ML13351A040: Difference between revisions

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| number = ML13351A040
| number = ML13351A040
| issue date = 12/11/2013
| issue date = 12/11/2013
| title = Supplement 1 and Response to Request for Additional Information Re Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E
| title = And Response to Request for Additional Information Re Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E
| author name = Sartain M
| author name = Sartain M
| author affiliation = Dominion Energy Kewaunee, Inc
| author affiliation = Dominion Energy Kewaunee, Inc
Line 14: Line 14:
| page count = 59
| page count = 59
| project =  
| project =  
| stage = Supplement
| stage = Response to RAI
}}
}}


=Text=
=Text=
{{#Wiki_filter:Dominion Energy Kewaunee, Inc.
{{#Wiki_filter:Dominion Energy Kewaunee, Inc.
5000 Dominion Boulevard, Glen Allen, VA 23060                       DoEEE     oflI Web Address: www.dom.com                     December 11, 2013 ATTN: Document Control Desk                                       Serial No. 13-390A U. S. Nuclear Regulatory Commission                               LIC/JG/RO Washington, DC 20555-0001                                         Docket No.: 50-305 License No.: DPR-43 DOMINION ENERGY KEWAUNEE, INC.
5000 Dominion Boulevard, Glen Allen, VA 23060 DoEEE oflI Web Address: www.dom.com December 11, 2013 ATTN: Document Control Desk Serial No.
13-390A U. S. Nuclear Regulatory Commission LIC/JG/RO Washington, DC 20555-0001 Docket No.: 50-305 License No.: DPR-43 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 By application dated July 31, 2013 (Reference 1), 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 By application dated July 31, 2013 (Reference 1), Dominion Energy Kewaunee, Inc.
(DEK) requested exemptions from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2),
(DEK) requested exemptions from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2),
and 10 CFR 50, Appendix E, Section IV, for Kewaunee Power Station (KPS). The requested exemptions would allow DEK to reduce emergency planning requirements and subsequently revise the KPS Emergency Plan consistent with the permanently defueled condition of the station.
and 10 CFR 50, Appendix E, Section IV, for Kewaunee Power Station (KPS).
Subsequently, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) regarding the proposed exemptions (Reference 2). The RAI questions and associated DEK response are provided in Attachment 1 to this letter.
The requested exemptions would allow DEK to reduce emergency planning requirements and subsequently revise the KPS Emergency Plan consistent with the permanently defueled condition of the station.
Subsequently, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) regarding the proposed exemptions (Reference 2).
The RAI questions and associated DEK response are provided in Attachment 1 to this letter.
In response to the staff's comments, DEK is revising the originally proposed exemption request. Attachment 2 to this letter provides a supplement to the proposed exemption request describing the revisions. The analyses and conclusions provided in Reference 1 are not changed by the proposed revisions. The conclusions of the no significant hazards consideration and the environmental considerations contained in Reference 1 are not affected by, and remain applicable to, this revised request.
In response to the staff's comments, DEK is revising the originally proposed exemption request. Attachment 2 to this letter provides a supplement to the proposed exemption request describing the revisions. The analyses and conclusions provided in Reference 1 are not changed by the proposed revisions. The conclusions of the no significant hazards consideration and the environmental considerations contained in Reference 1 are not affected by, and remain applicable to, this revised request.
The July 31, 2014 requested approval date for the submittal remains unchanged.
The July 31, 2014 requested approval date for the submittal remains unchanged.
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: 1. Letter from A. J. Jordan (DEK) to NRC Document Control Desk, "Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E," dated July 31, 2013.
: 1. Letter from A. J. Jordan (DEK) to NRC Document Control Desk, "Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E," dated July 31, 2013.
: 2. Email from Dr. Karl D. Feintuch (NRC) to Margaret Earle, Jack Gadzala, Craig Sly, et al (DEK), "MF2567 Kewaunee Emergency Plan Requests for Exemption MF2567-RAII-ORLT-Norris-001 to -014 8 October 2013," dated October 8, 2013.
: 2. Email from Dr. Karl D. Feintuch (NRC) to Margaret Earle, Jack Gadzala, Craig Sly, et al (DEK), "MF2567 Kewaunee Emergency Plan Requests for Exemption MF2567-RAII-ORLT-Norris-001 to -014 8 October 2013," dated October 8, 2013.
Commitments made by this letter: None cc:   Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Dr. Karl D. Feintuch Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-D15 11555 Rockville Pike Rockville, MD 20852-2738 Mr. Theodore Smith Project Manager U.S. Nuclear Regulatory Commission Two White Flint North, Mail Stop 08-F5 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station
Commitments made by this letter: None cc:
Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Dr. Karl D. Feintuch Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-D15 11555 Rockville Pike Rockville, MD 20852-2738 Mr. Theodore Smith Project Manager U.S. Nuclear Regulatory Commission Two White Flint North, Mail Stop 08-F5 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station


Serial No. 13-390A ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:
Serial No. 13-390A ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:
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10 CFR 50.47(c)(2), AND 10 CFR 50, APPENDIX E, SECTION IV KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.
10 CFR 50.47(c)(2), AND 10 CFR 50, APPENDIX E, SECTION IV KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.


Serial No. 13-390A Attachment 1 Page 1 of 22 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:
Serial No. 13-390A Page 1 of 22 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:
REQUEST FOR EXEMPTIONS FROM PORTIONS OF 10 CFR 50.47(b),
REQUEST FOR EXEMPTIONS FROM PORTIONS OF 10 CFR 50.47(b),
10 CFR 50.47(c)(2), AND 10 CFR 50, APPENDIX E, SECTION IV By application dated July 31, 2013 (Reference 1), Dominion Energy Kewaunee, Inc.
10 CFR 50.47(c)(2), AND 10 CFR 50, APPENDIX E, SECTION IV By application dated July 31, 2013 (Reference 1), Dominion Energy Kewaunee, Inc.
(DEK) requested exemptions from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2),
(DEK) requested exemptions from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2),
and 10 CFR 50, Appendix E, Section IV, for Kewaunee Power Station (KPS). The requested exemptions would allow DEK to reduce emergency planning requirements and subsequently revise the KPS Emergency Plan consistent with the permanently defueled condition of the station.
and 10 CFR 50, Appendix E, Section IV, for Kewaunee Power Station (KPS). The requested exemptions would allow DEK to reduce emergency planning requirements and subsequently revise the KPS Emergency Plan consistent with the permanently defueled condition of the station.
Subsequently, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) regarding the proposed exemptions (Reference 2). The RAI questions and associated DEK responses are provided below.
Subsequently, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) regarding the proposed exemptions (Reference 2).
In the NRC RAI questions, the specific portion of the requirement within the regulation from which exemption is being requested is depicted in emphasized (boldlunderlined) font. The table numbers refer to the tables contained in Reference 1. In the tables below, the column titled "Kewaunee Request Wording" indicates DEK's originally requested exemption as contained in Reference 1. The right column (titled "Past Precedence Wording") indicates exemptions (i.e., exempted wording) as previously granted by NRC for the associated regulation.
The RAI questions and associated DEK responses are provided below.
NRC Question MF2567-RAII-ORLOB-Norris-001 Table 1 Kewaunee Request Wording                   Past Precedence Wording 50.47(b)(1)   ... State and local organizations within ... State and local organizations within the the Emergency Planning Zones-....       Emergency Planning Zones....
In the NRC RAI questions, the specific portion of the requirement within the regulation from which exemption is being requested is depicted in emphasized (boldlunderlined) font. The table numbers refer to the tables contained in Reference 1. In the tables below, the column titled "Kewaunee Request Wording" indicates DEK's originally requested exemption as contained in Reference 1.
The right column (titled "Past Precedence Wording") indicates exemptions (i.e., exempted wording) as previously granted by NRC for the associated regulation.
NRC Question MF2567-RAII-ORLOB-Norris-001 Table 1 Kewaunee Request Wording Past Precedence Wording 50.47(b)(1)  
... State and local organizations within  
...State and local organizations within the the Emergency Planning Zones-....
Emergency Planning Zones....
Although formal offsite emergency plans have typically been exempted for decommissioning sites, State and local organizations continue to be relied upon for firefighting, law enforcement, ambulance and medical services. Please explain why this requirement would not be applicable.
Although formal offsite emergency plans have typically been exempted for decommissioning sites, State and local organizations continue to be relied upon for firefighting, law enforcement, ambulance and medical services. Please explain why this requirement would not be applicable.


