 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,.....e_e_t __ _ Pnnt Name Sign Name 1 Exelon Reviewer: 25 Tristan McDonald t """ Date _lf)-2-()-1 /, Pnnt Name Sign Name Date Independent 3'd Party Review Reqd? 20 Yes D 4'./ I ' Exelon Approver: 27 .5tE.vf. /w:ff/,..,. Lt/J Print Name I Sign Nan'le _ U_ ..... /r/-.)tJ .* JO* 2n* Date ATTACHMENT 2 CC-AA-1 03-1 003 Revision 12 Owner's Acceptance Review Checklist for External Design Analyses Page 1 of3 Design Analysis No.: _______ _ Rev: _o;.._ __ Page: -=2:.....-_ Contract#: _____ _ Release #: 128 No Question Instructions and Guidance 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?
ATTACHMENT 2 CC-AA-1 03-1003 Revision 12 Owner's Acceptance Review Checklist for External Design Analyses Page 2 of3 Design Analysis No.: ..... F ....... C;;,;;;0;,;;:;8....,1 0-..4------------Rev: .. o;...._ __ Page: ..... 3 __ No Question Instructions and Guidance Yes I No/ 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 ATTACHMENT 2 CC-AA-1 03-1 003 Revision 12 Owner's Acceptance Review Checklist for External Design Page 3 of 3 Design Analysis No.: _F_c_oo.;;:;s.:..1 Rev:...;:o;...._ __ Page: ..:4:.....--_ No Question Instructions and Guidance Yes I No I 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: --------
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. OVOID DATE FOR REV.: 10/19/2016 PREPARER: Helmut R. Kopke DATE: )CL/?/2.0/(p REVIEWER: DanielS. Elegant DATE: /olf:i /2.cJI6 APPROVER: Robert J. Peterson DATE: [cJ-JCI-/6 IDENTIFICATION OF PAGES ADDEDIREVISED/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: 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 Omaha Public Power District Fort Calhoun Station S&L Calculation 2016-10694 {FC08104) 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)
Omaha Public Power District Fort Calhoun Station 1.0 Purpose S&L Calculation 2016-10694 (FC08104) Revision 0 Page7 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 1 0 hours for uncovered spent fuel assemblies to heat-up to 900&deg;C (1652&deg;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&deg;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.
Omaha Public Power District Fort Calhoun Station 2.0 Inputs 2.1 Maximum Fuel Cladding Temperature S&L Calculation 2016-10694 (FC08104) Revision 0 PageS 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 1 0 hours is available from the time cooling is lost before the cladding temperature reaches 900&deg;C. Therefore, 900&deg;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&#x17d; 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&#x17d; 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-&deg;F Btu/lbm-&deg;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&deg;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&deg;C (86&deg;F). Both documents state that the analysis starts at the time of fuel uncovery. Using 140&deg;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&deg;F to 900&deg;C than is required to heat the assembly from 140&deg;F to 900&deg;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 Omaha Public Power District Fort Calhoun Station S&L Calculation 2016-10694 {FC08104) 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) T bl 2 2 F I A a e -: ue ssem bl G IY t eometry 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&#x17d; 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/cm3 {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.
Omaha Public Power District Fort Calhoun Station S&L Calculation 2016-10694 (FC08104) Revision 0 Page 10 Table 2-3: Uranium Dioxide Fuel Heat Capacity Temperature Volumetric Heat Capacity Mass Based Heat Capacity OF oc Btu/ft3-&deg;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 (OL T). 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&deg;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.
Omaha Public Power District Fort Calhoun Station 2.6 Heat Load S&L Calculation 2016-10694 (FC08104) Revision 0 Page 11 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&deg;C, although it is not a significant heat source below 600&deg;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.
Omaha Public Power District Fort Calhoun Station 3.0 Assumptions S&L Calculation 2016-10694 (FC08104) Revision 0 Page 12 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&deg;/o uranium 235 (pp. 48 and 108 of Ref. 4.9).
Omaha Public Power District Fort Calhoun Station 4.0 References S&L Calculation 2016-10694 (FC08104) Revision 0 Page 13 4.1 NSIRIDPR-ISG-02 Interim Staff Guidance, "Emergency Planning Exemption Requests for Decommissioning Nuclear Power Plants,'' dated May 11, 2015. ADAMS Accession No. ML 14106A057. 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. ML01 0430066. 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 Bumup PWR and BWR Reactor Models for the ORIGEN2 Computer Code," published December 1989.
Omaha Public Power District Fort Calhoun Station 5.0 Identification of Computer Programs S&L Calculation 2016-10694 (FC08104) Revision 0 Page 14 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 1 0734 under the Windows 7 Enterprise (Build 7601) operating system with Service Pack 1.
Omaha Public Power District Fort Calhoun Station 6.0 Method of Analysis 6.1 Adiabatic Heat-Up Methodology S&L Calculation 2016-10694 (FC08104) Revision 0 Page 15 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 q=p* *C P t Where: q heat generation rate (Btu/hr) p material density (lbm/ft3) V material volume (fe) cp material specific heat (Btu/lbm-&deg;F) T temperature increase (&deg;F) t heat-up time (hr) Equation 6-1 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 will be the same for both materials. Where: Xu signifies the property is for yranium dioxide Xz signifies the property is for alloy Equation 6-2 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 1 0 hour heat-up time since the heat capacity is temperature dependent. The heat-up time is calculated for discrete temperature increments of 1 00&deg;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 Omaha Public Power District Fort Calhoun Station S&L Calculation 2016-10694 (FC08104) 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&deg;C). The volume of uranium dioxide is given below. V cyl = (1t . _DP_2 ) . Nh . L . _1_ft_2_ u, 4 r 144 in2 Equation 6-4 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 = (1t . D c,o 2 -D c/ ) . N . L . 1 ft2 z,c 4 hr 144 in2 Equation 6-5 V = (1t . D t,o 2 -D u 2 J . N . L . 1 ft2 z,t 4 t 144 in2 Equation 6-6 Equation 6-7 Where: V z,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 {8 T) for this analysis is from the initial temperature of the pool, 140&deg;F (Input 2.3), to the cladding failure temperature of interest, 1652&deg;F (Section 1 and Input 2.1).
Omaha Public Power District Fort Calhoun Station S&L Calculation 2016-10694 (FC08104) 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 1 0 hours for the fuel to heat up to 900&deg;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&deg;C to 900&deg;C while NUREG-1738 shows that a 10 hour adiabatic heat up time from 30&deg;C to 900&deg;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.
Omaha Public Power District Fort Calhoun Station 7.0 Numeric Analysis S&L Calculation 2016-10694 (FC08104) Revision 0 Page 18 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.
Omaha Public Power District Fort Calhoun Station 8.0 Results and Conclusions S&L Calculation 2016-10694 (FC08104) Revision 0 Page 19 The heat-up times to 900&deg;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&deg;C (1652&deg;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&deg;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 168 performed a generic analysis that found that for PWRs, 2.5 years is expected to be the decay time needed to reach a 1 0 hour heat-up time from 30&deg;C to 900&deg;C. NUREG-1738 shows that a 10 hour heat up time to 900&deg;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&deg;C as a function of decay time.
Omaha Public Power District Fort Calhoun Station S&L Calculation 2016-10694 (FC08104) Revision 0 Page 20 Final 20 18 16 14 0 :5. p 12 0 0 en .s 10 CD E i= 8 Q. :::>> = 6 % 4 2 0 Figure 8-1: Heat-Up Time vs. Decay Time 900&deg;C (1652&deg;F) 1---& 1 0 hr to soooc ,""' , , , , , , , ,,l 'f', _, ,' , , , _4 , ,o-.,A-' I ,tr 1.453, 10.0 , r--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.
Omaha Public Power District Fort Calhoun Station A 8 1 I 2 Heated Rods per Assembl 3 Number of Guide/Instrument Tubes 4 Heated Length of Fuel 5 Heated Length of Fuel 6 7 Diameter of U02 Fuel Pellets! B Volume of U02 Fuel 9 10 Outer Diameter of Claddin 11 Inner Diameter of Claddin 12 Volume of Zirconium Alloy in Heated Rods 13 Outer Diameter of Guide/Instrument Tubes 14 Inner Diameter of Guidellnstrument Tubes 15 Volume of Zirconium Alloy in Gil Tubes 16 Total Volume of Zirconium Alloy 17 18 U02 Fuel Stack Density (* Fuel Density) 19 Density of Zirconium Alloy! 20 21 Initial Temperature 22 Final Temperature at 9oo*c 23 Total temperature Increase to 9oo*c 24 25 26 27 -30 31 33 fJ4 -"35 -36 37 38 39 Attachment 1: Adiabatic Spent Fuel Heat-Up Computation c D E F G H I I ZrAIIoy uo, 176 Rods Input 2.4 Temperature (*F) c;,,, (Btunbm-*F) c;,,u (Btunbm-*F) 5 Rods Input 2.4 31.7 0.0666 129.3 in Input 2.4 100 0.0680 0.0575 10.8 feet -in+ 12 in/ft 140 0.0689 0.0591 200 0.0701 0.0616 0.3805 jinches jlnput 2.4 300 0.0722 0.0645 1.497 ft Equation 6-4 400 0.0743 0.0667 500 0.0764 0.0683 0.440 inches Input 2.4 600 0.0785 0.0696 0.387 inches Input 2.4 700 0.0806 0.0707 0.453 n' Equation 6-5 BOO 0.0827 0.0716 1.115 inches Input 2.4 900 0.0848 0.0724 1.035 inches Input 2.4 1000 0.0869 0.0731 0.051 n' Equation 6-6 1100 0.0890 0.0737 0.504 n' Equation 6-7 1200 0.0911 0.0743 1300 0.0932 0.0748 647.6 lbm/rt* Input 2.5 1400 0.0953 0.0753 408.9 l1bm/ft3 !Input 2.2 1500 0.0974 0.0758 1520.3 0.0978 0.0759 140 *F Input 2.3 1592.3 0.2169 0.0762 1652 *F Input 2.1 1600 0.2116 0.0762 1512 *F = Initial
* Final 1652 0.1755 0.0765 1700 0.1423 0.0767 1790.3 0.0797 J 0.25 0.2 e .., 0.15 e .. .. :I: 0.1 u <;: *;:; .. 0.05 0 K ...... U02 ---0 500 L S&L Calculation 2016-10694 (FCOB104) Revision 0 Page 1 of 14 M N -----t -\ ---\ ,-c----1000 1500 2000 ---Temperature (*F) ---l l I I I I Italicized, bold, blue values are interpolated. Other values are taken from Input 2.2 (Zr Alloy) or 2.5 (UO:z). I l J Specific Heats Over Each Tem erature Increment T1 (*F) 140 200 300 400 500 600 700 800 900 T2 (*F 200 300 400 500 600 700 800 900 1000 6T(.F) 60 100 100 100 100 100 100 100 100 c,,, at T1 (Btu/lbm-*F) 0.0689 0.0701 0.0722 0.0743 0.0764 0.0785 0.0806 0.0827 0.0848 c,_, at T2 (Btu/lbm-*F) 0.0701 0.0722 0.0743 0.0764 0.0785 0.0806 0.0827 0.0848 0.0869 --c**'*"'" (Btunbm-*F) 0.0689 0.0701 0.0722 0.0743 0.0764 0.0785 0.0806 0.0827 0.0848 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 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 C,.u.mn (Btunbm-*F) 0.0591 0.0616 0.0645 0.0667 0.0683 0.0696 0.0707 0.0716 0.0724 I I I I Heat-Up Omaha Public Power District Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation A B c I D E F G H I J 40 1 W ., 1 J/sec
* 3600 seclhr + 1055.06 J/Btu CooUng Decay 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 41 (years) (Watts) (Btu/hr) (hours) (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 43 13 1.0833 4,764 16,255 8.0 0.3 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 45 15 1.25 4,291 14,641 8.9 0.3 0.5 0.5 0.5 0.6 46 16 1.33 4,086 13,942 9.4 0.3 0.5 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 48 18 1.5 3,726 12,714 10.3 0.3 0.6 0.6 0.6 0.6 49 24 2.0 2,921 9,967 13.1 0.4 0.7 0.8 0.8 0.8 50 30 2.5 2,380 8,121 16.1 0.5 0.9 1.0 1.0 1.0 51 36 3.0 2,002 6,831 19.1 0.6 1.1 1.1 1.2 1.2 52 --53 Determine the Cooling Time at which Heat-Up to Temperature of Interest= 10 hrs Heat-Up Time Interpolated Cooling Time (years) to900*c Years Days 54 (hours) 55 I 1.453 10.0 1 165 Heat-Up K Heat-Up Time fromT1 toT2 (hrs) 0.5 0.5 0.5 0.6 0.6 0.6 0.7 0.8 1.0 1.2 I ---------r L S&L Calculation 2016-10694 (FC08104) Revision o Page2 of14 M N I Heat-Up Time Heat-Up Time Heat-Up Time fromT1 toT2 from T1 toT2 fromT1 toT2 (hrs) (hrs) (hrs) 0.5 0.5 0.5 0.5 0.5 0.5 0.6 0.6 0.6 0.6 0.6 0.6 0.6 0.6 0.6 0.6 0.7 0.7 0.7 0.7 0.7 0.9 0.9 0.9 1.0 1.1 1.1 1.2 1.3 1.3 I I (I Omaha Public Power District Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation 0 p a R s T u 2 3 4 ------------------------*-*------------------1 5 6-----***-----7 8 11 12 13 I * -----------=-=----=-. -=--. ]::= __ . 1f ----------------I 18 I 19 I W I I 22 ---------------*-------------+------------------j------t-------25 1 I I 26 .g -------t------jt-28 --------------+---29 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 S&L Calculation 2016-10694 (FC08104) Revision 0 Page 3 of 14 Omaha Public Power District Fort Calhoun Station 0 p 40 Heat-Up Time Heat-Up Time fromT1 toT2 fromT1 toT2 41 (hrs) (hrs) 42 0.5 0.5 43 0.5 0.6 44 0.6 0.6 45 0.6 0.6 46 0.6 0.6 47 0.7 0.7 48 0.7 0.7 49 0.9 0.9 50 1.1 1.1 51 1.3 1.3 52 53 54 55 a Heat-Up Time from T1 toT2 (hrs) 0.5 0.6 0.6 0.6 0.7 0.7 0.7 0.9 1.1 1.3 I R Heat-Up Time fromT1 toT2 (hrs) 0.5 0.6 0.6 0.6 0.7 0.7 0.7 0.9 1.1 1.3 I I I Attachment 1: Adiabatic Spent Fuel Heat-Up Computation s T u Heat-Up Time Heat-Up Time Heat-Up Time from T1 toT2 fromT1 toT2 from T1 to T2 (hrs) (hrs) (hrs) 0.5 0.5 0.3 0.6 0.6 0.4 0.6 0.6 0.4 0.6 0.6 0.4 0.7 0.7 0.4 0.7 0.7 0.4 0.7 0.7 0.5 0.9 0.9 0.6 1.1 1.2 0.7 1.4 1.4 0.8 1 Heat-Up S&l Calculation 2016-10694 (FC08104) Revision 0 Page4 of 14 Omaha Public Power District Fort Calhoun Station Attachment 1: Adiabatic Spent Fuel Heat-Up Computation A B c D E 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()*C711214*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 1 OA2-C 11112)/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 26 27 I I 28 I 29 I Specific Heats Over Each Temperature I 30 I T1 (*F) 31 I T2 (*F) 32 6T (&deg;F) 33 Cp.z at T1 (Btu/Ibm-*F) 34 Cp.z at T2 (Btullbm-*F) 35 Cp,z.mn (BtuJlbm-*F) 36 Cp,u at T1 (Btullbm-*F) 37 Cp,u at T2 (Btu/ibm-*F) 36 Cp,u.mn (Btullbm-*F) 39 Heat-Up (Eqs) F --I I 140 200 :::F31-F30 :::VLOOKUP(F30,$G$3:$1$25,2) :::VLOOKUP(F31,$G$3:$1$25,2) =MIN(F33:F34) :::VLOOKUP(F30,$G$3:$1$25,3) :::VLOOKUP(F31,$G$3:$1$25,3) =MIN(F36:F37) S&L Calculation 2016-10694 (FC08104) Revision 0 PageS of 14 **-------------
Omaha Public Power District Fort Calhoun Station A 40 Cooling Time (months) 41 42 12 =A42/12 43 =A42+1 =A43112 44 =A43+1 =A44/12 45 =A44+1 =A45112 46 =A45+1 =A46/12 47 =A46+1 =A47/12 48 ::::A47+1 =A48/12 49 24 =A49/12 50 30 ::::A50f12 51 36 =A51/12 52 53 Determlnet 54 55 B Cooling Time (years) Attachment 1: Adiabatic Spent Fuel Heat-Up Computation c D E 1 W = 1 J/sec
* 3600 seclhr + 1055.06 Heat-Up Time to Decay Heat (Watts) Decay Heat (Btu/hr) 900"C (hours) 5043 =C42*3600/1 055.06 =SUM F42:U42 4764 =C43*3600/1 055.06 =SUM F43:U43 4515 =C44*3600/1055.06 =SUM F44:U44 4291 =C45*3600/1 055.06 =SUM F45:U45 4086 =C46*3600/1055.0S =SUM F46:U46 3898 =C47*3600/1055.06 =SUM F47:U47 3726 =C48*3600/1 055.08 =SUM F46:U48 2921 =C49*3600/1 055.06 =SUM F49:U49 2380 =C50*3600/1 055.08 =SUM F50:U50 2002 =C51*3600/1 055.06 :::SUM F51:U51 ----------Interpolated Cooling Time (years) Heat-Up Time to 900"C (hours) =B47+(B48-B47)/(E48-E47)*(E55-E47) 10 Heat-Up (Eqs) =F$32* =F$32* =F$32* =F$32* =F$32* =F$32* =F$32* =F$32*
=F$32* "'F$32* S&L Calculation 2016-10694 (FC08104) RevlsionO Page 6of14 F Heat-Up Time from T1 to T2 (hrs) $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D42 $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D43 $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D44 $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D45 $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D46 $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$047 $C$18*$C$8*F$38+$C$19*$C$16*F$35 1$048 $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D49 $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D50 $C$18*$C$8*F$38+$C$19*$C$16*F$35 /$D51 Years =FLOOR(D55,1)
Omaha Public Power District Fort Calhoun Station G I 1 2 Temperature ('F) 3 31.7 4 100 5 140 6 200 7 =G6+100 8 =G7+100 9 =G8+100 10 =G9+100 11 =G10+100 12 =G11+100 13 =G12+100 14 =G13+100 15 =G14+100 16 =G15+100 17 =G16+100 18 =G17+100 19 =G18+100 20 1520.3 21 1592.3 22 =G19+100 23 1652 24 =G22+100 25 1790.3 26 Italicized, bold, blue values are interpolated. Other values a r4 28 29 30 =F31 31 =G30+100 32 =G31-G30 33 =VLOOKUP(G30.SG$3:SIS25.2) 34 =VLOOKUP(G31,$G$3:$1$25.2) 35 =MIN(G33:G34) 36 =VLOOKUP(G30.SG$3:SIS25,3) 37 =VLOOKUP(G31,$GS3:SIS25,3) 38 =MIN(G36:G37) 39 Attachment 1: Adiabatic Spent Fuel Heat-Up Computation H I ZrAIIoy uo, C,.z (Btunbm-'F) _C,u (Btunbm-'F) 0.0666 =HS3+(HS20-HS3)/(G$20-G$3) '(G4-G $3) 0.0575 =H$3+(H$20-H$3)/(G$20.G$3
* G5-G$3 -14+(16-14 I(G6-G4)'(G5-G4 =HS3+(H$20-H$3)/(G$20-G$3) '(G6-GS3) 0.0616 =H$3+(H$20-H$3)1(G$20.G$3)'(G7-G$3) 0.0645 =H$3+(HS20-H$3)/(GS20.GS3) '(G8-G$3) 0.0667 -H$3+(H$20-H$3)/(G$20-G$3)'(G9-G$3) 0.0683 =H$3+(H$20-H$3/(GS20-G$3)'(G10-G$3) 0.0696 -HS3+{H$20-H$3)1(GS20-G$3)'(G11*GS3) 0.0707 =H$3+(H$20.H$3)/(G$20.G$3) '(G12-G$3) 0.0716 =H$3+(H$20.H$3)/(G$20-G$3)'(G13-G$3) 0.0724 -H$3+(H$20-H$3)/(G$20.G$3)'(G14--G$3) 0.0731 =H$3+(H$20-H$3)/(G$20.G$3) '(G 15-G$3) 0.0737 =H$3+(H$20-H$3)/(G$20-G$3)'(G16-G$3) 0.0743 =H$3+(H$20-H$3)/(G$20-G$3)'(G17-G$3) 0.0748 =HS3+(H$20-H$3)/(GS20.G$3) '(G 18-G $3) 0.0753 =H$3+(H$20.H$3)/(G$20.G$3) '(G19-G$3) 0.0758 0.0978 =1$19+ 1$22-1$19/(G$22-G$19 '(G20.G$19 0.2169 -I$19+(1$22-1$19)/(G$22-G$19)'(G21*G$19) H$21+(H$25-H$21)/(G$25-G$21 '(G22-G$21) 0.0762 =HS21+(H$25-H$21 I(G$25-G$21 '(G23-G$21 =122+(124--122 G24--G22 ' G23-G22 -HS21+(H$25-H$21)1(G$25-G$21)'{G24--GS21) 0.0767 0.0797 ---1--=G31 =H31 =H30+100 =130+100 =H31-H30 =131-130 =VLOOKUP(H30.$G$3:$1$25,2) =VLOOKUP(I30,SG$3:$1$25,2) =VLOOKUP(H31,$G$3:$1$25,2) =VLOOKUP(I31.$GS3:SI$25,2) =MIN(H33:H34) =MIN(/33:134) =VLOOKUP(H30,$GS3:SIS25,3) =VLOOKUP(I30,SGS3:SIS25,3) =VLOOKUP(H31.SGS3:SI$25.3) =VLOOKUP(I31,SGS3:$1$25,3) =MIN(H36:H37) =MIN(I36:137) I Heat-Up (Eqs) --I I S&L Calculation 2016-10694 (FC08104) Revision 0 Page 7 of 14 J ---------0.25 0.2 e .D 0.15 e .. .. :t: 0.1 u "' *c .. lit-.ll-0.05 0 0 -----=131 =J30+100 =J31-J30 =VLOOKUP(J30,SG$3:$1$25,2) =VLOOKUP(J31,$G$3:$1$25,2) =MIN(J33:J34) =VLOOKUP(J30.SG$3:SIS25,3) =VLOOKUP(J31,$G$3:SI$25,3) =MIN(J36:J37)
Omaha Public Power District Fort Calhoun Station G 40 41 Heat-Up Time from T1 to T2 (hrs) 42 $C$18*$C$8*G$38+$C$19*$C$16*G$35 43 $C$18*$C$8*G$38+$C$19*$C$16*G$35 44 =G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 45 =G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 48 =G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 47 =G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 48 =G$32* $C$18*$C$B*G$38+$C$19*$C$16*G$35 49 =G$32* $C$18*$C$B*G$3B+$C$19*$C$16*G$35 50 =G$32* $C$18*$C$8*G$38+$C$19*$C$16*G$35 51 =G$32* $C$18*$C$B*G$3B+$C$19*$C$16*G$35 52 53 Days 54 55 =(D55-F55)*365 1$042 /$043 =H$32* /$044 =H$32* /$045 =H$32* /$046 /$047 =H$32* /$048 ::H$32* /$049 /$050 =H$32* /$051 =H$32* -Attachment 1: Adiabatic Spent Fuel Heat-Up Computation H I Heat-Up Time from T1 to T2 (hrs) Heat-Up Time from T1 to T2 (hrs) $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 $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 $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 $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 $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 $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 $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 $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 $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 $C$18*$C$B*H$38+$C$19*$C$16*H$35)/$051 $C$18*$C$8*1$38+$C$19*$C$16*1$35 Heat-Up (Eqs) /$D42 ::J$32* /$D43 ;::J$32* /$D44 ;::J$32* /$D45 =J$32* /$046 =J$32* /$047 =J$32* /$048 =J$32* /$049 =J$32* /$D50 =J$32* /$D51 =J$32* S&L Calculation 2016-10694 (FC08104) Revision 0 PageS of 14 J Heat-Up Time from T1 to T2 (hrs) $C$18*$C$8* J$38+$C$19*$C$16* J$35 /$042 $C$18*$C$8* J$38+$C$19*$C$16* J$35 /$043 $C$18*$C$8* J$38+$C$19*$C$16* J$35 /$044 $C$18*$C$8* J$38+$C$19*$C$16* J$35 /$045 $C$18*$C$B* J$38+$C$19*$C$16* J$35 /$046 $C$18*$C$B* J$38+$C$19*$C$16* J$35 /$047 $C$18*$C$8*J$38+$C$19*$C$16*J$35 /$048 $C$18*$C$8* J$38+$C$19*$C$16* J$35 /$049 $C$18*$C$8* J$38+$C$19*$C$16* J$35 /$050 $C$18*$C$8* J$38+$C$19*$C$16* J$35 /$051 Omaha Public Power District Fort Calhoun Station K 1 2 3 4 r-4---7 r----ZrAIIoy 8 r-g* f-10 111 !-'-"-r& 13 15 r-!Z-18 200 400 22 r# 24 25 26 -28 29 30 =J31 31 =K30+100 32 =K31-K30 33 -VLOOKUP(K30,$G$3:SI$25,2) 34 =VLOOKUP(K31 .SG$3:$1$25.2) 35 -MIN(K33:K34) 36 =VLOOKUP(K30,SGS3:SIS25,3) 37 -VLOOKUP(K31 .SG$3:$1$25,3) 38 =M/N(K36:K37) 39 ------U02 I 600 Attachment 1: Adiabatic Spent Fuel Heat-Up Computation L M 1--I ---,.._ I " I "' ". 800 1000 1200 1400 1600 1800 Temperature ('F) ------j ---=K31 =L31 =L30+100 =M30+100 =L31-L30 =M31-M30 =VLOOKUP(l30,SG$3:$1$25,2) =VLOOKUP(M30,$G$3:$1$25.2) =VLOOKUP(l31 ,SG$3:$1$25,2) =VLOOKUP(M31 .SG$3:$1$25,2) =M/N(L33:L34) -MIN(M33:M34) =VLOOKUP(l30,SGS3:SIS25,3) =VLOOKUP(M30,SGS3:SIS25,3) =VLOOKUP(l31 ,SGS3:SIS25,3) =VLOOKUP(M31 ,SGS3:SI$25,3) =MIN(L36:L37) =MIN(M36:M37) (i Heat-Up (Eqs) S&L Calculation 2016-10694 (FC08104) Revision 0 Page 9 of 14 N -------'-----------2000 ------=M31 =N30+100 =N31-N30 =VLOOKUP(N30,$G$3:$1$25.2) =VLOOKUP(N31 .SG$3:$1$25,2) =M/N(N33:N34) =VLOOKUP(N30.SGS3:SIS25,3) =VLOOKUP(N31 ,SG$3:$1$25,3) =MIN(N36:N37)
Omaha Public Power District Fort Calhoun Station 40 K :e Time from T1 to T2 * ,.,.., , ........ , ........ 54 55 e. 1:;: 1:::: 1:::: lc 1:;: 1:;: 1:::: , ... ., ... lc Attachment 1: Adiabatic Spent Fuel Heat-Up Computation T M Time from T1 to T2 le Time from T1 to T2 S) !l:"l.l;i\ S35\J !l:"l.J>\1 ----------* *-----Heat-Up (Eqs) S&L Calculation 2016-10694 (FC08104) Revision 0 Page 10 of 14 N rime from toT2 --------
Omaha Public Power District Fort Calhoun Station 0 1 2 3 4 ...;.. 6 7 8 9 10 "'i1 -------12 13 14 ...!! -----------------------------17 18 19 20 21 22 -*--24 25 26 27 28 29 30 =N31 31 1100 32 =031-030 33 =VLOOKUP(030,$G$3:$1$25,2) 34 =VLOOKUP(031,$G$3:$1$25,2) 35 =MIN(033:034} 36 =VLOOKUP(030,$G$3:$1$25,3) 37 =VLOOKUP(031,$G$3:$1$25,3) 38 =MIN(036:037} 39 Attachment 1: Adiabatic Spent Fuel Heat-Up Computation p a I ----------------------------------------------.. -------------*--------*-------* -I *-------------I I =031 =P31 =P30+100 =030+100 "'P31-P30 =031-030 =VLOOKUP(P30,$G$3:$1$25,2) =VLOOKUP(Q30,$G$3:$1$25,2) =VLOOKUP(P31,$G$3:$1$25,2) =VLOOKUP(Q31,$G$3:$1$25,2) ""MIN(P33:P34} =MIN(Q33:Q34) =VLOOKUP(P30,$G$3:$1$25,3) =VLOOKUP(Q30,$G$3:$1$25,3) =VLOOKUP(P31,$G$3:$1$25,3) =VLOOKUP(Q31,$G$3:$1$25,3) =MIN(P36:P37} =MIN(Q36:Q37} Heat-Up (Eqs) S&L Calculation 2016-10694 (FC08104) Revision 0 Page 11 of 14 R --------------=031 =R30+100 =R31-R30 =VlOOKUP(R30,$G$3:$1$25,2) =VlOOKUP(R31,$G$3:$1$25,2) =MIN(R33:R34} =VlOOKUP(R30,$G$3:$1$25,3) "'VlOOKUP(R31,$G$3:$1$25,3) =MIN(R36:R37}
Omaha PubUc Power District Fort Calhoun Station 0 40 41 Heat-Up Time from T1 to T2 (hrs) 42 =0$32* 43 =0$32* 44 =0$32* $C$18*$C$8*0$38+$C$19*$C$16*0$35 45 =0$32* 46 =0$32* 47 =0$32* 48 =0$32* $C$18*$C$8*0$38+$C$19*$C$16*0$35 49 =0$32* $C$18*$C$8*0$38+$C$19*$C$16*0$35 50 :;:Q$32* $C$18*$C$8*0$38+$C$19*$C$16*0$35 51 =0$32* $C$18*$C$8*0$38+$C$19*$C$16*0$35 52 53 54 55 /$042 /$043 /$044 /$045 /$046 1$047 /$048 /$049 /$050 /$051 Attachment 1: Adiabatic Spent Fuel Heat-Up Computation p a Heat-Up Time from T1 to T2 (hrs) Heat-Up Time from T1 to T2 (hrs) =P$32* /$042 =Q$32* =P$32* /$043 =Q$32* =P$32* /$044 =Q$32* $C$18*$C$8*Q$38+$C$19*$C$16*Q$35 =P$32* /$045 cQ$32* =P$32* /$046 =Q$32* $C$18*$C$8*Q$38+$C$19*$C$16*Q$35 =P$32* /$047 =Q$32* $C$18*$C$8*Q$38+$C$19*$C$16*Q$35 =P$32* /$048 =Q$32* "'P$32* /$049 =Q$32* =P$32* /$050 =Q$32* "'P$32* /$051 =Q$32* $C$18*$C$8*Q$38+$C$19*$C$16*Q$35 *---*----.. ... Heat-Up (Eqs) /$042 =R$32* /$043 =R$32* 11$044 cR$32* /$045 =R$32* /$046 =R$32* /$047 =R$32* /$048 =R$32* /$049 =R$32* /$050 "'R$32* /$051 =R$32* S&L Calculation 2016-10694 (FC08104) Revision 0 Page 12 of 14 R Heat-Up Time from T1 to T2 (hrs) /$042 /$043 $C$18*$C$8*R$38+$C$19*$C$16*R$35 /$044 /$045 /$048 /$047 $C$18*$C$8*R$38+$C$19*$C$16*R$35 /$048 /$049 $C$18*$C$8*R$38+$C$19*$C$16*R$35 /$050 $C$18*$C$8*R$38+$C$19*$C$16*R$35 /$051 ----
Omaha PubUc Power Dlstrld Fort Calhoun Station s 1 2 3 --4 5 6 7 8 9 -11 12 13 14 15 16 -------------17 18 19 20 21 22 23 -----* 24 25 26 27 28 29 30 ::R31 31 :::530+100 32 :::531-830 33 ::VLOOKUP(S30,$G$3:$1$25,2) 34 :::VLQOKUP(S31 ,$G$3:$1$25,2) 35 ::M/N(S33:S34) 36 =VLOOKUP(S30,$G$3:$1$25,3) 37 =VLOOKUP(S31 ,$G$3:$1$25,3) 38 ::M/N(S36:S37) 39 I I I I I I Attachment 1: Adiabatic Spent Fuel Heat-Up Computation T u --------------**--* -------------------------------------------------*---. ----------::531 :::f31 :::f3Q+100 1652 =T31-T30 :::U31-U30 =VLOOKUP(T30,$G$3:$1$25,2) =VLOOKUP(U30,$G$3:$1$25,2) :::VLOOKUP(T31 ,$G$3:$1$25,2) ::VLOOKUP(U31 ,$G$3:$1$25,2) =MIN(T33:T34) ::M/N(U33:U34) =VLOOKUP(T30,$G$3:$1$25,3) =VLOOKUP(U30,$G$3:$1$25,3) :::VLOOKUP(T31 ,$G$3:$1$25,3) =VLOOKUP(U31 ,$G$3:$1$25,3) =MIN(T36:T37) =MIN(U36:U37) Heat-Up (Eqs) S&L Calculation 2016-10894 (FC08104) RevlslonO Page 13 of 14 Omaha Public Power District Fort Calhoun Station 40 s Time from T1 to T2 4 ..... M 55 I"' I'" I* 1= I"' I"' I"' I"' I Attachment 1: Adiabatic Spent Fuel Heat-Up Computation T u Time from T1 to T2 h s) Time from T1 to T2 I= 1=1 1"'1 1::1 1::1 j1" 1=1 1::1 -----* . -----{} Heat-Up (Eqs) S&L Calculation 2016-10694 (FC08104) RevlsionO Page 14of14 Omaha Public Power District Fort Calhoun Station Attachment 2: OPPD Letter NED-16-048 NFA/ TOOl 16-048 NFA S&L Calculation 2016-10694 (FC08104) Revision 0 Page 1 of 8 Omaha Public Power Oisiiict MEMORANDUM Date: From: To:
 
==Subject:==
 
==Reference:==
July 8, 2016 NED-16-048 NFA Carol Waszak Russell J. Pleskunas Transmittal of Fuels-Related Data for Adiabatic Spent Fuel Heat-Up Analysis 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 With 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 this letter is transmit 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, lJru-; At-Carol Waszak Supervisor-Nuclear Fuels and Analysis CLW/TPL cc: NED Correspondence Omaha Public Power District Fort Calhoun Station Attachment 2: OPPD Letter NED-16-048 NFA I TODI16-048 NFA Transmittal of Design Information S&L Calculation 2016-10694 (FC08104) Revision 0 Page 2 of 8 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 5) FC06800, Section 4 1 ) 08-9062240-000 6) Fort Calhoun USAR, Section 9.6.1 2) EA14-037 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 Distribution: Per RM-AA-1 01 Project File (Thomas P.luedeke) Date 8-July-2016 Use TOOl Number in any subsequent correspondence concerning input Omaha Public Power District Fort Calhoun Station Attachment 2: OPPD Letter NED-16-048 NFA I TODI16-048 NFA S&L Calculation 2016-10694 (FC08104) Revision 0 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 1, Table 3-1 water holes for guide/instrument configuration Engineering tubes. USAR Figure 3-1.2 has a depiction. ---Number of active full-length fuel rods 176 1, Table 3-1 -In assembly Number of active partial-length fuel 0 NIA Fort Calhoun does not utilize part-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 instrument 1, Table 3-1 powered rods (other than the rods tube/assembly water holes) This is a fresh fuel pellet outer diameter, and does not account Fuel pellet 0.3805 inches 1, Table 3-1 for dimensional changes due to diameter exposure. Applicable to U02, blanket, and gadolinia-bearing pellets. Cladding inner 0.387 inches This is a fresh fuel dimension, and diameter 1, Table 3-1 does not account for changes with exposure such as creep. Cladding outer 0.440 inches 1, Table 3-1 This is a fresh fuel dimension, and diameter does not account for changes with exposure such as creep.
Omaha Public Power District Fort Calhoun Station Parameter Length of active fuel for full-length rods in assembly, length of active fuel for part-length rods in assembly Inner diameter of non-powered rods Outer diameter of non-powered rods Cladding wall material Non-powered rod wall material Fuel density Fuel specific heat Cladding density Attachment 2: OPPD Letter NE0-16-048 NFA I TOOl 16-048 NFA Value Reference 129.3 inches 1, Table 3-1 -**-** --1.035 Inches 1, Table 3-1 1.115 inches 1, Table 3-1 M5TM Zirconium 1, Table 3-1 Alloy N/A -7, Table 5.1.1, 10.37 43 glee p.30 96% of theoretical 1, Table 3-1 ---See Table 2 6,p. 139 6.55 g/cm3 2, Section 4 S&L Calculation 2016-10694 (FC08104) Revision 0 Page4 of8 July 8, 2016 NED-16-048 NFA Comments Applicable to rods with and without axial blankets. Fort Calhoun does not use part-length rods. Value corresponds to both guide and instrument tubes. Fort Calhoun does not use non-powered rods. Value corresponds to both guide and Instrument tubes. Fort calhoun does not use non-powered rods. -Fort Calhoun does not use non-powered rods. Corresponds to the stack density for central zone fuel with no gadollnla (U02 only). Pellet density specification. Theoretical density = 10.96 glee, per p. 7 of Reference 5. This U02 volumetric heat capacity can be converted to a specific heat using the density of the material. This corresponds to Zr density. Per Attachment A of Reference 4, zirconium is an extremely high percentage of the alloy.
Omaha Public Power District Fort Calhoun Station Parameter Cladding specific heat Maximum single assembly heat load as a function of time following shutdown Spent fuel pool maximum normal operating temperature Attachment 2: OPPD Letter NED-16-048 NFA/ TODI16-048 NFA Value Reference See Table 3 e. p. 142-143 Not available N/A 140&deg;F 3, Section 9.6.1 S&L Calculation 2016-10694 (FC08104) Revision 0 Page 5 of8 JulyS, 2016 NED-16-048 NFA Comments This MS volumetric heat capacity can be converted to a specific heat using the density of the material. Parameter Is not currently available and will need to be generated. -**.--Table 2: Heat Capacity for U02 Fuel Temperature (&deg;F) Volumetric Heat Capacity (Btuift3-&deg;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 Omaha Public Power District Fort Calhoun Station Attachment 2: OPPD Letter NED-16-048 NFA I TODI16-048 NFA S&L Calculation 2016-10694 (FC08104) Revision 0 Page 6 of8 July a. 2016 NED-16-048 NFA Temperature (&deg;F) Volumetric Heat Capacity (Btuift3-&deg;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 Omaha Public Power District Fort Calhoun Station Attachment 2: OPPD Letter NED-16-048 NFA I TOOl 16-048 NFA S&L Calculation 2016-10694 (FC08104) Revision 0 Page 7 of8 JulyS, 2016 NED-16-048 NFA Temperature (&deg;F) Volumetric Heat Capacity (Btulft3-&deg;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 Omaha Public Power District Fort Calhoun Station Attachment 2: OPPD Letter NED-16-048 NFA I TODI16-048 NFA Table 3: Volumetric Heat Capacity for M5&#x17d; Cladding S&L Calculation 2016-10694 (FC08104) Revision 0 Page 8 of8 July 8, 2016 NED-16-048 NFA Temperature COF) Volumetric Heat Capacity (Btufft3-&deg;F) 31.7 27.23 1520.3 39.98 1592.3 88.69 1790.3 32.57 2420.3 35.87 Omaha Public Power District Fort Calhoun Station Attachment 3: OPPD TODI16-059 NFA Transmittal of Design Information S&L Calculation 2016-10694 (FC08104) Revision 0 Page 1 of2 CC-FC-31 0-F-01 Rev.O Transmittal of Design lnfonnation (TOOl) Refer to CC-AA-31 0 for requirements TOOl No.: Revision:.=--0 __ _ __ _ 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 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 TOOl Approver OJM...f. rJlAkfJ_. Distribution: Per RM-AA-101 Project File ----Date \O I\" / "Z.. '" Date \1) I Jo { 2.Dt\e Use TOOl Number in any subsequent correspondence concerning input  : Omaha Public Power District Fort Calhoun Station OPPD TODI16-059 NFA October 10, 2016 NED-16-059 NFA Mr. Robert Peterson Senior Manager* Nuclear Plant Analysis Sargent & Lundy 55 East Monroe Street Chicago, IL 60603 S&L Calculation 2016-10694 (FC08104) Revision 0 Page 2 of 2
 
==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 
 
, 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 r-rom From Is this Design Analysis Safeguards Information? 16 YesO No. If yes, see SY-AA-101-106 Does this Design Analysis contain Unverified Assumptions? 17 YesO No. If yes, ATI/AR#: 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. 0 cill &JsS_IdJnaA.-Ia /_rfJ.CJ IL Print Name / Sign Namll'=' Oat<# Method of Review: 21 Detailed Review [gl D Testing D Reviewer: 22 Steve Gebers, CHP /1/zbNL Print Name Sign Name' Date Review Notes: '' Independent Peer review 13 (For External Analyses Only) External Approver: " Print Name Sign Name Date Exelon Reviewer: " Print Name Sign Name Date Independent J*d Party Review Reqd? ,. YesD Wo.HtL--1d Exelon Approver:" Ca. (b I Wa_514L Print Name Sian Name L Date 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 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 Attachment 8: 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 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 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 Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 6 of269 Table 7-3 ............................................................................................................................................. 87 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).
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 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 Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 1 0 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 1 083FT 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 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 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 1 b. 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.
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/cm3 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/cm3) and 224 lbs/ft3 for type C concrete (equivalent to 3.588 grams/cm3). 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/cm3 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.
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 1 025' 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."
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 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 1 0-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.
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 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 ).
Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 18 of 269 5.3 Identification of Computer Code MCNP6&#x17d; 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 year effort to merge the MCNP5&#x17d; [X-503] and MCNPX&#x17d; [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.
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) / / / $ / / *c. / I v4tr I "ilcc .ff' / -9c / * &'/ / / / / // ,....._ SITE BOUNOERY ( I I I l 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 1 080' 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 1 038' (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). I 1 I I I I ' I I Figure 6.1.1-2 FCS Plant Layout with EAB SAR F ure 1.2-1, Reference 14. : i ;.. _______ _ 1 II L .. 11 ___ .J '.: II 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 Fort Calhoun Station RoomT/ (Control Room) Room69 CALCULATION SHEET Figure 6.1.1-4 FCS Primary Operating Floor Plan (FCS Drawing 11405-A-8 Reference 14.4) FC08513 Revision 0 Page 22 of 269 Rm2S 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 dose rate conversion factors from Reference 15 and presented in Reference 1 are used. The values have been converted from Rem/hr/(p/cm2-s) to mRem/hr/(p/cm2-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 c 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 MeV MeV 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 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 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/cm3 (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 cm3) for both Class A and B concrete not accounting for the rebar and 225 lbs/ft3 (3.6032 gl cm3) 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/cm3 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.
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) Element (MCNP ID) Model Composition (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/cm2) 2.35 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) Element (MCNP ID) Model Composition (Atom Density) N (7014) 0.000039 0 (8016) 0.000011 Density (g/cm2) 0.001205 Materials M4 and M5 The interior door of the railroad siding, door 1 004-1 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 1 004-1 A construction is of 20-guage steel and Door 1 004-1 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) Element (MCNP ID) Model Composition (Atom Density) c (6000) 0.001960 Fe (26000) 0.083907 Density (g/cm2) 7.82 Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 28 of 269 The specification of insulation for Door 1 004-1A 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) Element (MCNP ID) Model Composition (Atom Density) H (1001) 0.001160 c (6000) 0.001740 N (7014) 0.000232 0 (8016) 0.000232 Density (g/cm2) 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.
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 Component Material Quantity Weight /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 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 Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 31 of 269 Table 6.1.2-6 Summary of Fuel Assembly Weights Material Weights 304L M5 Alloy X-750 CF3 Alloy 718 Zircaloy-uo2 Region (kg) (kg) (kg) (kg) (kg) 4 (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.
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 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.