===Response===
===Response===
The intent of the originally requested exemption was to continue to rely on State and local organizations for firefighting, law enforcement, ambulance and medical services as needed for events at the site, but without an expected need for these organizations for offsite events.         However, the past precedence wording also meets this intent.
The intent of the originally requested exemption was to continue to rely on State and local organizations for firefighting, law enforcement, ambulance and medical services as needed for events at the site, but without an expected need for these organizations for offsite events.
However, the past precedence wording also meets this intent.
Therefore, DEK is revising the originally requested exemption from portions of 10 CFR
Therefore, DEK is revising the originally requested exemption from portions of 10 CFR


Serial No. 13-390A Attachment 1 Page 2 of 22 50.47(b)(1) in Reference 1, to read as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.
Serial No. 13-390A Page 2 of 22 50.47(b)(1) in Reference 1, to read as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.
Memoranda of Understanding (MOU) establish agreements for assistance of the local organizations for firefighting, law enforcement, ambulance and medical services. The MOU continue to be required per 10 CFR 50.47(b)(3) even after approval of the proposed exemption to a portion of that regulation. Applicable details are contained in the Permanently Defueled Emergency Plan.
Memoranda of Understanding (MOU) establish agreements for assistance of the local organizations for firefighting, law enforcement, ambulance and medical services. The MOU continue to be required per 10 CFR 50.47(b)(3) even after approval of the proposed exemption to a portion of that regulation. Applicable details are contained in the Permanently Defueled Emergency Plan.
NRC Question MF2567-RAII-ORLOB-Norris-002 Table 1 Kewaunee Request Wording                   Past Precedence Wording 50.47(b)(7)   Information is made available to the     Information is made available to the public on a periodic basis on how they public on a periodic basis on how they will be notified and what their initial will be notified and what their initial actions should be in an emergency         actions should be in an emergency (e.g., listening to a local broadcast     (e.g., listening to a local broadcast station and remaining indoors), the     station and remaining indoors), [T]he principal points of contact with the     principal points of contact with the news news media for dissemination of           media for dissemination of information information during an emergency         during an emergency (including the (including the physical location or       physical location or locations) are locations) are established in advance,   established in advance, and procedures for and procedures for coordinated           coordinated dissemination of information to dissemination of information to the     the public are established.
NRC Question MF2567-RAII-ORLOB-Norris-002 Table 1 Kewaunee Request Wording Past Precedence Wording 50.47(b)(7)
I public are established.                 II The regulations in 10 CFR 72.32(a)(16) states, "Arrangements made for providing information to the public." Although Kewaunee has a general licensed ISFSI, the staff informed the previous exemption granted with the regulations in Part 72 for a specific licensed ISFSI. Please describe how information would be disseminated to the public should an event occur at the KPS site.
Information is made available to the Information is made available to the public on a periodic basis on how they public on a periodic basis on how they will be notified and what their initial will be notified and what their initial actions should be in an emergency actions should be in an emergency (e.g., listening to a local broadcast (e.g., listening to a local broadcast station and remaining indoors), the station and remaining indoors), [T]he principal points of contact with the principal points of contact with the news news media for dissemination of media for dissemination of information information during an emergency during an emergency (including the (including the physical location or physical location or locations) are locations) are established in advance, established in advance, and procedures for and procedures for coordinated coordinated dissemination of information to dissemination of information to the the public are established.
I public are established.
II The regulations in 10 CFR 72.32(a)(16) states, "Arrangements made for providing information to the public." Although Kewaunee has a general licensed ISFSI, the staff informed the previous exemption granted with the regulations in Part 72 for a specific licensed ISFSI. Please describe how information would be disseminated to the public should an event occur at the KPS site.


===Response===
===Response===
Dominion Resources, Inc. (Dominion), the parent company for DEK, maintains a corporate communications organization, which includes a media relations group. News media contacts for the KPS location continue to be maintained. Should an event occur at the KPS site, information would be disseminated to the public and briefings with pertinent media organizations would be conducted per Dominion corporate communication protocols. 10 CFR 50.72(b)(2)(xi) requires that the NRC be notified, via the Emergency Notification System, within 4 hours of certain events for which a news release is planned. The process for complying with this requirement is procedurally implemented.
Dominion Resources, Inc. (Dominion), the parent company for DEK, maintains a corporate communications organization, which includes a media relations group. News media contacts for the KPS location continue to be maintained. Should an event occur at the KPS site, information would be disseminated to the public and briefings with pertinent media organizations would be conducted per Dominion corporate communication protocols. 10 CFR 50.72(b)(2)(xi) requires that the NRC be notified, via the Emergency Notification System, within 4 hours of certain events for which a news release is planned.
The process for complying with this requirement is procedurally implemented.


Serial No. 13-390A Attachment 1 Page 3 of 22 Since there are no longer any pre-planned actions that the public needs to take as a result of an anticipated emergency at KPS, it is no longer necessary to pre-plan dissemination of emergency information to the public.
Serial No. 13-390A Page 3 of 22 Since there are no longer any pre-planned actions that the public needs to take as a result of an anticipated emergency at KPS, it is no longer necessary to pre-plan dissemination of emergency information to the public.
The intent of the originally requested exemption was to discontinue specific emergency response organizational requirements for major interactions with news media.
The intent of the originally requested exemption was to discontinue specific emergency response organizational requirements for major interactions with news media.
However, the past precedence wording also meets this intent. Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50.47(b)(7) in Reference 1, to read as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.
However, the past precedence wording also meets this intent.
Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50.47(b)(7) in Reference 1, to read as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.
Applicable details regarding how information would be disseminated to the public should an event occur at the KPS site are contained in the Permanently Defueled Emergency Plan.
Applicable details regarding how information would be disseminated to the public should an event occur at the KPS site are contained in the Permanently Defueled Emergency Plan.
NRC Question MF2567-RAII-ORLOB-Norris-003 Table 1 Kewaunee Request Wording                 Past Precedence Wording 50.47(b)(9) Adequate methods, systems, and         Adequate methods, systems, and equipment for assessing and           equipment for assessing and monitoring monitoring actual or potential offsite actual or potential offsite consequences consequences of a radiological         of a radiological emergency condition are emergency condition are in use.       in use.
NRC Question MF2567-RAII-ORLOB-Norris-003 Table 1 Kewaunee Request Wording Past Precedence Wording 50.47(b)(9)
The regulations in 10 CFR 72.32(a)(4) states, "Detection of accidents. Identification of the means of detecting an accident condition." Previous exemptions were granted for only the "offsite" assessment and monitoring. Please provide specific justification for exempting this requirement.
Adequate methods, systems, and Adequate methods, systems, and equipment for assessing and equipment for assessing and monitoring monitoring actual or potential offsite actual or potential offsite consequences consequences of a radiological of a radiological emergency condition are emergency condition are in use.
in use.
The regulations in 10 CFR 72.32(a)(4) states, "Detection of accidents. Identification of the means of detecting an accident condition." Previous exemptions were granted for only the "offsite" assessment and monitoring.
Please provide specific justification for exempting this requirement.


===Response===
===Response===
The intent of the originally requested exemption was to discontinue only those requirements associated with "offsite" assessment and monitoring. However, the past precedence wording also meets this intent. Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50.47(b)(9) in Reference 1, as shown in of this submittal. The revised request is consistent with the past precedence wording shown above.
The intent of the originally requested exemption was to discontinue only those requirements associated with "offsite" assessment and monitoring. However, the past precedence wording also meets this intent. Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50.47(b)(9) in Reference 1, as shown in of this submittal.
The revised request is consistent with the past precedence wording shown above.
Applicable details regarding methods, systems, and equipment for assessing and monitoring actual or potential onsite consequences of a radiological emergency condition at the KPS site are contained in the Permanently Defueled Emergency Plan.
Applicable details regarding methods, systems, and equipment for assessing and monitoring actual or potential onsite consequences of a radiological emergency condition at the KPS site are contained in the Permanently Defueled Emergency Plan.


Serial No. 13-390A Attachment 1 Page 4 of 22 NRC Question MF2567-RAII-ORLOB-Norris-004 Table 1 Kewaunee Request Wording                       Past Precedence Wording 50.47(b)(14) Periodic exercises are (will be)           Periodic exercises are (will be) conducted conducted to evaluate maior               to evaluate major portions of emergency portions of emerciency response           response capabilities, periodic drills are (will capabilities, periodic drills are (will   be) conducted to develop and maintain key be) conducted to develop and maintain     skills, and deficiencies identified as a result key skills, and deficiencies identified as of exercises or drills are (will be) corrected.
Serial No. 13-390A Page 4 of 22 NRC Question MF2567-RAII-ORLOB-Norris-004 Table 1 Kewaunee Request Wording Past Precedence Wording 50.47(b)(14)
Periodic exercises are (will be)
Periodic exercises are (will be) conducted conducted to evaluate maior to evaluate major portions of emergency portions of emerciency response response capabilities, periodic drills are (will capabilities, periodic drills are (will be) conducted to develop and maintain key be) conducted to develop and maintain skills, and deficiencies identified as a result key skills, and deficiencies identified as of exercises or drills are (will be) corrected.
a result of exercises or drills are (will be) corrected.
a result of exercises or drills are (will be) corrected.
10 CFR 72.32(a)(12) states:
10 CFR 72.32(a)(12) states:
Line 90: Line 110:
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises


Serial No. 13-390A Attachment 1 Page 5 of 22 is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.
Serial No. 13-390A Page 5 of 22 is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.
KPS plans to conduct biennial exercises to test the adequacy of timing and content of implementing procedures and methods; to test emergency equipment and communication networks; and to ensure that emergency personnel are familiar with their duties. KPS plans to invite offsite response organizations to participate in the exercises.
KPS plans to conduct biennial exercises to test the adequacy of timing and content of implementing procedures and methods; to test emergency equipment and communication networks; and to ensure that emergency personnel are familiar with their duties.
KPS plans to invite offsite response organizations to participate in the exercises.
For alternating years, a drill would be conducted for the purpose of testing, developing, and maintaining the proficiency of on-site emergency responders.
For alternating years, a drill would be conducted for the purpose of testing, developing, and maintaining the proficiency of on-site emergency responders.
Exercise and drill scenarios would include, at a minimum, the following:
Exercise and drill scenarios would include, at a minimum, the following:
Line 100: Line 121:
Critiques would evaluate the performance of the organization.
Critiques would evaluate the performance of the organization.
Applicable details regarding specific requirements for exercises and drills at KPS are contained in the Permanently Defueled Emergency Plan.
Applicable details regarding specific requirements for exercises and drills at KPS are contained in the Permanently Defueled Emergency Plan.
NRC Question MF2567-RAII-ORLOB-Norris-005 Table 2 Kewaunee Request Wording                     Past Precedence Wording IOCFR50       ... and outside the site boundary to       ...... and outside the site boundary to Appendix     protect health and safety. The emergency     protect health and safety. The emergency E IV.B.1     action levels shall be based on in-plant     action levels shall be based on in-plant conditions and instrumentation in addition   conditions and instrumentation in addition to onsite and offsite monitoring. By June   to onsite and offsite monitoring. By June 20, 2012. for nuclear power reactor         20, 2012, for nuclear power reactor licensees, these action levels must         licensees, these action levels must include hostile action that may             include hostile action that may adversely affect the nuclear power plant,   adversely affect the nuclear power The initial emergency action levels shall be plant. The initial emergency action levels discussed and agreed on by the applicant     shall be discussed and agreed on by the or licensee and state and local             applicant or licensee and state and local governmental authorities, and approved by   governmental authorities, and approved by the NRC. Thereafter, emergency action       the NRC. Thereafter, [E]mergency action levels shall be reviewed with the State     levels shall be reviewed with the State and and local governmental authorities on       local governmental authorities on an an annual basis.                             annual basis.
NRC Question MF2567-RAII-ORLOB-Norris-005 Table 2 Kewaunee Request Wording Past Precedence Wording IOCFR50  
... and outside the site boundary to  
...... and outside the site boundary to Appendix protect health and safety. The emergency protect health and safety. The emergency E IV.B.1 action levels shall be based on in-plant action levels shall be based on in-plant conditions and instrumentation in addition conditions and instrumentation in addition to onsite and offsite monitoring. By June to onsite and offsite monitoring. By June 20, 2012. for nuclear power reactor 20, 2012, for nuclear power reactor licensees, these action levels must licensees, these action levels must include hostile action that may include hostile action that may adversely affect the nuclear power plant, adversely affect the nuclear power The initial emergency action levels shall be plant. The initial emergency action levels discussed and agreed on by the applicant shall be discussed and agreed on by the or licensee and state and local applicant or licensee and state and local governmental authorities, and approved by governmental authorities, and approved by the NRC. Thereafter, emergency action the NRC. Thereafter, [E]mergency action levels shall be reviewed with the State levels shall be reviewed with the State and and local governmental authorities on local governmental authorities on an an annual basis.
annual basis.