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 Fort Calhoun Station CALCULATION SHEET Box Number SS Box Volume (in "3) Bora I&#x17d; Volume (in "3) SS Wrapper Volume (in "3) Box Number SS Box Volume (in "3) BoraiTM Volume (in "3) SS Wrapper Volume (in "3) Box Number SS Box Volume (in "3) Bora I&#x17d; Volume (in "3) SS Wrapper Volume (in "3) Box Number SS Box Volume (in "3) BoraiTM Volume (in "3) Table 6.1.2-8 Region I Rack Component Volume by parts (Holtec Drawings 1 001 & 1 002) Box 1 Box2 Box3 Box4 408.618 408.618 408.618 408.618 278.4 278.4 278.4 278.4 88.595 88.595 137.13375 88.595 Box6 Box7 Box8 Box9 408.618 408.618 408.618 408.618 278.4 278.4 278.4 278.4 88.595 137.13375 137.13375 88.595 Box 11 Box 12 Box 13 Box 14 408.618 408.618 408.618 408.618 278.4 278.4 278.4 278.4 137.13375 137.13375 137.13375 137.13375 Box 16 Box 17 Box 18 Box 19 408.618 408.618 408.618 408.618 278.4 278.4 278.4 278.4 FC08513 Revision 0 Page 34 of 269 Box5 408.618 278.4 88.595 Box 10 408.618 278.4 137.13375 Box 15 408.618 278.4 137.13375 Box20 408.618 278.4 Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 35 of 269 SS Wrapper 137.13375 185.6725 185.6725 185.6725 185.6725 Volume (in "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 Joints Total Volume by 88.523 51.667 4.979 3.229 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 in3 (3) Boral&#x17d; Volume is calculated by multiplying the sheet dimensions listed in (4) Table 6.1.2-7: Boral&#x17d; Volume= 4 Sheets per Cell* Width* Height* Thickness = 4
* 7.25
* 128
* 0.075 = 278.4in3 (5) SS Wrapper Volume is calculated by multiplying the wrapper dimensions listed in (6) Table 6. 1.2-7: 16 561 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 in3 (Boxes 1, 2, 4, 5, 6 & 9) = 137.13375 in3 (Boxes 3, 7, 8, 10, 11-16) = 185.6725 in3 (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 in3 9b SS Joint Volumes = Number per joint (4) *Width
* Height* Thickness Fort Calhoun Station CALCULATION SHEET = 4
* 1.0
* 8.0
* 0.0897 = 2.8704 in3 9c SS Joint Volumes= Number per joint (1) *Width* Height* Thickness = 1
* 1.542
* 3.0
* 0.0897 = 0.4150 in3 9d SS Joint Volumes= Number per joint (1) *Width* Height* Thickness = 1 *1.0
* 3.0
* 0.0897 = 0.2691 in3 (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 in3 ss Total Volume (Boxes 1-20, Joints 9a-9d & SS comers) in3 Boral Total Volume (Boxes 1-20, Joints 9a-9d & SS comers) in3 46,109.43 22,272.00 FC08513 Revision 0 Page 36 of 269 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./ in3, (Reference 1 0) 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&#x17d;. The density of 8oral&#x17d; is 2.70 grams/cm3 or 0.09754421bs./ in3 (Reference 16). Using the volume of 8oral&#x17d; from Table 6.1.2-8, and the 8oral&#x17d; density determined mass of the 8oral&#x17d;, 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 R . IR k C tM eg1on ac s omponen ass Mass Mass Total SS Volume (lbs) (grams) x0.29 X 453.59 46,109.43 I in3 13371.733 6065284.597 Total 8oral Volume X 0.09754 X 453.59 22,272.00 J in3 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&#x17d;. Table 6.1.2-10 shows each rack by name and overall cell count. Table 6.1.2-10 (Holtec Drawin 1 000) 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 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 1 003 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 Rack Rows Bora I Shims Designation X Panels Sheathing length* Corners 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 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 Rack East-West South Panel All Panels Panel Panels X Designation Length Length Volume (in"3) Volume Columns (inches) (inches) (in"3) (in"3) (in"3) Rack 81 12 X 9 103.824 77.868 1253.67 16297.77 940.26 9402.56 Rack 82 12 X 9 103.824 77.868 1253.67 16297.77 940.26 9402.56 Rack G1 10 X 9 95.172 77.868 1149.20 12641.22 940.26 9402.56 Rack G2 10 X 9 86.52 77.868 1044.73 11492.02 940.26 9402.56 Rack C 11 X 9 95.172 77.868 1149.20 13790.42 940.26 9402.56 Rack D 11 X 8 95.172 69.216 1149.20 13790.42 835.78 7522.05 Rack E 10 X 10 86.52 86.52 1044.73 11492.02 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 12 103.824 86.52 1253.67 16297.77 1044.73 11492.02 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 in3 = 1253.67
* 13 (Panels per Row/Column, Cell Number 12+1) = 16297.77 in3 SS North-South Volume = North-South Length
* Height
* Thickness = 77.868
* 161
* 0.075 = 940.26 in3 = 940.26
* 10 (Panels per Row/Column, Cell Number 10+1) = 9402.56 in3 (2) Boral&#x17d; Volume is calculated by multiplying the sheet dimensions listed in Table 6.1.2-11 by the number of Boral&#x17d; panels per Rack shown in Table 6.1.2-12. Boral&#x17d; Volume = *Width
* Height* Thickness = 7.25
* 128
* 0.075 = 69.6 in3 Fort Calhoun Station CALCULATION SHEET = 69.6
* 1663 (Panels per Rack summed, 1663) = 1157 44.80 in3 FC08513 Revision 0 Page 40 of 269 (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 in3 = 69.6
* 1663 (Panels per Rack summed, 1663) = 60533.20 in3 (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 in2 = 0.675
* 640 in (Shims total length, 640") = 432.00 in3 (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 in3 = 35.0625
* 72 (8 Corners per Rack summed, 72) = 2524.50 in3 SS Total Volume (inA3) 128397.19 SS from Table 5.1.3-10 60533.20 Total SS Region II (inA3) Total Boral Region II (inA3) Region II Rack Mass 89010.91 432.00 2524.50 280897.80 115744.80 Using the density of stainless steel 304L of 0.29 lbs./ in3, (Reference 1 0) 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&#x17d;. The density of Boral&#x17d; is 2.70 grams /cm3 or 0.0975442 lbs./ in3 (Reference 10). Using the volume of Boral&#x17d; from Table 6.1.2-13Table 6.1.2-14, and the Boral&#x17d; density, determined mass of the Boral&#x17d;. Table 6.1.2-14 shows the results of multiplying the volume by the Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 41 of 269 density for both SS 304L and Bora I&#x17d;. 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 in3 81,460.363 36,949,606.24 Total Bora I rM Volume X 0.09754 x453.59 115744.80 in3 11,290.234 5,121 '137.204 Table 6.1.2-15 Total Spent Fuel Racks SS Component Mass Boral&#x17d; 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 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&#x17d;, 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 & 1 0) Element MCNP Nuclide 10 Nominal Composition Value Used (%) (%) 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%.
Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 43 of 269 Table 6.1.3-2 CF3-SS (ASTM A743, References 7 & 1 0) Element MCNP Nuclide ID Nominal Composition Value Used (%) (%) 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 1 00% 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 00-19-10 -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 & 1 0) 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.
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 & 1 0) 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 B-10 (Boron) B-11 (Boron) Boron is 0.006%, therefore A/0 19.61 80.39 B-1 0 = 0.18158
* 0.006 = 0.0011 B-11 = 0.81842
* 0.006 = 0.0049 AMU 10.0129388 11.0093053 Weight 1.963537 8.850381 10.813918 Wt.% 18.158% 81.842% 
 
Fort Calhoun Station CALCULATION SHEET (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/cm3 This gives a total natural boron density of 0.079265 + 0.353877 = 0.433143 g/cm3 FC08513 Revision 0 Page 46 of 269 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/cm3 Since the boral is largely aluminum, it will be assumed that it has an overall density of aluminum of 2.70 g/cm3. The density of aluminum then is: 2.70-0.433143-0.120305 = 2.146552 g/cm3 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.
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) Element MCNP SS-304L CF3 Alloy 718 Nuclide ID Density (g/cm3) 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 ---
 
Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 49 of 269 Table 6.1.3-11 Material M3 Percent by Nuclide (Materials: Boral&#x17d;, U02 & Air) Element MCNP Bora I&#x17d; U02 Air Nuclide ID Density (g/cm3) 2.700 388.6 (KgU) 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 ---
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 6,1 06,563.364 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. 
 
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&#x17d;, UOz, Air & M3 Composite) Element MCNP Zircaloy-4 Bora I U02 Air Material M3 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 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 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 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.
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 t t ORIGEN2 t t U[pu group s rue 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 Fort Calhoun Station 44 Group 45 44 43 42 41 40 39 38 37 36 35 34 33 32 31 30 29 28 27 26 25 24 23 22 21 20 19 18 17 16 15 14 13 12 11 10 9 8 7 6 5 4 3 2 1 CALCULATION SHEET Table 6.1.4-2 ORIGEN ENDF5 Max Energy (Mev) eV 1.00E-11 1.00E-05 3.00E-09 3.00E-03 7.50E-09 7.50E-03 1.00E-08 1.00E-02 2.53E-08 2.53E-02 3.00E-08 3.00E-02 4.00E-08 4.00E-02 5.00E-08 5.00E-02 7.00E-08 7.00E-02 1.00E-07 1.00E-01 1.50E-07 1.50E-01 2.00E-07 2.00E-01 2.25E-07 2.25E-01 2.50E-07 2.50E-01 2.75E-07 2.75E-01 3.25E-07 3.25E-01 3.50E-07 3.50E-01 3.75E-07 3.75E-01 4.00E-07 4.00E-01 6.25E-07 6.25E-01 1.00E-06 1.00E+OO 1.77E-06 1.77E+OO 3.00E-06 3.00E+OO 4.75E-06 4.75E+OO 6.00E-06 6.00E+OO 8.10E-06 8.10E+OO 1.00E-05 1.00E+01 3.00E-05 3.00E+01 1.00E-04 1.00E+02 5.50E-04 5.50E+02 3.00E-03 3.00E+03 1.70E-02 1.70E+04 2.50E-02 2.50E+04 1.00E-01 1.00E+05 4.00E-01 4.00E+05 9.00E-01 9.00E+05 1.40E+OO 1.40E+06 1.85E+OO 1.85E+06 2.35E+OO 2.35E+06 2.48E+OO 2.48E+06 3.00E+OO 3.00E+06 4.80E+OO 4.80E+06 6.43E+OO 6.43E+06 8.19E+OO 8.19E+06 2.00E+01 2.00E+07 FC08513 Revision 0 Page 56 of 269 Output 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 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) Energy Total Gamma Spectrum 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 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 Spectrum Output Group 811 Assemblies 133 Assemblies (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.
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 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 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) Energy Old Fuel Core Offload Total Gamma Spectrum (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 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: curies units of time(lst column): time (years} nuclides time br82 br83 il29 il30 i131 il32 il33 il34 il35 kr83m kr85 kr85m l.OOOE-03 1.431E+03 2.695E+03 1.844E-02 7.273E+03 3.088E+05 4.244E+05 4.791E+05 2.340E+03 2.385E+05 7.242E+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 2.671E+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-02 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 5.622E+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 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 of measure: curies units of time(lst column): time (years) nuclides time kr87 kr88 xel3lm xel33 xel33m xe135 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 2.611E+02 4.602E+03 6.006E+05 1.757E+04 1.354E+05 2.469E-02 O.OOOE+OO 1.827E-18 3.743E+03 2.327E+05 1.833E+03 1.265E-01 7.408E-02 O.OOOE+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 6.671E-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.
Fort Calhoun Station CALCULATION SHEET FC08513 1()8 1()4 1()2 100 10'2 1()-4 1()-6 1o--* iJ1D-12 1D-14 1D-18 1D-18 1Q-20 10'22 10'24 Revision 0 Page 63 of 269 Figure 6.1.5-2 Batch DD Curies by Isotope versus Time Cycle 3 Down -DD Assembly -----------total br82 br83 1129 1130 1131 1132 1133 1134 1135 kr83m kr85 kr85m kr87 0.00 0.25 0.50 0 .. 75 tlme(yan) 1.00 125 1.50 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 -
Fort Calhoun Station CALCULATION SHEET Figure 6.1.5-3 MCNP Model Auxiliary Building Slice (Y-Z) -Looking towards plant West -FC08513 Revision 0 Page 65 of 269 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 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. 6.2.1 Model Geometry Figure 6.1.5-4 MCNP Model Plant Cutaway Slice (Y-Z) -Looking towards plant North -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 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. c 3230 3260 c pz pz 365.76 0.00 c Containment Building c c Figure 6.2.1-1 Containment Geometry Cards $ Top of Containment Basemat (991'0") $ Bottom of Containment Basemat (979'0") 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/cm3. Cell 220 is the air within containment made up of material M1 with density of 0.001205 grams/cm3. c c C Containment Building c Figure 6.2.1-2 Containment Cell Cards 200 1 -2.35 5100 -5000 $ Containment Structure $ Volume in Containment 220 2 -0.001205 -5100 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:
Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 69 of 269 Table 6.2.1.2-1 S t F I P I 0 o* ;pen ue 00 verv1ew 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 F I P ue oo an dT t C IS rans er ana tructure o* 1mens1ons Description of Surface Dimensions from DWG Numerical Numerical 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 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.
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/cm3, 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/cm3. 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.
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 Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 73 of 269 Figure 6.2.1-6 MCNP Images of Spent Fuel Pool Cells Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 7 4 of 269 Figure 6.2.1-7 MCNP Images of Spent Fuel Pool Cells 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 1 025' 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 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 Fort Calhoun Station Room 77 Roof (1057') Room 77 Thickness (1'6") Room 69 Roof (1044') Room 69 Thickness (6") Room 77 Floor (1036') Room 77 Thickness (0'6") 1000 1050 1100 1150 1300 1350 1400 1450 1500 1550 1600 1650 1700 1750 1800 1850 c py 4084.60 py 4038.60 py 2133 0 60 py 2087.88 py -1844.04 py -1889.76 px -3810.00 px -3764.28 px -1889.76 px -1844.04 py 1988.82 py 1950.72 px 1135.38 px 1181.10 px 3550.92 px 3596.64 C Aux Building Z Planes c 3000 3010 3011 3012 3013 3014 3015 3020 3030 3040 3050 3060 3070 c 3100 3110 c pz pz pz pz pz pz pz pz pz pz pz pz pz pz pz 3169.92 3167.38 3164.84 3162.30 3159.76 3157.22 3154.68 2377.44 2331.72 1737.36 1722.12 1981.20 1965.96 1402.08 1386.84 CALCULATION SHEET = + 1057'0" -979'0" 78.00 = + 1057'0" -979'0" -1'6" 76.50 = + 1044'0" -979'0" 65.00 = + 1044'0" -979'0" -0'6" 64.50 = + 1036'0" -979'0" 57.00 = + 1036'0" -979'0" -0'6" 56.50 Figure 6.2.1-9 Room 69 and Control Room Surface Cards $ North Outer Wall (Rms. 69 & 77) $ North Inner Wall (Rms. 69 & 77) $ North Outer Wall (Rms. 3 & 69) $ North Inner Wall (Rms. 3 & 69) $ South Inner Wall (Rms. 3 & 25) $ South Outer Wall (Rms. 3 & 25) $ West Outer Wall (Rms. 3 & 25) $ West Inner Wall (Rms. 3 & 25) $ East Inner Wall (Rms. 3 & 25) $ East Outer Wall (Rms. 3 & 25) $ South Inner CR Wall (Rm. 77) $ South Outer CR Wall (Rm. 77) $ West Outer CR Wall (Rm. 77) $ West Inner CR Wall (Rm. 77) $ East Inner CR Wall (Rm. 77) $ East Outer CR Wall (Rm. 77) 2377.44 2331.72 1981.20 1965.96 1737.36 1722.12 $ Aux Bldg top of Roof Room 3 & 25 (1083') $ Aux Bldg roof -one inch, 3 & 25 (1083') $ Aux Bldg roof -two inches, 3 & 25 (1083') $ Aux Bldg roof-three inches, 3 & 25 (1083') $ Aux Bldg roof -four inches, 3 & 25 (1083') $ Aux Bldg roof -five inches, 3 & 25 (1083') $ Aux Bldg roof-six inches, 3 & 25 (1083') $ Aux Bldg top of Roof Room 77 (1057') $ Aux Bldg ceiling of Roof Room 77 (1057') $ Aux Bldg floor slab Room 77 (1036') $ Aux Bldg floor slab thickness Room 77 (1036') $ Aux Bldg top of Roof Room 69 (1044') $ Aux Bldg ceiling of Roof Room 69 (1044') $ Aux Bldg Room 69 Walkway Floor (1025') $ Aux Bldg Room 69 Walkway Ceiling (1025') FC08513 Revision 0 Page 77 of 269 11405-S-63 11405-S-59 11405-S-63 11405-S-59 11405-S-63 11405-S-59 Fort Calhoun Station CALCULATION SHEET FC08513 10 15 20 25 30 32 35 40 401 402 403 404 45 50 501 55 600 601 602 603 604 605 65 68 69 115 116 117 118 130 135 137 140 142 143 Revision 0 Page 78 of 269 The cell cards define the relationship 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/cm3, 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/cm3. Figure 6.2.1-10 Room 69 and Control Room Cell Cards 1 -2.35 1400 -1550 1350 -1000 3250 -3220 $ Aux Bldg Basemat 1 -2.35 1400 -1550 1150 -1100 3060 -3015 $ North wall Rm 3 1 -2.35 1400 -1550 1050 -1000 3220 -3070 $ North wall Rm 69 1 -2.35 1700 -1850 1050 -1000 3100 -3030 $ North wall Rm 77 1 -2.35 1700 -1850 1650 -1600 3100 -3030 $ South wall Rm 77 1 -2.35 1800 -1850 1600 -1050 3100 -3030 $ East wall Rm 77 1 -2.35 1400 -1550 1350 -1300 3220 -3015 $ South wall Rm 3 1 -2.35 1400 -1450 1150 -1050 3220 -3070 $ West wall rm 69 1 -2.35 1400 -1450 3310 -1150 3220 -3015 $ West N wall rms 3 & 69 1 -2.35 1400 -1450 3320 -3310 3300 -3015 $ West wall above door rm 25 1 -2.35 1400 -1450 3320 -3310 3220 -7000 $ West wall below door rm 25 1 -2.35 1400 -1450 1300 -3320 3220 -3015 $ West S wall rm 25 1 -2.35 1700 -1750 1600 -1050 3100 -3030 $ West wall rm 77 1 -2.35 1500 -1550 1300 -2600 3220 -3015 $ East wall Rm 3 1 -2.35 1500 -1550 2600 -1150 3070 -3015 $ East wall Rm 69 above 1025' 1 -2.35 1500 -1550 2600 -1050 3220 -3100 $ East wall Rm 69 below 1025' 1 -2.35 1400 -1550 1350 -1100 3010 -3000 $ Roof Segment Rm 3 1083'-1 1 -2.35 1400 -1550 1350 -1100 3011 -3010 $ Roof Segment Rm 3 1083'-2 1 -2.35 1400 -1550 1350 -1100 3012 -3011 $ Roof Segment Rm 3 1083'-3 1 -2.35 1400 -1550 1350 -1100 3013 -3012 $ Roof Segment Rm 3 1083'-4 1 -2.35 1400 -1550 1350 -1100 3014 -3013 $ Roof Segment Rm 3 1083'-5 1 -2.35 1400 -1550 1350 -1100 3015 -3014 $ Roof Segment Rm 3 1083'-6 1 -2.35 1400 -1550 1150 -1000 3070 -3060 $ Roof Segment 1044A' 1 -2.35 1750 -1800 1600 -1050 3050 -3040 $ Floor Segment 1036' 1 -2.35 1700 -1850 1650 -1000 3030 -3020 $ Roof Segment 1057' 1 -2.35 1550 -1750 2600 -1000 3110 -3100 $ Rm 69 Floor #200 #220 1 -2.35 1550 -1700 2600 -1000 3070 -3060 $ Rm 69 Ceiling #200 #220 1 -2.35 1550 -1700 1050 -1000 3100 -3070 $ Rm 69 North Wall CR 1 -2.35 1700 -1750 2600 -1650 3100 -3070 $ Rm 69 East Wall CR #200 #220 1 -2.35 2350 -2900 2750 -2450 3220 -3200 $ Rm 25A West wall 1 -2.35 2900 -1500 2750 -2700 3220 -3200 $ Rm 25A South wall 1 -2.35 1450 -1500 1300 -2750 3220 -7000 $ RR Siding Floor 1 -2.35 1450 -2350 2750 -1050 3110 -3100 $ Rm 67-69W-1025' elevation 1 -2.35 2350 -1500 2150 -1050 3110 -3100 $ Rm 69N-1025' elevation 1 -2.35 2200 -1500 2600 -2150 3110 -3100 $ Rm 69N-1025' elevation 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.
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 c c rpp -3764.37525 -3764.28 -1753.87 -1266.190 777.24 1432.56 c Vertical lift Door 1004-1A c 3340 3350 c 405 4 407 4 408 5 rpp -3789.90 777.24 rpp -3789.70156 777.43844 -7.82 -3330 -7.82 -3340 3350 -0.0482 -3350 -3779.74 1432.56 -3779.93844 1432.36156 -1753.87 -1266.190 -1753.67156 -1266.38844 $ Railroad Siding Door 1004-1C $ Railroad Siding Door 1004-1A $ Railroad Siding Door 1004-1A insul Figure 6.2.1-12 Railroad Siding Doors cutaway at Concrete Wall 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/cm3 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 Description of Surface Dimensions from DWG Numerical Numerical 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 Fort Calhoun Station CALCULATION SHEET FC08513 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 cm3 X*Y*Z = [(-2179.32)-(-2770.0224)]
* 588,007,031.474 [(1024.128)-(76.20)] * [(929.9448)-(521.02512)] = 228,971 ,853.30 Density (gmlcm3) 2.568032 grams I cm3 Fuel Source Density =Total Mass of Source I Volume of Source Figure 6.2.1-13 Spent Fuel Surface and Cell Cards Revision 0 Page 82 of 269 2050 2250 6000 6050 6200 6300 py 76.20 px -2179.32 pz 929.945 pz 521.025 px -2770.022 py 1024.128 $ SFP South Inner wall $ SFP East Inner wall $ Top of Fuel Racks $ Bottom of Fuel Racks $ West Boundary of Fuel Racks $ North Boundary of Fuel Racks c C Spent Fuel Pool c 300 3 -2.568032 6200 -2250 2050 -6300 6050 -6000 c $ Fuel Mixture volume 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 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. c Figure 6.2.1-14 Spent Fuel Source Term Definition (18 Month 44 Energy Group Neutron) c Source Definition c sdef si1 sp1 si2 sp2 si3 sp3 c c X=d1 y=d2 z=d3 par=1 H -2770.022 -2179.320 D 0 1 H 76.200 1024.128 D 0 1 H 521.025 929.945 D 0 1 erg=d4 $ Surface 6200 $ surface 6400 $ Surface 6050 to 2250 to 6300 to 6000 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 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).
Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 85 of 269 Table 7-1 G TIIR It am rna a1y esu s Histories EAB Dose CR Dose EAB Dose CR Dose (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 8.053063 7E+08 4.8715E-04 2.7896E-04 4.0254E-04 1.2079596E+09 4.2242E-04 1.9971E-04 3.4907E-04 1.6106127E+09 3.9640E-04 1.8288E-04 3.2760E-04 2.0132659E+09 1.4730E-03 1.5396E-04 1.2176E-03 2.4159191E+09 1.2988E-03 1.3851E-04 1.0736E-03 2.8185723E+09 1.1754E-03 1.2464E-04 9.7174E-04 3.2212255E+09 1.1539E-03 1.2634E-04 9.5482E-04 3.6238787E+09 1.0903E-03 1.1738E-04 9.0215E-04 4.0265318E+09 1.2041E-03 1.3338E-04 9.9591E-04 4.4291850E+09 1.1089E-03 1.2901E-04 9.1718E-04 4.8318382E+09 1.0650E-03 1.2521E-04 8.8082E-04 S.OOOOOOOE+09 1.0343E-03 1.2267E-04 8.5547E-04 5.6371446E+09 1.1640E-03 1.1937E-04 9.6275E-04 6.4424509E+09 1.3709E-03 1.1075E-04 1.1354E-03 7.2477573E+09 1.2599E-03 1.1823E-04 1.0420E-03 8.0530637E+09 1.1613E-03 1.1826E-04 9.6048E-04 8.8583700E+09 1.1584E-03 1.1429E-04 9.5782E-04 9.6636764E+09 1.0900E-03 1.0969E-04 9.0125E-04 l.OOOOOOOE+ 10 1.0590E-03 1.0735E-04 8.7565E-04 Average Tally 1.0416E-03 1.3630E-04 8.6144E-04 Table 7-2 Neutron Tally Results Histories EAB Dose CR Dose Histories EAB Dose (nps) (14mon) (14mon) (nps) (18mon) mRem/hr mRem/hr mRem/hr 3.1457280E+06 5.5506E-06 1.946E-03 5.0331648E+07 1.0360E-05 6.2914560E+06 5.5102E-06 1.879E-03 1.0066330E+08 1.1342E-05 6.2914560E+06 5.5102E-06 1.879E-03 1.5099494E+08 1.3747E-05 9.4371840E+06 5.3999E-06 1.760E-03 2.0132659E+08 1.3273E-05 1.2582912E+07 6.0547E-06 1.554E-03 2.5165824E+08 1.3310E-05 1.5728640E+07 1.1182E-05 1.834E-03 3.0198989E+08 1.3783E-05 1.8874368E+07 1.0251E-05 1.733E-03 3.5232154E+08 1.4124E-05 7.8906E-05 2.3080E-04 1.6521E-04 1.5128E-04 1.2733E-04 1.1457E-04 1.0312E-04 1.0345E-04 9.6153E-05 1.0938E-04 1.0581E-04 1.0275E-04 1.0068E-04 9.8168E-05 9.1104E-05 9.7243E-05 9.7341E-05 9.4124E-05 9.0353E-05 8.8403E-05 1.1231E-04 CR Dose (18mon) mRem/hr 1.5929E-03 1.4876E-03 1.6460E-03 1.6499E-03 1.7010E-03 1.7342E-03 1.7089E-03 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.
Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 87 of 269 Table 7-3 Point Detector Gamma Dose Rates (b) Decay I Dose Type I Detector Location Mean Dose Relative Error Upper Boundary (EAB or Control Rate 95% Confidence Tally Number Room) (mRemlhr) (fraction) 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 EAB 1.5595E-05 0.1158 1.91 E-05 (c) 14 Mo./ Neutron I 25 Control Room 1.8085E-03 0.0907 2.13E-03 (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.
Fort Calhoun Station Attachment A: MCNP Input Files CALCULATION SHEET FC08513 Revision 0 Page 88 of 269 Four different sets of MCNP models were run: Table A-1 presents the run log for this analysis. 1) FCS_14MOg_EABCR.inp 2) FCS _18M0g_ EABCR.inp 3) FCS _14MOn_EABCR.inp 4) FCS_18M0n_EABCR.inp CASE Input 14 Month decay using Gamma source term. 18 Month decay using Gamma source term. 14 Month decay using Neutron source term. 18 Month decay using Neutron source term. Table A-1. MCNP Run Log Output (*) Tape 14-Month Gamma FCS_14MOg_EABCR.inp FCS_14MOg_EABCR.out 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 FCS_14MOgEABCR.tap 18-Month Gamma FCS_18MOg_EABCR.inp 14-Month Neutron FCS_14MOn_EABCR.inp 18-Month Neutron FCS_18MOn_EABCR.inp FCS_18MOg_EABCR.out FCS_18MOg_EABCRca.out FCS 14MOn EABCR.out --FCS 14MOn EABCRca.out --FCS 14MOn EABCRcb.out --FCS 14MOn EABCRcc.out --FCS 14MOn EABCRcd.out --FCS 14MOn EABCRce.out --FCS 18MOn EABCR.out --FCS_18MOg_EABCRca.out FCS 18MOn EABCRcb.out --FCS 18MOn EABCRcc.out --FCS_18MOg_EABCR.tap FCS_14MOn_EABCR.tap FCS_18MOn_EAB.tap
* 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.
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 15 1 20 1 25 1 30 1 32 1 35 1 40 1 401 1 402 1 403 1 404 1 405 4 407 4 408 5 45 1 50 1 501 1 55 1 600 1 601 1 602 1 603 1 604 1 605 1 65 1 68 1 69 1 70 1 80 1 90 1 100 1 105 1 110 1 112 1 114 1 115 1 116 1 117 1 118 1 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35
-2.35 -2.35 -2.35
-2.35 -2.35 -2.35 -7.82 -7.82 1400 1400 1400 1700 1700 1800 1400 1400 1400 1400 1400 1400 -3330 -3340 -0.0482 -3350 -2.35 1700 -2.35 1500 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35
-2.35
-2.35 -2.35 -2.35 1500 1500 1400 1400 1400 1400 1400 1400 1400 1750 1700 2350 2350 2250 2300 -1550 -1550
-1550 -1850 -1850 -1850
-1550 -1450 -1450
-1450 -1450 -1450 3350 -1750 -1550 -1550 -1550 -1550 -1550 -1550 -1550 -1550 -1550 -1550 -1800 -1850
-2200 -2300 -2200 -2850 -2.35 2800 -2250 -2.35 2850 -2800 -2.35 2200 -1500 -2.35 1550 -1750 #200 #220 -2.35 1550 -1750 #200 #220 -2.35 1550 -1700 #200 #220 -2.35 1550 -1700 -2.35 1700 -1750 #200 #220 1350 1150 1050 1050 1650 1600 1350 1150 3310 3320 3320 1300 1600 1300 2600 2600 1350 1350 1350 1350 1350 1350 1150 1600 1650 2100 2450 2000 2000 2000 2000 2000 2000 2600 2600 1050 2600 120 1 -2.35 2550 125 1 -2.35 2300 130 1 -2.35 2350 -1500 2450 -2550 2450 -2900 2750 -1000 -1100 -1000 -1000 -1600 -1050 -1300 -1050 -1150
-3310 -3310 -3320 -1050 -2600 -1150 -1050 -1100 -1100 -1100 -1100 -1100 -1100 -1000 -1050 -1000
-2150 -2100 -2100 -2050 -2050 -2050
-2600 -2600 -1000 -1000 -1000 -1650 3250 3060 3220 3100 3100 3100 3220 3220 3220 3300 3220 3220 3100 3220 3070 3220 3010 3011 3012 3013 3014 3015 3070 3050 3030 3220 3220 3220 3220 3220 3220 3220 3220 3110 3070 3100 3100 -2000 3220 -2400 3220 -2450 3220 -3220 -3015 -3070 -3030 -3030 -3030 -3015 -3070 -3015 -3015 -7000 -3015 -3030 -3015 -3015 -3100 -3000
-3010 -3011 -3012 -3013 -3014 -3060 -3040 -3020 -3200
-3200
-3200
-3200 -3200 -3210 -3200 -3060 -3100 -3060 -3070 -3070 $ Aux Bldg Basemat $ North wall Rm 3 $ North wall Rm 69 $ North wall Rm 77 $ South wall Rm 77 $ East wall Rm 77 $ South wall Rm 3 $ West wall rm 69 $ West N wall rms 3 & 69 $ West wall above door rm 25 $ West wall below door rm 25 $ West S wall rm 25 $ Railroad Siding Door 1004-1C $ Railroad Siding Door 1004-1A $ Railroad Siding Door 1004-1A insul $ West wall rm 77 $ East wall Rm 3 $ East wall Rm 69 above 1025' $ East wall Rm 69 below 1025' $Roof Segment Rm 3 1083'-1 $Roof Segment Rm 3 1083'-2 $Roof Segment Rm 3 1083'-3 $Roof Segment Rm 3 1083'-4 $Roof Segment Rm 3 1083'-5 $Roof Segment Rm 3 1083'-6 $ Roof Segment 1044A' $ Floor Segment 1036' $ Roof Segment 1057' $ SFP North wall $ SFP West wall $ SFP East wall $ SFP South wall W $ SFP South wall E $ SFP South wall Weir $ TC Northeast wall $ TC Northeast extension $ Rm 69 Floor $ Rm 69 Ceiling $ Rm 69 North Wall CR $ Rm 69 East Wall CR -3200 $ TC East wall -3200 $ TC South wall -3200 $ Rm 25A West wall Fort Calhoun Station CALCULATION SHEET FC08513 135 1 137 1 140 1 142 1 143 1 150 2 151 2 155 2 156 2 160 2 161 2 162 2 164 2 165 2 166 2 170 2 175 2 185 2 190 2 c -2.35 -2.35 -2.35 -2.35 -2.35 2900 1450 1450 2350 2200 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 -1500
-1500 -2350 -1500