Serial No. 13-390A Attachment 1 Page 6 of 22 Maintaining the requirement for the offsite response organizations (OROs) to review the EALs on an annual basis will ensure the proper awareness by OROs of applicable emergency classifications and will also ensure that communications with the proper authorities are maintained. Please provide specific justification for exempting this requirement.
Serial No. 13-390A Page 6 of 22 Maintaining the requirement for the offsite response organizations (OROs) to review the EALs on an annual basis will ensure the proper awareness by OROs of applicable emergency classifications and will also ensure that communications with the proper authorities are maintained.
Please provide specific justification for exempting this requirement.


===Response===
===Response===
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Based on the significantly reduced scope of EALs for the permanently defueled facility, the scope of the annual review of EALs with State and local governmental authorities (to ensure proper awareness of applicable emergency classifications) is expected to be commensurately reduced (e.g., informational mailings, etc.).
Based on the significantly reduced scope of EALs for the permanently defueled facility, the scope of the annual review of EALs with State and local governmental authorities (to ensure proper awareness of applicable emergency classifications) is expected to be commensurately reduced (e.g., informational mailings, etc.).
Applicable details regarding review of EALs with State and local governmental authorities are contained in the Permanently Defueled Emergency Plan.
Applicable details regarding review of EALs with State and local governmental authorities are contained in the Permanently Defueled Emergency Plan.
NRC Question MF2567-RAII-ORLOB-Norris-006 Table 2 Kewaunee Request Wording           Past Precedence Wording IOCFR50     In addition, a radiological orientation Appendix     training program shall be made E IV.F.1     available to local services personnel; e.g., local emergency services/Civil Defense, local law enforcement personnel, local news media persons.
NRC Question MF2567-RAII-ORLOB-Norris-006 Table 2 Kewaunee Request Wording Past Precedence Wording IOCFR50 In addition, a radiological orientation Appendix training program shall be made E IV.F.1 available to local services personnel; e.g., local emergency services/Civil Defense, local law enforcement personnel, local news media persons.
Previous exemption requests did not include an exemption for this requirement.
Previous exemption requests did not include an exemption for this requirement.
Firefighting, local law enforcement and ambulance/medical facilities are still expected to play a role in some onsite emergencies; therefore they should have some basic knowledge about radiation. 10 CRF 50.47(b)(15) requires that radiological emergency response training is provided to those who may be called on to assist in an emergency.
Firefighting, local law enforcement and ambulance/medical facilities are still expected to play a role in some onsite emergencies; therefore they should have some basic knowledge about radiation. 10 CRF 50.47(b)(15) requires that radiological emergency response training is provided to those who may be called on to assist in an emergency.
10 CFR 72.32(a)(8) requires a commitment to and a brief description of the means to promptly notify offsite response organizations and request offsite assistance, including medical assistance for the treatment of contaminated injured onsite workers when
10 CFR 72.32(a)(8) requires a commitment to and a brief description of the means to promptly notify offsite response organizations and request offsite assistance, including medical assistance for the treatment of contaminated injured onsite workers when


Serial No. 13-390A Attachment 1 Page 7 of 22 appropriate. 10 CFR 72.32(a)(10) requires training including any special instructions or orientation tours the licensee would offer to fire, police medical and other emergency personnel.     Additionally, 10 CFR 72.32(a)(16) requires arrangements made for providing information to the public. Please provide justification for exempting this requirement.
Serial No. 13-390A Page 7 of 22 appropriate. 10 CFR 72.32(a)(10) requires training including any special instructions or orientation tours the licensee would offer to fire, police medical and other emergency personnel.
Additionally, 10 CFR 72.32(a)(16) requires arrangements made for providing information to the public.
Please provide justification for exempting this requirement.


===Response===
===Response===
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Applicable details regarding the extent of the radiological orientation training program available to local services personnel are contained in the Permanently Defueled Emergency Plan.
Applicable details regarding the extent of the radiological orientation training program available to local services personnel are contained in the Permanently Defueled Emergency Plan.


Serial No. 13-390A Attachment 1 Page 8 of 22 NRC Question MF2567-RAII-ORLOB-Norris-007 Table 2 Kewaunee Request Wording                     Past Precedence Wording 10CFR50     The plan shall describe provisions for the The plan shall describe provisions for the Appendix     conduct of emergency preparedness         conduct of emergency preparedness
Serial No. 13-390A Page 8 of 22 NRC Question MF2567-RAII-ORLOB-Norris-007 Table 2 Kewaunee Request Wording Past Precedence Wording 10CFR50 The plan shall describe provisions for the The plan shall describe provisions for the Appendix conduct of emergency preparedness conduct of emergency preparedness
  *EIV.F.2     exercises as follows: Exercises shall     exercises as follows: Exercises shall test the test the adequacy of timing and content   adequacy of timing and content of of implementing procedures and             implementing procedures and methods, test methods, test emergency equipment and     emergency equipment and communications communications networks, test the         networks, test the public alert and public alert and notification system,     notification system, and ensure that and ensure that emergency organization     emergency organization personnel are personnel are familiar with their duties. familiar with their duties.
*E IV.F.2 exercises as follows: Exercises shall exercises as follows: Exercises shall test the test the adequacy of timing and content adequacy of timing and content of of implementing procedures and implementing procedures and methods, test methods, test emergency equipment and emergency equipment and communications communications networks, test the networks, test the public alert and public alert and notification system, notification system, and ensure that and ensure that emergency organization emergency organization personnel are personnel are familiar with their duties.
familiar with their duties.
See RAI 4, above.
See RAI 4, above.
Previous exemptions did not grant an exemption as requested.                           Additionally, the regulations in 10 CFR 30, 10 CFR 40, 10 CFR 70, and 10 CFR 72 related to emergency plans require licensees to conduct a biennial exercise.                     Please provide specific justification for exempting this requirement.
Previous exemptions did not grant an exemption as requested.
Additionally, the regulations in 10 CFR 30, 10 CFR 40, 10 CFR 70, and 10 CFR 72 related to emergency plans require licensees to conduct a biennial exercise.
Please provide specific justification for exempting this requirement.


===Response===
===Response===
Line 134: Line 164:
Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.
Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.