-1500 2750 1300 2750 2150 2600 -2700 -2750 -1050 -1050 -2150 1450 -2350 2750 -1050 1450 -2350 2750 -1050 1450 -1500 2750 -1150 1500 -1550 2600 -1050 2350 -1500 2150 -1050 2350 -1500 2150 -1050 1450 -1500 1150 -1050 2200 -1500 2600 -2150 -0.001205 1550 -1700 2600 -1050 #200 #220 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 2200 -1500 2600 -2150 1450 -1500 1300 -2750 2900 -1500 2700 -2450 2850 -2800 2000 -2050 2300 -2550 2400 -2000 3220 -3200 3220 -7000 3110 -3100 3110 -3100 3110 -3100 7000 -3110 3100 -3200 3200 -3015 3100 -3070 7000 -3110 3100 -3200 3200 -3070 3220 -3110 3100 -3070 3100 -3200 7000 -3015 7000 -3200 3210 -3200 3220 -3200 $ Rm 25A South wall $ RR Siding Floor Revision 0 Page 90 of 269 $ Rm 67-69W-1025' elevation $ Rm 69N-1025' elevation $ Rm 69N-1025' elevation $ Area West of SFP below 1025' $Area West of SFP above 1025' $Area Above SFP 1038'6" $Area East of SFP below 1025' $Area North of SFP below 1025' $Area North of SFP above 1025' $ Upper Area North of SFP $Area East of SFP below 1025' $ Area in Rm 69 East $Area East of SFP above 1025' $ RR Siding South of Rm 25A $ Area in Rm 25A $ Area in Weir Gate $ Area in TC Gate 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 220 2 -0.001205 -5100 $ Containment Structure $ Volume in Containment c C Spent Fuel Pool c 300 3 -2.568032 6200 -2250 2050 -6300 6050 -6000 310 2 -0.001205 2300 -2250 2050 -2100 6000 -3200 315 2 -0.001205 2300 -2250 2050 -2100 3220 -6050 320 2 -0.001205 2300 -6200 2050 -2100 6050 -6000 325 2 -0.001205 6200 -2250 6300 -2100 6050 -6000 c c Ground Plane Region c $ Fuel Mixture volume $ Area Above Fuel Racks $ Area Below Fuel Racks $ Area West of Fuel Racks $ Area North of Fuel Racks 8000 1 -2.35 (-7000 3260 -10000) #10 #20 #35 #40 #50 #55 #70 #114 #120 #125 #130 #135 #137 #150 #220 #300 #315 #320 #325 #401 #403 $ Plant Grade Region #80 #90 #100 #105 #110 #112 #160 #164 #170 #175 #190 #200 #404 8100 2 -0.001205 7100 7000 -7200 #15 #20 #25 #30 #32 #35 #69 #70 #80 #90 #100 #105 #125 #130 #135 #140 #142 #143 #165 #166 #170 #175 #185 #190 #501 #401 #402 #404 #405 #407 -10000 #40 #45 #50 #110 #112 #114 #150 #151 #155 #195 #200 #220 #408 $ 1004'6" Elevation Region #55 #501 #65 #68 #115 #116 #117 #118 #120 #156 #160 #161 #162 #164 #300 #310 #320 #325 8200 1 -2.35 7000 -10000 -7100 -7200 -10000 $ 1080' Elevation Region 8300 2 -0.001205 7200 -10000 $Air Above 1080' c 9100 0 c #15 #35 #50 #600 #601 #602 #603 #604 #605 #155 #170 #200 #220 #501 #401 #402 #404 -3260 -10000 $ Universe Below Basemat Fort Calhoun Station c void c 9999 0 10000 c C surface Cards c C Aux Building XY Planes c 1000 1050 1100 1150 1300 1350 1400 1450 1500 1550 1600 1650 1700 1750 1800 1850 c C SFP c 2000 2050 2100 2150 2200 2250 2300 2350 2400 2450 2550 2600 2700 2750 2800 2850 2900 c py 4084.60 py 4038.60 py 2133.60 py 2087.88 py -1844.04 py -1889.76 px -3810.00 px -3764.28 px -1889.76 px -1844.04 py 1988.82 py 1950.72 px 1135.38 px 1181.10 px 3550.92 px 3596.64 py -45.72 py 76.20 py 1089.66 py 1257.30 px -2011.68 px -2179.32 px -2806.70 px -2974.34 py -198.12 py -350.52 px -1969.77 py 175.26 py -1211.58 py -1257.30 px -2219.96 px -2418.08 px -2928.62 C Aux Building z Planes c 3000 3010 3011 3012 3013 3014 3015 3020 pz pz pz pz pz pz pz pz 3169.92 3167.38 3164.84 3162.30 3159.76 3157.22 3154.68 2377.44 CALCULATION SHEET $ North Outer Wall (Rms. 69 & 77) $ North Inner Wall (Rms. 69 & 77) $ North Outer Wall (Rms. 3 & 69) $North Inner Wall (Rms. 3 & 69) $ South Inner Wall (Rms. 3 & 25) $ South Outer Wall (Rms. 3 & 25) $ West Outer Wall (Rms. 3 & 25) $ West Inner Wall (Rms. 3 & 25) $ East Inner Wall (Rms. 3 & 25) $ East Outer Wall (Rms. 3 & 25) $ South Inner CR Wall (Rm. 77) $ South Outer CR Wall (Rm. 77) $ West Outer CR Wall (Rm. 77) $ West Inner CR Wall (Rm. 77) $ East Inner CR Wall (Rm. 77) $ East Outer CR Wall (Rm. 77) $ SFP South Outer wall $ SFP South Inner wall $ SFP North Inner wall $ SFP North Outer wall $ SFP East Outer wall $ SFP East Inner wall $ SFP West Inner wall $ SFP West Outer wall $ TC South Inner wall $ TC South Outer wall $ TC East Inner wall $ TC North Outer wall $ Rm 25A South Inner wall $ Rm 25A South Outer wall $ Weir Gate East Edge $ Weir Gate West Edge $ Rm 25A West Inner wall $ Aux Bldg top of Roof Room 3 & 25 (1083') $ Aux Bldg roof -one inch, 3 & 25 (1083') $ Aux Bldg roof -two inches, 3 & 25 (1083') $ Aux Bldg roof-three inches, 3 & 25 (1083') $ Aux Bldg roof-four inches, 3 & 25 (1083') $ Aux Bldg roof-five inches, 3 & 25 (1083') $ Aux Bldg roof -six inches, 3 & 25 (1083') $ Aux Bldg top of Roof Room 77 (1057') FC08513 Revision 0 Page 91 of 269 Fort Calhoun Station CALCULATION SHEET 3030 3040 3050 3060 3070 c 3100 3110 c 3200 3210 3220 3250 c c 3300 3310 3320 c pz pz pz pz pz pz pz pz pz pz pz pz py py 2331.72 1737.36 1722.12 1981.20 1965.96 1402.08 1386.84 1798.32 899.16 502.92 137.16 1432.56 -1266.19 -1753.87 c Rollup Door 1004-1C c $ Aux Bldg ceiling of Roof Room 77 (1057') $ Aux Bldg floor slab Room 77 (1036') $ Aux Bldg floor slab thickness Room 77 (1036') $ Aux Bldg top of Roof Room 69 (1044') $ Aux Bldg ceiling of Roof Room 69 (1044') $ Aux Bldg Room 69 Walkway Floor (1025') $ Aux Bldg Room 69 Walkway Ceiling (1025') $Top of SFP Walkway (1038'6") $ Bottom of Weir Gate (1008'6") $ Bottom of SFP (995'6") $ Aux Bldg Basemat(983'6") $ Aux Bldg Door Top of Opening $ Aux Bldg Door North side of Opening $ Aux Bldg Door South side of Opening 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 -3789.90 777.24 3350 rpp -3789.70156 777.43844 c 3230 pz 365.76 $ 3260 pz 0.00 $ c C Containment Building c c 5000 5100 c c rcc rcc C Fuel Racks c c 6000 6050 6200 6300 c pz pz px py 0. 0. 0. 0. 0. 365.76 929.945 521.025 -2770.022 1024.128 c Optional planes of more c c 6100 py -2183.343 c 6200 py -2804.932 c 6300 px 1081.522 c 6400 px 83.911 c 6500 px -2559.538 -3779.74 -1753.87 -1266.190 1432.56 -3779.93844 -1753.67156 -1266.38844 1432.36156 Top of Containment Basemat (991'0") Bottom of Containment Basemat (979'0") 0. 0. 4343.07 1795.15 0. 0. 3785.87 1677.35 $ CAN Outer $ CAN Inner $ Top of Fuel Racks $ Bottom of Fuel Racks $ West Boundary of Fuel Racks $ North Boundary of Fuel Racks detial of the CPA $ East Boundary of Fuel Racks $ West Boundary of Fuel Racks $ North Boundary of Fuel Racks $ South Boundary of Fuel Racks $ East Boundary of CPA FC08513 Revision 0 Page 92 of 269 Fort Calhoun Station CALCULATION SHEET c 6600 c px 836.127 $ South Boundary of CPA 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 c c Universe c 3078.48 $ 1080' Elevation c 10000 10000 so 1448410.0 $ LPZ at 9 miles 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 c c c c c Air c c m2 c 1001 0.011914 6000 0.005899 8016 0.041881 12000 0.001408 13027 0.001892 14000 0.007311 16000 0.000131 19000 0.000061 20000 0.008719 26000 0.000280 0.001205 g/cm3 7014 8016 0.000039 0.000011 $ H atoms/barn-em $ c atoms/barn-em $ 0 atoms/barn-em $ Mg atoms/barn-em $ Al atoms/barn-em $ Si atoms/barn-em $ S atoms/barn-em $ K atoms/barn-em $ Ca atoms/barn-em $ Fe atoms/barn-em $ N atoms/barn-em $ 0 atoms/barn-em C Fuel Region Mixture c 2.568032 g/cm3 M3 13027 -8.344615E-03 $ Al Weight 18000 -4.242021E-06 $ Ar Weight 5010 -3.049262E-04 $ B10 Weight 5011 -1.361160E-03 $ B11 Weight 6000 -5.023089E-04 $ c Weight 17000 -5.747414E-09 $ Cl Weight 27059 -3.298115E-05 $ Co Weight 24000 -2.048669E-02 $ Cr Weight Percent Percent Percent Percent Percent Percent Percent Percent FC08513 Revision 0 Page 93 of 269 Fort Calhoun Station CALCULATION SHEET 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 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 Weight c Polyisocyanurate (PIR, density: 0.048200 g/cm3) c M5 c 1001 6000 7014 8016 0.00116 0.00174 0.000232 0.000232 C Variance Reduction c c imp:p 1 73r 0 0 c c c Source Definition c sdef si1 sp1 si2 sp2 si3 sp3 c c X=d1 y=d2 z=d3 par=2 H -2770.022 -2179.320 D 0 1 H 76.200 1024.128 D 0 1 H 521.025 929.945 D 0 1 erg=d4 $ Surface 6200 to 2250 $ surface 6400 to 6300 $ Surface 6050 to 6000 C Gamma Spectrum -Total Source Strength is 7.20E+18 C Table 8.6 Gamma Source Term (Fourteen Month) Percent FC08513 Revision 0 Page 94 of 269 Fort Calhoun Station CALCULATION SHEET C by 18 Energy Group (FC08514, Page 45) c si4 H sp4 D c c O.OOE+OO 1.00E-01 1.00E+00 4.00E+00 O.OOE+OO 2.53E+17 4.18E+17 9.97E+12 c Tally Definition c c 2.00E-02 3.00E-02 4.50E-02 1.50E-01 3.00E-01 4.50E-01 1.50E+00 2.00E+00 2.50E+00 6.00E+00 8.00E+00 1.10E+01 2.02E+18 4.25E+17 5.14E+17 3.08E+17 2.33E+17 1.15E+17 7.48E+16 5.63E+15 5.65E+15 1.26E+10 1.45E+09 1.67E+08 c Point Detector Particle Fluence Tallies c c 7.00E-02 7.00E-01 3.00E+00 3.63E+17 2.46E+18 1.10E+14 f15:p f25:p c -90000.00 -8950.00 3078.00 50 2964.18 2758.44 1920.24 50 $ EAB SSWest $ Control Room Proper fc15 Gamma Dose Rate at EAB SSWest fc25 Gamma Dose Rate at interior of CR (Rm 77) c fm15 7.20E+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 7.59E-4 8.78E-4 9.85E-4 1.08E-3 1.52E-3 1.68E-3 1.98E-3 2.51E-3 4.41E-3 4.83E-3 5.23E-3 5.60E-3 7.11E-3 7.66E-3 8.77E-3 1.03E-2 c c C Peripheral cards c c PRDMP 4J 3 c c c Problem cutoff c c nps 5.0E09 c c 2.83E-4 1.17E-3 2.99E-3 5.80E-3 1.18E-2 3.79E-4 5.01E-4 6.31E-4 1.27E-3 1.36E-3 1. 44E-3 3.42E-3 3.82E-3 4.01E-3 6.01E-3 6.37E-3 6.74E-3 1.33E-2 FC08513 Revision 0 Page 95 of 269 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 15 1 20 1 25 1 30 1 32 1 35 1 40 1 401 1 402 1 403 1 404 1 405 4 407 4 408 5 45 1 50 1 501 1 55 1 600 1 601 1 602 1 603 1 604 1 605 1 65 1 68 1 69 1 70 1 80 1 90 1 100 1 105 1 110 1 112 1 114 1 115 1 116 1 117 1 118 1 120 1 125 1 130 1 135 1 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -7.82 -7.82 1400 1400 1400 1700 1700 1800 1400 1400 1400 1400 1400 -1550 -1550 -1550 -1850 -1850 -1850 -1550 -1450 -1450 -1450 -1450 1400 -1450 -3330 -3340 3350 -0.0482 -3350 -2.35 1700 -1750 -2.35 1500 -1550 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35
-2.35
-2.35 -2.35
-2.35 -2.35 1500 1500 1400 1400 1400 1400 1400 1400 1400 1750 1700 2350 2350 2250 2300 2800 2850 2200 -1550
-1550 -1550 -1550 -1550 -1550 -1550 -1550 -1550 -1800 -1850 -2200 -2300
-2200
-2850 -2250
-2800 -1500 -2.35 1550 -1750 #200 #220 -2.35 1550 -1750 #200 #220 -2.35 1550 -1700 #200 #220 -2.35 1550 -1700 -2.35 1700 -1750 #200 #220 -2.35
-2.35
-2.35 -2.35 2550 2300 2350 2900 -1500 -2550 -2900
-1500 1350 1150 1050 1050 1650 1600 1350 1150 3310 3320 3320 1300 1600 1300 2600 2600 1350 1350 1350 1350 1350 1350 1150 1600 1650 2100 2450 2000 2000 2000 2000 2000 2000 2600 2600 1050 2600 2450 2450 2750 2750 -1000 -1100 -1000 -1000 -1600 -1050 -1300 -1050
-1150 -3310 -3310 -3320 -1050 -2600 -1150 -1050 -1100 -1100 -1100 -1100 -1100 -1100 -1000 -1050 -1000 -2150 -2100 -2100 -2050 -2050 -2050 -2600 -2600 -1000 -1000 -1000 -1650 -2000 -2400 -2450 -2700 3250 3060 3220 3100 3100 3100 3220 3220 3220 3300 3220 3220 3100 3220 3070 3220 3010 3011 3012 3013 3014 3015 3070 3050 3030 3220 3220 3220 3220 3220 3220 3220 3220 3110 3070 3100 3100 3220 3220 3220 3220 -3220 -3015 -3070 -3030 -3030 -3030 -3015 -3070 -3015 -3015 -7000 -3015 -3030 -3015 -3015 -3100 -3000 -3010 -3011 -3012 -3013 -3014 -3060 -3040 -3020 -3200 -3200 -3200 -3200 -3200 -3210 -3200
-3060 -3100
-3060 -3070 -3070 -3200 -3200 -3200
-3200 $ Aux Bldg Basemat $ North wall Rm 3 $ North wall Rm 69 $ North wall Rm 77 $ South wall Rm 77 $ East wall Rm 77 $ South wall Rm 3 $ West wall rm 69 $ West N wall rms 3 & 69 $ West wall above door rm 25 $ West wall below door rm 25 $ West S wall rm 25 $ Railroad Siding Door 1004-lC $ Railroad Siding Door 1004-lA $ Railroad Siding Door 1004-lA insul $ West wall rm 77 $ East wall Rm 3 $East wall Rm 69 above 1025' $ East wall Rm 69 below 1025' $Roof Segment Rm 3 1083'-1 $Roof Segment Rm 3 1083'-2 $Roof Segment Rm 3 1083'-3 $Roof Segment Rm 3 1083'-4 $Roof Segment Rm 3 1083'-S $Roof Segment Rm 3 1083'-6 $ Roof Segment 1044A' $ Floor Segment 1036' $Roof Segment 1057' $ SFP North wall $ SFP West wall $ SFP East wall $ SFP South wall W $ SFP South wall E $ SFP South wall Weir $ TC Northeast wall $ TC Northeast extension $ Rm 69 Floor $ Rm 69 Ceiling $ Rm 69 North Wall CR $ Rm 69 East Wall CR $ TC East wall $ TC South wall $ Rm 25A West wall $ Rm 25A South wall Fort Calhoun Station CALCULATION SHEET FC08513 137 1 140 1 142 1 143 1 150 2 151 2 155 2 156 2 160 2 161 2 162 2 164 2 165 2 166 2 170 2 175 2 185 2 190 2 c -2.35 1450 -1500 1300 -2750 -2.35 1450 -2350 2750 -1050 -2.35 2350 -1500 2150 -1050 -2.35 2200 -1500 2600 -2150 -0.001205 1450 -2350 2750 -1050 -0.001205 1450 -2350 2750 -1050 -0.001205 1450 -1500 2750 -1150 -0.001205 1500 -1550 2600 -1050 -0.001205 2350 -1500 2150 -1050 -0.001205 2350 -1500 2150 -1050 -0.001205 1450 -1500 1150 -1050 -0.001205 2200 -1500 2600 -2150 -0.001205 1550 -1700 2600 -1050 #200 #220 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 2200 -1500 2600 -2150 1450 -1500 1300 -2750 2900 -1500 2700 -2450 2850 -2800 2000 -2050 2300 -2550 2400 -2000 3220 -7000 3110 -3100 3110 -3100 3110 -3100 7000 -3110 3100 -3200 3200 -3015 3100 -3070 7000 -3110 3100 -3200 3200 -3070 3220 -3110 3100 -3070 3100 -3200 7000 -3015 7000 -3200 3210 -3200 3220 -3200 Revision 0 Page 97 of 269 $ RR Siding Floor $ Rm 67-69W-1025' elevation $ Rm 69N-1025' elevation $ Rm 69N-1025' elevation $ Area West of SFP below 1025' $ Area West of SFP above 1025' $Area Above SFP 1038'6" $Area East of SFP below 1025' $Area North of SFP below 1025' $Area North of SFP above 1025' $ Upper Area North of SFP $ Area East of SFP below 1025' $ Area in Rm 69 East $ Area East of SFP above 1025' $ RR Siding South of Rm 25A $ Area in Rm 25A $ Area in Weir Gate $ Area in TC Gate 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 220 2 -0.001205 -5100 $ Containment Structure $ Volume in Containment c C Spent Fuel Pool c 300 3 -2.568032 6200 -2250 2050 -6300 6050 -6000 310 2 -0.001205 2300 -2250 2050 -2100 6000 -3200 315 2 -0.001205 2300 -2250 2050 -2100 3220 -6050 320 2 -0.001205 2300 -6200 2050 -2100 6050 -6000 325 2 -0.001205 6200 -2250 6300 -2100 6050 -6000 c C Ground Plane Region c $ Fuel Mixture volume $ Area Above Fuel Racks $ Area Below Fuel Racks $ Area West of Fuel Racks $ Area North of Fuel Racks 8000 1 -2.35 (-7000 3260 -10000) $ Plant Grade Region 8100 #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 2 -0.001205 7100 7000 -7200 -10000 #15 #20 #25 #30 #32 #35 #40 #45 #50 #69 #70 #80 #90 #100 #105 #110 #112 #114 #125 #130 #135 #140 #142 #143 #150 #151 #155 #165 #166 #170 #175 #185 #190 #195 #200 #220 #501 #401 #402 #404 #405 #407 #408 $ 1004'6" #55 #501 #115 #116 #156 #160 #300 #310 Elevation Region #65 #68 #117 #118 #120 #161 #162 #164 #320 #325 8200 1 -2.35 7000 -10000 -7100 -7200 -10000 $ 1080' Elevation Region 8300 2 -0.001205 7200 -10000 $Air Above 1080' c 9100 0 c C void #15 #35 #50 #600 #601 #602 #603 #604 #605 #155 #170 #200 #220 #501 #401 #402 #404 -3260 -10000 $ Universe Below Basemat Fort Calhoun Station c 9999 0 10000 c C surface Cards c C Aux Building XY Planes c 1000 1050 1100 1150 1300 1350 1400 1450 1500 1550 1600 1650 1700 1750 1800 1850 c C SFP c 2000 2050 2100 2150 2200 2250 2300 2350 2400 2450 2550 2600 2700 2750 2800 2850 2900 c py 4084.60 py 4038.60 py 2133.60 py 2087.88 py -1844.04 py -1889.76 px -3810.00 px -3764.28 px -1889.76 px -1844.04 py 1988.82 py 1950.72 px 1135.38 px 1181.10 px 3550.92 px 3596.64 py -45.72 py 76.20 py 1089.66 py 1257.30 px -2011.68 px -2179.32 px -2806.70 px -2974.34 py -198.12 py -350.52 px -1969.77 py 175.26 py -1211.58 py -1257.30 px -2219.96 px -2418.08 px -2928.62 C Aux Building Z Planes c 3000 3010 3011 3012 3013 3014 3015 3020 3030 pz pz pz pz pz pz pz pz pz 3169.92 3167.38 3164.84 3162.30 3159.76 3157.22 3154.68 2377.44 2331.72 CALCULATION SHEET $ North Outer Wall (Rms. 69 & 77) $ North Inner Wall (Rms. 69 & 77) $ North Outer Wall (Rms. 3 & 69) $ North Inner Wall (Rms. 3 & 69) $ South Inner Wall (Rms. 3 & 25) $ South Outer Wall (Rms. 3 & 25) $ West Outer Wall (Rms. 3 & 25) $ West Inner Wall (Rms. 3 & 25) $ East Inner Wall (Rms. 3 & 25) $ East Outer Wall (Rms. 3 & 25) $ South Inner CR Wall (Rm. 77) $ South Outer CR Wall (Rm. 77) $ West Outer CR Wall (Rm. 77) $ West Inner CR Wall (Rm. 77) $ East Inner CR Wall (Rm. 77) $ East Outer CR Wall (Rm. 77) $ SFP South Outer wall $ SFP South Inner wall $ SFP North Inner wall $ SFP North Outer wall $ SFP East Outer wall $ SFP East Inner wall $ SFP West Inner wall $ SFP West Outer wall $ TC South Inner wall $ TC South Outer wall $ TC East Inner wall $ TC North Outer wall $ Rm 25A South Inner wall $ Rm 25A South Outer wall $ Weir Gate East Edge $ Weir Gate West Edge $ Rm 25A West Inner wall $ Aux Bldg top of Roof Room 3 & 25 (1083') $ Aux Bldg roof-one inch, 3 & 25 (1083') $ Aux Bldg roof -two inches, 3 & 25 (1083') $ Aux Bldg roof -three inches, 3 & 25 (1083') $ Aux Bldg roof-four inches, 3 & 25 (1083') $ Aux Bldg roof-five inches, 3 & 25 (1083') $ Aux Bldg roof -six inches, 3 & 25 (1083') $ Aux Bldg top of Roof Room 77 (1057') $ Aux Bldg ceiling of Roof Room 77 (1057') FC08513 Revision 0 Page 98 of 269 Fort Calhoun Station CALCULATION SHEET 3040 3050 3060 3070 c 3100 3110 c 3200 3210 3220 3250 c c 3300 3310 3320 c pz pz pz pz pz pz pz pz pz pz pz py py 1737.36 1722.12 1981.20 1965.96 1402.08 1386.84 1798.32 899.16 502.92 137.16 1432.56 -1266.19 -1753.87 $ Aux Bldg floor slab Room 77 (1036') $ Aux Bldg floor slab thickness Room 77 (1036') $ Aux Bldg top of Roof Room 69 (1044') $ Aux Bldg ceiling of Roof Room 69 (1044') $ Aux Bldg Room 69 Walkway Floor (1025') $ Aux Bldg Room 69 Walkway Ceiling (1025') $Top of SFP Walkway (1038'6") $Bottom of Weir Gate (1008'6") $ Bottom of SFP (995'6") $ Aux Bldg Basemat(983'6") $ Aux Bldg Door Top of Opening $ Aux Bldg Door North side of Opening $ Aux Bldg Door South side of Opening 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 -3789.90 777.24 3350 rpp -3789.70156 777.43844 c 3230 pz 365.76 $ 3260 pz 0.00 $ c C Containment Building c c -3779.74 -1753.87 -1266.190 1432.56 -3779.93844 -1753.67156 -1266.38844 1432.36156 Top of Containment Basemat (991'0") Bottom of Containment Basemat (979'0") 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 6050 pz 6200 px 6300 py c 929.945 521.025 -2770.022 1024.128 c Optional planes of more c c 6100 py -2183.343 c 6200 py -2804.932 c 6300 px 1081.522 c 6400 px 83.911 c 6500 px -2559.538 c 6600 px 836.127 $ Top of Fuel Racks $ Bottom of Fuel Racks $ West Boundary of Fuel Racks $ North Boundary of Fuel Racks detial of the CPA $ East Boundary of Fuel Racks $ West Boundary of Fuel Racks $ North Boundary of Fuel Racks $ South Boundary of Fuel Racks $ East Boundary of CPA $ South Boundary of CPA FC08513 Revision 0 Page 99 of 269 Fort Calhoun Station CALCULATION SHEET c c Ground Plane c 7000 pz 777.24 $ Plant Grade Elevation 1004'6" 7100 px c 7100 -46918.96 px -46939.20 $ Distance to West Side of Bridge $ Distance to EAB 7200 pz 3078.48 $ 1080' Elevation c c Universe c c 10000 10000 so 1448410.0 $ LPZ at 9 miles 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 c c c c c Air c c m2 c 1001 0.011914 6000 0.005899 8016 0.041881 12000 0.001408 13027 0.001892 14000 0. 007311 16000 0.000131 19000 0.000061 20000 0. 008719 26000 0.000280 0.001205 g/cm3 7014 8016 0.000039 0. 000011 $ H atoms/barn-em $ c atoms/barn-em $ 0 atoms/barn-em $ Mg atoms/barn-em $ Al atoms/barn-em $ Si atoms/barn-em $ S atoms/barn-em $ K atoms/barn-em $ Ca atoms/barn-em $ Fe atoms/barn-em $ N atoms/barn-em $ 0 atoms/barn-em C Fuel Region Mixture c 2.568032 g/cm3 M3 13027 -8.344615E-03 $ Al Weight 18000 -4.242021E-06 $ Ar Weight 5010 -3.049262E-04 $ B10 Weight 5011 -1. 361160E-03 $ B11 Weight 6000 -5.023089E-04 $ c Weight 17000 -5.747414E-09 $ Cl Weight 27059 -3. 298115E-05 $ Co Weight 24000 -2.048669E-02 $ Cr Weight 29000 -6.209134E-05 $ Cu Weight Percent Percent Percent Percent Percent Percent Percent Percent Percent FC08513 Revision 0 Page 1 00 of 269 Fort Calhoun Station CALCULATION SHEET 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 si1 sp1 si2 sp2 si3 sp3 c c X=d1 y=d2 z=d3 par=2 H -2770.022 -2179.320 D 0 1 H 76.200 1024.128 D 0 1 H 521.025 929.945 D 0 1 erg=d4 $ Surface 6200 to 2250 $ surface 6400 to 6300 $ Surface 6050 to 6000 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 Revision 0 Page 101 of 269 Fort Calhoun Station CALCULATION SHEET c si4 H sp4 D c c O.OOE+OO 1.00E-01 1.00E+00 4.00E+00 O.OOE+OO 2.10E+17 3.46E+17 8.24E+12 c Tally Definition c c 2.00E-02 3.00E-02 4.50E-02 1.50E-01 3.00E-01 4.50E-01 1.50E+00 2.00E+00 2.50E+00 6.00E+00 8.00E+00 1.10E+01 1.67E+18 3.51E+17 4.25E+17 2.55E+17 1.93E+17 9.51E+16 6.19E+16 4.65E+15 4.67E+15 1.04E+10 1.20E+09 1.38E+08 c Point Detector Particle Fluence Tallies c c 7.00E-02 7.00E-01 3.00E+00 3.00E+17 2.04E+18 9.10E+13 $ EAB SSWest f15:p f25:p c -90000.00 -8950.00 3078.00 50 2964.18 2758.44 1920.24 50 $ Control Room Proper 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 0.6 0.65 0.7 5.0 5.25 5.75 dfO c c 3.96E-3 7.59E-4 1. 52E-3 4.41E-3 7.11E-3 5.82E-4 8.78E-4 1. 68E-3 4.83E-3 7.66E-3 C Peripheral cards c c PRDMP 4J 3 c c C Problem cutoff c c nps 5.0E09 c cc 0.07 0.1 0.8 1.0 6.25 6.75 2.90E-4 9.85E-4 1.98E-3 5.23E-3 8.77E-3 0.15 0.2 0.25 0.3 0.35 0.4 0.45 0.5 1.4 1.8 2.2 2.6 2.8 3.25 3.75 4.25 7.5 9.0 11.0 13.0 15.0 2.58E-4 2.83E-4 3.79E-4 5.01E-4 6.31E-4 1.08E-3 1.17E-3 1.27E-3 1.36E-3 1.44E-3 2.51E-3 2.99E-3 3.42E-3 3.82E-3 4.01E-3 5.60E-3 5.80E-3 6.01E-3 6.37E-3 6.74E-3 1.03E-2 1.18E-2 1. 33E-2 FC08513 Revision 0 Page 1 02 of 269 0.55 4.75 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 c cell cards C Fuel Handling Building c 10 1 15 1 20 1 25 1 30 1 32 1 35 1 40 1 401 1 402 1 403 1 404 1 405 4 407 4 408 5 45 1 50 1 501 1 55 1 600 1 601 1 602 1 603 1 604 1 605 1 65 1 68 1 69 1 70 1 80 1 90 1 100 1 105 1 110 1 112 1 114 1 115 1 116 1 117 1 118 1 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 1400 1400 1400 1700 1700 1800 1400 1400 1400 1400 -1550 -1550 -1550 -1850 -1850 -1850 -1550 -1450 -1450 -1450 -2.35 1400 -1450 -2.35 1400 -1450 -7.82 -3330 -7.82 -3340 -0.0482 -3350 -2.35 1700 -2.35 1500 -2.35
-2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 1500 1500 1400 1400 1400 1400 1400 1400 1400 1750 1700 2350 2350 2250 2300 2800 2850 2200 3350 -1750 -1550 -1550 -1550 -1550 -1550 -1550 -1550
-1550 -1550 -1550 -1800 -1850 -2200 -2300 -2200 -2850 -2250 -2800 -1500 -2.35 1550 -1750 #200 #220 -2.35 1550 -1750 #200 #220 -2.35 1550 -1700 #200 #220 -2.35 1550 -1700 -2.35 1700 -1750 #200 #220 1350 1150 1050 1050 1650 1600 1350 1150 3310 3320 3320 1300 1600 1300 2600 2600 1350 1350 1350 1350 1350 1350 1150 1600 1650 2100 2450 2000 2000 2000 2000 2000 2000 2600 2600 1050 2600 120 1 -2.35 2550 125 1 -2.35 2300 130 1 -2.35 2350 135 1 -2.35 2900 -1500 2450 -2550 2450 -2900 2750 -1500 2750 -1000 -1100 -1000 -1000 -1600 -1050 -1300 -1050 -1150 -3310 -3310
-3320 -1050 -2600 -1150
-1050 -1100 -1100 -1100 -1100 -1100
-1100 -1000 -1050
-1000 -2150 -2100 -2100
-2050 -2050
-2050 -2600 -2600 -1000 -1000 -1000 -1650 3250 3060 3220 3100 3100 3100 3220 3220 3220 3300 3220 3220 3100 3220 3070 3220 3010 3011 3012 3013 3014 3015 3070 3050 3030 3220 3220 3220 3220 3220 3220 3220 3220 3110 3070 3100 3100 -2000 3220 -2400 3220 -2450 3220 -2700 3220 -3220 -3015 -3070 -3030 -3030 -3030
-3015 -3070 -3015 -3015 -7000 -3015 -3030 -3015 -3015 -3100 -3000
-3010 -3011 -3012
-3013
-3014 -3060 -3040
-3020 -3200 -3200 -3200
-3200 -3200
-3210 -3200
-3060 -3100 -3060 -3070 -3070 $ Aux Bldg Basemat $ North wall Rm 3 $ North wall Rm 69 $ North wall Rm 77 $ South wall Rm 77 $ East wall Rm 77 $ South wall Rm 3 $ West wall rm 69 $ West N wall rms 3 & 69 $ West wall above door rm 25 $ West wall below door rm 25 $ West S wall rm 25 $ Railroad Siding Door 1004-1C $ Railroad Siding Door 1004-1A $ Railroad Siding Door 1004-1A insul $ West wall rm 77 $ East wall Rm 3 $East wall Rm 69 above 1025' $East wall Rm 69 below 1025' $Roof Segment Rm 3 1083'-1 $Roof Segment Rm 3 1083'-2 $Roof Segment Rm 3 1083'-3 $Roof Segment Rm 3 1083'-4 $Roof Segment Rm 3 1083'-5 $Roof Segment Rm 3 1083'-6 $ Roof Segment 1044A' $ Floor Segment 1036' $Roof Segment 1057' $ SFP North wall $ SFP West wall $ SFP East wall $ SFP South wall W $ SFP South wall E $ SFP South wall Weir $ TC Northeast wall $ TC Northeast extension $ Rm 69 Floor $ Rm 69 Ceiling $ Rm 69 North Wall CR $ Rm 69 East Wall CR -3200 $ TC East wall -3200 $ TC South wall -3200 $ Rm 25A West wall -3200 $ Rm 25A South wall Fort Calhoun Station CALCULATION SHEET FC08513 137 1 140 1 142 1 143 1 150 2 151 2 155 2 156 2 160 2 161 2 162 2 164 2 165 2 166 2 170 2 175 2 185 2 190 2 c 195 2 c -2.35 1450 -1500 1300 -2750 -2.35 1450 -2350 2750 -1050 -2.35 2350 -1500 2150 -1050 -2.35 2200 -1500 2600 -2150 -0.001205 1450 -2350 2750 -1050 -0.001205 1450 -2350 2750 -1050 -0.001205 1450 -1500 2750 -1150 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 1500 -1550 2600 -1050 2350 -1500 2150 -1050 2350 -1500 2150 -1050 1450 -1500 1150 -1050 2200 -1500 2600 -2150 -0.001205 1550 -1700 2600 -1050 #200 #220 -0.001205 -0.001205
-0.001205 -0.001205 -0.001205 -0.001205 2200 -1500 2600 -2150 1450 -1500 1300 -2750 2900 -1500 2700 -2450 2850 -2800 2000 -2050 2300 -2550 2400 -2000 1750 -1800 1600 -1050 C Containment Building c 3220 -7000 3110 -3100 3110 -3100 3110 -3100 7000 -3110 3100 -3200 3200 -3015 3100 -3070 7000 -3110 3100 -3200 3200 -3070 3220 -3110 3100 -3070 3100 -3200 7000 -3015 7000 -3200 3210 -3200 3220 -3200 3040 -3030 Revision 0 Page 1 04 of 269 $ RR Siding Floor $ Rm 67-69W-1025' elevation $ Rm 69N-1025' elevation $ Rm 69N-1025' elevation $Area West of SFP below 1025' $Area West of SFP above 1025' $Area Above SFP 1038'6" $Area East of SFP below 1025' $Area North of SFP below 1025' $Area North of SFP above 1025' $ Upper Area North of SFP $ Area East of SFP below 1025' $ Area in Rm 69 East $Area East of SFP above 1025' $ RR Siding South of Rm 25A $ Area in Rm 25A $ Area in Weir Gate $ Area in TC Gate $ Area in Room77 200 1 -2.35 220 2 -0.001205 c 5100 -5000 -5100 $ Containment Structure $ Volume in Containment c Spent Fuel Pool c 300 3 -2.568032 6200 -2250 2050 -6300 6050 -6000 310 2 -0.001205 2300 -2250 2050 -2100 6000 -3200 315 2 -0.001205 2300 -2250 2050 -2100 3220 -6050 320 2 -0.001205 2300 -6200 2050 -2100 6050 -6000 325 2 -0.001205 6200 -2250 6300 -2100 6050 -6000 c C Ground Plane Region c $ Fuel Mixture volume $ Area Above Fuel Racks $ Area Below Fuel Racks $ Area West of Fuel Racks $ Area North of Fuel Racks 8000 1 -2.35 (-7000 3260 -10000) #10 #20 #35 #40 #50 #55 #114 #120 #125 #130 #135 #137 #220 #300 #315 #320 #325 #401 $ Plant Grade Region #70 #80 #90 #100 #105 #110 #112 #150 #160 #164 #170 #175 #190 #200 #403 #404 8100 2 -0.001205 7100 7000 -7200 -10000 #15 #20 #25 #30 #32 #35 #40 #45 #50 #69 #70 #80 #90 #100 #105 #110 #112 #114 #125 #130 #135 #140 #142 #143 #150 #151 #155 #165 #166 #170 #175 #185 #190 #195 #200 #220 #501 #401 #402 #404 #405 #407 #408 $ 1004'6" Elevation Region #55 #501 #65 #68 #115 #116 #117 #118 #120 #156 #160 #161 #162 #164 #300 #310 #320 #325 8200 1 -2.35 7000 -10000 -7100 -7200 -10000 $ 1080' Elevation Region 8300 2 -0.001205 7200 -10000 $Air Above 1080' c 9100 0 c C void c #15 #35 #50 #600 #601 #602 #603 #604 #605 #155 #170 #200 #220 #501 #401 #402 #404 -3260 -10000 $ Universe Below Basemat Fort Calhoun Station 9999 0 10000 c C surface Cards c C Aux Building XY Planes c 1000 1050 1100 1150 1300 1350 1400 1450 1500 1550 1600 1650 1700 1750 1800 1850 c C SFP c 2000 2050 2100 2150 2200 2250 2300 2350 2400 2450 2550 2600 2700 2750 2800 2850 2900 c py 4084.60 py 4038.60 py 2133.60 py 2087.88 py -1844.04 py -1889.76 px -3810.00 px -3764.28 px -1889.76 px -1844.04 py 1988.82 py 1950.72 px 1135.38 px 1181.10 px 3550.92 px 3596.64 py -45.72 py 76.20 py 1089.66 py 1257.30 px -2011.68 px -2179.32 px -2806.70 px -2974.34 py -198.12 py -350.52 px -1969.77 py 175.26 py -1211.58 py -1257.30 px -2219.96 px -2418.08 px -2928.62 C Aux Building Z Planes c 3000 3010 3011 3012 3013 3014 3015 3020 3030 3040 pz pz pz pz pz pz pz pz pz pz 3169.92 3167.38 3164.84 3162.30 3159.76 3157.22 3154.68 2377.44 2331.72 1737.36 CALCULATION SHEET $ North Outer Wall (Rms. 69 & 77) $ North Inner Wall (Rms. 69 & 77) $ North Outer Wall (Rms. 3 & 69) $ North Inner Wall (Rms. 3 & 69) $ South Inner Wall (Rms. 3 & 25) $ South Outer Wall (Rms. 3 & 25) $ West Outer Wall (Rms. 3 & 25) $ West Inner Wall (Rms. 3 & 25) $ East Inner Wall (Rms. 3 & 25) $ East Outer Wall (Rms. 3 & 25) $ South Inner CR Wall (Rm. 77) $ South Outer CR Wall (Rm. 77) $ West Outer CR Wall (Rm. 77) $ West Inner CR Wall (Rm. 77) $ East Inner CR Wall (Rm. 77) $ East Outer CR Wall (Rm. 77) $ SFP South Outer wall $ SFP South Inner wall $ SFP North Inner wall $ SFP North Outer wall $ SFP East Outer wall $ SFP East Inner wall $ SFP West Inner wall $ SFP West Outer wall $ TC South Inner wall $ TC South Outer wall $ TC East Inner wall $ TC North Outer wall $ Rm 25A South Inner wall $ Rm 25A South Outer wall $ Weir Gate East Edge $ Weir Gate West Edge $ Rm 25A West Inner wall $ Aux Bldg top of Roof Room 3 & 25 (1083') $ Aux Bldg roof -one inch, 3 & 25 (1083') $ Aux Bldg roof -two inches, 3 & 25 (1083') $ Aux Bldg roof -three inches, 3 & 25 (1083') $ Aux Bldg roof -four inches, 3 & 25 (1083') $ Aux Bldg roof-five inches, 3 & 25 (1083') $ Aux Bldg roof -six inches, 3 & 25 (1083') $ Aux Bldg top of Roof Room 77 (1057') $ Aux Bldg ceiling of Roof Room 77 (1057') $ Aux Bldg floor slab Room 77 (1036') FC08513 Revision 0 Page 1 05 of 269 Fort Calhoun Station CALCULATION SHEET 3050 3060 3070 c 3100 3110 c 3200 3210 3220 3250 c c 3300 3310 3320 c pz pz pz pz pz pz pz pz pz pz py py 1722.12 1981.20 1965.96 1402.08 1386.84 1798.32 899.16 502.92 137.16 1432.56 -1266.19 -1753.87 $ Aux Bldg floor slab thickness Room 77 (1036') $ Aux Bldg top of Roof Room 69 (1044') $ Aux Bldg ceiling of Roof Room 69 (1044') $ Aux Bldg Room 69 Walkway Floor (1025') $ Aux Bldg Room 69 Walkway Ceiling (1025') $Top of SFP Walkway (1038'6") $Bottom of Weir Gate (1008'6") $Bottom of SFP (995'6") $ Aux Bldg Basemat(983'6") $ Aux Bldg Door Top of Opening $ Aux Bldg Door North side of Opening $ Aux Bldg Door South side of Opening 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 -3789.90 777.24 3350 rpp -3789.70156 777.43844 c 3230 pz 365.76 $ 3260 pz 0.00 $ c C Containment Building c c -3779.74 -1753.87 -1266.190 1432.56 -3779.93844 -1753.67156 -1266.38844 1432.36156 Top of Containment Basemat (991'0") Bottom of Containment Basemat (979'0") 5000 5100 rcc 0. 0. 0. 0. 0. 4343.07 1795.15 rcc 0. 0. 365.76 0. 0. 3785.87 1677.35 $ CAN Outer $ CAN Inner c C Fuel Racks c c 6000 6050 6200 6300 c pz pz px py 929.945 521.025 -2770.022 1024.128 c Optional planes of more c c 6100 py -2183.343 c 6200 py -2804.932 c 6300 px 1081.522 c 6400 px 83.911 c 6500 px -2559.538 c 6600 px 836.127 c c Ground Plane $ Top of Fuel Racks $ Bottom of Fuel Racks $ West Boundary of Fuel Racks $ North Boundary of Fuel Racks detial of the CPA $ East Boundary of Fuel Racks $ West Boundary of Fuel Racks $ North Boundary of Fuel Racks $ South Boundary of Fuel Racks $ East Boundary of CPA $ South Boundary of CPA FC08513 Revision 0 Page 1 06 of 269 Fort Calhoun Station CALCULATION SHEET c 7000 7100 pz 777024 $ Plant Grade Elevation 1004'6" px -46918096 $ Distance to West Side of Bridge c 7100 7200 px -46939020 $ Distance to EAB pz c C Universe c 3078048 $ 1080' Elevation c 10000 10000 so 144841000 $ LPZ at 9 miles so 16100000 $ LPZ at 1 mile c Data cards c c Physics Definition c mode n c c Material Definition c C ordinary concrete = 2035 g/cm3 C 95 Concrete, Ordinary (NBS 03) c m1 c c c c c Air c c m2 c 1001 0 0 011914 6000 Oo005899 8016 Oo041881 12000 00001408 13027 00001892 14000 0 0 007311 16000 Oo000131 19000 Oo000061 20000 Oo008719 26000 00000280 Oo001205 g/cm3 7014 8016 Oo000039 Oo000011 $ H atoms/barn-em $ c atoms/barn-em $ 0 atoms/barn-em $ Mg atoms/barn-em $ Al atoms/barn-em $ Si atoms/barn-em $ S atoms/barn-em $ K atoms/barn-em $ Ca atoms/barn-em $ Fe atoms/barn-em $ N atoms/barn-em $ 0 atoms/barn-em c Fuel Region Mixture c 20568032 g/cm3 M3 13027 -8o344615E-03 $ Al Weight 18000 -4o242021E-06 $ Ar Weight 5010 -3 0 049262E-04 $ B10 Weight 5011 -10 361160E-03 $ B11 Weight 6000 -5o023089E-04 $ c Weight 17000 -5o747414E-09 $ Cl Weight 27059 -3 0 298115E-05 $ Co Weight 24000 -2o048669E-02 $ Cr Weight 29000 -6 o 209134E-05 $ Cu Weight 26000 -6o672541E-02 $ Fe Weight 25055 -10966967E-03 $ Mn Weight 42000 -4o783920E-05 $ Mo Weight Percent Percent Percent Percent Percent Percent Percent Percent Percent Percent Percent Percent FC08513 Revision 0 Page 1 07 of 269 Fort Calhoun Station CALCULATION SHEET 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 c 1001 6000 7014 8016 0.00116 0.00174 0.000232 0.000232 C Variance Reduction c c imp:n 1 73r 0 0 c c c Source Definition c sdef si1 sp1 si2 sp2 si3 sp3 c X=d1 y=d2 z=d3 par=1 erg=d4 H -2770.022 -2179.320 $ D 0 1 H 76.200 1024.128 $ D 0 1 H 521.025 929.945 $ D 0 1 c Surface 6200 to 2250 surface 6400 to 6300 Surface 6050 to 6000 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 Revision 0 Page 1 08 of 269 Fort Calhoun Station CALCULATION SHEET 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 f25:n c -90000.00 -8950.00 3078.00 50 2964.18 2758.44 1920.24 50 $ EAB SSWest $ Point X Control Room Proper 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 Revision 0 Page 1 09 of 269 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 15 1 20 1 25 1 30 1 32 1 35 1 40 1 401 1 402 1 403 1 404 1 405 4 407 4 408 5 45 1 50 1 501 1 55 1 600 1 601 1 602 1 603 1 604 1 605 1 65 1 68 1 69 1 70 1 80 1 90 1 100 1 105 1 110 1 112 1 114 1 115 1 116 1 117 1 118 1 120 1 125 1 130 1 135 1 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 1400 1400 1400 1700 1700 1800 1400 1400 1400 1400 -1550 -1550 -1550 -1850 -1850 -1850 -1550 -1450 -1450 -1450 -2.35 1400 -1450 -2.35 1400 -1450 -7.82 -3330 -7.82 -3340 3350 -0.0482 -3350 -2.35 -2.35 -2.35 -2.35
-2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35 -2.35
-2.35 1700 1500 1500 1500 1400 1400 1400 1400 1400 1400 1400 1750 1700 2350 2350 2250 2300 2800 2850 2200 -1750 -1550 -1550 -1550 -1550 -1550 -1550 -1550 -1550 -1550 -1550 -1800 -1850 -2200 -2300 -2200 -2850 -2250 -2800 -1500 -2.35 1550 -1750 #200 #220 -2.35 1550 -1750 #200 #220 -2.35 1550 -1700 #200 #220 -2.35 1550 -1700 -2.35 1700 -1750 #200 #220 -2.35 2550 -1500 -2.35 2300 -2550 -2.35 2350 -2900 -2.35 2900 -1500 1350 1150 1050 1050 1650 1600 1350 1150 3310 3320 3320 1300 1600 1300 2600 2600 1350 1350 1350 1350 1350 1350 1150 1600 1650 2100 2450 2000 2000 2000 2000 2000 2000 2600 2600 1050 2600 2450 2450 2750 2750 -1000 -1100 -1000 -1000 -1600 -1050 -1300 -1050 -1150 -3310 -3310 -3320 -1050 -2600 -1150 -1050 -1100 -1100 -1100 -1100 -1100 -1100 -1000 -1050 -1000 -2150 -2100 -2100 -2050 -2050
-2050 -2600 -2600 -1000 -1000 -1000 -1650 -2000 -2400 -2450 -2700 3250 3060 3220 3100 3100 3100 3220 3220 3220 3300 3220 3220 3100 3220 3070 3220 3010 3011 3012 3013 3014 3015 3070 3050 3030 3220 3220 3220 3220 3220 3220 3220 3220 3110 3070 3100 3100 3220 3220 3220 3220 -3220 -3015 -3070 -3030 -3030 -3030 -3015 -3070 -3015 -3015 -7000 -3015 -3030 -3015 -3015 -3100 -3000 -3010 -3011 -3012
-3013 -3014 -3060 -3040 -3020 -3200 -3200 -3200 -3200 -3200 -3210 -3200
-3060 -3100 -3060 -3070 -3070
-3200 -3200 -3200 -3200 $ Aux Bldg Basemat $ North wall Rm 3 $ North wall Rm 69 $ $ $ $ $ $ $ $ $ $ $ $ $ $
$ $ $ $ $ $ $ $ $ $ $ $ $