Serial No. 13-390A Attachment 1 Page 9 of 22 NRC Question MF2567-RAII-ORLOB-Norris-008 Table 2 Kewaunee Request Wording                       Past Precedence Wording IOCFR50     Each licensee at each site shall conduct a  Each licensee at each site shall conduct a Appendix  subsequent exercise of its onsite            subsequent exercise of its onsite emergency E IV.F.2.b emergency plan every 2 years. Nuclear        plan every 2 years. Nuclear power reactor power reactor licensees shall submit        licensees shall submit exercise scenarios exercise scenarios under F450.4 at least    under 4 50.4 at least 60 days before use 60 days before use in an exercise            in an exercise required by this paragraph required by this paragraph 2.b. The          2.b. The exercise may be included in the exercise may be included in the full         full participation biennial exercise participation biennial exercise required     required by paragraph 2.c. of this by paragraph 2.c. of this section. In        Section. In addition, the licensee shall take addition, the licensee shall take actions    actions necessary to ensure that adequate necessary to ensure that adequate            emergency response capabilities are emergency response capabilities              maintained during the interval between are maintained during the interval          biennial exercises by conducting drills, between biennial exercises by                including at least one drill involving a conducting drills, including at least one    combination of some of the principal drill involving a combination of some of the functional areas of the licensee's onsite principal functional areas of the licensee's emergency response capabilities. The onsite emergency response capabilities.      principal functional areas of emergency The principal functional areas of            response include activities such as emergency response include activities        management and coordination of emergency such as management and coordination of      response, accident assessment, event emergency response, accident                classification, notification of offsite assessment, event classification,            authorities, assessment of the onsite and notification of offsite authorities,        offsite impact of radiological releases, assessment of the onsite and offsite        protective action recommendation impact of radiological releases, protective  development, protective action decision action recommendation development,          making, plant system repair and mitigative protective action decision making, plant    action implementation. During these drills, system repair and mitigative action          activation of all of the licensee's emergency implementation. During these drills,        response facilities (Technical Support activation of all of the licensee's          Center (TSC), Operations Sutport Center emergency response facilities (Technical    (OSC), and the Emergency Operations Support Center (TSC), Operations            Facility (EOF)) would not be necessary, Support Center (OSC), and the               licensees would have the opportunity to Emergency Operations Facility (EOF))         consider accident management strategies, would not be necessary, licensees would     supervised instruction would be permitted, have the opportunity to consider accident   operating staff in all participating facilities management strategies, supervised           would have the opportunity to resolve instruction would be permitted, operating   problems (success paths) rather than have staff in all participating facilities would controllers intervene, and the drills may have the opportunity to resolve problems     focus on the onsite exercise training (success paths) rather than have             objectives.
Serial No. 13-390A Page 9 of 22 NRC Question MF2567-RAII-ORLOB-Norris-008 Table 2 Kewaunee Request Wording Past Precedence Wording IOCFR50 Appendix E IV.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 F4 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.
controllers intervene, and the drills may focus on the onsite exercise training objectives.
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, 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.
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 4 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, 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 Sutport 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.
See RAI 4, above.
See RAI 4, above.


Serial No. 13-390A Attachment 1 Page 10 of 22 Previous exemptions did not grant an exemption as requested. Additionally, the regulations in 10 CFR 30, 10 CFR 40, 10 CFR 70, and 10 CFR 72 related to emergency plans require licensees to conduct a biennial exercise.               Please provide specific justification for exempting this requirement.
Serial No. 13-390A Page 10 of 22 Previous exemptions did not grant an exemption as requested.
Additionally, the regulations in 10 CFR 30, 10 CFR 40, 10 CFR 70, and 10 CFR 72 related to emergency plans require licensees to conduct a biennial exercise.
Please provide specific justification for exempting this requirement.


===Response===
===Response===
As discussed in the response to Question 7 above, the intent of the originally requested exemption was to continue testing the adequacy of the emergency response organization, except to perform these tests during the conduct of drills. However, the past precedence wording also meets this intent. Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.F.2.b in Reference 1, as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.
As discussed in the response to Question 7 above, the intent of the originally requested exemption was to continue testing the adequacy of the emergency response organization, except to perform these tests during the conduct of drills. However, the past precedence wording also meets this intent.
Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.F.2.b in Reference 1, as shown in Attachment 2 of this submittal.
The revised request is consistent with the past precedence wording shown above.
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.
Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.
Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.
NRC Question MF2567-RAII-ORLOB-Norris-009 Table 2 Kewaunee Request Wording                 Past Precedence Wording IOCFR5O     Licensees shall enable any State or   Licensees shall enable any State or local Appendix     local government located within the   government located within the plume E IV.F.2.e   plume exposure Pathway EPZ to         exposure Pathway EPZ to participate in the participate in the licensee's drills licensee's drills when requested by such State when requested by such State or local or local government.
NRC Question MF2567-RAII-ORLOB-Norris-009 Table 2 Kewaunee Request Wording Past Precedence Wording IOCFR5O Licensees shall enable any State or Licensees shall enable any State or local Appendix local government located within the government located within the plume E IV.F.2.e plume exposure Pathway EPZ to exposure Pathway EPZ to participate in the participate in the licensee's drills licensee's drills when requested by such State when requested by such State or local or local government.
I government.
I government.
Previous exemption requests did not include an exemption for this requirement. 10 CFR 72,32(a)(10) requires training, including any special instructions or orientation tours the licensee would offer to fire, police medical and other emergency personnel.
Previous exemption requests did not include an exemption for this requirement.
10 CFR 72,32(a)(10) requires training, including any special instructions or orientation tours the licensee would offer to fire, police medical and other emergency personnel.
Additionally, 10 CFR 72.32(a)(12)(i) requires the licensee to invite offsite response organizations to participate in the biennial exercise. Additionally, CFR 72.32(a)(12) (ii) states in part, that participation of offsite response organizations in biennial exercises,
Additionally, 10 CFR 72.32(a)(12)(i) requires the licensee to invite offsite response organizations to participate in the biennial exercise. Additionally, CFR 72.32(a)(12) (ii) states in part, that participation of offsite response organizations in biennial exercises,


Serial No. 13-390A Attachment 1 Page 11 of 22 although recommended, is not required. Please provide justification for exempting this requirement.
Serial No. 13-390A Page 11 of 22 although recommended, is not required. Please provide justification for exempting this requirement.