$ $ $ $ $ $ North wall Rm 77 South wall Rm 77 East wall Rm 77 South wall Rm 3 West wall rm 69 West N wall rms 3 & 69 West wall above door rm 25 West wall below door rm 25 West S wall rm 25 Railroad Siding Railroad Siding Railroad Siding West wall rm 77 East wall Rm 3 Door 1004-1C Door 1004-1A Door 1004-1A insul East wall Rm 69 above 1025' East wall Rm 69 below 1025' Roof Segment Rm 3 1083'-1 Roof Segment Rm 3 1083'-2 Roof Segment Rm 3 1083'-3 Roof Segment Rm 3 1083'-4 Roof Segment Rm 3 1083'-5 Roof Segment Rm 3 1083'-6 Roof Segment 1044A' Floor Segment 1036' Roof Segment 1057' SFP North wall SFP West wall SFP East wall SFP South wall W SFP South wall E SFP South wall Weir TC Northeast wall TC Northeast extension $ Rm 69 Floor $ Rm 69 Ceiling $ Rm 69 North Wall CR $ Rm 69 East Wall CR $ TC East wall $ TC South wall $ Rm 25A West wall $ Rm 25A South wall Fort Calhoun Station CALCULATION SHEET FC08513 137 1 140 1 142 1 143 1 150 2 151 2 155 2 156 2 160 2 161 2 162 2 164 2 165 2 166 2 170 2 175 2 185 2 190 2 c 195 2 c c -2.35 1450 -2.35 1450 -2.35 2350 -2.35 2200 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 -1500 1300 -2750 -2350 2750 -1050 -1500 2150 -1050 -1500 2600 -2150 1450 -2350 2750 -1050 1450 -2350 2750 -1050 1450 -1500 2750 -1150 1500 -1550 2600 -1050 2350 -1500 2150 -1050 2350 -1500 2150 -1050 -0.001205 1450 -1500 1150 -1050 -0.001205 2200 -1500 2600 -2150 -0.001205 1550 -1700 2600 -1050 #200 #220 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 -0.001205 2200 -1500 2600 -2150 1450 -1500 1300 -2750 2900 -1500 2700 -2450 2850 -2800 2000 -2050 2300 -2550 2400 -2000 1750 -1800 1600 -1050 C Containment Building c 3220 -7000 3110 -3100 3110 -3100 3110 -3100 7000 -3110 3100 -3200 3200 -3015 3100 -3070 7000 -3110 3100 -3200 3200 -3070 3220 -3110 3100 -3070 3100 -3200 7000 -3015 7000 -3200 3210 -3200 3220 -3200 3040 -3030 Revision 0 Page 111 of 269 $ RR Siding Floor $ Rm 67-69W-1025' elevation $ Rm 69N-1025' elevation $ Rm 69N-1025' elevation $Area West of SFP below 1025' $Area West of SFP above 1025' $Area Above SFP 1038'6" $Area East of SFP below 1025' $Area North of SFP below 1025' $ Area North of SFP above 1025' $ Upper Area North of SFP $Area East of SFP below 1025' $ Area in Rm 69 East $Area East of SFP above 1025' $ RR Siding South of Rm 25A $ Area in Rm 25A $ Area in Weir Gate $ Area in TC Gate $ Area in Room77 200 1 -2.35 220 2 -0.001205 c 5100 -5000 -5100 $ Containment Structure $ Volume in Containment C Spent Fuel Pool c 300 3 -2.568032 6200 -2250 2050 -6300 6050 -6000 310 2 -0.001205 2300 -2250 2050 -2100 6000 -3200 315 2 -0.001205 2300 -2250 2050 -2100 3220 -6050 320 2 -0.001205 2300 -6200 2050 -2100 6050 -6000 325 2 -0.001205 6200 -2250 6300 -2100 6050 -6000 c C Ground Plane Region c $ Fuel Mixture volume $ Area Above Fuel Racks $ Area Below Fuel Racks $ Area West of Fuel Racks $ Area North of Fuel Racks 8000 1 -2.35 (-7000 3260 -10000) $ Plant Grade Region 8100 #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 2 -0.001205 7100 7000 -7200 -10000 #15 #20 #25 #30 #32 #35 #40 #45 #50 #69 #70 #80 #90 #100 #105 #110 #112 #114 #125 #130 #135 #140 #142 #143 #150 #151 #155 #165 #166 #170 #175 #185 #190 #195 #200 #220 #501 #401 #402 #404 #405 #407 #408 $ 1004'6" #55 #501 #115 #116 #156 #160 #300 #310 Elevation Region #65 #68 #117 #118 #120 #161 #162 #164 #320 #325 8200 1 -2.35 7000 -10000 -7100 -7200 -10000 $ 1080' Elevation Region 8300 2 -0.001205 7200 -10000 $Air Above 1080' c 9100 0 c C void #15 #35 #SO #600 #601 #602 #603 #604 #605 #155 #170 #200 #220 #501 #401 #402 #404 -3260 -10000 $ Universe Below Basemat Fort Calhoun Station c 9999 0 10000 c C surface Cards c C Aux Building XY Planes c 1000 1050 llOO l150 1300 1350 1400 1450 1500 1550 1600 1650 1700 1750 1800 1850 c C SFP c 2000 2050 2100 2150 2200 2250 2300 2350 2400 2450 2550 2600 2700 2750 2800 2850 2900 c py 4084.60 py 4038.60 py 2133.60 py 2087.88 py -1844.04 py -1889.76 px -3810.00 px -3764.28 px -1889.76 px -1844.04 py 1988.82 py 1950.72 px l135. 38 px l181. 10 px 3550.92 px 3596.64 py -45.72 py 76.20 py 1089.66 py 1257.30 px -2011.68 px -2179.32 px -2806.70 px -2974.34 py -198.12 py -350.52 px -1969.77 py 175.26 py -12l1.58 py -1257.30 px -2219.96 px -2418.08 px -2928.62 C Aux Building Z Planes c 3000 3010 3011 3012 3013 3014 3015 3020 3030 pz pz pz pz pz pz pz pz pz 3169.92 3167.38 3164.84 3162.30 3159.76 3157.22 3154.68 2377.44 2331.72 CALCULATION SHEET $ North Outer Wall (Rms. 69 & 77) $ North Inner Wall (Rms. 69 & 77) $ North Outer Wall (Rms. 3 & 69) $ North Inner Wall (Rms. 3 & 69) $ South Inner Wall (Rms. 3 & 25) $ South Outer Wall (Rms. 3 & 25) $ West Outer Wall (Rms. 3 & 25) $ West Inner Wall (Rms. 3 & 25) $ East Inner Wall (Rms. 3 & 25) $ East Outer Wall (Rms. 3 & 25) $ South Inner CR Wall (Rm. 77) $ South Outer CR Wall (Rm. 77) $ West Outer CR Wall (Rm. 77) $ West Inner CR Wall (Rm. 77) $ East Inner CR Wall (Rm. 77) $ East Outer CR Wall (Rm. 77) $ SFP South Outer wall $ SFP South Inner wall $ SFP North Inner wall $ SFP North Outer wall $ SFP East Outer wall $ SFP East Inner wall $ SFP West Inner wall $ SFP West Outer wall $ TC South Inner wall $ TC South Outer wall $ TC East Inner wall $ TC North Outer wall $ Rm 25A South Inner wall $ Rm 25A South Outer wall $ Weir Gate East Edge $ Weir Gate West Edge $ Rm 25A West Inner wall $ Aux Bldg top of Roof Room 3 & 25 (1083') $ Aux Bldg roof-one inch, 3 & 25 (1083') $ Aux Bldg roof -two inches, 3 & 25 (1083') $ Aux Bldg roof -three inches, 3 & 25 (1083') $ Aux Bldg roof -four inches, 3 & 25 (1083') $ Aux Bldg roof -five inches, 3 & 25 (1083') $ Aux Bldg roof-six inches, 3 & 25 (1083') $ Aux Bldg top of Roof Room 77 (1057') $ Aux Bldg ceiling of Roof Room 77 (1057') FC08513 Revision 0 Page 112 of 269 Fort Calhoun Station CALCULATION SHEET 3040 3050 3060 3070 c 3100 3110 c 3200 3210 3220 3250 c c 3300 3310 3320 c pz pz pz pz pz pz pz pz pz pz pz py py 1737.36 1722.12 1981.20 1965.96 1402.08 1386.84 1798.32 899.16 502.92 137.16 1432.56 -1266.19 -1753.87 c Rollup Door 1004-1C c $ Aux Bldg floor slab Room 77 (10361) $ Aux Bldg floor slab thickness Room 77 (10361) $ Aux Bldg top of Roof Room 69 (10441) $ Aux Bldg ceiling of Roof Room 69 (10441) $ Aux Bldg Room 69 Walkway Floor (10251) $ Aux Bldg Room 69 Walkway Ceiling (10251) $Top of SFP Walkway (103816") $Bottom of Weir Gate (100816") $ Bottom of SFP (99516") $ Aux Bldg Basemat(98316") $ Aux Bldg Door Top of Opening $ Aux Bldg Door North side of Opening $ Aux Bldg Door South side of Opening 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 -3789.90 777.24 3350 rpp -3789.70156 777.43844 c 3230 pz 365.76 $ 3260 pz 0.00 $ c c Containment Building c c -3779.74 -1753.87 -1266.190 1432.56 -3779.93844 -1753.67156 -1266.38844 1432.36156 Top of Containment Basemat (991 I 011) Bottom of Containment Basemat (979 I 010) 5000 5100 c rcc 0. 0. 0. 0. 0. 4343.07 1795.15 rcc 0. 0. 365.76 0. 0. 3785.87 1677.35 $ CAN Outer $ CAN Inner c c Fuel Racks c c 6000 6050 6200 6300 c pz pz px py 929.945 521.025 -2770.022 1024.128 c Optional planes of more c c 6100 PY -2183.343 c 6200 py -2804.932 c 6300 px 1081.522 c 6400 px 83.911 c 6500 px -2559.538 c 6600 px 836.127 $ Top of Fuel Racks $ Bottom of Fuel Racks $ West Boundary of Fuel Racks $ North Boundary of Fuel Racks detial of the CPA $ East Boundary of Fuel Racks $ West Boundary of Fuel Racks $ North Boundary of Fuel Racks $ South Boundary of Fuel Racks $ East Boundary of CPA $ South Boundary of CPA FC08513 Revision 0 Page 113 of 269 Fort Calhoun Station CALCULATION SHEET c c Ground Plane c 7000 pz 777.24 $ Plant Grade Elevation 1004'6" 7100 px c 7100 -46918.96 px -46939.20 $ Distance to West Side of Bridge $ Distance to EAB 7200 pz 3078.48 $ 1080' Elevation c C Universe c c 10000 10000 so 1448410.0 $ LPZ at 9 miles 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 c c C Air c c m2 c 1001 0.011914 6000 0.005899 8016 0.041881 12000 0.001408 13027 0.001892 14000 0. 007311 16000 0.000131 19000 0.000061 20000 0.008719 26000 0.000280 0.001205 g/cm3 7014 8016 0.000039 0. 000011 $ H atoms/barn-em $ c atoms/barn-em $ 0 atoms/barn-em $ Mg atoms/barn-em $ Al atoms/barn-em $ Si atoms/barn-em $ S atoms/barn-em $ K atoms/barn-em $ Ca atoms/barn-em $ Fe atoms/barn-em $ N atoms/barn-em $ 0 atoms/barn-em C Fuel Region Mixture c 2.568032 g/cm3 M3 13027 -8.344615E-03 $ Al Weight 18000 -4.242021E-06 $ Ar Weight 5010 -3.049262E-04 $ B10 Weight 5011 -1.361160E-03 $ B11 Weight 6000 -5.023089E-04 $ c Weight 17000 -5.747414E-09 $ Cl Weight 27059 -3. 298115E-05 $ Co Weight 24000 -2.048669E-02 $ Cr Weight 29000 -6. 209134E-05 $ Cu Weight 26000 -6. 672541E-02 $ Fe Weight 25055 -1.966967E-03 $ Mn Weight Percent Percent Percent Percent Percent Percent Percent Percent Percent Percent Percent FC08513 Revision 0 Page 114 of 269 Fort Calhoun Station CALCULATION SHEET 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 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 si1 sp1 si2 sp2 si3 sp3 c X=d1 y=d2 z=d3 par=1 erg=d4 H -2770.022 -2179.320 $ D 0 1 H 76.200 1024.128 $ D 0 1 H 521.025 929.945 $ D 0 1 c Surface 6200 to 2250 surface 6400 to 6300 Surface 6050 to 6000 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 Revision 0 Page 115 of 269 Fort Calhoun Station CALCULATION SHEET 3.00E-08 4.00E-08 5.00E-08 7.00E-08 1.00E-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+10 1.12E+10 3.57E+09 1.23E+09 c c c Tally Definition c c c Point Detector Particle Fluence Tallies c c f15:n f25:n c fc15 fc25 c fm15 fm25 c -90000.00 -8950.00 3078.00 50 2964.18 2758.44 1920.24 50 Gamma Dose Rate at EAB SSWest $ EAB SSWest $ Point X Control Room Proper Gamma Dose Rate at interior of CR (Rm 77) 2. 36E+ll 2.36E+ll 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 Problem cutoff c c nps 1.0E09 c FC08513 Revision 0 Page 116 of 269 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 4.0E-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 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 1.5128E-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 1.0736E-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 2.2E-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 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 1.0275E-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 1.0420E-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 Fort Calhoun Station CALCULATION SHEET FC08513 3) FCS_14MOn_EABCRci.txt nps 3145728 6291456 9437184 12582912 15728640 18874368 22020096 25165824 28311552 31457280 34603008 35000000 nps 6291456 12582912 18874368 25165824 31457280 37748736 44040192 50331648 56623104 62914560 69206016 75497472 81788928 88080384 94371840 100000000 tally mean error vov 5.5506E-06 0.1782 0.2265 5.5102E-06 0.1345 0.2032 5.3999E-06 0.1133 0.1195 6.0547E-06 0.1287 0.1876 1.1182E-05 0.2539 0.4204 1.0251E-05 0.2316 0.4149 1.1553E-05 0.2057 0.2612 1.1146E-05 0.1876 0.2556 1.0529E-05 0.1771 0.2523 1.0482E-05 0.1629 0.2361 1.0182E-05 0.1531 0.2316 1.0176E-05 0.1517 0.2306 mean tally error vov 5.5102E-06 0.1345 0.2032 6.0547E-06 0.1287 0.1876 1.0251E-05 0.2316 0.4149 1.1146E-05 0.1876 0.2556 1.0482E-05 0.1629 0.2361 1.0632E-05 0.1410 0.1952 1.0550E-05 0.1259 0.1720 1.0192E-05 0.1162 0.1599 1.0035E-05 0.1093 0.1388 1.0128E-05 0.1010 0.1219 1.0046E-05 0.0955 0.1099 1.1073E-05 0.1085 0.1862 1.1576E-05 0.1098 0.1423 1.1358E-05 0.1043 0.1397 1.1220E-05 0.0992 0.1362 1.0897E-05 0.0966 0.1354 4) FCS_18MOn_EABCRcc.txt nps 50331648 100663296 150994944 201326592 251658240 301989888 352321536 402653184 452984832 503316480 553648128 603979776 654311424 704643072 725000000 mean tally error vov 1.0360E-05 0.1388 0.1777 1.1342E-05 0.1126 0.2250 1.3747E-05 0.1038 0.1065 1.3273E-05 0.0871 0.0872 1.3310E-05 0.0761 0.0675 1.3783E-05 0.0693 0.0569 1.4124E-05 0.0646 0.0438 1.7055E-05 0.1888 0.8688 1.7312E-05 0.1679 0.8186 1.7236E-05 0.1529 0.7933 1.7087E-05 0.1410 0.7764 1.6951E-05 0.1313 0.7520 1.6618E-05 0.1241 0.7412 1.6364E-05 0.1177 0.7246 1.6178E-05 0.1158 0.7235 Revision 0 Page 119 of 269 14 Month decay using Neutron source term. 15 slope fom 15 1.8 5.1E-02 2.0 5.1E-02 2.2 4.8E-02 2.1 2.8E-02 1.9 5.7E-03 2.0 5.7E-03 1. 9 6. 2E-03 2.0 6.5E-03 2.0 6.4E-03 2.0 6.7E-03 2.1 6.8E-03 2.1 6.9E-03 slope fom 2.0 5.1E-02 2.1 2.8E-02 2.0 5.7E-03 2.0 6.5E-03 2.0 6.7E-03 2.1 5.1E-03 2.1 3.6E-03 2.1 2.9E-03 2.1 2.5E-03 2.1 2.4E-03 2.1 2.2E-03 2.0 1.5E-03 2. 0 1. 3E-03 2.1 1. 3E-03 2.3 1.3E-03 2.3 1.2E-03 tally mean error vov 1.9457E-03 0.3880 0.2959 1.8790E-03 0.2889 0.1686 1.7601E-03 0.2381 0.1173 1.5535E-03 0.2275 0.1117 1.8337E-03 0.2185 0.1641 1.7333E-03 0.2030 0.1407 1.6227E-03 0.1889 0.1321 1.5546E-03 0.1776 0.1194 1.5055E-03 0.1679 0.1080 1.4315E-03 0.1610 0.1031 1.3783E-03 0.1534 0.0994 1.3642E-03 0.1532 0.0994 tally mean error vov 1.8790E-03 0.2889 0.1686 1.5535E-03 0.2275 0.1117 1.7333E-03 0.2030 0.1407 1.5546E-03 0.1776 0.1194 1.4315E-03 0.1610 0.1031 1.3475E-03 0.1467 0.0925 1.4353E-03 0.1362 0.0762 1.4246E-03 0.1302 0.0697 1.4061E-03 0.1216 0.0623 1.3655E-03 0.1155 0.0570 1.5473E-03 0.1415 0.2487 1.5188E-03 0.1335 0.2390 1.4821E-03 0.1278 0.2281 1.4562E-03 0.1221 0.2187 1.3839E-03 0.1200 0.2179 1.4437E-03 0.1161 0.1751 25 slope fom 25 1.2 1.1E-02 1. 3 1.1E-02 1.3 1.1E-02 1.3 8.8E-03 1.3 7.7E-03 1.4 7.4E-03 1.5 7.3E-03 1.6 7.2E-03 1.6 7.2E-03 1.7 6.9E-03 1.8 6.8E-03 1.8 6.7E-03 slope fom 1.3 1.1E-02 1.3 8.8E-03 1.4 7 .4E-03 1.6 7.2E-03 1.7 6.9E-03 1.8 4.8E-03 1.9 3.1E-03 2.0 2.3E-03 2.2 2.0E-03 2.3 1.8E-03 2.3 l.OE-03 2.3 9.9E-04 2.5 9.5E-04 2.6 9.3E-04 2.7 8.7E-04 2.8 8.5E-04 18 Month decay using Neutron source term. 15 slope fom 2.1 8.4E-04 2.1 6.3E-04 2.1 5.0E-04 2.3 5.3E-04 2.7 5.6E-04 2.9 5.6E-04 2.9 5.5E-04 2.6 5.6E-05 2.5 6.3E-05 2.5 6.9E-05 2.5 7.4E-05 2.6 7.8E-05 2.7 8.0E-05 2.8 8.3E-05 2.8 8.3E-05 tally mean error vov 1.5929E-03 0.1891 0.2831 1.4876E-03 0.1208 0.1464 1.6460E-03 0.0955 0.0781 1.6499E-03 0.0794 0.0546 1.7010E-03 0.0709 0.0421 1.7342E-03 0.0640 0.0326 1.7089E-03 0.0581 0.0278 1.9534E-03 0.1418 0.7906 1.9429E-03 0.1281 0.7568 1.9028E-03 0.1184 0.7387 1.8580E-03 0.1106 0.7281 1.8157E-03 0.1041 0.7187 1.7817E-03 0.0983 0.7095 1.7739E-03 0.0934 0.6596 1.7783E-03 0.0907 0.6531 25 slope fom 2.0 4.5E-04 2.7 5.5E-04 2.6 5.9E-04 2.8 6.4E-04 3.2 6.4E-04 3.5 6.6E-04 3.9 6.8E-04 3.2 l.OE-04 3.3 1.1E-04 3.1 1.1E-04 3.0 1.2E-04 3.2 1.2E-04 3.1 1.3E-04 3.0 1.3E-04 3.0 1.4E-04 Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 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 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 The licensee concluded, "Based on calculated direct and composition, dose conversion. scattered dose rates from spent fuel assemblies in a SONGS SFP following drain down. it is concluded that the maximum Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 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 "The source terms for neutron and gamma radiation in spent respectively and by energy group spectrum 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 1 0 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 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 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." 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.
Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 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. (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, this 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. states "see basis for 50.47(b)(1 0)." Please provide specific Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 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 (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. (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 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. 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 1 0 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 *the information in the fuel management records (mentioned power distributions were not explicitly used, 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 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 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 " the source and value used for the specific heat of the rates for the limiting fuel assembly were 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 fuel source. The value of% TD is the lowest stack density shown in vendor specifications.
Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 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 EAL-RAI-01 will be maintained in accordance with 10 CFR 50.54(q). Technical Bases Manual. FC08513 and 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 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 EAL-RAI-02 not included, as stated in the endorsed guidance, or revise Technical Bases Manual. FC08513 and 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 warranted. strategies.
Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 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 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&deg;C? Please provide a copy of this analysis and the existing analysis for the Partial Loss of Cooling Water Inventory with No Air Cooling.
Fort Calhoun Station CALCULATION SHEET Attachment D: Excel Spreadsheets. Attachment 01 M3 Calculation20161014.xlsx Attachment 02 FCS MCNP Source lnputs.xlsx Attachment 03 FCS Geo Inputs 20161017.xlsx Attachment 04 FC08514 ext enr.xlsx Attachment 05 tallies_gamm20161017.xlsx Attachment 06 tallies neut20131017.xlsx FC08513 Revision 0 Page 129 of 269 Material M3 Calculation Source Density Calculation MCNP Geometry Inputs U-235 Average Enrichment Gamma Tally Results Neutron Tally Results M3 Calculation20161014.xlsx Element MCNP Nuclide ID AI 13027 Ar 18000 B 5000 B-10 5010 B-11 5011 c 6000 Cl 17000 Co 27059 ' Cr 24000 Cu 29000 Fe 26000 Mn 25055 Mo 42000 N 7014 Na 11023 Nb 41093 Ni 28000 0 8016 p 15031 Pb 82000 s 16000 Si 14000 Sn 50000 Ta 73181 Ti 22000 U-234 92234 U-235 92235 U-236 92236 U-238 92238 v 23000 Zr 40000 Attachment D SpreadSheets Material Calculation FRACTION of TOTAL MASS Weight Fractions 8.344615E-03 $ AI weight percent 4.242021E-06 $ Ar weight percent -$ B weight percent 3.049262E-04 $ BlO weight percent 1.361160E-03 $ B11 weight percent 5.023089E-04 $ C weight percent 5.747414E-09 $ Cl weight percent 3.298115E-05 $ Co weight percent 2.048669E-02 $ Cr weight percent 6.209134E-05 $ Cu weight percent 6.672541E-02 $ Fe weight percent 1.966967E-03 $ Mn weight percent 4. 783920E-05 $ Mo weight percent 2.698222E-04 $ N weight percent 5.747414E-09 $ Na weight percent 1.902357E-03 $ Nb weight percent 1.868235E-02 $ Ni weight percent 8.421467E-02 $ 0 weight percent 4.292521E-05 $ P weight percent 3. 735819E-08 $ Pb weight percent 3.621267E-05 $ S weight percent 1.214038E-03 $ Si weight percent 4.166875E-06 $ Sn weight percent 6.522834E-06 $ Ta weight percent 3.010457E-04 $ Ti weight percent 1.946356E-04 $ U-234 weight percent 2.183528E-02 $ U-235 weight percent 1.005027E-04 $ U-236 weight percent 6.017375E-01 $ U-238 weight percent 1.436854E-08 $ V weight percent 1.696187E-01 $ Zr weight percent l.OOOOOOE+OO FC08513 Page 130 of 269 FCS MCNP Source lnputs.xlsx One Assembly Volume (cc) TOP Region 304l 361.27481 M5 157.42358 X-750 375.50684 CF3 530.32650 Fuel Plenum 304l 22.42395 M5 837.23515 X-750 ------------Clad Volume 2076.23760 Fuel Region U02 42404.46089 304l 34.00966 M5 15384.77238 A718 118.84360 BOT Region 304l 10.09078 M5 23.52130 X-750 284.39747 CF3 883.87750 Zircaloy-4 27.28616 Attachment D SpreadSheets Source Density Calculation Reference FC07586 Total Volume 304l 427.799 M5 1424.53 18479.190 A718 118.844 X-750 is internal to the plenum X-750 2935.90 659.904 CF3 1414.204 Zircaloy-4 57942.09 27.286 U02 42404.461 Total Assembly Volume (cm3) 63531.688 1229.17 944 Assembly Volume (cm3) 59,973,913.63 Rack Material Volume 2,261,690.13 5,358,688.42 FC08513 Page 131 of 269 One Assemb TOP Region Fuel Plenum Fuel Region BOT Region Tab FCS FUEL (Cells A62-P124) Boral (cm3) SS 304 (cm3) Total Volume (cm3)
FCS MCNP Source lnputs.xlsx Volume of Source Mass U02 ss 304 Bora I M5 A-718 X-750 CF3 Zircaloy-4 Air Attachment D SpreadSheets Source Density Calculation 228,971,853.30 cm3 416,171,245.28 g 46,256,586.84 g 6,106,563.36 g 100,721,968.00 g 922,288.00 g 6,748,656.00 g 10,716,288.00 g 168,976.00 g 194,459.96 g Total Volume (Materials) 2.568032 gm/cc 67,594,292.19 I FC08513 228,971,853.30 I Page 132 of 269 Remaining Volume (air) 161,377,561.12 FCS MCNP Source lnputs.xlsx 944 Assembly Mass (g) ly Mass (kg) Mass (kg) Mass (g) 304L MS X-750 CF3 304L MS X-750 13.5 w/o U02 304L MS A718 304L MS X-750 CF3 Zircaloy-4 2.900 1.024 3.087 4.257 0.180 5.446 1.724 440.859 0.273 100.074 0.977 0.081 0.153 2.338 7.095 0.179 2,737,600 966,656 2,914,128 4,018,608 169,920 5,141,024 1,627,456 416,171,245 257,712 94,469,856 922,288 76,464 144,432 2,207,072 6,697,680 168,976 Attachment D SpreadSheets Source Density Calculation
 
==Reference:==
Document No. 51-Mass (g) 9124707-000 I 304L I 3,241,696 MS Fuel U02 100,721,968 Volume 2587.679 A718 I 42404.46 X-750 6,748,656 CF3 10,716,288 Zircaloy-4 168,976 U02 416,171,245 Total Assembly mass (gms) 538691117.280 FC08513 Page 133 of 269 pins diameter Stack Height 176 0.3805 129.3 0.224 0.205 FCS MCNP Source lnputs.xlsx Volume based on Defined MCNP source ('FCS Fuei'!X3) Air Mass 0.001205 g/cm3 194,459.961grams SFP Racks Mass Attachment D SpreadSheets Source Density Calculation
 
==Reference:==
PNNL-15870 Rev. 1 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/cm3 Mass (g) Boral 6,106,563.361 Sum of "FCS Racks" cells (G62 + J62)
* Density of Bora I -2.70 gm/cm3 364,972.691 1,896. 717.45 1 FC08513 Page 134 of 269 FCS MCNP Source lnputs.xlsx 5.0 CALCULATIONS Attachment D SpreadSheets Source Density Calculation FC08513 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 FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets Source Density Calculation FC08513 I Notes: The atomic masses of U-234, U-235, U-236, and U-238 are 234.040947, 235.043924, 236.045563, and Page 136 of 269 238.050785, 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.
FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets Source Density Calculation Table 5-1 Uranium Blend Atomic Masses as a Function of U-235 Enrichment F 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 age 137 of 269 FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets Source Density Calculation FC08513 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 Wt% U02 {glassy} {glassy} {glassy} {glassy} 2.0 440849.1989 69.1708 7772.0000 35.7512 2.5 440852.5901 86.4635 9715.0000 44.6890 3.0 440855.9813 103.7562 11658.0000 53.6268 3.5 440859.3726 121.0489 13601.0000 62.5646 4.0 440862.7638 138.3416 15544.0000 71.5024 4.5 440866.1550 155.6343 17487.0000 80.4402 5.0 440869.5462 172.9270 19430.0000 89.3780 U-238 {glassy} 380723.0780 378753.8475 376784.6170 374815.3865 372846.1560 370876.9255 368907.6950 FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets Source Density Calculation Table 5-4 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 2.0 0.0157% 1.7630% 0.0081% 86.3613% 2.5 0.0196% 2.2037% 0.0101% 85.9139% 3.0 0.0235% 2.6444% 0.0122% 85.4666% 3.5 0.0275% 3.0851% 0.0142% 85.0193% 4.0 0.0314% 3.5258% 0.0162% 84.5719% 4.5 0.0353% 3.9665% 0.0182% 84.1246% 5.0 0.0392% 4.4072% 0.0203% 83.6773% U02 Density as a function U-235 Enrichment U-235 Enrichment Uranium Weight Properites Wt% U02 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 Page 139 of 2 5 9 0 11.8519% 11.8526% 11.8533% 11.8540% 11.8547% 11.8553%
11.8560%
FCS MCNP Source lnputs.xlsx Region I Racks Cell Dimensions (SS Box) Width 8.46" Cell Dimensions (SS Box) Height 161" Cell Dimensions (SS Box) Thickness 0.075" Boral sheet -Width 7.25" Boral sheet -Height 128" Boral sheet -Thickness 0.075" Boundary SS Wrapper sheet -Width 7.25" Boundary SS Wrapper sheet -Height 128" Boundary 55 Wrapper sheet -Thickness 0.075" Inner 55 Wrapper sheet -Thickness 0.0235" 9a Dimensions-Width 1.542" 9a Dimensions -Height 8" 4 times 9a Dimensions-Thickness 0.0897" 9b Dimensions-Width 1" 9b Dimensions -Height 8" 4 times 9b Dimensions -Thickness 0.0897" 9c Dimensions-Width 1.542" 9c Dimensions-Height 3" 9c Dimensions-Thickness 0.0897" 9d Dimensions -Width 1 9d Dimensions-Height 3" 9d Dimensions -Thickness 0.0897" Region II Racks Cell Dimensions (SS Box) Width 8.46" Cell Dimensions (SS Box) Height 161" Cell Dimensions (SS Box) Thickness 0.075" Boral sheet -Width 7.25" Bora I sheet -Height 128" Attachment D SpreadSheets Source Density Calculation 8.46 161.00 0.075 7.25 128.00 0.075 7.25 130.00 0.075 0.0235 1.542 8.00 0.0897 1.00 8.00 0.0897 1.542 3.00 0.0897 1.00 3.00 0.0897 8.46 161.00 0.075 7.25 128.00 Box Number SS Box Volume 4 (in"3) Boral Box Volume 3 (in"3) 55 Wrapper 2 Volume (in"3) Rack Designation RxC Rack B1 12 X 9 Rack B2 12 X 9 Rack G1 10 X 9 Rack G2 10x9 Holtec Drawin Box 1 Box 2 408.618 408.618 278.4 278.4 88.595 88.595 Box 10 Box 11 408.618 408.618 278.4 278.4 137.13375 137.13375 Box 19 Box 20 408.618 408.618 278.4 278.4 185.6725 185.6725 Total 55 46,109.43 in"3 Total Boral 22,272.00 in"3 Table 5.1.3-10 AI-B4C ss Boral Panels Sheating 195 195 195 195 178 178 157 157 FC08513 Page 140 of 269 FCS MCNP Source lnputs.xlsx Boral sheet -Thickness 0.075" Boundary 55 Wrapper sheet -Width 8.00" Boundary 55 Wrapper sheet -Height 130" Boundary 55 Wrapper sheet -Thickness 0.075" 55 Wrapper sheet -Thickness 0.035" Shim -Thickness 0.075" Shim -width 9" Corner -Thickness 3/16" Corner -length 22" Corner -Width 8.5" Holtec Drawing 1000 Rack B1 12 X 9 Rack B2 12 X 9 Rack G1 10x9 Rack G2 10 X 9 Rack C llx 9 Rack D 11 X 8 Rack E 10 X 10 Rack F1 10 X 12 Rack F2 10 X 12 Region I Total Attachment D SpreadSheets Source Density Calculation 0.075 8.00 130.00 0.075 0.035 0.075 9.000 0.1875 22.000 108 108 99 88 100 90 90 120 120 923 160 1083 8.500 Rack C Rack D Rack E Rack F1 Rack F2 FC08513 11 x9 180 180 11 X 8 161 161 Page 141 of 269 10 X 10 161 161 10 X 12 218 218 10 X 12 218 218 I Panel Total 1663 1663 Panel Volume (in"3 69.6 36.4 !Total Volume 115744.80 60533.20 Region I FCS MCNP Source lnputs.xlsx 1gs 1001 & 1002 Box3 Box4 Box5 Box 6 408.618 408.618 408.618 408.618 278.4 278.4 278.4 278.4 137.13375 88.595 88.595 88.595 Box 12 Box 13 Box 14 Box 15 408.618 408.618 408.618 408.618 278.4 278.4 278.4 278.4 137.13375 137.13375 137.13375 137.13375 9a 9b 9c 4.4262 2.8704 0.4150 20 18 12 88.523 51.667 4.979 Mass 13371.733 6065284.6 2171.520 984979.8 ss ss shims length corners Rack Dimensions 70 8 103.824 70 8 103.824 70 8 95.172 70 8 86.52 Attachment D SpreadSheets Source Density Calculation Box 7 Box 8 408.618 408.618 278.4 278.4 137.13375 137.13375 Box 16 Box 17 408.618 408.618 278.4 278.4 137.13375 185.6725 9d SS corners 0.2691 35.0625 12 16 3.229 561 overhang 16ths 77.868 77.868 77.868 77.868 FC08513 Box9 Page 142 of 269 408.618 278.4 88.595 Box 18 408.618 278.4 185.6725 E-W N-S volume panel all panels volume panel all panels 1253.67 16297.77 940.26 9402.56 1253.67 16297.77 940.26 9402.56 1149.20 12641.22 940.26 9402.56 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 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 cm3 Region II cm3 Overall cm3 755,598.11 4,603,090.31 5,358,688.42 Total volume AI-Total volume AI-Total volume AI-B4C cm3 B4C cm3 Overall B4C cm3 364,972.69 1,896,717.45 2,261,690.13 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 FCS MCNP Source lnputs.xlsx Attachment D SpreadSheets FC08513 Source Density Calculation Reference FC08514, Reference FC08514, Page 46 Page47 Page 145 of 269 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 FCS MCNP Source lnputs.xlsx 40 41 42 43 44 Total Reference FC08514, Page46 Origen Inputs to MCNP FCS Source 18 Month Decay neutrons/sf Mev mev 3.00E+OO 2.18E+10 4.80E+OO 3.95E+10 6.43E+OO 1.12E+10 8.19E+OO 3.57E+09 2.00E+Ol 1.23E+09 2.36E+ll Attachment D SpreadSheets Source Density Calculation 40 41 42 43 44 Total Reference FC08514, Page47 Origen Inputs to MCNP FCS Source 14 Month Decay neutrons/sf Mev mev 3.00E+OO 2.25E+10 4.80E+OO 4.09E+10 6.43E+OO 1.16E+l0 8.19E+OO 3.69E+09 2.00E+Ol 1.27E+09 FC08513 Page 146 of 269 FCS MCNP Source lnputs.xlsx Surface Number MCNP 6000 pz 929.945 $Top of Fuel Racks 6050 pz 521.025 $ Bottom of Fuel Rae Not Used Not Used Not Used Not Used Not Used Not Used 2250 px -2179.32 $ SFP East Inner wal 6200 px -2770.022 $West Boundary of 6300 py 1024.128 $ North Boundary o 2050 py 76.20 $ SFP South Inner w FCS Fuel Source Attachment D SpreadSheets Source Density Calculation Fuel Racks Plus Fuel Top of Fuel Racks Bottom of Fuel Racks East Edge of Fuel Racks West Edge of Fuel Racks North Edge of Fuel Racks South Edge of Fuel Racks South Edge of CPA East Edge of CPA Alternate dimensions East Edge of Fuel Racks West Edge of Fuel Racks North Edge of Fuel Racks South Edge of Fuel Racks 944 -2179.3200 -2770.0220 76.2000 1024.1280 929.9450 521.0250 = (995'6" -979'0") + 161" + 7.125" = (995'6" -979'0") + 7.125" = -(60'6" + 5'6" + 5'6" + 1.578" = -(60'6" + 5'6" + 5'6" + 20'7") + 0.698" = + 2'6" + 33'3"-3.203" = + 2'6" + 3.042" = + 2'6" + 33'3"-96"-3.813" = -(60'6" + 5'6" + 5'6" + 20'7") + 1.313" = -(60'6" + 5'6" + 5'6" = -(60'6" + 5'6" + 5'6" + 19.38) = + 2'6" + 33'3"-3.203" = + 2'6" + 3.042" 71.50 90.88 2.50 33.60 30.51 17.09 FC08513 17.094 -71.632 -92.025 35.483 2.753 27.432 -83.974 -71.500 -90.880 33.600 2.500 Fuel + Rack SQ Ft 667.47 602.66 19.38 31.10 19.38 31.10 13.42 47 of 269 FCS MCNP Source lnputs.xlsx 929.94480 11405-S-61 + 1002, 1004 521.02512 11405-S-61 + 1002, 1004 -2183.34336 11405-S-61 -2804.93206 11405-S-61 1081.52184 11405-S-61 83.91144 11405-S-61 836.12736 11405-S-61 -2559.53758 TDB--2179.32000 11405-S-61 -2770.02240 11405-S-61 1024.12800 11405-S-61 76.20000 11405-S-61 0.95021 602.65509 Fuel Volume (Fuel+ Racks+ Air) 228,971,853.30 cm3 8086.1 ft3 z z X X y y y X X X y y Attachment D SpreadSheets Source Density Calculation 168.125 7.125 1.578 0.698 3.203 3.042 3.813 1.313 FC08513 14.010 Page 148 of 269 0.594 0.132 0.058 0.267 0.253 0.318 0.109 FCS Geo Inputs 20161017.xlsx Elevation 995.5-1007 SFP outer width outer length Wall thinckness Wall thinckness Canal end Top of Pool Bottom of Pool Pool basemat Distance from Containment toAUX SFP Weir gate @1008.5' Width @1039' Width =5.5+20'7"+5.5 =5.5+33'3"+4 5'6" 4'0" 1038'6" 995.5' 10'+989'0" =55'6" + 5'0" 5' 7'8" feet 31.5833 42.75 5.5 4.0 1038.5 43.0 16.5 60.5 Attachment D SpreadSheets MCNP Geometry Inputs OPPD Drawing 11405-S-51 11405-S-51 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-60 11405-S-48 11405-S-61 11405-S-61 Material Concrete FC08513 Page 149 of 269 FCS Geo Inputs 20161017.xlsx Surface Number MCNP 2200 px -2011.* 2250 px -2179. 2300 px -2806. 2350 px -2974.3 2000 py -45.7..! 2050 py 76.20 2100 py 1089.6 2150 py 1257.3 2400 py -198.1 2450 py -350.5 1550 px -1844.C Not Used 2550 px -1969. 2600 py 175.2 2800 px -2219.S 2850 px -2418.C East Exterior Wall of SFP East Interior Wall of SFP West Interior Wall of SFP West Exterior Wall of SFP South Exterior Wall of SFP South Interior Wall of SFP North Interior Wall of SFP North Exterior Wall of SFP South Interior Wall ofTC South Exterior Wall ofTC East Exterior Wall of TC East Interior Wall of TC at bottom East Interior Wall of TC overall North Exterior Wall of TC East Edge of Weir Gate West Edge of Weir Gate Attachment D SpreadSheets MCNP Geometry Inputs feet = 60'6" + 5'6" -66.00 = 60'6" + 5'6" + 5'6" -71.50 = 60'6" + 5'6" + 5'6" -20'7" -92.08 = 60'6" + 5'6" + 5'6" + 20'7" + 5'6" -97.58 = -1'6" -1.50 = + 2'6" 2.50 = + 2'6" + 33'3" 35.75 = + 2'6" + 33'3" + 5'6" 41.25 =-1'6"-5'0" -6.50 =-1'611-5'0" -5'0" -11.50 = 60'6" -60.50 = 60'6" + 2" -62.50 = 60'6" + 2'0" + 2'0" + 1.5" -64.63 = + 1'6" + 4'3" 5.75 =-(60'6" + 2'6" + 13'1" + Avg (4' + 2'6") -72.83 =-{60'6" + 2'6" + 13'1"-Avg (4' + 2'6") -79.33 em -2011.680 -2179.320 -2806.700 -2974.340 -45.720 76.200 1089.660 1257.300 -198.120 -350.520 -1844.040 -1905.000 -1969.770 175.260 -2219.960 -2418.080 Dimensional OPPD Drawing Attribute 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 X X X X y y y y y y X X X y X X FC08513 Page 150 of 269 FCS Geo Inputs 20161017.xlsx Surface Number MCNP 1500 px -South Exterior Wall of Aux (Room 3) 1550 px -South Interior Wall of Aux (Room 3) 1150 PY 2 North Interior Wall of Aux (Room 3) 1100 py 2 North Exterior Wall of Aux (Room 3) 1050 py 4 North Interior Wall of Aux (Room 69, below 1044') 1000 py 4 North Exterior Wall of Aux (Room 69, below 1044') Not Used South Interior Wall of East of SFP (Room 69, below 1044') Not Used South Exterior Wall of East of SFP (Room 69, below 1044') 1300 py -East Exterior Wall (Room 25A) 1350 py -East Interior Wall (Room 25A) 2900 px -West Interior Wall (Room 25A) 2350 px -West Exterior Wall (Room 25A) 2700 py -South Interior Wall (Room 25A) 2750 py -South Exterior Wall (Room 25A) 1650 py 1 South Exterior Wall of CR (Room 77) 1600 py 1 South Interior Wall of CR (Room 77) 1850 px 3 East Exterior Wall of CR (Room 77) 1800 px 3 East Interior Wall of CR (Room 77) 1750 px 1 West Interior Wall of CR (Room 77) 1700 px 1 West Exterior Wall of CR (Room 77) 1550 px -East Exterior Wall of Aux (Room 3) 1500 px -1 East Interior Wall of Aux (Room 3) 1450 px -3 West Interior Wall of Aux (Room 3) 1400 px -3 West Exterior Wall of Aux (Room 3) 3100 pz 1 Room 67, 68, 69 Floor (1025') 3110 pz 1 Room 67, 68, 69 Thickness (6") 3000 pz 3 Room 3 Roof (1083') 3015 pz 3 Room 3 Thickness (6") 3020 pz 2 Room 77 Roof (1057') 3030 pz 2 Room 77 Thickness (1'6") =-62' Attachment D SpreadSheets MCNP Geometry Inputs =-62' + 1'6" = +68'6" = + 68'6" + 1'6" = + 134'0" -1'6" = + 134'0" = + 21'6"-4'3" = + 21'6" + 1'6"-4'3" =-60'6" =-62'-1'6" = -(60'6" + 5'6" + 5'6" + 20'7" + 5'6" -1.5) =-(60'6" + 5'6" + 5'6" + 20'7" + 5'6") =-(62' + 21'11" + 0'9") =-(62' + 21'11" -0'9") = 29'0" + 36'0" -1'0" = 29'0" + 36'0" + 0'3" = 118'0" = 118'0"-1'6" = (1'0" + 18'6" + 18'6" + 0'9") = (1'0" + 18'6" + 18'6"-0'9") =-60'6" = -60'6" -1'6" = -60'6"-64'6" + 1'6" = -60'6" -64'6" = + 1025'0" -979'0" = + 1025'0"-979'0"-0'6" = + 1083'0" -979'0" = + 1083'0" -979'0"-0'6" = + 1057'0"-979'0" = + 1057'0"-979'0" -1'6" feet -62.00 -60.50 68.50 70.00 132.50 134.00 -17.25 -18.75 -60.50 -62.00 -96.08 -97.58 -39.75 -41.25 64.00 65.25 118.00 116.50 38.75 37.25 -60.50 -62.00 -123.50 -125.00 46.00 45.50 104.00 103.50 78.00 76.50 em -1889.76 -1844.04 2087.88 2133.60 4038.60 4084.32 -525.78 -571.50 -1844.04 -1889.76
-2928.62 -2974.34 -1211.58 -1257.30 1950.72 1988.82 3596.64 3550.92 1181.10 1135.38 -1844.04 -1889.76 -3764.28 -3810.00 1402.08 1386.84 3169.92 3154.68 2377.44 2331.72 DWG 11405-S-51 11405-S-51 11405-S-57 11405-S-57 11405-S-55 11405-S-55 11405-S-55 11405-S-55 11405-S-51 11405-S-51 11405-S-57 11405-S-57 11405-S-57 11405-S-57 11405-S-55 11405-S-55 11405-S-55 11405-S-55 11405-S-55 11405-S-55 11405-S-51 11405-S-51 11405-S-51 11405-S-51 11405-S-63 11405-S-59 11405-S-63 11405-S-59 11405-S-63 11405-S-59 FC08513 Page 151 of 269 MCNP cartesian coordinate Dimensi on Perspect ive y y y y y y y y X X X X y y y y X X X X X X X
X z FCS Geo Inputs 20161017.xlsx 3060 pz 1 Room 69 Roof {1044') 3070 pz 1 Room 69 Thickness {6") 3040 pz 1 Room 77 Floor {1036') 3050 pz 1 Room 77 Thickness {0'6") Not Used Containment top Not Used Containment Dome Inside elevation Not Used Top of Cylinder 1099' Elevation Not Used Containment Basemat Top Elevation Not Used Containment Basemat Bottom Elevation 3200 pz 1 SFP Walkway 3210 pz ! Bottom of Weir Gate 3220 pz ' Pool Bottom Not Used Aux Bldg Basemat Top Elevation 3250 pz Aux Bldg Basemat Bottom Elevation 7000 pz Plant Grade Elevation 7200 pz 3 EAB Elevation Hwy 75 7100 px -4 EAB Elevation Hwy 75 Step 10000 so LPZ Not Used LPZ Containment Bu1ldmg Not Used Liner Interior Radius Not Used Liner Thickness 5100 rcc 0 Containment Interior Radius 5000 rcc 0 Containment Exterior Radius Not Used Containment Dome Thickness Not Used Containment Dome radius Not Used Containment Dome Inside elevation Not Used Containment Dome radius at 1099' elevation Not Used Containment radius at 1099' elevation for dome shape Railroad Siding Doors Not Used Elevation at botton {1004'0", Use Plant Grade level 1004'6") 3330,3340, Top of Door Opening {22' above 1004'0") 3330,3340, Door Opening width including Frame{16'6") 3330, 3340, South Door Opening from CL of Containemt {16'6") Attachment D SpreadSheets MCNP Geometry Inputs = + 1044'0" -979'0" = + 1044'0"-979'0" -0'6" = + 1036'0" -979'0" = + 1036'0" -979'0" -0'6" = + 1128'4.5"-979'0"+ 0'3" = + 1128'4.5"-979'0" = + 1099'0" -979'0" = + 991'0" -979'0" = + 979'0" -979'0" = + 1038'6" -979'0" = + 1008'6" -979'0" -1'6" = + 995'6"-979'0" = + 989'0" -979'0" = + 983'6" -979'0" = + 1004'6"-979'0" = + 1080'0" -979'0" = -1540 = 1 Mile (5280) = 9 Miles {9*5280) 55'0" 0'0.25" 58'10.75" 3' 90'0" 1128'4.5" 20'0" 35'0" = + 1004'6" -979'0" = + 1004'0" + 22'0" -979'0" = 16' =-62' + 1'6" + 2'11.5" 65.00 64.50 57.00 56.50 149.63 149.38 120.00 12.00 0.00 59.00 29.50 16.50 10.00 4.50 25.50 101.00 -1540.00 5280.00 47520.00 55.00 0.021 58.90 3.00 90.00 149.38 20.00 35.00 25.50 47.00 16.00 -57.54 1981.20 1965.96 1737.36 1722.12 4560.57 4552.95 3657.60 365.76 0.00 1798.32 899.16 502.92 304.80 137.16 777.24 3078.48 -46939.20 160934.40 1448409.60 1676.40 0.63 1677.035 1795.145 2743.20 4552.95 777.24 1432.56 487.68 -1753.87 o.,.,...,. 11:;.' 11405-S-63 "' 11405-S-59 11405-S-63 11405-S-59 11405-S-63 11405-S-59 11405-S-59 11405-S-63 11405-S-12 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-61 11405-S-59 11405-S-251-1:12 11405-S-270 11405-S-55 11405-S-55 11405-S-55 11405-S-2 11405-S-2 11405-S-2 11405-S-2 11405-S-2 11405-S-51 11405-S-51 & 11405-S-74 11405-S-51 & 11405-S-74 11405-S-51 & 11405-S-74 FC08513 of269 z z z z X so so cz cz cz s s s s z z y y FCS Geo Inputs 20161017.xlsx 3330,3340,: North Door Opening from CL of Containemt (16'6") 3330 rpp -Door 1004-1A design thickness (20 ga) 3340 rpp -Door 1004-1C Sheathing design thickness (14 ga) 3350 rpp -Door 1004-1C design insulation thickness Not Used Door "B" Frame Dimension (Ignore Security Gate) Attachment D SpreadSheets MCNP Geometry Inputs =-62' + 1'6" + 2'11.5" 16' = 20 guage steel (.0375") = 14 guage steel (.078125") =Foam insulation (3.75") Use remainder of door thickness 0.078125*2)=3.84375 N/A FC08513 o .... ,.,,.,. 11:. ,...f'>ao .... -41.54 -1266.19 11405-S-51 & 11405-S-74 y 0.03750 0.0953 232552 X 0.07813 0.1984 G-576 X 3.84375 9.7631 G-576 X (4.0 FCS Geo Inputs 20161017.xlsx AUX Building Elevation North Outerwall to Containment Centerline 995.5-1007 North Wall Thickness West Wall Thickness South Wall Thickness East Wall Thickness East Outerwall to Containment Centerline Roof Elevation Roof Thickness West Wall to East Wall North Wall to South Wall Containment Building Liner Interior Radius Liner Thickness Exterior Radius Basemat Top Elevation Containment Basemat Bottom Elevation Aux Bldg Basemat Bottom Elevation SFP Bottom Elevation Containment Dome Thickness Containment Dome radius Containment Dome Inside elevation Containment Dome radius at 1099' elevation Containment radius at 1099' elevation for dome shape Containment Cylinder Elevation Top of Dome to 1099' Elevation Railroad Siding Doors Attachment D SpreadSheets MCNP Geometry Inputs 134' 1.5' 2.0' 2.0' 2.0' 60.5' 1083' 64.5' 132.0' 55'0" 0'0.25" 58'10.75" 989'0" 979'0" 983'6" 995'6" 3' 90'0" 1128'4.5" 20'0" 35'0" 1099'0" 29'4.5" Insulation Thickness FC08513 Page 154 of 269 feet OPPD Drawin Dwg Loc 134 11405-S-51 1.5 11405-S-67 2 11405-S-67 2 11405-S-67 2 11405-S-67 60.5 11405-S-67 11405-S-60 64.5 11405-S-60 132 11405-S-60 55.000 11405-S-55 0.021 11405-S-55 58.896 11405-S-55 989.0 11405-S-60 979.0 11405-S-64 983.5 11405-S-64 995.5 11405-S-64 3.00 11405-5-2 90.00 11405-S-2 1128.375 11405-S-2 20.00 11405-S-2 35.00 11405-S-2 1099.00 11405-S-2 29.375 11405-S-2 FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets MCNP Geometry Inputs x-direction Steel y-direction 14 gauge Steel y-direction Steel z-direction 14 gauge Steel z-direction x-directionl2o gauge Steel Gauge Measure (inches) 14 20 5/64 3/80 FC08513 3764.375251 3764.280001 0.095251 numerical Centimeters 0.078125 1.984375 0.078125 0.0375 0.9525 0.037500 FCS Geo Inputs 20161017.xlsx Scatch pad Area c Rollup Door 1004-1C c Attachment D SpreadSheets MCNP Geometry Inputs 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 405 4 407 4 408 5 -7.82 -7.82 -3330 -3340 3350 $ Railroad Siding Door 1004-1C $ Railroad Siding Door 1004-1A -0.0482 -3350 $ Railroad Siding Door 1004-1A insul 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 Opening 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" Scatch pad Area Volume of Door Area 30440.6 238045.571 .:5L46998.409 3115""6) 3 4 131236.0951 1026266.26 3115762.314 24365261.3 14611492.84 11334053 p 13657.5349 FC08513 Page 156 of 269 1.755004 g/cm3 FCS Geo Inputs 20161017.xlsx OPPD Drawings: 11405-S-02, 11405-S-08, 11405-S-08 degrees radians X dome Y dome edge. edge. (Not used) (Not used) 120.042 0 0.000 55.00 120.04 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 0.017 0.035 0.052 0.070 0.087 0.105 0.122 0.140 0.157 0.175 0.192 0.209 0.227 0.244 0.262 0.279 0.297 0.314 0.332 0.349 0.367 0.384 0.401 0.419 0.436 0.454 0.471 0.489 0.506 0.524 55.00 54.99 54.97 54.95 54.92 54.89 54.85 54.81 54.75 54.70 54.63 54.56 54.49 54.41 54.32 54.23 54.13 54.02 53.91 53.79 53.67 53.54 53.41 53.27 53.13 52.98 52.82 52.66 52.49 52.32 120.39 120.74 121.09 121.44 121.79 122.13 122.48 122.83 123.17 123.51 123.86 124.20 124.54 124.88 125.22 125.55 125.89 126.22 126.55 126.88 127.21 127.53 127.86 128.18 128.49 128.81 129.12 129.43 129.74 130.04 60.667 Attachment D SpreadSheets MCNP Geometry Inputs degrees radians 0 0.000 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 0.017 0.035 0.052 0.070 0.087 0.105 0.122 0.140 0.157 0.175 0.192 0.209 0.227 0.244 0.262 0.279 0.297 0.314 0.332 0.349 0.367 0.384 0.401 0.419 0.436 0.454 0.471 0.489 0.506 0.524 X 58.90 58.90 58.89 58.87 58.84 58.81 58.77 58.73 58.68 58.62 58.55 58.48 58.40 58.31 58.22 58.12 58.01 57.90 57.77 57.65 57.51 57.37 57.23 57.07 56.91 56.75 56.57 56.39 56.21 56.02 55.82 y 120.04 120.44 120.84 121.25 121.65 122.05 122.45 122.84 123.24 123.64 124.04 124.43 124.82 125.22 125.61 125.99 126.38 126.77 127.15 127.53 127.91 128.28 128.66 129.03 129.40 129.76 130.12 130.48 130.84 131.19 131.54 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 0.000 0.017 0.035 0.052 0.070 0.087 0.105 0.122 0.140 0.157 0.175 0.192 0.209 0.227 0.244 0.262 0.279 0.297 0.314 0.332 0.349 0.367 0.384 0.401 0.419 0.436 0.454 0.471 0.489 0.506 0.524 X 90.00 89.99 89.95 89.88 89.78 89.66 89.51 89.33 89.12 88.89 88.63 88.35 88.03 87.69 87.33 86.93 86.51 86.07 85.60 85.10 84.57 84.02 83.45 82.85 82.22 81.57 80.89 80.19 79.47 78.72 77.94 FC08513 Page 157 of 269 y 59.38 60.95 62.52 64.09 65.65 67.22 68.78 70.34 71.90 73.45 75.00 76.55 78.09 79.62 81.15 82.67 84.18 85.69 87.19 88.68 90.16 91.63 93.09 94.54 95.98 97.41 98.83 100.23 101.63 103.01 104.37 FCS Geo Inputs 20161017.xlsx 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 0.541 0.559 0.576 0.593 0.611 0.628 0.646 0.663 0.681 0.698 0.716 0.733 0.750 0.768 0.785 0.803 0.820 0.838 0.855 0.873 0.890 0.908 0.925 0.942 0.960 0.977 0.995 1.012 1.030 1.047 1.065 1.082 1.100 1.117 1.134 1.152 1.169 1.187 52.14 130.34 51.96 130.64 51.77 130.93 51.58 131.23 51.38 131.51 51.18 131.80 50.97 132.08 50.76 132.36 50.54 132.63 50.32 132.90 50.09 133.16 49.86 133.42 49.63 133.68 49.39 133.94 49.14 134.18 48.89 134.43 48.64 134.67 48.38 134.90 48.12 135.14 47.86 135.36 47.59 135.58 47.31 135.80 47.04 136.01 46.76 136.22 46.47 136.43 46.18 136.62 45.89 136.82 45.60 137.00 45.30 137.19 45.00 137.36 44.70 137.53 44.39 137.70 44.08 137.86 43.77 138.02 43.45 138.17 43.13 138.31 42.81 138.45 42.49 138.59 Attachment D SpreadSheets MCNP Geometry Inputs 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 0.541 0.559 0.576 0.593 0.611 0.628 0.646 0.663 0.681 0.698 0.716 0.733 0.750 0.768 0.785 0.803 0.820 0.838 0.855 0.873 0.890 0.908 0.925 0.942 0.960 0.977 0.995 1.012 1.030 1.047 1.065 1.082 1.100 1.117 1.134 1.152 1.169 1.187 55.61 131.89 55.41 132.23 55.19 132.57 54.97 132.90 54.74 133.23 54.51 133.56 54.27 133.88 54.02 134.20 53.77 134.52 53.52 134.83 53.26 135.13 52.99 135.43 52.72 135.73 52.44 136.02 52.16 136.31 51.88 136.59 51.59 136.86 51.29 137.13 50.99 137.40 50.68 137.66 50.37 137.92 50.06 138.17 49.74 138.41 49.42 138.65 49.09 138.88 48.76 139.11 48.43 139.33 48.09 139.55 47.75 139.76 47.40 139.96 47.05 140.16 46.70 140.35 46.34 140.54 45.98 140.71 45.62 140.89 45.25 141.05 44.89 141.21 44.52 141.37 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 0.541 0.559 0.576 0.593 0.611 0.628 0.646 0.663 0.681 0.698 0.716 0.733 0.750 0.768 0.785 0.803 0.820 0.838 0.855 0.873 0.890 0.908 0.925 0.942 0.960 0.977 0.995 1.012 1.030 1.047 1.065 1.082 1.100 1.117 1.134 1.152 1.169 1.187 77.15 76.32 75.48 74.61 73.72 72.81 71.88 70.92 69.94 68.94 67.92 66.88 65.82 64.74 63.64 62.52 61.38 60.22 59.05 57.85 56.64 55.41 54.16 52.90 51.62 50.33 49.02 47.69 46.35 45.00 43.63 42.25 40.86 39.45 38.04 36.61 35.17 33.71 FC08513 Page 158 of 269 105.73 107.07 108.39 109.70 111.00 112.28 113.54 114.78 116.01 117.23 118.42 119.60 120.75 121.89 123.01 124.12 125.20 126.26 127.30 128.32 129.32 130.30 131.25 132.19 133.10 133.99 134.86 135.70 136.52 137.32 138.09 138.84 139.57 140.27 140.94 141.59 142.22 142.82 FCS Geo Inputs 20161017.xlsx 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 1.204 1.222 1.239 1.257 1.274 1.292 1.309 1.326 1.344 1.361 1.379 1.396 1.414 1.431 1.449 1.466 1.484 1.501 1.518 1.536 1.553 1.571 1.588 1.606 1.623 1.641 1.658 1.676 1.693 1.710 1.728 1.745 1.763 1.780 1.798 1.815 1.833 1.850 42.17 41.84 41.51 41.18 40.85 40.51 40.18 39.84 39.50 39.16 38.82 38.47 38.13 37.78 37.44 37.09 36.74 36.40 36.05 35.70 35.35 35.00 -35.35 -35.70 -36.05 -36.40 -36.74 -37.09 -37.44 -37.78 -38.13 -38.47 -38.82 -39.16 -39.50 -39.84 -40.18 -40.51 138.71 138.84 138.95 139.06 139.17 139.27 139.36 139.45 139.53 139.60 139.67 139.74 139.80 139.85 139.89 139.93 139.97 139.99 140.01 140.03 140.04 140.04 140.04 140.03 140.Q1 139.99 139.97 139.93 139.89 139.85 139.80 139.74 139.67 139.60 139.53 139.45 139.36 139.27 Attachment D SpreadSheets MCNP Geometry Inputs 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 1.204 1.222 1.239 1.257 1.274 1.292 1.309 1.326 1.344 1.361 1.379 1.396 1.414 1.431 1.449 1.466 1.484 1.501 1.518 1.536 1.553 1.571 1.588 1.606 1.623 1.641 1.658 1.676 1.693 1.710 1.728 1.745 1.763 1.780 1.798 1.815 1.833 1.850 44.14 43.77 43.39 43.01 42.62 42.24 41.85 41.46 41.07 40.68 40.29 39.89 39.50 39.10 38.70 38.30 37.90 37.50 37.10 36.70 36.30 35.90 35.50 35.10 34.70 34.30 33.90 33.50 33.10 32.70 32.30 31.91 31.51 31.12 30.73 30.34 29.95 29.56 141.51 141.65 141.79 141.92 142.04 142.15 142.26 142.36 142.45 142.54 142.62 142.69 142.76 142.82 142.87 142.92 142.95 142.99 143.01 143.03 143.04 143.04 143.04 143.03 143.01 142.99 142.95 142.92 142.87 142.82 142.76 142.69 142.62 142.54 142.45 142.36 142.26 142.15 69 70 71 72 73 74 75 76 77 78 79 80 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 105 106 1.204 1.222 1.239 1.257 1.274 1.292 1.309 1.326 1.344 1.361 1.379 1.396 1.414 1.431 1.449 1.466 1.484 1.501 1.518 1.536 1.553 1.571 1.588 1.606 1.623 1.641 1.658 1.676 1.693 1.710 1.728 1.745 1.763 1.780 1.798 1.815 1.833 1.850 32.25 30.78 29.30 27.81 26.31 24.81 23.29 21.77 20.25 18.71 17.17 15.63 14.08 12.53 10.97 9.41 7.84 6.28 4.71 3.14 1.57 0.00 -1.57 -3.14 -4.71 -6.28 -7.84
-9.41 -10.97 -12.53 -14.08 -15.63 -17.17 -18.71 -20.25 -21.77 -23.29 -24.81 FC08513 Page 159 of 269 143.40 143.95 144.47 144.97 145.44 145.89 146.31 146.70 147.07 147.41 147.72 148.01 148.27 148.50 148.70 148.88 149.03 149.16 149.25 149.32 149.36 149.37 149.36 149.32 149.25 149.16 149.03 148.88 148.70 148.50 148.27 148.01 147.72 147.41 147.07 146.70 146.31 145.89 FCS Geo Inputs 20161017.xlsx 107 108 109 110 111 112 113 114 115 116 117 118 119 120 121 122 123 124 125 126 127 128 129 130 131 132 133 134 135 136 137 138 139 140 141 142 143 144 1.868 1.885 1.902 1.920 1.937 1.955 1.972 1.990 2.007 2.025 2.042 2.059 2.077 2.094 2.112 2.129 2.147 2.164 2.182 2.199 2.217 2.234 2.251 2.269 2.286 2.304 2.321 2.339 2.356 2.374 2.391 2.409 2.426 2.443 2.461 2.478 2.496 2.513 -40.85 139.17 -41.18 139.06 -41.51 138.95 -41.84 138.84 -42.17 138.71 -42.49 138.59 -42.81 138.45 -43.13 138.31 -43.45 138.17 -43.77 138.02 -44.08 137.86 -44.39 137.70 -44.70 137.53 -45.00 137.36 -45.30 137.19 -45.60 137.00 -45.89 136.82 -46.18 136.62 -46.47 136.43 -46.76 136.22 -47.04 136.01 -47.31 135.80 -47.59 135.58 -47.86 135.36 -48.12 135.14 -48.38 134.90 -48.64 134.67 -48.89 134.43 -49.14 134.18 -49.39 133.94 -49.63 133.68 -49.86 133.42 -50.09 133.16 -50.32 132.90 -50.54 132.63 -50.76 132.36 -50.97 132.08 -51.18 131.80 Attachment D SpreadSheets MCNP Geometry Inputs 107 108 109 110 111 112 113 114 115 116 117 118 119 120 121 122 123 124 125 126 127 128 129 130 131 132 133 134 135 136 137 138 139 140 141 142 143 144 1.868 1.885 1.902 1.920 1.937 1.955 1.972 1.990 2.007 2.025 2.042 2.059 2.077 2.094 2.112 2.129 2.147 2.164 2.182 2.199 2.217 2.234 2.251 2.269 2.286 2.304 2.321 2.339 2.356 2.374 2.391 2.409 2.426 2.443 2.461 2.478 2.496 2.513 29.18 28.79 28.41 28.03 27.66 27.28 26.91 26.55 26.18 25.82 25.46 25.10 24.75 24.40 24.05 23.71 23.37 23.04 22.71 22.38 22.06 21.74 21.43 21.12 20.81 20.51 20.21 19.92 19.64 19.36 19.08 18.81 18.54 18.28 18.03 17.78 17.53 17.29 142.04 141.92 141.79 141.65 141.51 141.37 141.21 141.05 140.89 140.71 140.54 140.35 140.16 139.96 139.76 139.55 139.33 139.11 138.88 138.65 138.41 138.17 137.92 137.66 137.40 137.13 136.86 136.59 136.31 136.02 135.73 135.43 135.13 134.83 134.52 134.20 133.88 133.56 107 108 109 110 111 112 113 114 115 116 117 118 119 120 121 122 123 124 125 126 127 128 129 130 131 132 133 134 135 136 137 138 139 140 141 142 143 144 1.868 1.885 1.902 1.920 1.937 1.955 1.972 1.990 2.007 2.025 2.042 2.059 2.077 2.094 2.112 2.129 2.147 2.164 2.182 2.199 2.217 2.234 2.251 2.269 2.286 2.304 2.321 2.339 2.356 2.374 2.391 2.409 2.426 2.443 2.461 2.478 2.496 2.513 -26.31 -27.81 -29.30 -30.78 -32.25 -33.71 -35.17 -36.61 -38.04 -39.45 -40.86 -42.25
-43.63 -45.00 -46.35 -47.69 -49.02 -50.33 -51.62 -52.90 -54.16 -55.41 -56.64 -57.85 -59.05 -60.22 -61.38 -62.52 -63.64 -64.74 -65.82 -66.88 -67.92 -68.94 -69.94 -70.92 -71.88 -72.81 FC08513 Page 160 of 269 145.44 144.97 144.47 143.95 143.40 142.82 142.22 141.59 140.94 140.27 139.57 138.84 138.09 137.32 136.52 135.70 134.86 133.99 133.10 132.19 131.25 130.30 129.32 128.32 127.30 126.26 125.20 124.12 123.01 121.89 120.75 119.60 118.42 117.23 116.01 114.78 113.54 112.28 FCS Geo Inputs 20161017.xlsx 145 146 147 148 149 150 151 152 153 154 155 156 157 158 159 160 161 162 163 164 165 166 167 168 169 170 171 172 173 174 175 176 177 178 179 180 2.531 2.548 2.566 2.583 2.601 2.618 2.635 2.653 2.670 2.688 2.705 2.723 2.740 2.758 2.775 2.793 2.810 2.827 2.845 2.862 2.880 2.897 2.915 2.932 2.950 2.967 2.985 3.002 3.019 3.037 3.054 3.072 3.089 3.107 3.124 3.142 -51.38 131.51 -51.58 131.23 -51.77 130.93 -51.96 130.64 -52.14 130.34 -52.32 130.04 -52.49 129.74 -52.66 129.43 -52.82 129.12 -52.98 128.81 -53.13 128.49 -53.27 128.18 -53.41 127.86 -53.54 127.53 -53.67 127.21 -53.79 126.88 -53.91 126.55 -54.02 126.22 -54.13 125.89 -54.23 125.55 -54.32 125.22 -54.41 124.88 -54.49 124.54 -54.56 124.20 -54.63 123.86 -54.70 123.51 -54.75 123.17 -54.81 122.83 -54.85 122.48 -54.89 122.13 -54.92 121.79 -54.95 121.44 -54.97 121.09 -54.99 120.74 -55.00 120.39 -55.00 120.04 Attachment D SpreadSheets MCNP Geometry Inputs 145 146 147 148 149 150 151 152 153 154 155 156 157 158 159 160 161 162 163 164 165 166 167 168 169 170 171 172 173 174 175 176 177 178 179 180 2.531 2.548 2.566 2.583 2.601 2.618 2.635 2.653 2.670 2.688 2.705 2.723 2.740 2.758 2.775 2.793 2.810 2.827 2.845 2.862 2.880 2.897 2.915 2.932 2.950 2.967 2.985 3.002 3.019 3.037 3.054 3.072 3.089 3.107 3.124 3.142 17.06 133.23 16.83 132.90 16.61 132.57 16.39 132.23 16.19 131.89 15.98 131.54 15.78 131.19 15.59 130.84 15.41 130.48 15.23 130.12 15.05 129.76 14.89 129.40 14.73 129.03 14.57 128.66 14.43 128.28 14.29 127.91 14.15 127.53 14.03 127.15 13.90 126.77 13.79 126.38 13.68 125.99 13.58 125.61 13.49 125.22 13.40 124.82 13.32 124.43 13.25 124.04 13.18 123.64 13.12 123.24 13.07 122.85 13.03 122.45 12.99 122.05 12.96 121.65 12.93 121.25 12.91 120.84 12.90 120.44 12.90 120.04 145 146 147 148 149 150 151 152 153 154 155 156 157 158 159 160 161 162 163 164 165 166 167 168 169 170 171 172 173 174 175 176 177 178 179 180 2.531 2.548 2.566 2.583 2.601 2.618 2.635 2.653 2.670 2.688 2.705 2.723 2.740 2.758 2.775 2.793 2.810 2.827 2.845 2.862 2.880 2.897 2.915 2.932 2.950 2.967 2.985 3.002 3.019 3.037 3.054 3.072 3.089 3.107 3.124 3.142 -73.72 -74.61 -75.48 -76.32 -77.15 -77.94 -78.72 -79.47 -80.19 -80.89 -81.57 -82.22 -82.85 -83.45 -84.02 -84.57 -85.10 -85.60 -86.07 -86.51 -86.93 -87.33 -87.69 -88.03 -88.35 -88.63 -88.89 -89.12 -89.33 -89.51 -89.66 -89.78 -89.88 -89.95 -89.99 -90.00 FC08513 Page 161 of 269 111.00 109.70 108.39 107.07 105.73 104.38 103.01 101.63 100.23 98.83 97.41 95.98 94.54 93.09 91.63 90.16 88.68 87.19 85.69 84.18 82.67 81.15 79.62 78.09 76.55 75.00 73.45 71.90 70.34 68.78 67.22 65.65 64.09 62.52 60.95 59.38 FCS Geo Inputs 20161017.xlsx X y 120.042 60.667 55.00 0.00 55.00 120.04 X y -55.00 0.00 -55.00 120.04 Height above Basemat 1128'4.5"-90' -979' 59.375 Attachment D SpreadSheets MCNP Geometry Inputs OPPD Drawings: 11405-S-02, 11405-S-08, 11405-S-08 Angle from Containment Centerline at feet above basemat See drawing 11405-S-X Position 02 from Degrees radians centerline 0 0.000 93.50 1 0.017 93.49 2 0.035 93.44 3 0.052 93.37 4 0.070 93.27 5 0.087 93.14 6 0.105 92.99 7 0.122 92.80 8 0.140 92.59 9 0.157 92.35 10 0.175 92.08 11 0.192 91.78 12 0.209 91.46 13 0.227 91.10 14 0.244 90.72 15 0.262 90.31 16 0.279 89.88 17 0.297 89.41 18 0.314 88.92 19 0.332 88.41 20 0.349 87.86 21 0.367 87.29 22 0.384 86.69 23 0.401 86.07 24 0.419 85.42 25 0.436 84.74 26 0.454 84.04 27 0.471 83.31 28 0.489 82.56 29 0.506 81.78 30 0.524 80.97 FC08513 Page 162 of 269 Y Position from centerline 93' radius 59.38 61.00 62.62 64.24 65.86 67.48 69.10 70.71 72.32 73.92 75.52 77.12 78.71 80.30 81.87 83.45 85.01 86.57 88.11 89.65 91.18 92.70 94.21 95.71 97.20 98.68 100.14 101.60 103.04 104.46 105.87 FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets FC08513 MCNP Geometry Inputs 31 0.541 80.15 107.27 Page 163 of 269 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 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 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 69 1.204 33.51 146.20 Page 164 of 269 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 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 -22.62 149.61 105 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 107 1.868 -27.34 148.31 Page 165 of 269 108 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 -50.92 137.37 124 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 -73.68 116.63 143 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 145 2.531 -76.59 112.72 Page 166 of 269 146 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 89.65 162 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 FCS Geo Inputs 20161017 .xlsx Attachment D SpreadSheets MCNP Geometry Inputs -63.13 X y 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. 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 FC08513 Page 167 of 269 0> <0 N -0 CX> <0 ...-Q) C> ro a.. ..... c QJ E c ro ..... c 0 u Vl u LL. -0 0 0 0 0.0 "'" N 0 ..-< ..-< ..-< ..-< ----0 0 0 0 00 0.0 "'" N 0 0 r--0 0.0 0 Lf'l 0 m 0 N 0 ..-< 0 0 '";' 0 'i' 0 'f 0 t;-
FCS Geo Inputs 20161017.xlsx Alternate calculation for Rack weight (Not USED in MCNP) OPPD Drawing 1000 (Holtec Racks) Rack Al 10x8 16000 12800 2159.11 Rack A2 10x8 16000 12800 2169.02 Rack Bl 12 X 9 15200 12160 2182.96 Rack B2 12 x9 15200 12160 2181.60 Rack Gl 10 X 9 12600 10080 1886.33 Rack G2 10 x9 12600 10080 1885.20 Rack C 11 X 9 13900 11120 2035.67 Rack D 11 x8 12400 9920 1854.59 Rack E lOx 10 14000 11200 2052.06 Rack Fl lOx 12 16800 13440 2183.99 Rack F2 10 X 12 16800 13440 2380.32 lsubracting out weight below rack base plate 181.5325 89.231 19.47542 51.75 Attachment D SpreadSheets MCNP Geometry Inputs 10640.89 396.71 10630.98 9977.04 9978.40 8193.67 8194.80 9084.33 8065.41 9147.94 11256.01 11059.68 48176.49 FC08513 Page 169 of 269 FCS Geo Inputs 20161017.xlsx Attachment D SpreadSheets MCNP Geometry Inputs 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 75.83 feet I second Latitude 0.016667 16.66667 101.23 longitude 75.83 latitude 6.074 4.550 Containment Bridge End Distance Hwy 75 Access Property Corner 12+90N 20+00W 1305.9 1516.7 10+30N 40+00W 1042.7 3033.3 263.2 1516.7 1539.3 feet 0.292 miles 10+30N 52+00W 1042.7 3943.3 263.2 2426.7 O+OON 51+69.24V 0.0 3920.0 1305.9 2403.3 feet 1539 feet 2440.9 0.462 feet 2735.2 0.518 FC08513 Page 170 of 269 meters em 469.19 46918.96 meters 743.99 74398.61 meters 833.70 83369.6 FC08514 ext enr.xlsx Assembly ID A001 A002 A003 A004 AOOS A006 A007 A008 A009 A010 A011 A012 A013 A014 A015 A016 A017 A018 A019 A020 A021 A022 A023 A024 A025 A026 A027 A028 A029 A030 A031 A032 A033 A034 A035 A036 A037 A038 A039 A040 A041 A042 A043 A044 A045 B001 B002 B003 B004 BOOS B006 Initial U-235 Enr 1.400 1.380 1.380 1.380 1.390 1.390 1.400 1.400 1.400 1.390 1.390 1.390 1.390 1.390 1.390 1.400 1.400 1.400 1.390 1.390 1.390 1.390 1.390 1.390 1.390 1.390 1.380 1.380 1.380 1.380 1.380 1.390 1.390 1.390 1.390 1.400 1.400 1.400 1.400 1.400 1.390 1.390 1.380 1.390 1.390 2.370 2.370 2.370 2.370 2.370 2.390 Burn up MWD/T 7913 8638 8638 15310 8494 8494 8494 8252 8494 7913 7913 15310 15667 9026 15310 7842 17050 7842 15310 9026 15667 7842 8768 8439 15667 9026 15310 15310 15310 15667 9026 15310 15667 8439 15667 8439 17050 17050 7913 17050 8439 7842 8439 15667 15667 27041 27041 27041 27041 27041 27041 Assembly ID 14654 14654 14654 13310 14654 14654 14654 14654 14654 14654 14654 13310 13310 14654 13310 14654 14054 14654 13310 14654 13310 14654 14654 14654 13310 14654 13310 13310 13310 13310 14654 13310 13310 14654 13310 14654 14054 14054 14654 14054 14654 14654 14654 13310 13310 13690 13690 13690 13690 13690 13690 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.37461 0.37165 0.37135 0.37155 0.37595 0.37684 0.37643 0.37611 0.37581 0.37517 0.37468 0.37314 0.37173 0.37247 0.37186 0.37169 0.37166 0.37192 0.37195 0.37055 0.37105 0.37279 0.37138 0.37217 0.37123 0.37017 0.37110 0.37082 0.37027 0.37163 0.37231 0.37090 0.37097 0.37143 0.37121 0.37080 0.37036 0.37034 0.37132 0.37207 0.37225 0.37298 0.37688 0.37051 0.37144 0.37214 0.35354 0.37214 0.37214 0.35448 0.37288 Assembly ID 2/8/1975 2/8/1975 2/8/1975 10/15/1978 2/8/1975 2/8/1975 2/8/1975 2/8/1975 2/8/1975 2/8/1975 2/8/1975 10/15/1978 10/15/1978 2/8/1975 10/15/1978 2/8/1975 10/1/1976 2/8/1975 10/15/1978 2/8/1975 10/15/1978 2/8/1975 2/8/1975 2/8/1975 10/15/1978 2/8/1975 10/15/1978 10/15/1978 10/15/1978 10/15/1978 2/8/1975 10/15/1978 10/15/1978 2/8/1975 10/15/1978 2/8/1975 10/1/1976 10/1/1976 2/8/1975 10/1/1976 2/8/1975 2/8/1975 2/8/1975 10/15/1978 10/15/1978 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 Assembly ID 1.400 1.380 1.380 1.380 1.390 1.390 1.400 1.400 1.400 1.390 1.390 1.390 1.390 1.390 1.390 1.400 1.400 1.400 1.390 1.390 1.390 1.390 1.390 1.390 1.390 1.390 1.380 1.380 1.380 1.380 1.380 1.390 1.390 1.390 1.390 1.400 1.400 1.400 1.400 1.400 1.390 1.390 1.380 1.390 1.390 2.370 2.370 2.370 2.370 2.370 2.390 Assembly ID 4/10/1975 4/10/1975 4/10/1975 12/10/1978 4/10/1975 4/10/1975 4/10/1975 4/10/1975 4/10/1975 4/10/1975 4/10/1975 12/10/1978 12/10/1978 4/10/1975 12/10/1978 4/10/1975 12/8/1976 4/10/1975 12/10/1978 4/10/1975 12/10/1978 4/10/1975 4/10/1975 4/10/1975 12/10/1978 4/10/1975 12/10/1978 12/10/1978 12/10/1978 12/10/1978 4/10/1975 12/10/1978 12/10/1978 4/10/1975 12/10/1978 4/10/1975 12/8/1976 12/8/1976 4/10/1975 12/8/1976 4/10/1975 4/10/1975 4/10/1975 12/10/1978 12/10/1978 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 FC08513 Page 171 of 269 FCOB514 ext enr.xlsx Assembly ID B007 BOOB B009 BlOl B102 B103 B104 B105 B106 B107 BlOB B109 B110 B111 B112 B113 B114 B115 B116 B117 B11B B119 B120 B121 B122 B123 B124 B201 B202 B203 B204 B205 B206 B207 B20B B209 B210 B211 B212 B213 B214 B215 B216 B217 B21B B219 B220 B221 B222 B223 B224 Initial U-235 Enr 2.3BO 2.3BO 2.3BO 2.370 2.370 2.370 2.370 2.370 2.410 2.3BO 2.3BO 2.3BO 2.3BO 2.370 2.370 2.370 2.370 2.3BO 2.3BO 2.3BO 2.3BO 2.370 2.370 2.400 2.370 2.370 2.370 2.410 2.370 2.370 2.370 2.370 2.370 2.370 2.3BO 2.3BO 2.3BO 2.3BO 2.3BO 2.3BO 2.3BO 2.3BO 2.390 2.3BO 2.3BO 2.3BO 2.410 2.410 2.410 2.400 2.390 Burn up MWD/T 27041 27041 2B040 22090 2B564 22090 200BO 2B564 2B661 200BO 2B564 22090 22090 22030 2B564 22030 22090 22090 22030 2B564 200BO 2B564 2B564 200BO 22090 22030 22090 22190 29014 29014 22B10 29014 29014 29014 29014 22190 22190 22190 22B10 29014 22B10 22B10 29014 22B10 22B10 22B10 22660 22660 22660 22660 22B10 Assembly ID 13690 13690 13310 14054 13690 14054 14054 13690 13690 14054 13690 14054 14054 14054 13690 14054 14054 14054 14054 13690 14054 13690 13690 14054 14054 14054 14054 14054 13690 13690 14054 13690 13690 13690 13690 14054 14054 14054 14054 13690 14054 14054 13690 14054 14054 14054 14054 14054 14054 14054 14054 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.372Bl 0.37071 0.37057 0.35420 0.35354 0.354B6 0.35570 0.3544B 0.36146 0.35431 0.35225 0.35332 0.35364 0.353B7 0.353BO 0.35400 0.35451 0.35552 0.35395 0.35345 0.35520 0.35347 0.35393 0.35329 0.35473 0.3543B 0.35479 0.35333 0.354B2 0.35409 0.3542B 0.35427 0.35445 0.35376 0.353B9 0.35501 0.35479 0.35466 0.35490 0.35423 0.35453 0.35425 0.35252 0.35219 0.35250 0.3533B 0.3544B 0.35374 0.352BO 0.35360 0.353BO Assembly ID 9/30/1977 9/30/1977 10/15/197B 10/1/1976 9/30/1977 10/1/1976 10/1/1976 9/30/1977 9/30/1977 10/1/1976 9/30/1977 10/1/1976 10/1/1976 10/1/1976 9/30/1977 10/1/1976 10/1/1976 10/1/1976 10/1/1976 9/30/1977 10/1/1976 9/30/1977 9/30/1977 10/1/1976 10/1/1976 10/1/1976 10/1/1976 10/1/1976 9/30/1977 9/30/1977 10/1/1976 9/30/1977 9/30/1977 9/30/1977 9/30/1977 10/1/1976 10/1/1976 10/1/1976 10/1/1976 9/30/1977 10/1/1976 10/1/1976 9/30/1977 10/1/1976 10/1/1976 10/1/1976 10/1/1976 10/1/1976 10/1/1976 10/1/1976 10/1/1976 Assembly ID 2.3BO 2.3BO 2.3BO 2.370 2.370 2.370 2.370 2.370 2.410 2.3BO 2.3BO 2.3BO 2.3BO 2.370 2.370 2.370 2.370 2.3BO 2.3BO 2.3BO 2.3BO 2.370 2.370 2.400 2.370 2.370 2.370 2.410 2.370 2.370 2.370 2.370 2.370 2.370 2.3BO 2.3BO 2.3BO 2.3BO 2.3BO 2.3BO 2.3BO 2.3BO 2.390 2.3BO 2.3BO 2.3BO 2.410 2.410 2.410 2.400 2.390 Assembly ID 11/16/1977 11/16/1977 12/10/197B 12/B/1976 11/16/1977 12/B/1976 12/B/1976 11/16/1977 11/16/1977 12/B/1976 11/16/1977 12/B/1976 12/B/1976 12/B/1976 11/16/1977 12/B/1976 12/B/1976 12/B/1976 12/B/1976 11/16/1977 12/B/1976 11/16/1977 11/16/1977 12/B/1976 12/B/1976 12/B/1976 12/B/1976 12/B/1976 11/16/1977 11/16/1977 12/B/1976 11/16/1977 11/16/1977 11/16/1977 11/16/1977 12/B/1976 12/B/1976 12/B/1976 12/B/1976 11/16/1977 12/B/1976 12/B/1976 11/16/1977 12/B/1976 12/B/1976 12/B/1976 12/B/1976 12/B/1976 12/B/1976 12/B/1976 12/B/1976 FCOB513 Page 172 of 269 FC08514 ext enr.xlsx Assembly ID COOl C002 C003 C004 coos C006 C007 coos C009 COlO C011 C012 C013 C014 C015 C016 ClOl Cl02 Cl03 Cl04 ClOS Cl06 C107 Cl08 Cl09 C110 Cl11 C112 C113 C114 C115 C116 DOOl D002 D003 D004 DOOS D006 D007 DOOS D009 DOlO DOll D012 D013 D014 D015 D016 D017 D018 D019 Initial U-235 Enr 3.180 3.190 3.190 3.190 3.190 3.190 3.190 3.180 3.190 3.180 3.190 3.180 3.190 3.180 3.180 3.190 3.180 3.180 3.190 3.190 3.190 3.190 3.190 3.180 3.180 3.170 3.170 3.160 3.170 3.160 3.160 3.160 2.960 2.960 2.960 2.950 2.950 2.940 2.960 2.960 2.960 2.960 2.960 2.960 2.960 2.960 2.960 2.950 2.960 2.950 2.960 Burn up MWD/T 35491 25442 25442 25442 27396 27396 27396 27642 27642 35491 35491 25442 35491 27642 27642 27396 29445 29445 29445 29445 29445 29445 29445 29245 29245 29245 29245 29445 29245 29245 29245 29245 28394 27388 27618 28394 51504 30218 28394 28394 30218 27608 27388 27388 27388 27608 27608 28394 28394 28394 27388 Assembly ID 13310 13690 13690 13690 13690 13690 13690 13690 13690 13310 13310 13690 13310 13690 13690 13690 13690 13690 13690 13690 13690 13690 13690 13690 13690 13690 13690 13690 13690 13690 13690 13690 13310 13310 13310 13310 11799 12849 13310 13310 12849 13310 13310 13310 13310 13310 13310 13310 13310 13310 13310 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.36999 0.37182 0.37283 0.37192 0.37279 0.37259 0.37338 0.37232 0.37041 0.37002 0.37138 0.37271 0.37046 0.37032 0.37119 0.37265 0.36389 0.36372 0.36322 0.36205 0.36194 0.36227 0.36274 0.36280 0.36253 0.36288 0.36408 0.36355 0.36323 0.36443 0.36479 0.36284 0.37157 0.37227 0.37879 0.37225 0.37199 0.37184 0.37083 0.37171 0.37125 0.37276 0.37310 0.37307 0.37310 0.37184 0.37191 0.37215 0.37159 0.37211 0.37283 Assembly ID 10/15/1978 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 10/15/1978 10/15/1978 9/30/1977 10/15/1978 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 9/30/1977 10/15/1978 10/15/1978 10/15/1978 10/15/1978 12/3/1982 1/18/1980 10/15/1978 10/15/1978 1/18/1980 10/15/1978 10/15/1978 10/15/1978 10/15/1978 10/15/1978 10/15/1978 10/15/1978 10/15/1978 10/15/1978 10/15/1978 Assembly ID 3.180 3.190 3.190 3.190 3.190 3.190 3.190 3.180 3.190 3.180 3.190 3.180 3.190 3.180 3.180 3.190 3.180 3.180 3.190 3.190 3.190 3.190 3.190 3.180 3.180 3.170 3.170 3.160 3.170 3.160 3.160 3.160 2.960 2.960 2.960 2.950 2.950 2.940 2.960 2.960 2.960 2.960 2.960 2.960 2.960 2.960 2.960 2.950 2.960 2.950 2.960 Assembly ID 12/10/1978 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 12/10/1978 12/10/1978 11/16/1977 12/10/1978 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 11/16/1977 12/10/1978 12/10/1978 12/10/1978 12/10/1978 2/18/1983 4/16/1980 12/10/1978 12/10/1978 4/16/1980 12/10/1978 12/10/1978 12/10/1978 12/10/1978 12/10/1978 12/10/1978 12/10/1978 12/10/1978 12/10/1978 12/10/1978 FC08513 Page 173 of 269 FC08514 ext enr.xlsx Assembly 10 0020 0021 0022 0023 0024 0025 0026 0027 0028 0029 0030 0031 0032 0033 0034 0035 0036 0037 0038 0039 0040 0041 0042 0043 0044 E001 E002 E003 E004 E005 E006 E007 E008 E009 E010 E011 E012 E013 E014 E015 E016 E017 E018 E019 E020 E021 E022 E023 E024 E101 E102 Initial U-235 Enr 2.960 2.960 2.960 2.960 2.960 2.960 2.960 2.960 2.970 2.970 2.970 2.960 2.970 2.970 2.970 2.970 2.970 2.970 2.970 2.970 2.970 2.970 2.970 2.990 2.970 3.030 3.030 3.030 3.050 3.030 3.030 3.030 3.030 3.030 3.030 3.030 3.030 3.040 3.040 3.040 3.040 3.040 3.030 3.030 3.030 3.030 3.030 3.040 3.030 2.720 2.720 Burn up MWO/T 27388 27608 27388 27608 27388 27608 27608 33748 33346 33748 33748 33748 33748 33346 33748 33346 33346 33748 33748 33346 33346 30218 33346 30218 33346 32830 29083 36880 30315 32830 32830 32830 32830 30324 32830 32830 26442 39667 36880 36880 26442 36880 39667 30324 26442 39667 32830 26442 29083 27677 27677 Assembly 10 13310 13310 13310 13310 13310 13310 13310 12849 12849 12849 12849 12849 12849 12849 12849 12849 12849 12849 12849 12849 12849 12849 12849 12849 12849 12241 12241 12241 12241 12241 12241 12241 12241 12241 12241 12241 12849 11799 12241 12241 12849 12241 11799 12241 12849 11799 12241 12849 12241 12849 12849 Attachment 0 SpreadSheets U-235 Average Enrichment Assembly 