===Response===
===Response===
Similar to the discussion contained in the response to Question 1 above, the intent of the originally requested exemption was to continue to rely on State and local organizations for firefighting, law enforcement, ambulance and medical services as needed for events at the site, but without an expected need for these organizations for offsite events.       However, the past precedence wording also meets this intent.
Similar to the discussion contained in the response to Question 1 above, the intent of the originally requested exemption was to continue to rely on State and local organizations for firefighting, law enforcement, ambulance and medical services as needed for events at the site, but without an expected need for these organizations for offsite events.
However, the past precedence wording also meets this intent.
Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.F.2.e in Reference 1, as shown in Attachment 2 of this submittal.
Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.F.2.e in Reference 1, as shown in Attachment 2 of this submittal.
The revised request is consistent with the past precedence wording shown above.
The revised request is consistent with the past precedence wording shown above.
Applicable details regarding offsite agency personnel participation in the licensee's drills are contained in the Permanently Defueled Emergency Plan.
Applicable details regarding offsite agency personnel participation in the licensee's drills are contained in the Permanently Defueled Emergency Plan.
NRC Question MF2567-RAII-ORLOB-Norris-010 Table 2 I       Kewaunee Request Wording         f           Past Precedence Wording 10CFR50     Remedial exercises will be required if     Remedial exercises will be required if the Appendix    the emergency plan is not satisfactorily  emergency plan is not satisfactorily tested E IV.F.2.f  tested during the biennial exercise,      during the biennial exercise, such that such that NRC. in consultation with        NRC, in consultation with FEMA. cannot FEMA, cannot (1) find reasonable          (1) find reasonable assurance that assurance that adequate protective        adequate protective measures can and will measures can and will be taken in the      be taken in the event of a radiological event of a radiological emergency or      emergency or (2) determine that the (2) determine that the Emergency          Emergency Response Organization (ERO)
NRC Question MF2567-RAII-ORLOB-Norris-010 Table 2 I
Response Organization (ERO) has           has maintained key skills specific to maintained key skills specific to         emergency response. The extent of State emergency response. The extent of          and local participation in remedial State and local participation in          exercises must be sufficient to show remedial exercises must be sufficient      that appropriate corrective measures to show that appropriate corrective        have been taken regarding the elements measures have been taken regarding        of the plan not properly tested in the the elements of the plan not properly     previous exercises.
Kewaunee Request Wording f
t~~t~d in the~ nr~vin.i~ ~
Past Precedence Wording 10CFR50 Appendix E IV.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 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.
_____    teste     th previous_____exercises___
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 t~~t~d in the~ nr~vin.i~ ~
in~~~
teste in~~~
th previous_____exercises___
Previous exemptions did not grant an exemption as requested. Biennial exercises are required and are subject to NRC inspection. 10 CFR 50.47(b)(14) states that 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. A remedial exercise ensures that, in the event that an exercise did not provide reasonable
Previous exemptions did not grant an exemption as requested. Biennial exercises are required and are subject to NRC inspection. 10 CFR 50.47(b)(14) states that 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. A remedial exercise ensures that, in the event that an exercise did not provide reasonable


Serial No. 13-390A Attachment 1 Page 12 of 22 assurance to the NRC that the license can and will take adequate protective measures in the event of a radiological emergency, the deficiencies are corrected. Please provide justification for exempting this requirement.
Serial No. 13-390A Page 12 of 22 assurance to the NRC that the license can and will take adequate protective measures in the event of a radiological emergency, the deficiencies are corrected. Please provide justification for exempting this requirement.


===Response===
===Response===
As discussed in the response to Questions 7 and 8 above, the intent of the originally requested exemption was to continue testing the adequacy of the emergency response organization, except to perform these tests during the conduct of drills. However, the past precedence wording also meets this intent. Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.F.2.f in Reference 1, as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.
As discussed in the response to Questions 7 and 8 above, the intent of the originally requested exemption was to continue testing the adequacy of the emergency response organization, except to perform these tests during the conduct of drills. However, the past precedence wording also meets this intent.
Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.F.2.f in Reference 1, as shown in Attachment 2 of this submittal.
The revised request is consistent with the past precedence wording shown above.
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.
Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.
Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.
NRC Question MF2567-RAII-ORLOB-Norris-O1l Table 2 Kewaunee Request Wording                   Past Precedence Wording IOCFR50     Licensees shall use drill and exercise Licensees shall use drill and exercise Appendix     scenarios that provide reasonable       scenarios that provide reasonable assurance E IV.F.2.i   assurance that anticipatory responses   that anticipatory responses will not result from will not result from preconditioning of preconditioning of participants. Such participants. Such scenarios for       scenarios for nuclear power reactor nuclear power reactor licensees must   licensees must include a wide spectrum of include a wide spectrum of             radiological releases and events, including radiological releases and events,       hostile action.
NRC Question MF2567-RAII-ORLOB-Norris-O1l Table 2 Kewaunee Request Wording Past Precedence Wording IOCFR50 Licensees shall use drill and exercise Licensees shall use drill and exercise Appendix scenarios that provide reasonable scenarios that provide reasonable assurance E IV.F.2.i assurance that anticipatory responses that anticipatory responses will not result from will not result from preconditioning of preconditioning of participants. Such participants. Such scenarios for scenarios for nuclear power reactor nuclear power reactor licensees must licensees must include a wide spectrum of include a wide spectrum of radiological releases and events, including radiological releases and events, hostile action.
including hostile action. Exercise and Exercise and drill scenarios as appropriate drill scenarios as appropriate must     must emphasize coordination among onsite emphasize coordination among onsite     and offsite response organizations.
including hostile action. Exercise and Exercise and drill scenarios as appropriate drill scenarios as appropriate must must emphasize coordination among onsite emphasize coordination among onsite and offsite response organizations.
and offsite response organizations.
and offsite response organizations.
Previous exemptions did not grant an exemption as requested. Biennial exercises are required and are subject to NRC inspection.               10 CFR 50.47(b)(14) that 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,
Previous exemptions did not grant an exemption as requested. Biennial exercises are required and are subject to NRC inspection.
10 CFR 50.47(b)(14) that 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,


Serial No. 13-390A Attachment 1 Page 13 of 22 and deficiencies identified as a result of exercises or drills are (will be) corrected. 10 CFR 72.32(a)(10) requires training including any special instructions or orientation tours the licensee would offer to fire, police medical and other emergency personnel.
Serial No. 13-390A Page 13 of 22 and deficiencies identified as a result of exercises or drills are (will be) corrected.
10 CFR 72.32(a)(10) requires training including any special instructions or orientation tours the licensee would offer to fire, police medical and other emergency personnel.
Additionally, 10 CFR 72.32(a)(12)(i) requires the licensee to invite offsite response organizations to participate in the biennial exercise. Additionally, CFR 72.32(a)(12) (ii) states in part, that participation of offsite response organizations in biennial exercises, although recommended, is not required. Please provide justification for exempting this requirement.
Additionally, 10 CFR 72.32(a)(12)(i) requires the licensee to invite offsite response organizations to participate in the biennial exercise. Additionally, CFR 72.32(a)(12) (ii) states in part, that participation of offsite response organizations in biennial exercises, although recommended, is not required. Please provide justification for exempting this requirement.


===Response===
===Response===
As discussed in the response to Questions 7, 8, and 10 above, the intent of the originally requested exemption was to continue testing the adequacy of the emergency response organization, except to perform these tests during the conduct of drills.
As discussed in the response to Questions 7, 8, and 10 above, the intent of the originally requested exemption was to continue testing the adequacy of the emergency response organization, except to perform these tests during the conduct of drills.
However, the past precedence wording also meets this intent. Therefore, DEK is revising the exemption originally requested from portions of 10 CFR 50, Appendix E, IV.F.2.i in Reference 1, as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.
However, the past precedence wording also meets this intent.
Therefore, DEK is revising the exemption originally requested from portions of 10 CFR 50, Appendix E, IV.F.2.i in Reference 1, as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.
Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.
Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.
NRC Question MF2567-RAII-ORLOB-Norris-012 The Executive Summary in NUREG-1738 states, in part, "the staffs 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 industry decommissioning commitments (IDCs) and staff decommissioning assumptions (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. Please explain if/how KPS meets each of these IDCs and SDAs.
NRC Question MF2567-RAII-ORLOB-Norris-012 The Executive Summary in NUREG-1738 states, in part, "the staffs 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 industry decommissioning commitments (IDCs) and staff decommissioning assumptions (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.
Please explain if/how KPS meets each of these IDCs and SDAs.