10 0.37303 0.37278 0.37309 0.37276 0.37288 0.37266 0.37265 0.37272 0.37285 0.37181 0.37216 0.37277 0.37205 0.37241 0.37214 0.37225 0.37273 0.37191 0.37177 0.37277 0.37223 0.37238 0.37217 0.37321 0.37251 0.36474 0.36477 0.36434 0.37324 0.36474 0.36524 0.36507 0.36472 0.36464 0.36478 0.36504 0.36395 0.36384 0.36313 0.36342 0.36378 0.36364 0.36448 0.36449 0.36425 0.36433 0.36472 0.36284 0.36467 0.36451 0.36451 Assembly 10 10/15/1978 10/15/1978 10/15/1978 10/15/1978 10/15/1978 10/15/1978 10/15/1978 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 1/18/1980 12/3/1982 9/18/1981 9/18/1981 1/18/1980 9/18/1981 12/3/1982 9/18/1981 1/18/1980 12/3/1982 9/18/1981 1/18/1980 9/18/1981 1/18/1980 1/18/1980 Assembly 10 2.960 2.960 2.960 2.960 2.960 2.960 2.960 2.960 2.970 2.970 2.970 2.960 2.970 2.970 2.970 2.970 2.970 2.970 2.970 2.970 2.970 2.970 2.970 2.990 2.970 3.030 3.030 3.030 3.050 3.030 3.030 3.030 3.030 3.030 3.030 3.030 3.030 3.040 3.040 3.040 3.040 3.040 3.030 3.030 3.030 3.030 3.030 3.040 3.030 2.720 2.720 Assembly 10 12/10/1978 12/10/1978 12/10/1978 12/10/1978 12/10/1978 12/10/1978 12/10/1978 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 4/16/1980 2/18/1983 12/1/1981 12/1/1981 4/16/1980 12/1/1981 2/18/1983 12/1/1981 4/16/1980 2/18/1983 12/1/1981 4/16/1980 12/1/1981 4/16/1980 4/16/1980 FC08513 Page 174 of 269 FC08514 ext enr.xlsx Assembly ID El03 El04 ElOS El06 El07 El08 El09 E110 E111 E112 El13 E114 EllS E116 Ell7 E118 E119 El20 FOOl F002 F003 F004 FOOS F006 F007 F008 F009 FOlO FOll F012 FlOl Fl02 Fl03 Fl04 FlOS Fl06 Fl07 Fl08 Fl09 F110 Fl11 F112 GOOl G002 G003 G004 GOOS G006 G007 G008 G009 Initial U-235 Enr 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 3.020 3.010 3.010 3.010 3.020 3.030 3.030 3.030 3.020 3.020 3.020 3.020 2.740 2.740 2.740 2.740 2.740 2.740 2.740 2.740 2.740 2.740 2.740 2.740 3.030 3.030 3.030 3.020 3.020 3.020 3.020 3.020 3.030 Burn up MWD/T 27677 30618 27932 27932 27932 27932 30618 27932 27677 30618 27932 27677 30618 27677 27677 27677 27932 27932 39667 32398 32398 32398 39667 39667 30324 29083 29083 39667 39667 32398 30618 30618 29637 29637 29637 29637 29637 29637 29637 29637 30618 30618 38488 38294 38294 33253 31973 42169 39051 38577 38577 Assembly ID 12849 12241 12849 12849 12849 12849 12241 12849 12849 12241 12849 12849 12241 12849 12849 12849 12849 12849 11799 12241 12241 12241 11799 11799 12241 12241 12241 11799 11799 12241 12241 12241 12241 12241 12241 12241 12241 12241 12241 12241 12241 12241 10244 11344 11344 11799 11799 11344 11344 11344 11344 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.36498 0.36457 0.36356 0.36334 0.36316 0.36419 0.36431 0.36416 0.36438 0.36393 0.36358 0.36472 0.36417 0.36451 0.36439 0.36478 0.36393 0.36312 0.36450 0.36468 0.36471 0.36427 0.36523 0.36497 0.36577 0.36660 0.36632 0.36437 0.36442 0.36418 0.36441 0.36443 0.36444 0.36506 0.36495 0.36568 0.36522 0.36493 0.36538 0.36531 0.36441 0.36392 0.36349 0.36334 0.36325 0.36317 0.36341 0.36305 0.36392 0.36429 0.36332 Assembly ID 1/18/1980 9/18/1981 1/18/1980 1/18/1980 1/18/1980 1/18/1980 9/18/1981 1/18/1980 1/18/1980 9/18/1981 1/18/1980 1/18/1980 9/18/1981 1/18/1980 1/18/1980 1/18/1980 1/18/1980 1/18/1980 12/3/1982 9/18/1981 9/18/1981 9/18/1981 12/3/1982 12/3/1982 9/18/1981 9/18/1981 9/18/1981 12/3/1982 12/3/1982 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 9/18/1981 3/7/1987 3/2/1984 3/2/1984 12/3/1982 12/3/1982 3/2/1984 3/2/1984 3/2/1984 3/2/1984 Assembly ID 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 2.720 3.020 3.010 3.010 3.010 3.020 3.030 3.030 3.030 3.020 3.020 3.020 3.020 2.740 2.740 2.740 2.740 2.740 2.740 2.740 2.740 2.740 2.740 2.740 2.740 3.030 3.030 3.030 3.020 3.020 3.020 3.020 3.020 3.030 Assembly ID 4/16/1980 12/1/1981 4/16/1980 4/16/1980 4/16/1980 4/16/1980 12/1/1981 4/16/1980 4/16/1980 12/1/1981 4/16/1980 4/16/1980 12/1/1981 4/16/1980 4/16/1980 4/16/1980 4/16/1980 4/16/1980 2/18/1983 12/1/1981 12/1/1981 12/1/1981 2/18/1983 2/18/1983 12/1/1981 12/1/1981 12/1/1981 2/18/1983 2/18/1983 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 12/1/1981 3/24/1987 6/18/1984 6/18/1984 2/18/1983 2/18/1983 6/18/1984 6/18/1984 6/18/1984 6/18/1984 FC08513 Page 175 of 269 FC08514 ext enr.xlsx Assembly ID GOlD GOll G012 G013 G014 G015 G016 G017 G018 G019 G020 G021 G022 G023 G024 G025 G026 G027 G028 G029 G030 G031 G032 G033 G034 G035 G036 G037 G038 G039 G040 G041 G042 G043 G044 HAOl HA02 HA03 HA04 HA05 HA06 HA07 HA08 HA09 HAlO HAll HA12 HA13 HA14 HA15 HA16 Initial U-235 Enr 3.020 3.030 3.020 3.030 3.030 3.020 3.020 3.020 3.020 3.020 3.020 3.020 3.030 3.020 3.020 3.020 3.020 3.020 3.020 3.020 3.020 3.010 3.010 3.010 3.010 3.020 3.020 3.010 3.010 3.030 3.010 3.010 3.030 3.030 3.030 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 Burn up MWD/T 42169 39985 33253 41606 39051 31973 31071 42169 42169 31973 41606 33253 38294 41606 31973 39051 31071 33253 31071 39051 41606 38488 38294 38488 38577 38577 36183 39815 39343 36183 39343 39343 36183 36183 38488 23264 23264 23264 23264 33270 33270 33270 39859 39859 36953 39859 33270 36266 25270 39859 37338 Assembly ID 11344 11344 11799 10244 11344 11799 11799 11344 11344 11799 10244 11799 11344 10244 11799 11344 11799 11799 11799 11344 10244 10244 11344 10244 11344 11344 10244 10769 10769 10244 10769 10769 10244 10244 10244 11344 11344 11344 11344 11344 11344 11344 10769 10769 9674 10769 11344 10244 11344 10769 10769 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.36285 0.36277 0.36285 0.36282 0.36293 0.36267 0.36346 0.36283 0.36286 0.36202 0.36346 0.36311 0.36352 0.36335 0.36319 0.36337 0.36397 0.36347 0.36382 0.36371 0.36351 0.36540 0.36494 0.36507 0.36488 0.36468 0.36470 0.36444 0.36479 0.36416 0.36510 0.36414 0.36435 0.36446 0.36403 0.35633 0.35704 0.35743 0.35675 0.35726 0.35674 0.35742 0.35810 0.35749 0.35751 0.35679 0.35679 0.35836 0.35769 0.35757 0.35759 Assembly ID 3/2/1984 3/2/1984 12/3/1982 3/7/1987 3/2/1984 12/3/1982 12/3/1982 3/2/1984 3/2/1984 12/3/1982 3/7/1987 12/3/1982 3/2/1984 3/7/1987 12/3/1982 3/2/1984 12/3/1982 12/3/1982 12/3/1982 3/2/1984 3/7/1987 3/7/1987 3/2/1984 3/7/1987 3/2/1984 3/2/1984 3/7/1987 9/28/1985 9/28/1985 3/7/1987 9/28/1985 9/28/1985 3/7/1987 3/7/1987 3/7/1987 3/2/1984 3/2/1984 3/2/1984 3/2/1984 3/2/1984 3/2/1984 3/2/1984 9/28/1985 9/28/1985 9/27/1988 9/28/1985 3/2/1984 3/7/1987 3/2/1984 9/28/1985 9/28/1985 Assembly ID 3.020 3.030 3.020 3.030 3.030 3.020 3.020 3.020 3.020 3.020 3.020 3.020 3.030 3.020 3.020 3.020 3.020 3.020 3.020 3.020 3.020 3.010 3.010 3.010 3.010 3.020 3.020 3.010 3.010 3.030 3.010 3.010 3.030 3.030 3.030 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 Assembly ID 6/18/1984 6/18/1984 2/18/1983 3/24/1987 6/18/1984 2/18/1983 2/18/1983 6/18/1984 6/18/1984 2/18/1983 3/24/1987 2/18/1983 6/18/1984 3/24/1987 2/18/1983 6/18/1984 2/18/1983 2/18/1983 2/18/1983 6/18/1984 3/24/1987 3/24/1987 6/18/1984 3/24/1987 6/18/1984 6/18/1984 3/24/1987 12/8/1985 12/8/1985 3/24/1987 12/8/1985 12/8/1985 3/24/1987 3/24/1987 3/24/1987 6/18/1984 6/18/1984 6/18/1984 6/18/1984 6/18/1984 6/18/1984 6/18/1984 12/8/1985 12/8/1985 10/30/1988 12/8/1985 6/18/1984 3/24/1987 6/18/1984 12/8/1985 12/8/1985 FC08513 Page 176 of 269 FC08514 ext enr.xlsx Assembly ID HA17 HA18 HA19 HA20 HA21 HA22 HA23 HA24 HA25 HA26 HA27 HA28 HA29 HA30 HA31 HA32 HA33 HA34 HA35 HA36 HA37 HA38 HA39 HA40 IA01 IA02 IA03 IA04 lAOS IA06 IA07 IA08 IA09 IA10 I All IA12 IA13 IA14 IA15 IA16 IA17 IA18 IA19 IA20 IA21 IA22 IA23 IA24 IA25 IA26 IA27 Initial U-235 Enr 3.480 3.480 3.430 3.480 3.480 3.490 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.470 3.510 3.510 3.510 3.470 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 Burn up MWD/f 37219 37219 37229 39859 33061 39487 39859 39859 37219 37219 37219 33061 33061 37219 33061 33387 33387 33387 38933 38933 37219 38933 33387 38933 27167 27167 38557 27167 27167 27167 27167 27167 34980 34917 34980 34980 34980 34897 34980 34980 32189 32189 32189 32189 23835 23835 23835 23835 36088 36088 36088 Assembly ID 10769 10769 10769 10769 10244 10769 10769 10769 10769 10769 10769 10244 10244 10769 10244 10244 10244 10244 10769 10769 10769 10769 10244 10769 10769 10769 9674 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 Attachment D SpreadSheets U-235 Average Enrichment Assembly 10 0.35684 0.35700 0.36486 0.35635 0.35758 0.36640 0.35767 0.35702 0.35724 0.35759 0.35792 0.35848 0.35816 0.35657 0.35823 0.35830 0.35798 0.35797 0.35755 0.35755 0.35698 0.35824 0.35729 0.35728 0.35686 0.35731 0.35735 0.35747 0.35769 0.35735 0.35776 0.35767 0.35758 0.36555 0.35782 0.35689 0.35754 0.36565 0.35736 0.35695 0.35748 0.35721 0.35745 0.35698 0.35773 0.35730 0.35730 0.35676 0.35702 0.35725 0.35695 Assembly ID 9/28/1985 9/28/1985 9/28/1985 9/28/1985 3/7/1987 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 3/7/1987 3/7/1987 9/28/1985 3/7/1987 3/7/1987 3/7/1987 3/7/1987 9/28/1985 9/28/1985 9/28/1985 9/28/1985 3/7/1987 9/28/1985 9/28/1985 9/28/1985 9/27/1988 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 Assembly ID 3.480 3.480 3.430 3.480 3.480 3.490 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.480 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.470 3.510 3.510 3.510 3.470 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 Assembly ID 12/8/1985 12/8/1985 12/8/1985 12/8/1985 3/24/1987 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 3/24/1987 3/24/1987 12/8/1985 3/24/1987 3/24/1987 3/24/1987 3/24/1987 12/8/1985 12/8/1985 12/8/1985 12/8/1985 3/24/1987 12/8/1985 12/8/1985 12/8/1985 10/30/1988 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 FC08513 Page 177 of 269 FC08514 ext enr.xlsx Assembly 10 IA28 IA29 IA30 IA31 IA32 IA33 IA34 IA35 IA36 IA37 IA38 IA39 IA40 JA01 JA02 JA03 JA04 JA05 JA06 JA07 JA08 JA09 JA10 JAll JA12 JA13 JA14 JA15 JA16 JA17 JA18 JA19 JA20 JA21 JA22 JA23 JA24 JA25 JA26 JA27 JA28 JA29 JA30 JA31 JA32 JA33 JA34 JA35 JA36 KA01 KA02 Initial U-235 Enr 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.520 3.490 Burn up MWD/T 36088 36124 36124 36124 36124 38091 38091 38091 38091 38091 38091 38091 38091 34439 34439 39620 39620 39570 39570 39570 39570 38792 38792 38972 38792 39669 39669 39669 39669 38879 38879 38879 38879 35805 35805 35805 35805 36464 36464 36464 36464 39834 38738 38792 39278 39106 39611 38788 38659 43910 42296 Assembly ID 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10769 10244 10244 9674 9674 9674 9674 9674 9674 10244 10244 9166 10244 10244 10244 10244 10244 10244 10244 10244 10244 10244 10244 10244 10244 10244 10244 10244 10244 9166 9166 10244 9166 9166 9166 9166 9166 9166 9674 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.35736 0.35722 0.35786 0.35738 0.35710 0.35711 0.35786 0.35695 0.35721 0.35740 0.35677 0.35764 0.35777 0.35533 0.35477 0.35616 0.35581 0.35538 0.35492 0.35524 0.35492 0.35513 0.35596 0.35592 0.35559 0.35565 0.35565 0.35570 0.35535 0.35586 0.35596 0.35599 0.35548 0.35564 0.35464 0.35514 0.35585 0.35575 0.35544 0.35540 0.35555 0.35582 0.35515 0.35592 0.35566 0.35513 0.35597 0.35509 0.35537 0.35459 0.35594 Assembly ID 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 9/28/1985 3/7/1987 3/7/1987 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 3/7/1987 3/7/1987 2/17/1990 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 3/7/1987 2/17/1990 2/17/1990 3/7/1987 2/17/1990 2/17/1990 2/17/1990 2/17/1990 2/17/1990 2/17/1990 9/27/1988 Assembly ID 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.510 3.520 3.490 Assembly ID 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 12/8/1985 3/24/1987 3/24/1987 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 3/24/1987 3/24/1987 3/16/1990 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/24/1987 3/12/1990 3/12/1990 3/24/1987 3/16/1990 3/16/1990 3/12/1990 3/12/1990 3/16/1990 3/14/1990 10/30/1988 FC08513 Page 178 of 269 FC08514 ext enr.xlsx Assembly ID KA03 KA04 KAOS KA06 KA07 KA08 KA09 KA10 KAll KA12 KA13 KA14 KA15 KA16 KA17 KA18 KA19 KA20 KA21 KA22 KA23 KA24 KA25 KA26 KA27 KA28 KA29 KA30 KA31 KA32 LA01 LA02 LA03 LA04 LAOS LA06 LA07 LAOS LA09 LA10 LAll LA12 LA13 LA14 LA15 LA16 LA17 LA18 LA19 LA20 LA21 Initial U-235 Enr 3.490 3.510 3.520 3.500 3.440 3.490 3.490 3.490 3.480 3.500 3.300 3.500 3.500 3.500 3.500 3.490 3.450 3.490 3.510 3.490 3.500 3.490 3.510 3.420 3.500 3.520 3.510 3.500 3.500 3.510 3.800 3.810 3.800 3.810 3.800 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.800 3.810 3.800 Burn up MWD/T 42296 42747 44202 44304 43206 42296 42296 41976 42174 42892 39620 45692 44281 45048 38557 25716 44254 33684 45048 39620 44281 33684 45564 45144 45692 44281 38557 44281 38557 45048 38724 38541 38555 38832 37296 37868 37075 37411 38089 37773 38033 38089 37793 37633 38000 38255 29735 29735 29735 38770 39159 Assembly ID 9674 9166 9166 9166 9166 9674 9674 9166 9166 9166 9674 9674 9674 9674 9674 10244 9674 10244 9674 9674 9674 10244 9674 9674 9674 9674 9674 9674 9674 9674 9166 9166 9166 9166 8452 8452 8452 8452 8452 8452 8452 8452 9166 9166 9166 9166 9674 9674 9674 9166 9166 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.35578 0.35450 0.35466 0.35644 0.36481 0.35558 0.35541 0.35615 0.36314 0.35616 0.35964 0.33982 0.33974 0.33999 0.33928 0.33977 0.34879 0.33940 0.33992 0.33939 0.33988 0.33970 0.34169 0.35795 0.34052 0.33741 0.33856 0.33972 0.33915 0.33988 0.35881 0.35804 0.35865 0.35817 0.35859 0.35758 0.35814 0.35776 0.35793 0.35758 0.35763 0.35745 0.35736 0.35758 0.35686 0.35753 0.35779 0.35775 0.35696 0.35652 0.35614 Assembly ID 9/27/1988 2/17/1990 2/17/1990 2/17/1990 2/17/1990 9/27/1988 9/27/1988 2/17/1990 2/17/1990 2/17/1990 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 3/7/1987 9/27/1988 3/7/1987 9/27/1988 9/27/1988 9/27/1988 3/7/1987 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 2/17/1990 2/17/1990 2/17/1990 2/17/1990 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/17/1990 2/17/1990 2/17/1990 2/17/1990 9/27/1988 9/27/1988 9/27/1988 2/17/1990 2/17/1990 Assembly ID 3.490 3.510 3.520 3.500 3.440 3.490 3.490 3.490 3.480 3.500 3.300 3.500 3.500 3.500 3.500 3.490 3.450 3.490 3.510 3.490 3.500 3.490 3.510 3.420 3.500 3.520 3.510 3.500 3.500 3.510 3.800 3.810 3.800 3.810 3.800 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.800 3.810 3.800 Assembly ID 10/30/1988 3/14/1990 3/16/1990 3/14/1990 3/16/1990 10/30/1988 10/30/1988 3/16/1990 3/16/1990 3/14/1990 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 3/24/1987 10/30/1988 3/24/1987 10/30/1988 10/30/1988 10/30/1988 3/24/1987 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 3/16/1990 3/12/1990 3/14/1990 3/16/1990 2/20/1992 2/22/1992 2/22/1992 2/20/1992 2/22/1992 2/20/1992 2/20/1992 2/22/1992 3/12/1990 3/12/1990 3/12/1990 3/12/1990 10/30/1988 10/30/1988 10/30/1988 3/12/1990 3/12/1990 FC08513 Page 179 of 269 FC08514 ext enr.xlsx Assembly ID LA22 LA23 LA24 LA25 LA26 LA27 LA28 LA29 LA30 LA31 LA32 LA33 LA34 LA35 LA36 LA37 LA38 LA39 LA40 LA41 LA42 LA43 LA44 MOOl M002 M003 M004 MOOS M006 M007 MOOS M009 MOlO MOll M012 M013 M014 MOlS M016 M017 M018 M019 M020 MlOl Ml02 Ml03 Ml04 MlOS Ml06 Ml07 Ml08 Initial U-235 Enr 3.810 3.810 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.800 3.780 3.770 3.800 3.770 3.780 3.780 3.770 3.770 3.790 3.790 3.800 3.800 3.810 3.800 3.810 3.800 3.800 3.800 Burn up MWD/T 39441 39504 39199 43394 43235 33078 33078 43276 43016 43655 43739 44194 43914 33078 33078 32681 32681 32681 32681 33512 33512 33512 33512 38167 44172 44041 44063 37978 37860 37671 43960 37397 38176 42819 42993 38264 42710 42816 42851 42671 42737 38530 42892 40557 40577 39214 39249 40593 39140 39613 45005 Assembly ID 9166 9166 9166 9166 9166 9674 9674 9166 9166 9166 9166 9166 9166 9674 9674 9674 9674 9674 9674 9674 9674 9674 9674 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 8452 5654 5654 5654 5654 5654 5654 5122 8452 Attachment D SpreadSheets U-235 Average Enrichment Assembly 10 0.35662 0.35664 0.35629 0.34066 0.33994 0.33987 0.34053 0.34115 0.33987 0.34166 0.34081 0.34070 0.34174 0.34118 0.34125 0.34055 0.34100 0.34051 0.34080 0.34045 0.34070 0.34073 0.34026 0.36404 0.36472 0.36453 0.36448 0.36411 0.36443 0.36426 0.36467 0.36413 0.36413 0.36520 0.36520 0.36449 0.36461 0.36418 0.36407 0.36413 0.36523 0.36512 0.36458 0.34872 0.34905 0.34823 0.34857 0.34831 0.34889 0.34825 0.34749 Assembly ID 2/17/1990 2/17/1990 2/17/1990 2/17/1990 2/17/1990 9/27/1988 9/27/1988 2/17/1990 2/17/1990 2/17/1990 2/17/1990 2/17/1990 2/17/1990 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 9/27/1988 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 2/1/1992 10/1/1999 10/1/1999 10/1/1999 10/1/1999 10/1/1999 10/1/1999 3/15/2001 2/1/1992 Assembly ID 3.810 3.810 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.800 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.810 3.800 3.780 3.770 3.800 3.770 3.780 3.780 3.770 3.770 3.790 3.790 3.800 3.800 3.810 3.800 3.810 3.800 3.800 3.800 Assembly ID 3/16/1990 3/14/1990 3/16/1990 3/16/1990 3/12/1990 10/30/1988 10/30/1988 3/12/1990 3/12/1990 3/16/1990 3/14/1990 3/16/1990 3/12/1990 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 10/30/1988 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/20/1992 2/22/1992 2/20/1992 2/20/1992 2/22/1992 10/12/1999 10/12/1999 10/13/1999 10/12/1999 10/13/1999 10/12/1999 3/31/2001 2/20/1992 FC08513 Page 180 of 269 FC08514 ext enr.xlsx Assembly ID M109 MllO Mlll M112 Mll3 M114 M115 M116 Mll7 M118 M119 M120 M121 M122 M123 M124 NOOl N002 N003 N004 NODS N006 N007 N008 N009 NOlO NOll N012 NOB N014 N015 N016 N017 N018 N019 N020 NlOl N102 N103 N104 NlOS N106 N107 N108 N109 N110 N111 N112 N113 N114 N115 Initial U-235 Enr 3.800 3.800 3.790 3.790 3.810 3.790 3.790 3.800 3.790 3.800 3.810 3.800 3.800 3.800 3.800 3.800 3.700 3.690 3.690 3.700 3.690 3.690 3.700 3.700 3.690 3.690 3.690 3.700 3.690 3.690 3.690 3.700 3.700 3.700 3.700 3.700 3.710 3.700 3.710 3.710 3.710 3.690 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 Burn up MWD/T 39535 38905 45650 48373 45904 39155 39628 38925 38914 39541 45901 40552 45926 46097 38919 46307 41449 44707 45098 45302 41417 45116 41469 45535 44813 45018 45105 45420 44797 45167 41404 45459 46040 46097 45864 45898 45737 41158 45727 45728 45750 38438 41292 41302 41334 41161 48376 48425 48430 48402 41341 Assembly ID 5122 5654 7851 5654 8452 5654 5122 5654 5654 5122 8452 5654 8452 8452 5654 5122 5654 7851 7851 5654 5654 7851 5654 5654 7851 7851 7851 5654 7851 7851 5654 5654 7851 7851 7851 7851 7851 7337 7851 7851 7851 5654 7337 5654 5654 7337 5654 5654 5654 5654 5654 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.34842 0.34847 0.34891 0.34880 0.34907 0.34937 0.34906 0.34815 0.34926 0.34883 0.34879 0.34895 0.34885 0.34913 0.34832 0.34664 0.36399 0.36373 0.36431 0.36472 0.36463 0.36449 0.36456 0.36506 0.36419 0.36415 0.36456 0.36494 0.36397 0.36405 0.36473 0.36518 0.36557 0.36522 0.36561 0.36577 0.34837 0.34891 0.34858 0.34826 0.34833 0.34778 0.34874 0.34820 0.34835 0.34887 0.34978 0.34966 0.34953 0.34937 0.34837 Assembly ID 3/15/2001 10/1/1999 9/25/1993 10/1/1999 2/1/1992 10/1/1999 3/15/2001 10/1/1999 10/1/1999 3/15/2001 2/1/1992 10/1/1999 2/1/1992 2/1/1992 10/1/1999 3/15/2001 10/1/1999 9/25/1993 9/25/1993 10/1/1999 10/1/1999 9/25/1993 10/1/1999 10/1/1999 9/25/1993 9/25/1993 9/25/1993 10/1/1999 9/25/1993 9/25/1993 10/1/1999 10/1/1999 9/25/1993 9/25/1993 9/25/1993 9/25/1993 9/25/1993 2/20/1995 9/25/1993 9/25/1993 9/25/1993 10/1/1999 2/20/1995 10/1/1999 10/1/1999 2/20/1995 10/1/1999 10/1/1999 10/1/1999 10/1/1999 10/1/1999 Assembly ID 3.800 3.800 3.790 3.790 3.810 3.790 3.790 3.800 3.790 3.800 3.810 3.800 3.800 3.800 3.800 3.800 3.700 3.690 3.690 3.700 3.690 3.690 3.700 3.700 3.690 3.690 3.690 3.700 3.690 3.690 3.690 3.700 3.700 3.700 3.700 3.700 3.710 3.700 3.710 3.710 3.710 3.690 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 Assembly ID 3/28/2001 10/12/1999 10/8/1993 10/11/1999 2/20/1992 10/11/1999 3/29/2001 10/13/1999 10/12/1999 3/28/2001 2/20/1992 10/13/1999 2/20/1992 2/20/1992 10/13/1999 3/28/2001 10/13/1999 10/8/1993 10/8/1993 10/11/1999 10/11/1999 10/8/1993 10/12/1999 10/11/1999 10/8/1993 10/8/1993 10/8/1993 10/13/1999 10/8/1993 10/8/1993 10/11/1999 10/12/1999 10/6/1993 10/8/1993 10/8/1993 10/8/1993 10/6/1993 3/4/1995 10/8/1993 10/8/1993 10/8/1993 10/12/1999 3/4/1995 10/12/1999 10/13/1999 3/4/1995 10/12/1999 10/12/1999 10/11/1999 10/11/1999 10/13/1999 FC08513 Page 181 of 269 FC08514 ext enr.xlsx Assembly ID N116 N117 N118 N119 N120 N121 N122 N123 N124 POOl P002 P003 P004 PODS P006 P007 P008 PlOl P102 P103 P104 PlOS P106 P107 P108 P109 P110 P111 P112 P113 P114 P115 P116 P117 P118 P119 P120 P121 P122 P123 P124 P125 P126 P127 P128 P129 P130 P131 P132 ROOl R002 Initial U-235 Enr 3.700 3.700 3.710 3.710 3.700 3.690 3.690 3.690 3.690 3.940 3.940 3.940 3.930 3.930 3.940 3.940 3.940 3.600 3.600 3.590 3.580 3.600 3.600 3.600 3.600 3.580 3.580 3.590 3.580 3.590 3.600 3.600 3.580 3.580 3.580 3.590 3.580 3.580 3.590 3.590 3.580 3.590 3.590 3.590 3.590 3.580 3.580 3.580 3.580 0.740 0.750 Burn up MWD/T 38424 38446 41170 41318 38435 40267 40240 40306 40231 45060 44521 44413 44350 44448 45229 45181 45145 37400 37744 41842 41566 37316 37809 37854 37750 41477 42487 42524 43425 41729 37346 37229 47520 47445 41552 41795 44724 41572 41923 41502 47536 41475 41587 44837 41979 43410 43466 43229 47473 13047 12952 Assembly ID 5654 5654 5654 7337 5654 5654 5654 5654 5654 7337 7337 7337 7337 7337 7337 7337 7337 7851 7851 7337 7337 7851 7851 7851 7851 7337 5654 5654 7337 7337 7851 7851 5654 5654 7337 7337 5654 7337 7337 7337 5654 7337 7337 5654 7337 7337 7337 7337 5654 6745 6745 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.34743 0.34759 0.34782 0.34786 0.34763 0.34748 0.34743 0.34721 0.34709 0.36528 0.36548 0.36551 0.36585 0.36596 0.36619 0.36596 0.36640 0.34975 0.34981 0.34980 0.34997 0.34972 0.34981 0.34999 0.34996 0.35030 0.34989 0.34975 0.34962 0.35006 0.34993 0.34975 0.34977 0.35002 0.34984 0.35009 0.35020 0.34990 0.34975 0.34975 0.34973 0.34962 0.34944 0.34962 0.34986 0.34996 0.34985 0.34964 0.34974 0.37585 0.37604 Assembly ID 10/1/1999 10/1/1999 10/1/1999 2/20/1995 10/1/1999 10/1/1999 10/1/1999 10/1/1999 10/1/1999 2/20/1995 2/20/1995 2/20/1995 2/20/1995 2/20/1995 2/20/1995 2/20/1995 2/20/1995 9/25/1993 9/25/1993 2/20/1995 2/20/1995 9/25/1993 9/25/1993 9/25/1993 9/25/1993 2/20/1995 10/1/1999 10/1/1999 2/20/1995 2/20/1995 9/25/1993 9/25/1993 10/1/1999 10/1/1999 2/20/1995 2/20/1995 10/1/1999 2/20/1995 2/20/1995 2/20/1995 10/1/1999 2/20/1995 2/20/1995 10/1/1999 2/20/1995 2/20/1995 2/20/1995 2/20/1995 10/1/1999 10/5/1996 10/5/1996 Assembly ID 3.700 3.700 3.710 3.710 3.700 3.690 3.690 3.690 3.690 3.940 3.940 3.940 3.930 3.930 3.940 3.940 3.940 3.600 3.600 3.590 3.580 3.600 3.600 3.600 3.600 3.580 3.580 3.590 3.580 3.590 3.600 3.600 3.580 3.580 3.580 3.590 3.580 3.580 3.590 3.590 3.580 3.590 3.590 3.590 3.590 3.580 3.580 3.580 3.580 0.740 0.750 Assembly ID 10/12/1999 10/13/1999 10/12/1999 3/4/1995 10/13/1999 10/13/1999 10/11/1999 10/12/1999 10/11/1999 3/4/1995 3/4/1995 3/4/1995 3/4/1995 3/4/1995 3/4/1995 3/4/1995 3/4/1995 10/8/1993 10/8/1993 3/4/1995 3/4/1995 10/8/1993 10/6/1993 10/8/1993 10/8/1993 3/4/1995 10/12/1999 10/13/1999 3/4/1995 3/4/1995 10/8/1993 10/6/1993 10/12/1999 10/11/1999 3/4/1995 3/4/1995 10/12/1999 3/4/1995 3/4/1995 3/4/1995 10/12/1999 3/4/1995 3/4/1995 10/13/1999 3/4/1995 3/4/1995 3/4/1995 3/4/1995 10/11/1999 10/20/1996 10/18/1996 FC08513 Page 182 of 269 FC08514 ext enr.xlsx Assembly ID R003 R004 RODS R006 R007 R008 R009 ROlO ROll R012 R013 R014 ROlS R016 R017 R018 R019 R020 R021 R022 R023 R024 R025 R026 R027 R028 R029 R030 R031 R032 R033 R034 R035 R036 R037 R038 R039 R040 R041 R042 R043 R044 R045 R046 R047 R048 R049 ROSO ROSl ROS2 SOOl Initial U-235 Enr 0.740 0.750 3.860 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.860 3.850 3.860 3.850 3.860 3.860 3.860 3.860 3.860 3.860 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.860 3.860 3.860 3.850 3.850 3.860 3.850 3.850 3.600 3.600 3.600 3.600 3.600 3.600 3.600 3.600 3.850 Burn up MWD/T 13078 12980 42402 42250 42348 42298 42111 41710 41911 41988 45732 45845 45857 45799 41381 38552 41447 38602 51770 51611 51751 51657 45998 45979 45957 46013 42479 42480 42492 42484 38477 38459 38507 38465 42224 42039 42216 42215 41768 41705 41757 41706 38374 38369 38359 38357 48573 48577 48643 48625 33851 Assembly ID 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6201 6201 6201 6201 6201 6201 6201 6201 6201 6201 6201 6201 6201 6201 6201 6201 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6745 6201 6201 6201 6201 6201 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.37594 0.37581 0.37420 0.37437 0.37442 0.37436 0.37412 0.37337 0.37347 0.37308 0.37421 0.37428 0.37408 0.37399 0.37380 0.37493 0.37449 0.37458 0.37467 0.37391 0.37444 0.37348 0.37476 0.37452 0.37483 0.37470 0.37516 0.37470 0.37453 0.37471 0.37454 0.37423 0.37438 0.37414 0.37352 0.37323 0.37432 0.37385 0.37391 0.37440 0.37449 0.37478 0.37501 0.37490 0.37539 0.37542 0.37509 0.37465 0.37509 0.37495 0.37324 Assembly ID 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 Assembly ID 0.740 0.750 3.860 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.860 3.850 3.860 3.850 3.860 3.860 3.860 3.860 3.860 3.860 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.850 3.860 3.860 3.860 3.850 3.850 3.860 3.850 3.850 3.600 3.600 3.600 3.600 3.600 3.600 3.600 3.600 3.850 Assembly ID 10/19/1996 10/19/1996 10/20/1996 10/19/1996 10/19/1996 10/20/1996 10/19/1996 10/18/1996 10/18/1996 10/18/1996 10/19/1996 10/18/1996 10/20/1996 10/19/1996 10/20/1996 10/19/1996 10/19/1996 10/20/1996 4/19/1998 4/19/1998 4/18/1998 4/18/1998 4/18/1998 4/18/1998 4/19/1998 4/19/1998 4/20/1998 4/20/1998 4/19/1998 4/19/1998 4/19/1998 4/17/1998 4/19/1998 4/20/1998 10/20/1996 10/19/1996 10/19/1996 10/20/1996 10/20/1996 10/17/1996 10/19/1996 10/18/1996 10/20/1996 10/20/1996 10/19/1996 10/19/1996 4/18/1998 4/19/1998 4/19/1998 4/18/1998 4/20/1998 FC08513 Page 183 of 269 FC08514 ext enr.xlsx Assembly ID S002 S003 S004 soos S006 S007 S008 S009 SOlO SOll S012 S013 S014 SOlS S016 S017 S018 S019 S020 S021 S022 S023 S024 S025 S026 S027 S028 S029 S030 S031 S032 S033 S034 S035 S036 S037 S038 S039 S040 S041 TOOl T002 T003 T004 TOOS T006 T007 T008 T009 TOlO TOll Initial U-235 Enr 3.850 3.850 3.850 3.840 3.850 3.850 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.350 3.360 3.360 3.360 3.360 3.360 3.360 3.350 3.360 3.350 3.350 3.350 3.350 3.350 3.350 3.350 3.360 3.350 3.360 3.350 3.350 3.360 3.350 3.350 3.350 3.350 3.350 4.160 4.160 4.170 4.160 4.160 4.170 4.160 4.160 4.160 4.150 4.150 Burn up MWD/T 33872 33847 33780 46261 46217 46244 46195 30264 30141 37446 29991 51971 30136 36858 36702 25321 25264 36777 38841 25244 25263 36056 36733 38839 36194 36245 38922 38852 38842 38847 35305 38828 48846 36839 36653 38889 48575 36580 48653 48849 43418 40362 40558 40400 40575 42678 43447 42817 41055 41057 40971 Assembly ID 6201 6201 6201 6201 6201 6201 6201 6745 6745 6745 6745 6201 6745 6745 6745 6745 6745 6745 6201 6745 6745 6745 6745 6201 6745 6745 6201 6201 6201 6201 6745 6201 6201 6745 6745 6201 6201 6745 6201 6201 4709 4709 4709 4709 4709 4709 4709 4709 4709 4709 4709 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.37275 0.37450 0.37303 0.37381 0.37386 0.37296 0.37351 0.37370 0.37412 0.37409 0.37442 0.37392 0.37349 0.37391 0.37378 0.37422 0.37397 0.37390 0.37430 0.37460 0.37484 0.37438 0.37399 0.37472 0.37447 0.37484 0.37491 0.37435 0.37487 0.37397 0.37441 0.37394 0.37366 0.37382 0.37383 0.37350 0.37371 0.37393 0.37376 0.37361 0.37475 0.37483 0.37461 0.37482 0.37457 0.37481 0.37510 0.37496 0.37436 0.37387 0.37426 Assembly ID 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 4/1/1998 10/5/1996 10/5/1996 10/5/1996 10/5/1996 4/1/1998 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 10/5/1996 4/1/1998 10/5/1996 10/5/1996 10/5/1996 10/5/1996 4/1/1998 10/5/1996 10/5/1996 4/1/1998 4/1/1998 4/1/1998 4/1/1998 10/5/1996 4/1/1998 4/1/1998 10/5/1996 10/5/1996 4/1/1998 4/1/1998 10/5/1996 4/1/1998 4/1/1998 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 Assembly ID 3.850 3.850 3.850 3.840 3.850 3.850 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.350 3.360 3.360 3.360 3.360 3.360 3.360 3.350 3.360 3.350 3.350 3.350 3.350 3.350 3.350 3.350 3.360 3.350 3.360 3.350 3.350 3.360 3.350 3.350 3.350 3.350 3.350 4.160 4.160 4.170 4.160 4.160 4.170 4.160 4.160 4.160 4.150 4.150 Assembly ID 4/18/1998 4/18/1998 4/20/1998 4/18/1998 4/19/1998 4/18/1998 4/19/1998 10/19/1996 10/20/1996 10/18/1996 10/19/1996 4/18/1998 10/20/1996 10/20/1996 10/18/1996 10/20/1996 10/19/1996 10/19/1996 4/20/1998 10/18/1996 10/19/1996 10/19/1996 10/19/1996 4/20/1998 10/18/1996 10/19/1996 4/19/1998 4/19/1998 4/20/1998 4/19/1998 10/20/1996 4/18/1998 4/19/1998 10/20/1996 10/18/1996 4/19/1998 4/19/1998 10/19/1996 4/18/1998 4/18/1998 5/12/2002 5/12/2002 5/12/2002 5/12/2002 5/12/2002 5/12/2002 5/12/2002 5/12/2002 5/12/2002 5/12/2002 5/12/2002 FC08513 Page 184 of 269 FC08514 ext enr.xlsx Assembly ID T012 T013 T014 T015 T016 T017 T018 T019 T020 T021 T022 T023 T024 T025 T026 T027 T028 T029 T030 T031 T032 T033 T034 T035 T036 T037 T038 T039 T040 T041 T042 T043 T044 T045 T046 T047 T048 U001 U002 U003 U004 U005 U006 U007 U008 U009 U010 UOll U012 U013 U014 Initial U-235 Enr 4.160 4.160 4.160 4.150 4.160 4.150 4.150 4.160 4.160 4.160 4.160 4.150 4.160 3.760 3.760 3.760 3.760 3.760 3.760 3.750 3.750 3.750 3.750 3.750 3.760 3.750 3.750 3.760 3.760 3.750 3.750 3.750 3.750 3.750 3.750 3.750 3.750 0.270 0.270 0.270 0.270 4.250 4.240 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 Burn up MWD/T 41048 38480 33582 42777 42780 38499 38433 38489 33594 42706 33617 42735 33611 32898 32859 34882 34802 40615 30273 30286 40662 34088 30276 40626 40681 30252 34080 34107 34086 40251 37417 40304 37448 40317 40267 37411 37551 10537 10650 10492 10630 46960 46926 46963 48955 46971 49176 49116 49026 40246 41735 Assembly ID 4709 4709 4709 4709 4709 4709 4709 4709 4709 4709 4709 4709 4709 6201 6201 6201 6201 4709 4709 4709 4709 4709 4709 4709 4709 4709 4709 4709 4709 4709 6201 4709 6201 4709 4709 6201 6201 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.37449 0.37412 0.37440 0.37436 0.37441 0.37407 0.37402 0.37415 0.37438 0.37430 0.37450 0.37420 0.37486 0.37413 0.37431 0.37436 0.37423 0.37386 0.37405 0.37393 0.37393 0.37468 0.37407 0.37361 0.37368 0.37431 0.37507 0.37490 0.37501 0.37460 0.37434 0.37485 0.37481 0.37454 0.37480 0.37473 0.37442 0.37500 0.37600 0.37500 0.37600 0.37431 0.37436 0.37455 0.37514 0.37414 0.37543 0.37465 0.37560 0.37532 0.37522 Assembly ID 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 4/1/1998 4/1/1998 4/1/1998 4/1/1998 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 5/3/2002 4/1/1998 5/3/2002 4/1/1998 5/3/2002 5/3/2002 4/1/1998 4/1/1998 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 Assembly ID 4.160 4.160 4.160 4.150 4.160 4.150 4.150 4.160 4.160 4.160 4.160 4.150 4.160 3.760 3.760 3.760 3.760 3.760 3.760 3.750 3.750 3.750 3.750 3.750 3.760 3.750 3.750 3.760 3.760 3.750 3.750 3.750 3.750 3.750 3.750 3.750 3.750 0.270 0.270 0.270 0.270 4.250 4.240 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 Assembly ID 5/12/2002 5/14/2002 5/14/2002 5/14/2002 5/14/2002 5/14/2002 5/13/2002 5/13/2002 5/14/2002 5/13/2002 5/14/2002 5/13/2002 5/14/2002 4/18/1998 4/19/1998 4/18/1998 4/19/1998 5/13/2002 5/14/2002 5/14/2002 5/14/2002 5/13/2002 5/14/2002 5/13/2002 5/14/2002 5/14/2002 5/13/2002 5/14/2002 5/14/2002 5/13/2002 4/19/1998 5/14/2002 4/18/1998 5/14/2002 5/13/2002 4/19/1998 4/18/1998 3/28/2001 3/29/2001 3/30/2001 3/30/2001 3/29/2001 3/29/2001 3/28/2001 3/30/2001 3/30/2001 3/29/2001 3/29/2001 3/30/2001 3/30/2001 3/29/2001 FC08513 Page 185 of 269 FC08514 ext enr.xlsx Assembly ID UOlS U016 U017 U018 U019 U020 U021 U022 U023 U024 U025 U026 U027 U028 U029 U030 U031 U032 U033 U034 U035 U036 U037 U038 U039 U040 U041 U042 U043 U044 WOOl W002 W003 W004 woos W006 W007 W008 W009 WOlO W011 W012 W013 W014 WOlS W016 W017 W018 W019 W020 W021 Initial U-235 Enr 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.260 4.250 4.250 4.260 4.260 4.260 4.250 4.260 4.250 4.260 4.260 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.210 4.210 4.210 4.210 4.220 4.210 4.220 4.210 4.220 4.220 4.220 4.220 3.600 3.600 3.600 3.600 3.600 3.600 3.600 3.600 3.600 Burn up MWD{f 40279 41789 41641 41632 41646 41644 43491 43474 43485 43491 44329 44339 44328 47457 47446 47401 44201 47301 41484 41644 41628 41469 44968 44978 45000 44891 42776 42863 42874 42780 35313 35242 35413 35515 25485 33783 25146 33740 33982 25180 25423 34026 36117 29926 36131 29958 36118 30005 35985 30007 37505 Assembly ID 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.37588 0.37550 0.37559 