Serial No. 13-390A Attachment 1 Page 14 of 22
Serial No. 13-390A Page 14 of 22


===Response===
===Response===
Results of a comparison of the KPS spent fuel pool against the IDCs and SDAs listed in tables 4.1-1 and 4.1-2, respectively, of NUREG-1738, are shown below.
Results of a comparison of the KPS spent fuel pool against the IDCs and SDAs listed in tables 4.1-1 and 4.1-2, respectively, of NUREG-1738, are shown below.
TABLE 1 Industry Decommissioning Commitments (IDCs) Comparison IDC       IDC Description                           KPSIDEK Alignment to IDC Description No.
TABLE 1 Industry Decommissioning Commitments (IDCs) Comparison IDC IDC Description KPSIDEK Alignment to IDC Description No.
: 1. Cask drop analyses will     KPS design aligns with this description. DEK controls the handling of heavy be performed or single     loads within the protected area by an administrative procedure designed to failure-proof cranes will   meet the guidance provided in NUREG-0612. Additionally, the auxiliary be in use for handling of   building crane, which is employed for the lifting and handling of heavy loads heavy loads (i.e., phase II in the vicinity of the spent fuel storage pool, is of a single failure proof of NUREG-0612 will be       design. The NRC Safety Evaluation Report (SER) for KPS License implemented).               Amendment 200 documents the single failure proof design of the auxiliary building crane (SER dated November 20, 2008 (ML082971079)).
: 1. Cask drop analyses will KPS design aligns with this description. DEK controls the handling of heavy be performed or single loads within the protected area by an administrative procedure designed to failure-proof cranes will meet the guidance provided in NUREG-0612. Additionally, the auxiliary be in use for handling of building crane, which is employed for the lifting and handling of heavy loads heavy loads (i.e., phase II in the vicinity of the spent fuel storage pool, is of a single failure proof of NUREG-0612 will be design. The NRC Safety Evaluation Report (SER) for KPS License implemented).
: 2. Procedures and training     DEK practices align with this description. Consistent with KPS Emergency of personnel will be in     Plan requirements, DEK has procedures in place to ensure that onsite and place to ensure that       offsite resources are available and personnel are trained on the access and onsite and offsite         use thereof during an event. The Permanently Defueled Emergency Plan resources can be brought   (PDEP) being submitted for NRC approval addresses these requirements.
Amendment 200 documents the single failure proof design of the auxiliary building crane (SER dated November 20, 2008 (ML082971079)).
to bear during an event. KPS also maintains Letters of Agreement (LOA) or Memoranda of Understanding (MOU) with offsite agencies to ensure additional resources are available if needed. The discussion in KPS/DEK Alignment to IDC Description #4 contains additional details.
: 2. Procedures and training DEK practices align with this description. Consistent with KPS Emergency of personnel will be in Plan requirements, DEK has procedures in place to ensure that onsite and place to ensure that offsite resources are available and personnel are trained on the access and onsite and offsite use thereof during an event. The Permanently Defueled Emergency Plan resources can be brought (PDEP) being submitted for NRC approval addresses these requirements.
: 3. Procedures will be in       DEK practices align with this description. Should severe weather or seismic place to establish         events occur that result in an Emergency Plan entry, procedures are in place communication between       that direct personnel to establish the necessary communications and make onsite and offsite         the appropriate notifications. For example, the Emergency Director (Shift organizations during       Manager) would direct notification of the ERO, applicable State and Counties severe weather and         officials, and the NRC. As decommissioning progresses, DEK anticipates seismic events,             maintaining procedures in place to require the appropriate communication between onsite and offsite organizations.
to bear during an event.
: 4. An offsite resource plan   DEK practices align with this description. Consistent with directions provided will be developed which     in the Emergency Plan, DEK maintains an Emergency Telephone Directory will include access to     (ETD). The ETD provides the information necessary to access necessary portable pumps and         offsite resources in a timely manner. Appropriate station personnel are emergency power to         trained to use the ETD to obtain offsite resources when needed to support supplement onsite           onsite resources.
KPS also maintains Letters of Agreement (LOA) or Memoranda of Understanding (MOU) with offsite agencies to ensure additional resources are available if needed. The discussion in KPS/DEK Alignment to IDC Description #4 contains additional details.
resources. The plan would principally identify ETD subsection ETD 02 lists contacts for government agencies, emergency organizations or suppliers equipment contacts (e.g., for fuel, electrical power, makeup water, where offsite resources     firefighting equipment). It also identifies private agencies that would be could be obtained in a     capable of providing resources if requested (such as INPO, NEI, Bartlett timely manner.             Nuclear, and Point Beach Nuclear Plant).
: 3. Procedures will be in DEK practices align with this description. Should severe weather or seismic place to establish events occur that result in an Emergency Plan entry, procedures are in place communication between that direct personnel to establish the necessary communications and make onsite and offsite the appropriate notifications. For example, the Emergency Director (Shift organizations during Manager) would direct notification of the ERO, applicable State and Counties severe weather and officials, and the NRC. As decommissioning progresses, DEK anticipates seismic events, maintaining procedures in place to require the appropriate communication between onsite and offsite organizations.
: 4. An offsite resource plan DEK practices align with this description. Consistent with directions provided will be developed which in the Emergency Plan, DEK maintains an Emergency Telephone Directory will include access to (ETD). The ETD provides the information necessary to access necessary portable pumps and offsite resources in a timely manner. Appropriate station personnel are emergency power to trained to use the ETD to obtain offsite resources when needed to support supplement onsite onsite resources.
resources. The plan would principally identify ETD subsection ETD 02 lists contacts for government agencies, emergency organizations or suppliers equipment contacts (e.g., for fuel, electrical power, makeup water, where offsite resources firefighting equipment). It also identifies private agencies that would be could be obtained in a capable of providing resources if requested (such as INPO, NEI, Bartlett timely manner.
Nuclear, and Point Beach Nuclear Plant).
ETD subsection ETD 03 has a section specifically for emergency contacts.
ETD subsection ETD 03 has a section specifically for emergency contacts.
It lists items such as a portable diesel driven pump, diesel fuel, construction
It lists items such as a portable diesel driven pump, diesel fuel, construction