0.37597 0.37556 0.37590 0.37535 0.37515 0.37521 0.37518 0.37502 0.37492 0.37469 0.37498 0.37518 0.37536 0.37526 0.37496 0.37470 0.37487 0.37515 0.37506 0.37449 0.37471 0.37479 0.37469 0.37517 0.37509 0.37495 0.37524 0.37416 0.37364 0.37405 0.37361 0.37451 0.37439 0.37455 0.37440 0.37394 0.37448 0.37470 0.37372 0.37210 0.37063 0.37164 0.37090 0.37182 0.37103 0.37216 0.37126 0.37192 Assembly ID 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 Assembly ID 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.260 4.250 4.250 4.260 4.260 4.260 4.250 4.260 4.250 4.260 4.260 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.210 4.210 4.210 4.210 4.220 4.210 4.220 4.210 4.220 4.220 4.220 4.220 3.600 3.600 3.600 3.600 3.600 3.600 3.600 3.600 3.600 Assembly ID 3/29/2001 3/30/2001 3/30/2001 3/30/2001 3/29/2001 3/29/2001 3/29/2001 3/30/2001 3/29/2001 3/30/2001 3/30/2001 3/30/2001 3/29/2001 3/30/2001 3/29/2001 3/28/2001 3/29/2001 3/29/2001 3/28/2001 3/29/2001 3/31/2001 3/30/2001 3/30/2001 3/29/2001 3/29/2001 3/30/2001 3/28/2001 3/30/2001 3/29/2001 3/28/2001 3/28/2001 3/28/2001 3/28/2001 3/30/2001 3/29/2001 3/28/2001 3/28/2001 3/28/2001 3/30/2001 3/28/2001 3/30/2001 3/29/2001 3/30/2001 3/30/2001 3/28/2001 3/28/2001 3/30/2001 3/29/2001 3/28/2001 3/30/2001 3/29/2001 FC08513 Page 186 of 269 FC08514 ext enr.xlsx Assembly ID W022 W023 W024 W025 W026 W027 W028 W029 W030 W031 W032 W033 W034 W035 W036 W037 W038 W039 W040 W041 W042 W043 W044 X001 X002 X003 X004 X005 X006 X007 X008 X009 X010 X011 X012 X013 X014 X015 X016 X017 X018 X019 X020 X021 X022 X023 X024 X025 X026 X027 X028 Initial U-235 Enr 3.600 3.600 3.600 3.600 3.600 3.600 3.610 3.600 3.600 3.600 3.600 3.610 3.610 3.610 3.610 3.610 3.610 3.610 3.610 4.240 4.230 4.240 4.240 3.510 3.510 3.510 3.510 4.170 4.170 4.170 4.170 4.170 4.170 4.170 4.170 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 Burn up MWD/f 31266 37423 31442 37489 36963 37583 31411 31459 37167 37539 37491 37289 37035 37147 37262 36860 36913 36777 36885 38275 38351 38247 38415 35800 35762 35709 35743 46151 46136 46203 45565 45561 45440 46286 45529 47767 39355 47533 47666 39317 39272 47604 39383 44473 44526 44249 44369 37726 37677 37659 37746 Assembly ID 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 5122 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.37254 0.37155 0.37165 0.37154 0.37168 0.37147 0.37125 0.37154 0.37125 0.37188 0.37127 0.37218 0.37216 0.37167 0.37144 0.37014 0.37026 0.37032 0.37072 0.37076 0.37087 0.37086 0.37108 0.37092 0.37077 0.37089 0.37086 0.37439 0.37561 0.37505 0.37406 0.37406 0.37435 0.37439 0.37391 0.37324 0.37286 0.37325 0.37313 0.37320 0.37311 0.37330 0.37320 0.37249 0.37250 0.37293 0.37273 0.37212 0.37206 0.37167 0.37158 Assembly ID 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 3/15/2001 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 Assembly ID 3.600 3.600 3.600 3.600 3.600 3.600 3.610 3.600 3.600 3.600 3.600 3.610 3.610 3.610 3.610 3.610 3.610 3.610 3.610 4.240 4.230 4.240 4.240 3.510 3.510 3.510 3.510 4.170 4.170 4.170 4.170 4.170 4.170 4.170 4.170 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 4.190 Assembly ID 3/30/2001 3/28/2001 3/29/2001 3/29/2001 3/29/2001 3/30/2001 3/29/2001 3/30/2001 3/30/2001 3/28/2001 3/30/2001 3/29/2001 3/30/2001 3/30/2001 3/29/2001 3/29/2001 3/30/2001 3/30/2001 3/28/2001 3/28/2001 3/28/2001 3/29/2001 3/30/2001 9/23/2003 9/26/2003 9/26/2003 9/22/2003 9/26/2003 9/23/2003 9/22/2003 9/22/2003 9/26/2003 9/26/2003 9/26/2003 9/22/2003 9/23/2003 9/23/2003 9/26/2003 9/22/2003 9/22/2003 9/22/2003 9/26/2003 9/26/2003 9/22/2003 9/23/2003 9/26/2003 9/26/2003 9/26/2003 9/22/2003 9/22/2003 9/23/2003 FC08513 Page 187 of 269 FC08514 ext enr.xlsx Assembly ID X029 X030 X031 X032 X033 X034 X035 X036 X037 X038 X039 X040 YOOl Y002 Y003 Y004 Y005 Y006 Y007 Y008 Y009 YOlO Y011 Y012 Y013 Y014 Y015 Y016 Y017 Y018 Y019 Y020 Y021 Y022 Y023 Y024 Y025 Y026 Y027 Y028 Y029 Y030 Y031 Y032 Y033 Y034 Y035 Y036 Y037 Y038 Y039 Initial U-235 Enr 4.200 4.200 4.200 4.200 4.200 4.200 4.200 4.200 4.200 4.200 4.200 4.200 0.340 0.350 0.340 0.340 3.740 3.740 3.740 3.740 3.740 3.740 3.740 3.740 3.960 3.960 3.960 3.960 3.960 3.960 3.960 3.960 3.960 3.960 3.910 3.910 3.910 3.910 3.920 3.910 3.910 3.910 4.050 4.050 4.060 4.050 4.030 4.030 4.030 4.040 4.020 Burn up MWD/T 39462 44486 40850 40853 44442 40681 40864 44499 39451 39499 39565 44510 11666 11671 11687 11658 39696 39687 49259 33047 39696 49401 39689 47084 44814 44723 44668 44661 41757 44681 43506 41776 45045 43489 43429 42418 42524 41708 42484 42532 43409 41687 55274 55233 55230 55269 44562 44544 44550 44568 44386 Assembly ID 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 4212 2530 2530 2530 2530 1968 1968 2530 4212 1968 1968 1968 3118 3118 3118 3118 3118 1968 3677 1445 1968 3677 1445 1445 3118 3118 1968 3118 3118 1445 1968 1968 1968 1968 1968 1968 1968 1968 1968 3677 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.37163 0.37131 0.37140 0.37135 0.37122 0.37144 0.37125 0.37100 0.37173 0.37161 0.37179 0.37133 0.37509 0.37520 0.37510 0.37515 0.37159 0.37172 0.37163 0.37143 0.37168 0.37159 0.37180 0.37167 0.37273 0.37267 0.37270 0.37260 0.37247 0.37246 0.37240 0.37248 0.37245 0.37263 0.37050 0.37110 0.37118 0.37117 0.37111 0.37125 0.37125 0.37131 0.37431 0.37440 0.37397 0.37408 0.37385 0.37373 0.37375 0.37380 0.37337 Assembly ID 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 9/12/2003 4/19/2008 4/19/2008 4/19/2008 4/19/2008 11/1/2009 11/1/2009 4/19/2008 9/12/2003 11/1/2009 11/1/2009 11/1/2009 9/9/2006 9/9/2006 9/9/2006 9/9/2006 9/9/2006 11/1/2009 2/26/2005 4/9/2011 11/1/2009 2/26/2005 4/9/2011 4/9/2011 9/9/2006 9/9/2006 11/1/2009 9/9/2006 9/9/2006 4/9/2011 11/1/2009 11/1/2009 11/1/2009 11/1/2009 11/1/2009 11/1/2009 11/1/2009 11/1/2009 11/1/2009 2/26/2005 Assembly ID 4.200 4.200 4.200 4.200 4.200 4.200 4.200 4.200 4.200 4.200 4.200 4.200 0.340 0.350 0.340 0.340 3.740 3.740 3.740 3.740 3.740 3.740 3.740 3.740 3.960 3.960 3.960 3.960 3.960 3.960 3.960 3.960 3.960 3.960 3.910 3.910 3.910 3.910 3.920 3.910 3.910 3.910 4.050 4.050 4.060 4.050 4.030 4.030 4.030 4.040 4.020 Assembly ID 9/26/2003 9/23/2003 9/26/2003 9/26/2003 9/22/2003 9/23/2003 9/23/2003 9/26/2003 9/23/2003 9/26/2003 9/23/2003 9/25/2003 5/1/2008 4/30/2008 4/30/2008 5/1/2008 11/11/2009 11/11/2009 4/30/2008 9/23/2003 11/10/2009 11/11/2009 11/11/2009 9/17/2006 9/17/2006 9/17/2006 9/18/2006 9/18/2006 11/11/2009 3/13/2005 4/23/2011 11/11/2009 3/12/2005 4/23/2011 4/22/2011 9/18/2006 9/18/2006 11/12/2009 9/19/2006 9/19/2006 4/24/2011 11/11/2009 11/11/2009 11/11/2009 11/10/2009 11/11/2009 11/10/2009 11/11/2009 11/11/2009 11/11/2009 3/12/2005 FC08513 Page 188 of 269 FC08514 ext enr.xlsx Assembly ID Y040 Y041 Y042 Y043 Y044 Y045 Y046 Y047 Y048 Y049 YOSO Y051 Y0 52 Y0 53 Z001 Z002 Z003 Z004 zoos Z006 Z007 zoos Z009 ZOlO Z011 Z012 Z013 Z014 Z015 Z016 Z017 Z018 Z019 Z020 Z021 Z022 Z023 Z024 Z025 Z026 Z027 Z028 Z029 Z030 Z031 Z032 Z033 Z034 Z035 Z036 Z037 Initial U-235 Enr 4.010 4.030 4.030 4.030 4.020 4.000 4.000 4.000 4.010 4.000 4.000 4.010 4.000 3.430 3.440 3.440 3.440 3.440 3.650 3.650 3.650 3.650 3.420 3.420 3.430 3.430 3.430 3.430 3.420 3.420 3.680 3.680 3.680 3.680 3.690 3.680 3.680 3.680 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 Burn up MWD/T 44410 44669 44687 44706 44687 42242 38532 42173 38671 42141 38740 38643 42367 46667 35649 35732 35893 35840 37298 37405 37387 37238 35805 35872 35908 36151 35918 35957 36019 36077 42323 42239 31456 41856 41851 31448 42219 42298 38797 37838 37923 38791 35729 35600 36268 36037 36163 35888 35591 35509 38817 Assembly ID 3677 1968 1968 1968 1968 3677 3118 3677 3118 3677 3118 3118 3677 3677 3677 3677 3677 3677 3118 3118 3118 3118 3677 3677 3677 3677 3677 3677 3677 3677 3118 3118 3677 0 0 3677 3118 3118 2530 3118 3118 2530 3677 3677 3677 3677 3677 3677 3677 3677 2530 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.37340 0.37365 0.37361 0.37344 0.37305 0.37256 0.37310 0.37241 0.37316 0.37313 0.37328 0.37295 0.37251 0.37309 0.37155 0.37150 0.37150 0.37138 0.37141 0.37141 0.37116 0.37136 0.37090 0.37075 0.37091 0.37092 0.37082 0.37091 0.37074 0.37072 0.37219 0.37229 0.37211 0.37211 0.37205 0.37221 0.37213 0.37229 0.37089 0.37074 0.37086 0.37126 0.37114 0.37103 0.37101 0.37083 0.37105 0.37081 0.37084 0.37080 0.37081 Assembly ID 2/26/2005 11/1/2009 11/1/2009 11/1/2009 11/1/2009 2/26/2005 9/9/2006 2/26/2005 9/9/2006 2/26/2005 9/9/2006 9/9/2006 2/26/2005 2/26/2005 2/26/2005 2/26/2005 2/26/2005 2/26/2005 9/9/2006 9/9/2006 9/9/2006 9/9/2006 2/26/2005 2/26/2005 2/26/2005 2/26/2005 2/26/2005 2/26/2005 2/26/2005 2/26/2005 9/9/2006 9/9/2006 2/26/2005 3/24/2015 3/24/2015 2/26/2005 9/9/2006 9/9/2006 4/19/2008 9/9/2006 9/9/2006 4/19/2008 2/26/2005 2/26/2005 2/26/2005 2/26/2005 2/26/2005 2/26/2005 2/26/2005 2/26/2005 4/19/2008 Assembly ID 4.010 4.030 4.030 4.030 4.020 4.000 4.000 4.000 4.010 4.000 4.000 4.010 4.000 3.430 3.440 3.440 3.440 3.440 3.650 3.650 3.650 3.650 3.420 3.420 3.430 3.430 3.430 3.430 3.420 3.420 3.680 3.680 3.680 3.680 3.690 3.680 3.680 3.680 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 3.630 Assembly ID 3/13/2005 11/12/2009 11/12/2009 11/11/2009 11/11/2009 3/12/2005 9/17/2006 3/13/2005 9/19/2006 3/13/2005 9/18/2006 9/18/2006 3/13/2005 3/13/2005 3/13/2005 3/12/2005 3/13/2005 3/13/2005 9/18/2006 9/18/2006 9/18/2006 9/17/2006 3/13/2005 3/12/2005 3/12/2005 3/12/2005 3/13/2005 3/13/2005 3/13/2005 3/12/2005 9/18/2006 9/18/2006 3/13/2005 3/24/2015 3/24/2015 3/12/2005 9/17/2006 9/18/2006 5/1/2008 9/18/2006 9/18/2006 5/1/2008 3/13/2005 3/13/2005 3/12/2005 3/13/2005 3/12/2005 3/13/2005 3/12/2005 3/13/2005 4/30/2008 FC08513 Page 189 of 269 FC08514 ext enr.xlsx Assembly ID Z038 Z039 Z040 AA01 AA02 AA03 AA04 AA05 AA06 AA07 AA08 AA09 AA10 AA11 AA12 AA13 AA14 AA15 AA16 AA17 AA18 AA19 AA20 AA21 AA22 AA23 AA24 AA25 AA26 AA27 AA28 AA29 AA30 AA31 AA32 AA33 AA34 AA35 AA36 AA37 AA38 AA39 AA40 AA41 AA42 AA43 AA44 BB01 BB02 BB03 BB04 Initial U-235 Enr 3.630 3.630 3.630 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 4.250 4.250 4.250 4.250 4.180 4.180 4.180 4.180 4.180 4.180 4.180 4.190 4.190 4.190 4.190 4.190 4.310 4.310 4.310 4.300 4.280 4.270 4.270 4.280 4.300 4.300 4.300 4.300 4.110 4.110 4.110 4.110 Burn up MWD/T 37915 37852 38829 46757 46798 41031 41035 44560 39436 44435 40317 40319 44490 39439 44494 40983 41047 46840 46860 51519 51197 51224 51519 43668 46703 43731 44256 44215 47706 47712 44216 44256 43754 46701 43693 44072 44094 44093 44074 48361 35667 50914 48362 46172 45624 45623 46172 49362 49356 49334 49378 Assembly ID 3118 3118 2530 2530 2530 2530 2530 2530 1968 2530 1968 1968 2530 1968 2530 2530 2530 2530 2530 2530 2530 2530 2530 2530 1968 2530 1968 1968 1968 1968 1968 1968 2530 1968 2530 1445 1445 1445 1445 0 3118 1445 0 0 0 0 0 1445 1445 1445 1445 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.37107 0.37107 0.37114 0.37135 0.37142 0.37155 0.37133 0.37125 0.37126 0.37127 0.37140 0.37133 0.37122 0.37135 0.37124 0.37109 0.37128 0.37140 0.37123 0.37206 0.37227 0.37232 0.37231 0.37099 0.37080 0.37108 0.37115 0.37084 0.37089 0.37093 0.37100 0.37098 0.37069 0.37062 0.37100 0.37307 0.37330 0.37339 0.37309 0.37240 0.37292 0.37270 0.37244 0.37296 0.37275 0.37280 0.37269 0.37353 0.37343 0.37363 0.37387 Assembly ID 9/9/2006 9/9/2006 4/19/2008 4/19/2008 4/19/2008 4/19/2008 4/19/2008 4/19/2008 11/1/2009 4/19/2008 11/1/2009 11/1/2009 4/19/2008 11/1/2009 4/19/2008 4/19/2008 4/19/2008 4/19/2008 4/19/2008 4/19/2008 4/19/2008 4/19/2008 4/19/2008 4/19/2008 11/1/2009 4/19/2008 11/1/2009 11/1/2009 11/1/2009 11/1/2009 11/1/2009 11/1/2009 4/19/2008 11/1/2009 4/19/2008 4/9/2011 4/9/2011 4/9/2011 4/9/2011 3/24/2015 9/9/2006 4/9/2011 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 4/9/2011 4/9/2011 4/9/2011 4/9/2011 Assembly ID 3.630 3.630 3.630 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 3.950 4.250 4.250 4.250 4.250 4.180 4.180 4.180 4.180 4.180 4.180 4.180 4.190 4.190 4.190 4.190 4.190 4.310 4.310 4.310 4.300 4.280 4.270 4.270 4.280 4.300 4.300 4.300 4.300 4.110 4.110 4.110 4.110 Assembly ID 9/19/2006 9/17/2006 4/30/2008 4/30/2008 4/30/2008 5/1/2008 4/29/2008 4/30/2008 11/11/2009 4/29/2008 11/12/2009 11/11/2009 5/1/2008 11/12/2009 5/1/2008 5/1/2008 4/30/2008 4/30/2008 4/29/2008 4/30/2008 4/30/2008 4/30/2008 4/30/2008 5/1/2008 11/12/2009 5/1/2008 11/12/2009 11/10/2009 11/10/2009 11/12/2009 11/12/2009 11/11/2009 4/30/2008 11/11/2009 4/29/2008 4/23/2011 4/24/2011 4/23/2011 4/24/2011 3/24/2015 9/18/2006 4/23/2011 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 4/23/2011 4/23/2011 4/23/2011 4/23/2011 FC08513 Page 190 of 269 FC08514 ext enr.xlsx Assembly ID BB05 BB06 BB07 BB08 BB09 BB10 BB11 BB12 BB13 BB14 BB15 BB16 BB17 BB18 BB19 BB20 BB21 BB22 BB23 BB24 BB25 BB26 BB27 BB28 BB29 BB30 BB31 BB32 BB33 BB34 BB35 BB36 BB37 BB38 BB39 BB40 BB41 BB42 BB43 BB44 CC01 CC02 CC03 CC04 CC05 CC06 CC07 CC08 CC09 CC10 CC11 Initial U-235 Enr 4.090 4.090 4.090 4.090 4.090 4.090 4.090 4.090 4.090 4.090 4.090 4.090 3.970 3.970 3.970 3.970 3.970 3.970 3.970 3.970 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.830 3.830 3.830 3.840 3.830 3.840 3.840 3.840 3.840 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 Burn up MWD/T 45301 45312 45757 45065 47731 46380 46385 47715 45066 45758 45314 45300 40337 43174 43175 40338 43289 40544 40546 43285 37656 39743 39731 46510 46530 41846 38079 38136 41858 37748 37720 41857 38060 37962 41847 46532 46506 39731 39744 37665 43912 43979 41020 40984 44983 45037 45036 44984 40986 41017 43979 Assembly ID 0 0 1968 1968 1445 1445 1445 1445 1968 1968 0 0 1445 1968 1968 1445 1968 1445 1445 1968 2530 1968 1968 1968 1968 0 2530 2530 0 2530 2530 0 2530 2530 0 1968 1968 1968 1968 2530 1445 1445 0 0 0 0 0
0 0
0 1445 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.37342 0.37340 0.37302 0.37327 0.37324 0.37327 0.37315 0.37326 0.37318 0.37323 0.37338 0.37347 0.37078 0.37074 0.37069 0.37075 0.37070 0.37070 0.37079 0.37076 0.37070 0.37064 0.37057 0.37075 0.37069 0.37059 0.37058 0.37076 0.37055 0.37082 0.37068 0.37103 0.37090 0.37092 0.37073 0.37089 0.37100 0.37092 0.37082 0.37073 0.38889 0.38880 0.38886 0.38919 0.38898 0.38921 0.38866 0.38881 0.38919 0.38933 0.38936 Assembly ID 3/24/2015 3/24/2015 11/1/2009 11/1/2009 4/9/2011 4/9/2011 4/9/2011 4/9/2011 11/1/2009 11/1/2009 3/24/2015 3/24/2015 4/9/2011 11/1/2009 11/1/2009 4/9/2011 11/1/2009 4/9/2011 4/9/2011 11/1/2009 4/19/2008 11/1/2009 11/1/2009 11/1/2009 11/1/2009 3/24/2015 4/19/2008 4/19/2008 3/24/2015 4/19/2008 4/19/2008 3/24/2015 4/19/2008 4/19/2008 3/24/2015 11/1/2009 11/1/2009 11/1/2009 11/1/2009 4/19/2008 4/9/2011 4/9/2011 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 4/9/2011 Assembly ID 4.090 4.090 4.090 4.090 4.090 4.090 4.090 4.090 4.090 4.090 4.090 4.090 3.970 3.970 3.970 3.970 3.970 3.970 3.970 3.970 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.840 3.830 3.830 3.830 3.840 3.830 3.840 3.840 3.840 3.840 3.980 3.990 3.990 3.990 3.990 3.990 3.980 3.990 3.980 3.990 3.990 Assembly ID 3/24/2015 3/24/2015 11/11/2009 11/10/2009 4/23/2011 4/23/2011 4/23/2011 4/23/2011 11/11/2009 11/11/2009 3/24/2015 3/24/2015 4/22/2011 11/11/2009 11/12/2009 4/24/2011 11/12/2009 4/23/2011 4/23/2011 11/10/2009 4/30/2008 11/11/2009 11/10/2009 11/11/2009 11/11/2009 3/24/2015 4/30/2008 4/30/2008 3/24/2015 4/30/2008 4/30/2008 3/24/2015 4/30/2008 4/30/2008 3/24/2015 11/11/2009 11/10/2009 11/11/2009 11/11/2009 4/29/2008 4/24/2011 4/23/2011 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 4/24/2011 FC08513 Page 191 of 269 FC08514 ext enr.xlsx Assembly ID CC12 CC13 CC14 CC15 CC16 CC17 CC18 CC19 CC20 CC21 CC22 CC23 CC24 CC25 CC26 CC27 CC28 CC29 CC30 CC31 CC32 CC33 CC34 CC35 CC36 CC37 CC38 CC39 CC40 CC41 CC42 CC43 CC44 DDOl 0002 0003 0004 0005 0006 0007 0008 0009 DOlO DOll 0012 0013 0014 0015 0016 0017 0018 Initial U-235 Enr 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 Burn up MWD/T 43914 42936 42941 42941 42937 39856 39882 42004 38805 38807 41959 38840 37079 38798 37165 37181 38797 37078 38841 41959 38806 38806 42005 39882 39856 43930 42849 42816 43962 43961 42815 42850 43930 48208 48253 48251 48204 48800 48882 48899 48733 33752 33730 33734 33769 40664 40565 40566 40665 49074 48980 Assembly ID 1445 1445 1445 1445 1445 1445 1445 1445 1968 1968 1445 1968 1445 1968 1445 1445 1968 1445 1968 1445 1968 1968 1445 1445 1445 1445 1445 1445 1445 1445 1445 1445 1445 0 0 0 0 0 0 0 0
0 0 0 0 0 0 0 0 0 0 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.38925 0.38902 0.38903 0.38893 0.38907 0.38649 0.38648 0.38652 0.38658 0.38648 0.38639 0.38651 0.38658 0.38668 0.38667 0.38680 0.38636 0.38628 0.38619 0.38620 0.38656 0.38639 0.38643 0.38627 0.38624 0.38675 0.38673 0.38664 0.38571 0.38644 0.38603 0.38604 0.38598 0.38861 0.38862 0.38847 0.38845 0.38834 0.38844 0.38829 0.38821 0.38837 0.38863 0.38827 0.38797 0.38798 0.38776 0.38856 0.38862 0.38552 0.38504 Assembly ID 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 11/1/2009 11/1/2009 4/9/2011 11/1/2009 4/9/2011 11/1/2009 4/9/2011 4/9/2011 11/1/2009 4/9/2011 11/1/2009 4/9/2011 11/1/2009 11/1/2009 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 Assembly ID 3.990 3.970 3.970 3.970 3.970 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.700 3.880 3.870 3.870 3.870 3.870 3.870 3.870 3.870 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.030 4.030 Assembly ID 4/22/2011 4/24/2011 4/22/2011 4/24/2011 4/23/2011 4/24/2011 4/24/2011 4/23/2011 11/11/2009 11/11/2009 4/24/2011 11/11/2009 4/23/2011 11/11/2009 4/23/2011 4/23/2011 11/10/2009 4/23/2011 11/11/2009 4/22/2011 11/10/2009 11/11/2009 4/23/2011 4/23/2011 4/22/2011 4/23/2011 4/24/2011 4/24/2011 4/24/2011 4/23/2011 4/23/2011 4/23/2011 4/24/2011 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 FC08513 Page 192 of 269 FC08514 ext enr.xlsx Assembly 10 0019 0020 0021 0022 0023 0024 0025 0026 0027 0028 0029 0030 0031 0032 0033 0034 0035 0036 0037 0038 0039 0040 0041 0042 0043 0044 EEOl EE02 EE03 EE04 EEOS EE06 EE07 EE08 EE09 EElO EE11 EE12 EE13 EE14 EElS EE16 EE17 EElS EE19 EE20 EE21 EE22 EE23 EE24 EE25 Initial U-235 Enr 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 Burn up MWO/T 48987 49060 38016 50296 37730 38026 33382 33599 33598 33383 45812 45889 45891 45812 31421 31364 31364 31420 43157 43099 43098 43156 41382 41297 41303 41378 32665 32575 32579 32666 27569 27476 27479 27565 33427 33426 33420 33432 32913 32862 32867 32907 34430 34376 34373 34434 34508 34478 34475 34507 34091 Assembly 10 0 0 1445 0 1445 1445 1445 1445 1445 1445 0 0 0 0 1445 1445 1445 1445 0 0 0 0 0 0 0 0 0
0 0
0 0
0 0
0 0
0 0 0 0 0 0
0 0 0 0 0 0 0 0 0 0 Attachment 0 SpreadSheets U-235 Average Enrichment Assembly 10 0.38508 0.38536 0.38501 0.38549 0.38546 0.38546 0.38540 0.38560 0.38547 0.38545 0.38564 0.38575 0.38563 0.38567 0.38560 0.38594 0.38633 0.38658 0.38673 0.38544 0.38620 0.38520 0.38541 0.38543 0.38538 0.38542 0.38957 0.38943 0.38947 0.38941 0.38929 0.38918 0.38913 0.38931 0.38923 0.38924 0.38927 0.38937 0.38937 0.38939 0.38912 0.38919 0.38630 0.38668 0.38665 0.38660 0.38663 0.38639 0.38616 0.38636 0.38635 Assembly 10 3/24/2015 3/24/2015 4/9/2011 3/24/2015 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 4/9/2011 3/24/2015 3/24/2015 3/24/2015 3/24/2015 4/9/2011 4/9/2011 4/9/2011 4/9/2011 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 Assembly 10 4.030 4.030 4.030 4.030 4.030 4.030 4.030 4.030 4.030 4.030 4.030 4.030 4.030 4.030 3.860 3.860 3.870 3.870 3.870 3.860 3.870 3.850 3.850 3.850 3.850 3.850 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.150 4.030 4.030 4.030 4.030 4.030 4.030 4.030 4.030 4.030 Assembly 10 3/24/2015 3/24/2015 4/23/2011 3/24/2015 4/23/2011 4/23/2011 4/23/2011 4/23/2011 4/24/2011 4/23/2011 3/24/2015 3/24/2015 3/24/2015 3/24/2015 4/23/2011 4/23/2011 4/22/2011 4/23/2011 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 FC08513 Page 193 of 269 FC08514 ext enr.xlsx Assembly ID EE26 EE27 EE28 EE29 EE30 EE31 EE32 EE33 EE34 EE35 EE36 FF01 FF02 FF03 FF04 FF05 FF06 FF07 FF08 FF09 FF10 FF11 FF12 FF13 FF14 FF15 FF16 FF17 FF18 FF19 FF20 FF21 FF22 FF23 FF24 FF25 FF26 FF27 FF28 FF29 FF30 FF31 FF32 FF33 FF34 FF35 FF36 FF37 FF38 FF39 FF40 Initial U-235 Enr 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 4.000 Burn up MWD/f 34048 34100 34076 32461 32033 32039 32462 34858 34827 34832 34854 14936 14877 14879 14936 15143 15154 15157 15140 15880 15879 15883 15878 15802 15817 15821 15799 16525 16490 16494 16524 17667 17769 17764 17670 16464 16488 16493 16459 16803 16908 16924 16792 17208 17214 17220 17210 17188 17196 17203 17178 Assembly ID 0 0
0 0 0 0 0 0 0 0 0
0 0 0 0
0 0 0 0
0 0 0 0 0 0 0
0 0
0 0
0 0 0 0
0 0
0 0 0 0 0 0
0 0 0 0 0 0 0 0 0 Attachment D SpreadSheets U-235 Average Enrichment Assembly ID 0.38640 0.38637 0.38621 0.38538 0.38537 0.38524 0.38500 0.38500 0.38521 0.38533 0.38532 0.38660 0.38675 0.38680 0.38686 0.38704 0.38702 0.38671 0.38672 0.38600 0.38574 0.38560 0.38588 0.38578 0.38584 0.38525 0.38598 0.38433 0.38430 0.38420 0.38444 0.38404 0.38404 0.38450 0.38438 0.38435 0.38432 0.38452 0.38460 0.38390 0.38432 0.38401 0.38359 0.38304 0.38298 0.38281 0.38288 0.38307 0.38311 0.38319 0.38300 Assembly ID 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 Assembly ID 4.030 4.030 4.030 3.860 3.860 3.860 3.860 3.860 3.860 3.860 3.860 4.280 4.280 4.280 4.280 4.280 4.280 4.280 4.280 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.250 4.020 4.020 4.020 4.020 4.020 4.020 4.030 4.030 4.030 4.030 4.030 4.030 4.030 4.020 4.030 4.020 3.860 3.860 3.860 3.860 3.860 3.860 3.860 3.860 Assembly ID 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 3/24/2015 FC08513 Page 194 of 269 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 tallies_gamm20161017.xlsx FCS 14MOg EAB Dose (14mog) nps mRem/hr EAB error 4.0265318E+08 6.591E-04 0.5942 8.0530637E+08 4.872E-04 0.4395 1.2079596E+09 4.224E-04 0.3470 1.6106127E+09 3.964E-04 0.2858 2.0132659E+09 1.473E-03 0.7517 2.4159191E+09 1.299E-03 0.7111 2.8185723E+09 1.175E-03 0.6738 3.2212255E+09 1.154E-03 0.6040 3.6238787E+09 1.090E-03 0.5698 4.0265318E+09 1.204E-03 0.4944 4.4291850E+09 1.109E-03 0.4880 4.8318382E+09 1.065E-03 0.4662 5.0000000E+09 1.034E-03 0.4639 5.6371446E+09 1.164E-03 0.3876 6.4424509E+09 1.371E-03 0.3440 7.2477573E+09 1.260E-03 0.3329 8.0530637E+09 1.161E-03 0.3251 8.8583700E+09 1.158E-03 0.3029 9.6636764E+09 1.090E-03 0.2951 l.OOOOOOOE+10 1.059E-03 0.2935 1.0416E-03 EAB vov 0.9264 0.6738 0.6077 0.5411 0.9861 0.9824 0.9805 0.9587 0.9478 0.7516 0.7516 0.7490 0.7490 0.6028 0.3791 0.3785 0.3782 0.3482 0.3478 0.3478 Attachment D SpreadSheets Gamma Tally Results CR Dose(14mog) EAB slope EAB fom mRem/hr 1.7 9.5E-03 9.5316E-05 1.8 8.6E-03 2.7896E-04 1.7 9.2E-03 1.9971E-04 1.7 l.OE-02 1.8288E-04 1.7 1.2E-03 1.5396E-04 1.8 1.1E-03 1.3851E-04 1.7 l.lE-03 1.2464E-04 1.7 1.2E-03 1.2634E-04 1.8 1.2E-03 1.1738E-04 1.8 1.4E-03 1.3338E-04 1.8 1.4E-03 1.2901E-04 1.8 1.4E-03 1.2521E-04 1.8 1.3E-03 1.2267E-04 1.7 1.7E-03 1.1937E-04 1.7 1.9E-03 1.1075E-04 1.7 1.8E-03 1.1823E-04 1.7 1.7E-03 1.1826E-04 1.7 1.8E-03 1.1429E-04 1.7 1.7E-03 1.0969E-04 1.7 1.7E-03 1.0735E-04 1.3630E-04 FC08513 EAB Dose Error (14mog) Error (14mog) CR error CR vov CR slope CR fom mRem/hr mRem/hr 0.4717 0.3369 1.6 1.5E-02 0.5776 0.5716 1.6 5.0E-03 0.5385 0.5692 1.5 3.8E-03 0.4575 0.4943 1.5 4.0E-03 0.4354 0.4912 1.6 3.6E-03 0.4058 0.4795 1.6 3.5E-03 0.3871 0.4764 1.6 3.3E-03 0.3440 0.4271 1.5 3.7E-03 0.3296 0.4244 1.5 3.6E-03 0.3214 0.3000 1.5 3.4E-03 0.3046 0.2907 1.5 3.5E-03 0.2913 0.2769 1.6 3.5E-03 0.2875 0.2764 1.6 3.5E-03 0.2660 0.2608 1.6 3.6E-03 0.2520 0.2561 1.6 3.5E-03 0.2324 0.1975 1.6 3.7E-03 0.2158 0.1769 1.6 3.9E-03 0.2038 0.1741 1.6 4.0E-03 0.1974 0.1655 1.6 3.9E-03 0.1950 0.1652 1.6 3.8E-03 1.64E-03 1.88E-04 tallies_gamm20161017.xlsx FCS 18MOg EAB Dose {18mog) nps mRem/hr EAB error EAB vov 4.02653E+08 5.4485E-04 0.5941 0.9262 8.05306E+08 4.0254E-04 0.4395 0.6738 1.20796E+09 3.4907E-04 0.3470 0.6077 1.61061E+09 3.2760E-04 0.2857 0.5411 2.01327E+09 1.2176E-03 0.7516 0.9861 2.41592E+09 1.0736E-03 0.7110 0.9824 2.81857E+09 9.7174E-04 0.6737 0.9805 3.22123E+09 9.5482E-04 0.6033 0.9588 3.62388E+09 9.0215E-04 0.5692 0.9478 4.02653E+09 9.9591E-04 0.4940 0.7519 4.42919E+09 9.1718E-04 0.4876 0.7519 4.83184E+09 8.8082E-04 0.4658 0.7493 5.00000E+09 8.5547E-04 0.4635 0.7493 5.63714E+09 9.6275E-04 0.3873 0.6029 6.44245E+09 1.1354E-03 0.3433 0.3792 7.24776E+09 1.0420E-03 0.3326 0.3785 8.05306E+09 9.6048E-04 0.3248 0.3783 8.85837E+09 9.5782E-04 0.3027 0.3484 9.66368E+09 9.0125E-04 0.2949 0.3480 1.00000E+10 8.7565E-04 0.2934 0.3480 8.614E-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 Attachment D SpreadSheets Gamma Tally Results CR Dose(18mog) EAB slope EAB fom mRem/hr 1.7 6.7E-04 7.8906E-05 1.8 6.0E-04 2.3080E-04 1.7 6.4E-04 1.6521E-04 1.7 7.1E-04 1.5128E-04 1.7 8.2E-05 1.2733E-04 1.8 7.7E-05 1.1457E-04 1.7 7.4E-05 1.0312E-04 1.7 8.0E-05 1.0345E-04 1.8 8.0E-05 9.6153E-05 1.8 9.6E-05 1.0938E-04 1.8 9.0E-05 1.0581E-04 1.8 9.0E-05 1.0275E-04 1.8 8.8E-05 1.0068E-04 1.7 l.lE-04 9.8168E-05 1.6 1.2E-04 9.1104E-05 1.7 1.2E-04 9.7243E-05 1.7 l.lE-04 9.7341E-05 1.7 1.2E-04 9.4124E-05 1.7 l.lE-04 9.0353E-05 1.7 l.lE-04 8.8403E-05 1.1231E-04 CR error 0.4716 0.5777 0.5386 0.4576 0.4356 0.4059 0.3872 0.3475 0.3328 0.3237 0.3066 0.2932 0.2893 0.2674 0.2532 0.2335 0.2168 0.2046 0.1981 0.1958 FC08513 EAB Dose e Error {18mog) Error {18mog) CR vov CR slope CR fom mRem/hr mRem/hr 0.3374 1.6 l.lE-03 0.5722 1.6 3.SE-04 0.5698 1.5 2.7E-04 0.4949 1.5 2.8E-04 0.4917 1.5 2.SE-04 0.4800 1.6 2.4E-04 0.4769 1.6 2.2E-04 0.4283 1.5 2.4E-04 0.4256 1.5 2.4E-04 0.3003 1.5 2.2E-04 0.2910 1.5 2.3E-04 0.2771 1.6 2.3E-04 0.2766 1.6 2.3E-04 0.2602 1.6 2.3E-04 0.2555 1.6 2.3E-04 0.1971 1.6 2.4E-04 0.1763 1.6 2.SE-04 0.1736 1.6 2.6E-04 0.1650 1.6 2.SE-04 0.1647 1.6 2.SE-04 1.36E-03 1.55E-04 
.... ..r::. ........ 3.0000E-04 2.SOOOE-04 2.0000E-04 E l.SOOOE-04 Ql a: E l.OOOOE-04 S.OOOOE-05 O.OOOOE+OO O.OOE+OO l.OOE+09 2.00E+09 Dose(14mog) mRem/hr EAB & CR Gamma Dose Rate (mRem/hr) 3.00E+09 4.00E+09 CR Dose(18mog) mRem/hr S.OOE+09 Total Particles 6.00E+09 7.00E+09 Dose (14mog) mRem/hr 8.00E+09 9.00E+09 Page 198 of 269 1.600E-03 1.3SOE-03 l.lOOE-03 8.SOOE-04 6.000E-04 3.SOOE-04 l.OOOE-04 l.OOE+lO EAB Dose (18mog) mRem/hr 0.70 0.60 0.50 I ;;:; 0 6 0.40 c 0 *.;::; u "' .... LL .... 0 .... .... w 0.30 0.20 0.10 0.00 O.OE+OO l.OE+09 2.0E+09 3.0E+09 -+-CRerror Relative Error 4.0E+09 5.0E+09 Total Particles 6.0E+09 CR error --Desired error (0.05) .,._EAB error 7.0E+09 8.0E+09 9.0E+09 EAB error Page 199 of 269 0.90 0.80 0.70 0.60 0.50 0.40 0.30 0.20 0.10 0.00 l.OE+10 1.10 1.00 0.90 0.80 0.70 0.60 ci v > 0 > 0.50 0.40 0.30 0.20 0.10 0.00 O.OOE+OO 1.00E+09 Variance of the Variance 2.00E+09 3.00E+09 4.00E+09 5.00E+09 Total Particles 6.00E+09 7.00E+09 (R VOV -Desired VOV (0.10) Page 200 of 269 8.00E+09 9.00E+09 l.OOE+10 0 q m v QJ a. 0 Vi u.. 0 0.. 3.20 3.00 2.80 2.60 2.40 2.20 2.00 1.80 1.60 1.40 O.OOE+OO 1.00E+09 2.00E+09 3.00E+09 4.00E+09 PDF Slope S.OOE+09 Total Particles .,._EAB slope .,._CR slope _._EAB slope Page 201 of 269 6.00E+09 7.00E+09 8.00E+09 9.00E+09 1.00E+10 CR slope --Desired slope (3.00) 1.6000E-02 1.4000E-02 1.2000E-02 l.OOOOE-02 8.0000E-03 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 Figure of Merit 4.00E+09 S.OOE+09 Total Particles _._EAB fom _._CR fom _._EAB fom Page 202 of 269 6.00E+09 7.00E+09 8.00E+09 9.00E+09 l.OOE+lO CR fom tallies_neut20131017.xlsx Attachment D SpreadSheets Neutron Tally Results Summary of Gamma and Neutron Results Neutron Gamma Total 14 Month EAB 1....-------' 14 Month CR 18 Month EAB FC08513 Page 203 of 269 18 Month CR tallies_neut20131017.xlsx FCS 14M0n EAB Dose nps (14mon) EAB error EAB VOV 3.1457280E+06 5.5506E-06 1.782E-01 2.265E-01 6.2914560E+06 5.5102E-06 1.345E-01 2.032E-01 6.2914560E+06 5.5102E-06 1.345E-01 2.032E-01 9.4371840E+06 5.3999E-06 1.133E-01 1.195E-01 1.2582912E+07 6.0547E-06 1.287E-01 1.876E-01 1.5728640E+07 1.1182E-05 2.539E-01 4.204E-01 1.8874368E+07 1.0251E-05 2.316E-01 4.149E-01 2.2020096E+07 1.1553E-05 2.057E-01 2.612E-01 2.5165824E+07 1.1146E-05 1.876E-01 2.556E-01 2.8311552E+07 1.0529E-05 1.771E-01 2.523E-01 3.1457280E+07 1.0482E-05 1.629E-01 2.361E-01 3.4603008E+07 1.0182E-05 1.531E-01 2.316E-01 3.5000000E+07 1.0176E-05 1.517E-01 2.306E-01 3. 77 48736E+07 1.0632E-05 1.410E-01 1.952E-01 4.4040192E+07 1.0550E-05 1.259E-01 1.720E-01 5.0331648E+07 1.0192E-05 1.162E-01 1.599E-01 5.6623104E+07 1.0035E-05 1.093E-01 1.388E-01 6.2914560E+07 1.0128E-05 1.010E-01 1.219E-01 6.9206016E+07 1.0046E-05 9.550E-02 1.099E-01 7.5497472E+07 1.1073E-05 1.085E-01 1.862E-01 8.1788928E+07 1.1576E-05 1.098E-01 1.423E-01 8.8080384E+07 1.1358E-05 1.043E-01 1.397E-01 9.4371840E+07 1.1220E-05 9.920E-02 1.362E-01 l.OOOOOOOE+08 1.0897E-05 9.660E-02 1.354E-01 7.2500000E+08 1.5595E-05 1.1580E-01 Attachment D SpreadSheets Neutron Tally Results CR EAB slope EAB fom Dose(14mon) 1.80 5.1E-02 1.946E-03 2.00 5.1E-02 1.879E-03 2.00 5.1E-02 1.879E-03 2.20 4.8E-02 1.760E-03 2.10 2.8E-02 1.554E-03 1.90 5.7E-03 1.834E-03 2.00 5.7E-03 1.733E-03 1.90 6.2E-03 1.623E-03 2.00 6.5E-03 1.555E-03 2.00 6.4E-03 1.506E-03 2.00 6.7E-03 1.432E-03 2.10 6.8E-03 1.378E-03 2.10 6.9E-03 1.364E-03 2.10 5.1E-03 1.348E-03 2.10 3.6E-03 1.435E-03 2.10 2.9E-03 1.425E-03 2.10 2.5E-03 1.406E-03 2.10 2.4E-03 1.366E-03 2.10 2.2E-03 1.547E-03 2.00 1.5E-03 1.519E-03 2.00 1.3E-03 1.482E-03 2.10 1.3E-03 1.456E-03 2.30 1.3E-03 1.384E-03 2.30 1.2E-03 1.4437E-03 1.8085E-03 -FC08513 CR error CR vov CR slope CR fom EAB DcHilQil lfii Plus Error Error (14mon) 3.880E-01 2.959E-01 1.20 l.lE-02 2.889E-01 1.686E-01 1.30 l.lE-02 2.889E-01 1.686E-01 1.30 l.lE-02 2.381E-01 1.173E-01 1.30 l.lE-02 2.275E-01 1.117E-01 1.30 8.8E-03 2.185E-01 1.641E-01 1.30 7.7E-03 2.030E-01 1.407E-01 1.40 7.4E-03 1.889E-01 1.321E-01 1.50 7.3E-03 1.776E-01 1.194E-01 1.60 7.2E-03 1.679E-01 1.080E-01 1.60 7.2E-03 1.610E-01 1.031E-01 1.70 6.9E-03 1.534E-01 9.940E-02 1.80 6.8E-03 1.532E-01 9.940E-02 1.80 6.7E-03 1.467E-01 9.250E-02 1.80 4.8E-03 1.362E-01 7.620E-02 1.90 3.1E-03 1.302E-01 6.970E-02 2.00 2.3E-03 1.216E-01 6.230E-02 2.20 2.0E-03 1.155E-01 5.700E-02 2.30 1.8E-03 1.415E-01 2.487E-01 2.30 1.0E-03 1.335E-01 2.390E-01 2.30 9.9E-04 1.278E-01 2.281E-01 2.50 9.5E-04 1.221E-01 2.187E-01 2.60 9.3E-04 1.200E-01 2.179E-01 2.70 8.7E-04 1.161E-01 1.751E-01 2.80 8.5E-04 9.0700E-02 1.91E-05 I 2.13E-03 tallies_neut20131017.xlsx FCS 18M0n EAB Dose nps (18m on) 5.0331648E+07 1.0360E-05 1.0066330E+08 1.1342E-05 1.5099494E+08 1.3747E-05 2.0132659E+08 1.3273E-05 2.5165824E+08 1.3310E-05 3.0198989E+08 1.3783E-05 3.5232154E+08 1.4124E-05 4.0265318E+08 1.7055E-05 4.5298483E+08 1.7312E-05 5.0331648E+08 1.7236E-05 5.5364813E+08 1.7087E-05 6.0397978E+08 1.6951E-05 6.5431142E+08 1.6618E-05 7.0464307E+08 1.6364E-05 7 .2500000E+08 1.6178E-05 1.4983E-05 Desired error nps O.OOOOOOOE+OO 1.0000000E+09 Ratio of 14 mo to 18 mo (0.05) 0.05 0.05 EAB Dose (14mon) tally EAB error 1.3880E-01 1.1260E-01 1.0380E-01 8.7100E-02 7.6100E-02 6.9300E-02 6.4600E-02 1.8880E-01 1.6790E-01 1.5290E-01 1.4100E-01 1.3130E-01 1.2410E-01 1.1770E-01 1.1580E-01 Desired vov (0.10) 0.10 0.10 nps 1.00000E+08 pseudo 7.25E+08 1.0897E-05 1.0408E+OO nps 1.00663E+08 7.25000E+08 1.559SE-OS EAB Dose (18mon) 1.1342E-05 1.4983E-05 EAB vov 1.7770E-01 2.2500E-01 1.0650E-Ol 8.7200E-02 6.7500E-02 5.6900E-02 4.3800E-02 8.6880E-Ol 8.1860E-Ol 7.9330E-01 7.7640E-01 7.5200E-Ol 7.4120E-01 7.2460E-01 7.2350E-01 Desired slope (3.00) 3.00 3.00 Attachment D SpreadSheets Neutron Tally Results 15 EAB slope 2.10 2.10 2.10 2.30 2.70 2.90 2.90 2.60 2.50 2.50 2.50 2.60 2.70 2.80 2.80 CR Dose(14mon ) EAB fom 8.4E-04 6.3E-04 5.0E-04 5.3E-04 5.6E-04 5.6E-04 5.5E-04 5.6E-05 6.3E-05 6.9E-05 7.4E-05 7.8E-05 8.0E-05 8.3E-05 8.3E-05 1.444E-03 1.0304E+OO l.BOSSE-03 CR Dose(18mon ) 1.4876E-03 1.7551E-03 CR Dose(18mon) 1.5929E-03 1.4876E-03 1.6460E-03 1.6499E-03 1.7010E-03 1.7342E-03 1.7089E-03 1.9534E-03 1.9429E-03 1.9028E-03 1.8580E-03 1.8157E-03 1.7817E-03 1.7739E-03 1.7783E-03 1.7551E-03 tally CR error 1.8910E-01 1.2080E-01 9.5500E-02 7.9400E-02 7.0900E-02 6.4000E-02 5.8100E-02 1.4180E-01 1.2810E-01 1.1840E-01 1.1060E-01 1.0410E-01 9.8300E-02 9.3400E-02 9.0700E-02 FC08513 25 CRvov CR slope CR fom EAB Plus Error Error (18mon) 2.8310E-01 2.00 4.5E-04 1.4640E-01 2.70 5.5E-04 7.8100E-02 2.60 5.9E-04 5.4600E-02 2.80 6.4E-04 4.2100E-02 3.20 6.4E-04 3.2600E-02 3.50 6.6E-04 2.7800E-02 3.90 6.8E-04 7.9060E-01 3.20 1.0E-04 7.5680E-01 3.30 l.lE-04 7.3870E-01 3.10 1.1E-04 7.2810E-Ol 3.00 1.2E-04 7.1870E-01 3.20 1.2E-04 7.0950E-01 3.10 1.3E-04 6.5960E-01 3.00 1.3E-04 6.5310E-01 3.00 1.4E-04 1.84E-05 I 2.07E-03 
.... ..c. ....... E Q) 0::: E 2.500E-03 2.250E-03 2.000E-03 1.750E-03 1.500E-03 1.250E-03 l.OOOE-03 7.500E-04 5.000E-04 O.OOE+OO EAB & CR Neutron Dose Rate (mRem/hr) l.OOE+08 2.00E+08 Dose(14mon) 3.00E+08 4.00E+08 Total Particles 5.00E+08 6.00E+08 CR Dose(18mon) EAB Dose (14mon) -.-EAB Dose (18mon) 7.00E+08 Page 206 of 269 2.500E-05 2.250E-05 2.000E-05 1.750E-05 1.500E-05 1.250E-05 l.OOOE-05 7.500E-06 5.000E-06 8.00E+08 Lil 0 c:i 0.45 0.40 0.35 0.30 :::!... 0.25 c:: 0 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 Relative Error 4.00E+08 Total Particles -EAB error -CR error -EAB error Page 207 of 269 5.00E+08 6.00E+08 7.00E+08 8.00E+08 CR error --Desired error (0.05) 0 .-I ci v > 0 > 1.00 0.90 0.80 0.70 I 0.60 0.50 0.40 0.30 0.20 0.10 0.00 O.OE+OO 1.0E+08 Variance of the Variance 2.0E+08 3.0E+08 4.0E+08 Total Particles S.OE+08 6.0E+08 CR vov --Desired vov (0.10) Page 208 of 269 7.0E+08 8.0E+08 0 q m v 4.20 3.80 3.40 3.00 2.60 0 Vi u.. 0 a.. 2.20 1.40 1.00 O.OOE+OO l.OOE+08 2.00E+08 3.00E+08 PDF Slope 4.00E+08 Total Particles slope slope Page 209 of 269 5.00E+08 6.00E+08 7.00E+08 8.00E+08 CR slope --Desired slope (3.00) 2.00E-02 l.SOE-02 l.OOE-02 LL S.OOE-03 O.OOE+OO O.OOE+OO l.OOE+08 2.00E+08 Figure of Merit 3.00E+08 4.00E+08 Total Particles fom fom fom Page 210 of 269 S.OOE+08 6.00E+08 7.00E+08 8.00E+08 CR fom tallies_neut20161017.xlsx Attachment D SpreadSheets Neutron Tally Results Summary of Gamma and Neutron Results Neutron Gamma Total 14 Month EAB L...------....1 14 Month CR 18 Month EAB FC08513 Page 211 of 269 18 Month CR tallies_neut20161017 .xlsx FCS 14MOn EAB Dose nps (14mon) EAB error EAB vov 3.1457280E+06 5.5506E-06 1.782E-01 2.265E-01 6.2914560E+06 5.5102E-06 1.345E-01 2.032E-01 6.2914560E+06 5.5102E-06 1.345E-01 2.032E-01 9.4371840E+06 5.3999E-06 1.133E-01 1.195E-01 1.2582912E+07 6.0547E-06 1.287E-01 1.876E-01 1.5728640E+07 1.1182E-05 2.539E-01 4.204E-01 1.8874368E+07 1.0251E-05 2.316E-01 4.149E-01 2.2020096E+07 1.1553E-05 2.057E-01 2.612E-01 2.5165824E+07 1.1146E-05 1.876E-01 2.556E-01 2.8311552E+07 1.0529E-05 1.771E-01 2.523E-01 3.1457280E+07 1.0482E-05 1.629E-01 2.361E-01 3.4603008E+07 1.0182E-05 1.531E-01 2.316E-01 3.5000000E+07 1.0176E-05 1.517E-01 2.306E-01 3.7748736E+07 1.0632E-05 1.410E-01 1.952E-01 4.4040192E+07 1.0550E-05 1.259E-01 1.720E-01 5.0331648E+07 1.0192E-05 1.162E-01 1.599E-01 5.6623104E+07 1.0035E-05 1.093E-01 1.388E-01 6.2914560E+07 1.0128E-05 1.010E-01 1.219E-01 6.9206016E+07 1.0046E-05 9.550E-02 1.099E-01 7.5497472E+07 1.1073E-05 1.085E-01 1.862E-01 8.1788928E+07 1.1576E-05 1.098E-01 1.423E-01 8.8080384E+07 1.1358E-05 1.043E-01 1.397E-01 9.4371840E+07 1.1220E-05 9.920E-02 1.362E-01 1.0000000E+08 1.0897E-05 9.660E-02 1.354E-01 7.2500000E+08 1.5595E-05 1.1580E-01 Attachment D SpreadSheets Neutron Tally Results CR EAB slope EAB fom Dose(14mon) 1.80 5.1E-02 1.946E-03 2.00 5.1E-02 1.879E-03 2.00 5.1E-02 1.879E-03 2.20 4.8E-02 1.760E-03 2.10 2.8E-02 1.554E-03 1.90 5.7E-03 1.834E-03 2.00 5.7E-03 1.733E-03 1.90 6.2E-03 1.623E-03 2.00 6.SE-03 1.555E-03 2.00 6.4E-03 1.506E-03 2.00 6.7E-03 1.432E-03 2.10 6.8E-03 1.378E-03 2.10 6.9E-03 1.364E-03 2.10 5.1E-03 1.348E-03 2.10 3.6E-03 1.435E-03 2.10 2.9E-03 1.425E-03 2.10 2.5E-03 1.406E-03 2.10 2.4E-03 1.366E-03 2.10 2.2E-03 1.547E-03 2.00 1.5E-03 1.519E-03 2.00 1.3E-03 1.482E-03 2.10 1.3E-03 1.456E-03 2.30 1.3E-03 1.384E-03 2.30 1.2E-03 1.4437E-03 1.8085E-03 -FC08513 CR error CR vov CR slope CR fom EAB DcH8gt Plus Error Error (14mon) 3.880E-01 2.959E-01 1.20 l.lE-02 2.889E-01 1.686E-01 1.30 l.lE-02 2.889E-01 1.686E-01 1.30 l.lE-02 2.381E-01 1.173E-01 1.30 l.lE-02 2.275E-01 1.117E-01 1.30 8.8E-03 2.185E-01 1.641E-01 1.30 7.7E-03 2.030E-01 1.407E-01 1.40 7.4E-03 1.889E-01 1.321E-01 1.50 7.3E-03 1.776E-01 1.194E-01 1.60 7.2E-03 1.679E-01 1.080E-01 1.60 7.2E-03 1.610E-01 1.031E-01 1.70 6.9E-03 1.534E-01 9.940E-02 1.80 6.8E-03 1.532E-01 9.940E-02 1.80 6.7E-03 1.467E-01 9.250E-02 1.80 4.8E-03 1.362E-01 7.620E-02 1.90 3.1E-03 1.302E-01 6.970E-02 2.00 2.3E-03 1.216E-01 6.230E-02 2.20 2.0E-03 1.155E-01 5.700E-02 2.30 1.8E-03 1.415E-01 2.487E-01 2.30 1.0E-03 1.335E-01 2.390E-01 2.30 9.9E-04 1.278E-01 2.281E-01 2.50 9.5E-04 1.221E-01 2.187E-01 2.60 9.3E-04 1.200E-01 2.179E-01 2.70 8.7E-04 1.161E-01 1.751E-01 2.80 8.5E-04 9.0700E-02 1.91E-05 I 2.13E-03 tallies_neut20161017.xlsx FCS 18M0n tally EAB Dose nps (18m on) EAB error EAB VOV 5.0331648E+07 1.0360E-05 1.3880E-01 1.7770E-01 1.0066330E+08 1.1342E-05 1.1260E-01 2.2500E-01 1.5099494E+08 1.3747E-05 1.0380E-01 1.0650E-01 2.0132659E+08 1.3273E-05 8.7100E-02 8.7200E-02 2.5165824E+08 1.3310E-05 7.6100E-02 6.7500E-02 3.0198989E+08 1.3783E-05 6.9300E-02 5.6900E-02 3.5232154E+08 1.4124E-05 6.4600E-02 4.3800E-02 4.0265318E+08 1.7055E-05 1.8880E-01 8.6880E-01 4.5298483E+08 1.7312E-05 1.6790E-01 8.1860E-01 5.0331648E+08 1.7236E-05 1.5290E-01 7.9330E-01 5.5364813E+08 1.7087E-05 1.4100E-01 7.7640E-01 6.0397978E+08 1.6951E-05 1.3130E-01 7.5200E-01 6.5431142E+08 1.6618E-05 1.2410E-01 7.4120E-01 7.0464307E+08 1.6364E-05 1.1770E-01 7.2460E-01 7 .2500000E+08 1.6178E-05 1.1580E-01 7.2350E-01 1.4983E-05 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 EAB Dose nps (14mon) l.OOOOOE+08 1.0897E-05 1.0408E+OO pseudo 7.25E+08 1.559SE-05 EAB Dose nps (18mon) 1.00663E+08 1.1342E-05 7.25000E+08 1.4983E-05 Attachment D SpreadSheets Neutron Tally Results 15 CR EAB slope EAB fom Dose(18mon) 2.10 8.4E-04 1.5929E-03 2.10 6.3E-04 1.4876E-03 2.10 5.0E-04 1.6460E-03 2.30 5.3E-04 1.6499E-03 2.70 5.6E-04 1.7010E-03 2.90 5.6E-04 1.7342E-03 2.90 5.5E-04 1.7089E-03 2.60 5.6E-05 1.9534E-03 2.50 6.3E-05 1.9429E-03 2.50 6.9E-05 1.9028E-03 2.50 7.4E-05 1.8580E-03 2.60 7.8E-05 1.8157E-03 2.70 B.OE-05 1.7817E-03 2.80 8.3E-05 1.7739E-03 2.80 8.3E-05 1.7783E-03 1.7551E-03 CR Dose(14mon ) 1.444E-03 1.0304E+OO 1.8085E-03 CR Dose(18mon ) 1.4876E-03 1.7551E-03 FC08513 tally 25 CR error CR vov CR slope CR fom EAB D<H8gt11Z11 Plus Error Error (18mon) 1.8910E-01 2.8310E-01 2.00 4.5E-04 1.2080E-01 1.4640E-01 2.70 5.5E-04 9.5500E-02 7.8100E-02 2.60 5.9E-04 7.9400E-02 5.4600E-02 2.80 6.4E-04 7.0900E-02 4.2100E-02 3.20 6.4E-04 6.4000E-02 3.2600E-02 3.50 6.6E-04 5.8100E-02 2.7800E-02 3.90 6.8E-04 1.4180E-01 7.9060E-01 3.20 1.0E-04 1.2810E-01 7.5680E-01 3.30 l.lE-04 1.1840E-01 7.3870E-01 3.10 1.1E-04 1.1060E-01 7.2810E-01 3.00 1.2E-04 1.0410E-01 7.1870E-01 3.20 1.2E-04 9.8300E-02 7.0950E-01 3.10 1.3E-04 9.3400E-02 6.5960E-01 3.00 1.3E-04 9.0700E-02 6.5310E-01 3.00 1.4E-04 1.84E-05 I 2.07E-03 
..... ..c. --E <IJ a:: E 2.500E-03 2.250E-03 2.000E-03 1.750E-03 I 1.500E-03 1.250E-03 l.OOOE-03 7.500E-04 5.000E-04 -O.OOE+OO EAB & CR Neutron Dose Rate (mRem/hr) l.OOE+OS 2.00E+08 -+-CR Dose(14mon) 3.00E+08 4.00E+08 Total Particles 5.00E+08 6.00E+08 CR Dose(18mon) -+-EAB Dose (14mon) -+-EAB Dose (18m on) 7.00E+08 Page 214 of 269 2.500E-05 2.250E-05 2.000E-05 1.750E-05 1.500E-05 1.250E-05 l.OOOE-05 7.500E-06 5.000E-06 8.00E+08 
;n 0 0 0.45 0.40 0.35 0.30 ::!... 0.25 c: 0 *.::; 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 Relative Error 4.00E+08 Total Particles -+-EAB error -+-CR error -+-EAB error Page 215 of 269 5.00E+08 6.00E+08 7.00E+08 8.00E+08 CR error --Desired error (0.05) 1.00 0.90 0.80 0.70 0.60 0 c:i v 0.50 > 0 > 0.40 0.30 0.20 0.10 O.OE+OO l.OE+08 Variance of the Variance 2.0E+08 3.0E+08 4.0E+08 Total Particles ......_. EAB vov ......_. CR vov ......_. EAB vov S.OE+08 6.0E+08 CR vov --Desired vov (0.10) Page 216 of 269 7.0E+08 8.0E+08 4.20 3.80 3.40 3.00 v 2.60 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 PDF Slope 4.00E+08 Total Particles -EAB slope -CR slope -EAB slope Page 217 of 269 S.OOE+08 6.00E+08 7.00E+08 8.00E+08 CR slope --Desired slope (3.00) 2.00E-02 l.SOE-02 l.OOE-02 u.. S.OOE-03 O.OOE+OO O.OOE+OO l.OOE+08 2.00E+08 Figure of Merit 3.00E+08 4.00E+08 Total Particles -+-EAB fom -+-CR fom EAB fom Page 218 of 269 5.00E+08 6.00E+08 7.00E+08 8.00E+08 CR fom Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 219 of 269 Attachment E: Additional References Attachment E1 Attachment E2 Attachment E3 Attachment E4 Attachment E5 Reference 5: Pages 107,108,240,241,281,287-290,320,321,350 & 351. Reference 10, All Pages. Reference 18, Page 941. DOW CHEMICAL CO TRYMER 9501-4. 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)
Attachment E1 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). FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Page 220 )f 269 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 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 Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Photons 1000 -0.008485 1000 0.149867 1000 221 pf 269 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 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 Attachment E1 MCNP Form I Weight Fractions I Atom Fractions Neutrons 1001 -0.143716 1001 0.666662 6000 -0.856284 6000 0.333338 Photons 1000 -0.143716 1000 0.666662 6000 -0.856284 6000 0.333338 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). I FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Page 222 pf 269 Atom Densities 1001 0.079855 6000 0.039929 1000 0.079855 6000 0.039929 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 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 Attachment E1 7000 -0.111941 7000 8000 -0.127866 8000 CEPXS Form: material H 0.040277 c 0.719916 N 0.111941 0 0.127866 matname Polyisocyanurate (PIR) density 0.048200 Comments and References 0.068966 0.068966 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. I Page 223 )f 269 7000 O.OOU232 8000 0.000232 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&c TG=1 &c TC= 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 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 Attachment E1 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) = FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Page 224 pf 269 -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 299 Steel, Stainless 304L Formula= -Density (g/cm3) = 8.000000 Attachment E1 Molecular weight (g/mole) = Total atom density (atoms/b-ern)= FC08513 PIET -43741-TM-963 PNNL-15870 Rev. 1 Page 225 of 269 -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 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 Attachment E 1 matname Steel, Stainless 304L density 8.000000 Comments and References Density= 8.00 g/cm3 and weight fractions from FC08513 PIET-43741-TM-963 1 Page 226 )f 269 http://www. matweb.com/search/DataSheet.aspx?MatGUID=e214 7b8f727343b0b0d51 efe02a6127 e (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 Attachment E1 28000 -0.120000 28000 42000 -0.025000 42000 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 0.113803 0.014504 FC08513 PIET -43741-TM-963 PNNL-15870 Rev. 1 Pa_ge 227 )f 269 28000 0.009850 42000 0.001255 http://www.matweb.com/search/DataSheet.aspx?MatGUI D=50f320bd 1 daf4fa 7965448c30d3114ad&ckck= 1 (Automation Creations 2010). Density= 8.03 g/cm3 and same weight fractions at http://www.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 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 Attachment E1 FC08513 PIET -43 7 41-TM-963 PNNL-15870 Rev. I D<:>no??R MCNP Form I Weight Fractions I Atom Fractions I Atom Densities v 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 f269 Attachment E1 FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 U-236 92236 92000 0.000125 0.000046 0.000004 Page 22S 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 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 of269 Attachment E1 FC08513 PIET -43741-TM-963 PNNL-15870 Rev. 1 of269 MCNP Form I Weight Fractions I Atom Fractions I Page 23C 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 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 Total MCNP Form I Weight Fractions Neutrons 8016 -0.001197 24000 -0.000997 26000 -0.000997 28000 -0.000499 40000 -0.982348 50000 -0.013962 Photons 8000 -0.001197 24000 -0.000997 26000 -0.000997 28000 -0.000499 40000 -0.982348 50000 -0.013962 CEPXS Form: material 0 Cr Fe Ni Zr Sn matname Zircaloy-2 density 6.560000 Comments and References See Attachment E1 1.000000 1.000000 I Atom Fractions 8016 0.006796 24000 0.001743 26000 0.001623 28000 0.000772 40000 0.978381 50000 0.010686 8000 0.006796 24000 0.001743 26000 0.001623 28000 0.000772 40000 0.978381 50000 0.010686 0.001197 0.000997 0.000997 0.000499 0.982348 0.013962 I FC08513 PIET-43741-TM-963 PNNL-15870 Rev. 1 Page 231 of 269 0.043481 Atom Densities 8016 0.000296 24000 0.000076 26000 0.000071 28000 0.000034 40000 0.042541 50000 0.000465 8000 0.000296 24000 0.000076 26000 0.000071 28000 0.000034 40000 0.042541 50000 0.000465 http://www. matweb.com/search/DataSheet. aspx?MatGU ID=eb 1 dad5ce 1 ad4a 1 f9e92f86d5b44 7 40d&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 MCNP Form I Weight Fractions Neutrons 8016 -0.001196 24000 -0.000997 26000 -0.001994 40000 -0.981858 50000 -0.013955 Photons 8000 -0.001196 24000 -0.000997 26000 -0.001994 40000 -0.981858 50000 -0.013955 CEPXS Form: material 0 Cr Fe Zr Sn matname Zircaloy-4 density 6.560000 Comments and References Attachment E1 I Atom Fractions 8016 0.006790 24000 0.001741 26000 0.003242 40000 0.977549 50000 0.010677 8000 0.006790 24000 0.001741 26000 0.003242 40000 0.977549 50000 0.010677 0.001196 0.000997 0.001994 0.981858 0.013955 I FC08513 PIET -43741-TM-963 PNNL-15870 Rev. I Page 232 of269 Atom Densities 8016 0.000295 24000 0.000076 26000 0.000141 40000 0.042520 50000 0.000464 8000 0.000295 24000 0.000076 26000 0.000141 40000 0.042520 50000 0.000464 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 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 Neutron reactions
* Attachment E3 Effective FC08513 energy Barns per decay,' Page 248 of 269 Element Mass Abundance Decay MeV 2200 m/s Reso-(Symbol) no., (a/o) or cross nance At. no., Z A Mass, amu half-life Type Percent Total 'Y Prod. section integral 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 .2 'Y 2390 4 4.00260312 99.999863 a/o 0 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 3 7 7.0160039 92.4368 afo 'Y 0.037 8 8.0224871 0.84 8 -,2a 100 8.49 0 Beryllium 9 9.0121855 100 a/o 'Y 0.0092 0.004 (Be) 10 10.0135344 2.5E6y 100 (0.22) 0 'Y <0.001 4 Boron 10.811 a 759 341 (B) 10 10.0129388 19.61 a/o a '3837 1722 5 11 11.0093053 80.39 a/o 'Y 0.0055 12 12.0143537 0.0203 s 100 6.381 0.058 Carbon 12.01115 'Y O.OQ34 0.0015 (C) 11 11.0114317 20.3 m + 100 (1.40) 1.022 't 6 12 12.0000000 98.893 a/o 'Y 0.0034 . 0.0015 ... 13 13.003354 1.107 a/o 'Y 0.0009 0;0013 j 14 14.00324197 5730 y 100 0.045 0 0 ..
1Q/1412016 Attachment E4 www.hazard.com/msds/12/cdrv'cdnsx.html FCOSS13 DOW CHEMICAL CO--TRYMER 9501-4 RIGID FOAM INSULATION, 02837 --9320-00N056603 ===================== Product Identification ===================== 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 Page 249 of 269 1/4 1011412016 Attachment E4 www.hazard.com/msds/12/cdr"Vcdnsx.html DENSE SMOKE & IRRITATING/TOX GASES. RIGID POLYURETHANE (lNG 4) RTECS #:9999999ZZ Ingred Name:ING 3:FOAMS AUTOIGNITE @ 650-800F(343-427C) & RIGID 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 FC08513 Page 250 of 269 214 10114/2016 Attachment E4 www.hazard.com/msdslf2/cdrv'cdnsx.html ===================== Hazards Identification ===================== LDSe LCSe Mixture:NONE SPECIFIED BY MANUFACTURER. 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 FC08513 Page 251 of 269 314 10/1412016 Attachment E4 www.hazard.com/msds/12/cdnlcdnsx.html 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 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/Materials 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 FC08513 Page 252 of 269 414 Thread Name & Version = MCNP5, 1.68 I I I <= 1-1 Attachment E BENCHMARK Case Results +---------------------------------------------------------------------+ 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 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 FC08513 Page 253 of 269 ************************************************************************* probid 87/14/11 13:11:48 n=hall185a tasks 4 1tally fluctuation charts tally nps mean error vov 8192888 2.3587E-14 8.4688 8.7587 16384888 2.2313E-14 8.2972 8.4899 24576888 3.7911E-14 8.2323 8.1475 32768888 4.6842E-14 8.2215 8.1219 48968888 3.8782E-14 8.2111 8.1213 49152888 3.5725E-14 8.1954 8.1113 4 slope fom 2.8 1.2E-81 1.9 1.4E-81 1.7 1.3E-81 1.6 1.8E-81 1.7 8.3E-82 1.6 7.9E-82 tally 34 mean error vov slope fom 1.9187E-81 8.5698 8.9589 1.9 8.3E-82 1.6214E-81 8.3552 8.7723 1.9 9.7E-82 1.7643E-81 8.2368 8.5616 1.8 1.2E-81 2.2123E-81 8.2497 8.4843 1.8 7.9E-82 1.9141E-81 8.2312 8.4821 1.9 6.9E-82 1.9475E-81 8.2898 8.2948 2.8 6.9E-82 Page 1 tally 44 mean error vov slope fom 3.2888E-82 8.1824 8.2469 2.6 8.1E-81 3.5885E-82 8.1231 8.1211 2.1 8.1E-81 3.6983E-82 8.1282 8.2169 2.3 4.7E-81 3.5676E-82 8.8979 8.1818 2.2 5.2E-81 3.3821E-82 8.8865 8.1515 2.3 4.9E-81 5.8388E-82 8.3341 8.9588 2.1 2.7E-82 57344eee 3.4564E-14 e.1816 e.e978 1.6 65536eee 5.2265E-14 9.2936 9.5667 1.6 73728eee 4.9845E-14 9.2753 9.5535 1.7 8192eeee 5.9239E-14 9.2599 9.5198 1.6 9e112eee 4.9269E-14 9.2352 9.4893 1.6 983e4eee 4.9743E-14 9.2195 9.4396 1.6 1eeeeeeee 4.9329E-14 9.2177 9.4392 1.6 1mesh-based weight window generator normalization: 9-group: 3.4962E-99 327691 particles escaped wwg mesh. 7.7E-e2 2.5E-e2 2.5E-e2 2.7E-e2 2.7E-e2 2.8E-e2 2.8E-e2 Attachment E BENCHMARK Case Results 2.e123E-e1 e.1971 e.2211 1.9 6.5E-92 2.2868E-91 9.1758 9.1552 1.8 7 .eE-e2 2.3626E-e1 e.1732 e.1441 1.8 6.4E-e2 2.473eE-e1 e.1622 e.1132 1.8 6.4E-e2 2.4227E-e1 e.1529 e.1e55 1.8 6.5E-e2 2.5897E-e1 e.15e5 e.e914 1.7 6.1E-e2 2.5684E-e1 e.1488 e.e911 1.8 6.1E-e2 print 1-group: 3.4962E-e9 4.9239E-e2 e.2949 e.9211 6.4263E-92 9.2965 9.4699 8.4795E-e2 e.3462 e.4963 8.7632E-92 9.3134 9.4393 8.2632E-e2 e.3922 e.43ee 7.9472E-e2 e.2882 e.4289 7.8821E-e2 e.2857 e.4288 table 199 *********************************************************************************************************************** dump no. 2 on file hall195ar nps 1eeeeeeee call = 883937219 ctm = 743.93 nrn = 11881778178 tally data written to file hall195am 23 warning messages so far. run terminated when 1eeeeeeee particle histories were done. computer time = 743.11 minutes Page 2 FC08513 2 .1 2. 9E 254 of 269 1.9 2.5E-92 2.e 1.6E-e2 2.e 1.7E-e2 2.e 1.7E-e2 2.e 1. 7E-e2 2.1 1.6E-e2 Thread Name & Version = MCNP5, 1.69 I I I I_) I Attachment E BENCHMARK Case Results +---------------------------------------------------------------------+ 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 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 FC08513 Page 255 of 269 ************************************************************************* probid 19/98/16 97:29:15 i=hall195a.txt o=hall195a.out5 1tally fluctuation charts tally 4 nps mean error vov slope fom 8192999 2.3455E-14 9.4683 9.7643 2.9 4.8E-91 16384999 2.2389E-14 9.2978 9.4967 1.9 6.2E-91 24576999 3.6196E-14 9.2313 9.1632 1.7 7.1E-91 32768999 4.5999E-14 9.2243 9.1397 1.7 5.7E-91 49969999 3.7943E-14 9.2136 9.1399 1.7 5.9E-91 49152999 3.5982E-14 9.1975 9.1191 1.6 4.8E-91 57344999 3.4994E-14 9.1833 9.1941 1.6 4.8E-91 65536999 4.6736E-14 9.3149 9.7383 1.7 1.4E-91 73728999 4.4898E-14 9.2939 9.7197 1.7 1.5E-91 81929999 4.5938E-14 9.2637 9.6599 1.6 1. 7E-91 tally 34 mean error vov slope fom 1.9219E-91 9.5696 9.9512 1.9 3.2E-91 1.6265E-91 9.3551 9.7719 1.9 4.4E-91 1.7533E-91 9.2377 9.5648 1.8 6.8E-91 2.2969E-91 9.2591 9.4915 1.8 4.6E-91 1.9999E-91 9.2315 9.3993 1.9 4.2E-91 1.9443E-91 9.2999 9.2926 2.9 4.3E-91 2.9996E-91 9.1972 9.2197 1.9 4.1E-91 2.2914E-91 9.1794 9.1799 1.8 4.4E-91 2.2859E-91 9.1767 9.1539 1.8 4.1E-91 2.4977E-91 9.1648 9.1196 1.8 4.2E-91 Page 1 tally 44 mean error vov slope fom 3.2856E-92 9.1821 9.2456 2.5 3.1E+99 3.5929E-92 9.1226 9.1194 2.1 3.7E+99 3.6927E-92 9.1299 9.2185 2.3 2.6E+99 3.5644E-92 9.9977 9.1821 2.2 3.9E+99 3.3788E-92 9.9864 9.1525 2.3 3.9E+99 5.9279E-92 9.3343 9.9519 2.2 1. 7E-91 4.9246E-92 9.2948 9.9212 2.1 1.9E-91 6.5887E-92 9.3974 9.4893 1.9 1. 5E-91 8.6181E-92 9.3478 9.4717 1.9 1.9E-91 8.8951E-92 9.3149 9.4119 1.9 1.2E-91 90112000 983e4eee 1eeeeeeee 4.5333E-14 0.2458 0.6301 1.6 1.7E-01 4.6108E-14 0.2283 9.5609 1.6 1.9E-01 4.5757E-14 0.2262 9.5604 1.6 1.9E-01 1mesh-based weight window generator Attachment E BENCHMARK Case Results 2.3628E-01 0.1552 0.1124 1.8 4.4E-01 8.3836E-02 0.3038 0.4197 2.5256E-e1 9.1526 9.9951 1.7 4.1E-91 8.e598E-92 e.2898 e.4097 2.5143E-01 9.1598 0.0948 1.7 4.2E-01 7.9930E-02 0.2873 0.4096 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. computer time = 105.29 minutes mcnp version 5 09292010 10/08/16 09:07:27 probid 10/08/16 07:20:15 Page 2 FC08513 2.e 256 of 269 2.0 1.2E-01 2.1 1.2E-01 Code Name & Version MCNP6, 1.0 _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 _I _I _I_!_! _I _I Attachment E BENCHMARK Case Results _I _1_1 _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 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 FC08513 Page 257 of 269 ************************************************************************* probid 10108116 00:02:59 i=hall105a.txt o=hall105a.out 1tally fluctuation charts tally 4 nps mean error vov slope fom 8192000 2.3455E-14 0.4683 0.7643 2.0 3.6E-01 16384000 2.2389E-14 0.2978 0.4067 1.9 4.7E-01 24576000 3.6196E-14 0.2313 0.1632 1.7 5.0E-01 32768000 4.5098E-14 0.2243 0.1307 1.7 3.8E-01 40960000 3.7943E-14 0.2136 0.1300 1.7 3.3E-01 49152000 3.5081E-14 0.1975 0.1191 1.6 3.2E-01 tally mean error vov 1.9210E-01 0.5696 0.9512 1.6265E-01 0.3551 0.7710 1.7533E-01 0.2377 0.5648 2.2068E-01 0.2501 0.4015 1.9099E-01 0.2315 0.3993 1.9443E-01 0.2099 0.2926 Page 1 34 slope fom 1.9 2.4E-01 1.9 3.3E-01 1.8 4. 7E-01 1.8 3.1E-01 1.9 2.8E-01 2.0 2.8E-01 tally 44 mean error vov slope fom 3.2856E-02 0.1821 0.2456 2.5 2.4E+00 3.5029E-02 0.1226 0.1194 2.1 2.8E+00 3.6927E-02 0.1200 0.2185 2.3 1.9E+00 3.5644E-02 0.0977 0.1821 2.2 2.0E+00 3.3788E-02 0.0864 0.1525 2.3 2.0E+00 5.0270E-02 0.3343 0.9510 2.2 1.1E-01 57344999 3.4994E-14 9.1834 9.1941 1.6 3.2E-91 65536999 4.6736E-14 9.3149 9.7383 1.7 9.3E-92 73728999 4.4898E-14 9.2939 9.7197 1.7 9.5E-92 81929999 4.5938E-14 9.2637 9.6599 1.6 1.9E-91 99112999 4.5333E-14 9.2458 9.6391 1.6 1.1E-91 98394999 4.6198E-14 9.2283 9.5699 1.6 1.2E-91 199999999 4.5757E-14 9.2262 9.5694 1.6 1. 2E-91 1mesh-based weight window generator Attachment E BENCHMARK Case Results 2.9996E-91 9.1972 9.2197 1.9 2.7E-91 2.2914E-91 9.1794 9.1799 1.8 2.9E-91 2.2859E-91 9.1767 9.1539 1.8 2.6E-91 2.4977E-91 9.1648 9.1196 1.8 2.7E-91 2.3628E-91 9.1552 9.1124 1.8 2.8E-91 2.5256E-91 9.1526 9.9951 1.7 2.6E-91 2.5143E-91 9.1598 9.9948 1.7 2.7E-91 print 4.9246E-92 9.2948 9.9212 6.5887E-92 9.3974 9.4893 8.6181E-92 9.3478 9.4717 8.8951E-92 9.3149 9.4119 8.3836E-92 9.3938 9.4197 8.9598E-92 9.2898 9.4997 7.9939E-92 9.2873 9.4996 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 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 Page 2 FC08513 2.1 Page 258 of 269 1.2E-91 1.9 9.8E-92 1.9 6.7E-92 1.9 7.4E-92 2.9 7.2E-92 2.9 7.3E-92 2.1 7.4E-92 Fort Calhoun Station CALCULATION SHEET 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 Revision 0 Page 259 of 269 Home Share v 1' > This PC I Program Files Program Files (x86) Recovery
* SWSetup Users I Default.migrated I DefaultAppPool 3D Objects (&#xa3;] Contacts
* Desktop Documents .. Downloads Favorites I Google Drive 5 items View > Local Disk (C:) > Users > Joe "' .... Name 14M0gamma 14MOneutron 18MOgamma 18MOneutron Benchmark Attachment F FC08513 Page 260 of 269 D X * > MCNP _RUNS > SFP > RUN v b Search RUN p Date mo.dified Type Size ---10/17/2016 9:S2 AM File folder 10/17/20161:02 PM File folder 10/17/2016 1:06PM File folder 10/17/2016 9:S4 AM File folder 10/17/2016 1:38PM File folder 14 items Attachment F -;-I 14MOgamma Home! Share! Vie!W y 1' << Local Disk (C:) ) Users ) Joe ) MCNP _RUNS ) SFP ) RUN ) 14MOgamma god iva RNP SFP
* metal ex
* RUN 14MOgamma
* 14M0neutron 18MOgamma 18M0neutron
* save
* case runs old
* Cases for reference FCS_imp
* FCS_WWbase OldM3 SG_Cases Name ]l FCS_14MOg_EABCR.inp J] FCS_14MOg_EABCR.out 0 FCS_14MOg_EABCR.tap .J] FCS_ 14MOg_EABCRca.out FCS_ 14MOg_EABCRcb.out ]l FCS_ 14MOg_EABCRcc.out FCS_ 14MOg_EABCRcd.out Jj FCS_14MOg_EABCRce.out .]) FCS_ 14MOg_EABCRd.out FCS_ 14MOg_EABCRcg.out FCS_ 14MOg_EABCRch.out ]I FCS_ 14MOg_EABCRci.out FCS_14MOg_EABCRci.txt 0 FCS_ 14MOgEABCR.tap Date modified 10/4/2016 8:36AM 10/4/2016 9:04AM 10/4/2016 9:03AM 10/17/2016 9:22AM 10/17/2016 9:22AM 10/17/2016 9:22AM 10/17/2016 9:22 AM 10/17/2016 9:22AM 10/17/2016 9:22AM 10/17/2016 9:22AM 10n12016 9:20PM 10/10/2016 7:33AM 10/10/2016 8:41AM 10/9/201612:24 AM Search 14MOgamma Type Size INP File 17 KB OUT File 118 KB TAP File 42,361 KB OUT File 3 KB OUT File 59 KB OUT File 10 KB OUT File 10 KB OUT File 10 KB OUT File 10 KB OUT File 10 KB OUT File 121 KB OUT File 126 KB Text Document 3 KB TAP File 294,147 KB FC08513 Page 261 of 269 0 X
* p Attachment F
* I 14MOneutron Home Share View .,. 1' * , Joe , MCNP _RUNS , SFP , RUN > 14M0neutron ...... Downloads Favorites
* Google Drive f4 Links MCNP_RUNS
* Benchmark
* examples_files god iva RNP FCS_SFP _H20
* metal ex RUN I14MOgamma 14MOneutron 18MOgamma 18M0neutron Benchmark
* save SCE Files .test 11 Music 10 items 1 item selected 117 KB Name Jj FCS_14MOn_EABCR.inp ,Jl FCS_14M0n_EABCR.out 0 FCS_ 14M0n_EABCR.tap Jj FCS_ 14MOn_EABCRca.out Jj FCS_ 14MOn_EABCRcb.out J FCS_ 14MOn_EABCRcc.out ] FCS_ 14MOn_EABCRcd.out Jl FCS_ 14Mon_EABCRce.out liJ FCS_ 14MOn_EABCRci:-txt-0 FCS_Cont Date modified 10/13/2016 5:21 PM 10/3/2016 5:13PM 10/14/201610:48 AM 10/6/2016 10:54 PM 10/8/20164:2.9 AM 10/17/2016 2:21PM 10/17/2016 2:22PM 10/17/2016 2:19PM 10/17/2016 2:28PM 10/17/2016 1:05AM FC08513 Page 262 of 269 0 X 'v
* Search 1 MOneutron jJ Type Size INP File 17 KB OUT File 173 KB TAP File 640,706 KB OUT File OUT File OUT File OUT File OUT File Text Document File 77 KB 116 KB 3 KB 85 KB 118 KB 3 KB 1 KB m--r,;] &sect;--I.!!!J 8 items Attachment F I
* I 18MOgamma Home Share View v ..,.. << Local Disk (C:) , Users , Joe , MCNP _RUNS , SFP , RUN , 18MOgamma I godiva RNP I SFP I metal ex I RUN I 14MOgamma I 14MOneutron 18MOgamma I 18MOneutron case runs old Cases for reference FCS_imp FCS_WWbase I OldM3 I SG_Cases rrr r:1 ..... Name J FCS_18MOg_EABCR.inp .) FCS_18MOg_EABCR.out D FCS_18MOg_EABCR.tap FCS_18MOg_EABCR.txt ]l FCS_18MOg_EABCRca.out FCS_18MOg_EABCRca.txt D FCS_Cont .]l FCS_Cont..dat Date modified 10/4/2016 8:36AM 10/7/2016 9:18PM 10/11/2016 5:04PM 10/10/2016 8:50AM 10/11/20167:51 PM 10/11/2016 5:33PM 10/10/2016 3:35PM 10/11/2016 7:51PM 5earch lSMOgamma Type Size INP File 17 KB OUT File 192 KB TAP File 440,819 KB Text Document 2 KB OUT File 134 KB Text Document 2 KB File 1 KB DAT File 1 KB FC08513 Page 263 of 269 D X v. p 9 items Attachment F
* I 18MOneutron Home Share View << Lo<:!!l Disl> Users > Joe > MCNP _RUNS , SFP , RUN , 18MOneutron metal ex
* RUN 14MOgamma
* 14MOneutron 18MOgamma 18MOneutron
* save case runs old Ca.ses for reference FCS_imp FCS_WWba.se OldM3 SG_Cases e-rr-r-:1 ** "' Name v Date modified 0 FCS_18MOn_EAB.tap 10/10/2016 3:09PM J FCS_18M0n_EABCR.inp 10/4/2016 8:37AM Jj FCS_18MOn_EABCR.out 10/10/2016 3:39PM ] FCS_ 18MOn_EABCRca.out 10/10/2016 3:38PM J FCS_18MOn_EABCRcb.out 10/10/2016 3:38PM Jl FCS_18MOn_EABCRcc.out 10/10/2016 3:39PM FCS_18MOn_EABCRcc.txt 10/10/2016 3:28PM 0 FCS_Cont 10/10/201612:22 ... FCS_Cont.dat 10/10/2016 3:38PM Search 18MOneutron Type Size TAP File 3,718, 112 KB INP File 17 KB OUT File 501 KB OUT File 3 KB OUT File 3 KB OUT File 116 KB Text Document 2 KB File 1 KB OAT File 1 KB FC08513 Page 264 of 269 D X -* p Attachment F FC08513 Page 265 of 269
* I Eenchmark D X Home Share View -* << Local Disk (C:) > Users ) Joe ) MCNP _RUNS > SFP > RUN > Benchmark Search Benchmark p Output 1<. Name Date modified Type Size
* examples_files ] ha1110Sa.out5 10/8/2016 9:07AM OUTS File 202 KB gociva ]] haii10Sa.out6 10/8/2016 2:49AM OUT6 File 218 KB > RNP haii10Sa.txt 7/8/2015 5:07 AM Text Document 13 KB ..,
* SFP E) haii10Sae.txt 7/8/2015 5:07 AM Text Document 126 KB
* metal ex v RUN 14MOgamma haii10Sao.txt 7/8/2015 5:07 AM Text Document 203 KB 0 metal 10/8/2016 2:49 AM File 7 KB 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 File 1,389 KB 18MOgamma 0 wwout 10/8/2016 2:49AM File 126 KB 18MOneutron D wwouu 10/8/2016 9:07AM File 126 KB Benchmark .., save
* case runs old Ca.ses for reference FCS_imp y U"AIL---11 items Fort Calhoun Station CALCULATION SHEET FC08513 Revision 0 Page 266 of 269 Attachment G: Reviewer Alternate Calculation Joe Willett From: Sent: To:
 
==Subject:==
Attachments: Attachment G Alternate Calculation Steve Gebers, CHP <gebers@centurylink.net> Wednesday, October 12, 2016 12:58 PM 'Jan Bostelman Wildblue'; 'Joseph E Willett' 18 Month EAB exposure rate Skyshine.doc 18 Month EAB exposure rate Skyshine.doc.docx FC08513 Page 267 of 269 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 Geometry Name File Time Group II 1 2 3 4 5 6 7 8 9 10 11 12 13 14 Totals: ...... Jl X Length Diameter Thickness One ........ AJr Water Concrete Iron Lead Zirconium Urania .... 0.100000 Attachment G Alternate Calculation C..lnlonullon Rectangular Volume Source Behind a Wall Rectangular Volume Source Behind a Wall FC08513 Page 268 of 269 C:\Users\Steve\CioudStation\Quantum\BE Wofk\Decommission FCS\mcnp shine\SFP Model for SB Dose.sky Wednesday, October 12,201612:49:03 PM ......... DaHRMullll Energy II Acdvly Radii I PtiCIIIon [ mR/hr (MeV) (PhofoMIMCt 0.100000 2.1200E+17 4.8576E-49 4.2464E-25 0.150000 2.5800E+17 1.1213E-50 1.1929E-26 0.300000 1.9500E+17 4.3329E-35 3.4640E-11 0.450000 5.5900E+16 2.2753E-32 5.2447E-09 0.700000 2.0400E+18 1.0500E-30 0.000009 1.000000 3.5100E+17 8.0622E-30 0.000012 1.500000 6.2200E+16 4.8060E-29 0.000012 2.000000 4.7000E+15 1.2203E-28 0.000002 2.500000 4.7500E+15 2.1335E-28 0.000004 3.000000 9.3000E+13 3.2346E-28 1.2404E-07 4.000000 8.3300E+12 5.2024E-28 1.7870E-08 6.000000 1.0400E+10 8.2986E-28 3.5588E-11 8.000000 1.2000E+09 9.7147E-28 4.8070E-12 10.000000 1.3800E+08 1.0380E-27 5.9067E-13 3.9700E+01 1r 3.1837E+18 4.0756E-27 0.000040 ca.om.Q ........... (Unlla:....., v.lue Jl ,.._ If Vlllue II ....... II Vllue 2.9528E+03 fl y 5.2400E+01 z O.OOOOE+OO 9.2025E+01 Vllidth 3.5483E+01 Height -4.2000E+01 10.000000 Radius One 0.041667 Radius Two 0.041667 1.600000 :i Thickness Two 4.9800E+01 .l llllildll ...... II Amllltnt AJr II Clwwlllb a.-SIIIeld [I Sounle'#a-0.001200 0.012000 2.350000 2.568020 Buildup...., llld AtiiMUIIOft Cotllal .... Blllldup...., ........................... Alit Clwwlllb 0 a.-r8MIIII II ... 1.000000 2.0411E+01 2.806908 6.8137E+02 7.0459E+02 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 
 
COMMONWEALTH OF VIRGINIA CITY OF LYNCHBURG AFFIDAVIT ss. 1. My name is Gayle Elliott. I am Deputy Director, Licensing & Regulatory Affairs, for AREVA NP Inc. (AREVA NP) and as such I am authorized to execute this Affidavit. 2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria. 3. I am familiar with the AREVA NP information contained in Calculation FC08513, entitled, "EAB Radiation Shine Dose 18 Months Post Shutdown with the SFP Drained," and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information. 4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary 'and confidential. 5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information." 
: 6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary: (a) The information reveals details of AREVA NP's research and development plans and programs or their results. (b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service. (c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP. (d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability. (e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP. The information in this Document is considered proprietary for the reasons set forth in paragraphs 6(c), 6(d) and 6(e) above. 7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information. 8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis. 
: 9. The foregoing statements are true and correct to the best of my knowledge, information, and belief. SUBSCRIBED before me this __ I to_th __ dayof J 2016. Sherry L. McFaden NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 10/31/18 Reg. # 7079129 
 
LIC-16-0109 Attachment 6 Page 1    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 NRC's information and are  commitments. Revise the USAR to include a description of how the FCS Spent Fuel Pool design and operational characteristics meets or compares with the NUREG-1738 Industry Decommissioning Commitments (IDCs) and Staff Decommissioning Assumptions (SDAs). Complete in accordance with next scheduled USAR update following exemption approval.}}

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