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Serial No. 13-390A Page 15 of 22 IOC IDC Description KPSIDEK Alignment to IDC Description No.
and lifting equipment, firefighting equipment, electrical power equipment, compressed gas and air and it identifies the Pooled Equipment Inventory Co., to which DEK subscribes.
and lifting equipment, firefighting equipment, electrical power equipment, compressed gas and air and it identifies the Pooled Equipment Inventory Co., to which DEK subscribes.
As decommissioning progresses, DEK intends to maintain an ETD and necessary subscriptions in place to facilitate the timely acquisition of outside resources if they are needed during Emergency Plan implementation activities.
As decommissioning progresses, DEK intends to maintain an ETD and necessary subscriptions in place to facilitate the timely acquisition of outside resources if they are needed during Emergency Plan implementation activities.
: 5. SFP instrumentation will   The KPS design aligns with the intent of this description. Spent fuel pool include readouts and       (SFP) instrumentation provides alarms in the control room for: Elevated SFP alarms in the control     Area Radiation Level (alert and high level setpoints), SFP Level Hi (two room (or where personnel   channels), SFP Level Lo (two channels), and SFP Temperature Hi (two are stationed) for SFP     channels). Also, the SFP area radiation level reads out in the control room.
: 5.
temperature, water level,   Additionally, the SFP instrumentation provides local readouts of SFP level and area radiation levels, and temperature. A local alarm to notify personnel of high area radiation levels is also in place. Additionally, DEK intends to improve SFP level monitoring instrumentation. After the planned modifications, KPS will have an additional level monitoring capability (radar). This added feature will provide two channels of pool level readout locally (in the SFP heat exchanger room from which mitigating action would be taken) and low level alarms in the control room.
SFP instrumentation will The KPS design aligns with the intent of this description. Spent fuel pool include readouts and (SFP) instrumentation provides alarms in the control room for: Elevated SFP alarms in the control Area Radiation Level (alert and high level setpoints), SFP Level Hi (two room (or where personnel channels), SFP Level Lo (two channels), and SFP Temperature Hi (two are stationed) for SFP channels). Also, the SFP area radiation level reads out in the control room.
: 6. SFP seals that could       The KPS design aligns with this description. The design of the SFP gates is cause leakage leading to   self limiting to prevent draining to a point where the fuel would be uncovered.
temperature, water level, Additionally, the SFP instrumentation provides local readouts of SFP level and area radiation levels, and temperature. A local alarm to notify personnel of high area radiation levels is also in place. Additionally, DEK intends to improve SFP level monitoring instrumentation. After the planned modifications, KPS will have an additional level monitoring capability (radar). This added feature will provide two channels of pool level readout locally (in the SFP heat exchanger room from which mitigating action would be taken) and low level alarms in the control room.
fuel uncovery in the event The relative elevation between the bottom of the SFP gate openings and the of seal failure shall be   top of spent fuel assemblies while stored within the spent fuel storage racks self limiting to leakage or ensures that the inadvertent drainage or leakage via spent fuel gate opening otherwise engineered so     cannot uncover the fuel - i.e., the bottom of the gate opening is above the that drainage cannot       top of stored spent fuel assemblies.
: 6. SFP seals that could The KPS design aligns with this description. The design of the SFP gates is cause leakage leading to self limiting to prevent draining to a point where the fuel would be uncovered.
fuel uncovery in the event The relative elevation between the bottom of the SFP gate openings and the of seal failure shall be top of spent fuel assemblies while stored within the spent fuel storage racks self limiting to leakage or ensures that the inadvertent drainage or leakage via spent fuel gate opening otherwise engineered so cannot uncover the fuel - i.e., the bottom of the gate opening is above the that drainage cannot top of stored spent fuel assemblies.
occur.
occur.
: 7. Procedures or               DEK practices align with this description. DEK employs a formal procedure administrative controls to to allow for pumping down specified volumes within the SFP. Although the reduce the likelihood of   method used does not have the capability to drain down the SFP rapidly, the rapid draindown events     procedure includes the key elements called out in IDC 7. Similarly, the KPS will include (1)           ISFSI equipment design is such that there are no ISFSI-related SFP prohibitions on the use of operations that have the potential to cause a rapid drain down event. The pumps that lack adequate   DEK procedure governing "procedure use and adherence" requires siphon protection or (2)   procedure users to conform to process requirements established to control controls for pump suction   specific activities. Moreover, work activities, whether performed under a and discharge points,       specific procedure designed to control that activity, or under the general The functionality of anti- process controls of the work control process, are subject to the DEK siphon devices will be     integrated risk management procedure, wherein appropriate measures are periodically verified,     employed to assess and manage the risk associated with such activities (e.g., address the affects upon SFP cooling availability).
: 7. Procedures or DEK practices align with this description. DEK employs a formal procedure administrative controls to to allow for pumping down specified volumes within the SFP. Although the reduce the likelihood of method used does not have the capability to drain down the SFP rapidly, the rapid draindown events procedure includes the key elements called out in IDC 7. Similarly, the KPS will include (1)
: 8. An onsite restoration plan DEK practices align with this description. DEK maintains in place will be in place to provide contingency work orders to support the repair and replacement of key SFP repair of the SFP cooling   cooling components. Similarly, the necessary work orders are maintained in systems or to provide       place to support the alignment of a Residual Heat Removal system heat access for makeup water     exchanger to allow for its use to cool the SFP should it be needed.
ISFSI equipment design is such that there are no ISFSI-related SFP prohibitions on the use of operations that have the potential to cause a rapid drain down event. The pumps that lack adequate DEK procedure governing "procedure use and adherence" requires siphon protection or (2) procedure users to conform to process requirements established to control controls for pump suction specific activities. Moreover, work activities, whether performed under a and discharge points, specific procedure designed to control that activity, or under the general The functionality of anti-process controls of the work control process, are subject to the DEK siphon devices will be integrated risk management procedure, wherein appropriate measures are periodically verified, employed to assess and manage the risk associated with such activities (e.g., address the affects upon SFP cooling availability).
to the SFP. The plan will   Additionally, there are procedures in place that provide for the use of a
: 8. An onsite restoration plan DEK practices align with this description. DEK maintains in place will be in place to provide contingency work orders to support the repair and replacement of key SFP repair of the SFP cooling cooling components. Similarly, the necessary work orders are maintained in systems or to provide place to support the alignment of a Residual Heat Removal system heat access for makeup water exchanger to allow for its use to cool the SFP should it be needed.
to the SFP. The plan will Additionally, there are procedures in place that provide for the use of a


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Serial No. 13-390A Page 16 of 22 IDC IDC Description KPSIDEK Alignment to IDC Description No.
provide for remote         backup means (beyond the NRC Safety Guide 13 capabilities required by alignment of the makeup the KPS design basis for fuel pool makeup) of SFP water makeup which can source to the SFP without be executed without requiring entry to the refuel floor.
provide for remote backup means (beyond the NRC Safety Guide 13 capabilities required by alignment of the makeup the KPS design basis for fuel pool makeup) of SFP water makeup which can source to the SFP without be executed without requiring entry to the refuel floor.
requiring entry to the refuel floor.
requiring entry to the refuel floor.
: 9. Procedures will be in       DEK practices align with this description. The KPS SFP design precludes an place to control SFP       operationally induced rapid decrease in SFP inventory. DEK employs a operations that have the   formal procedure to allow for pumping down specified volumes within the potential to rapidly       SFP. Similarly, the KPS ISFSI equipment design is such that there are no decrease SFP inventory. ISFSI-related SFP operations that have the potential to cause a rapid drain These administrative       down event. The DEK process governing "procedure use and adherence" controls may require       requires procedure users to conform to process requirements established to additional operations or   control specific activities. Moreover, work activities, whether performed management review,         under a specific procedure designed to control that activity, or under the management physical         general process controls of the work control process, are subject to the DEK presence for designated     integrated risk management procedure wherein appropriate measures are operations or               employed to assess and manage the risk associated with such activities. For administrative limitations example, the integrated risk management procedure requires DEK such as restrictions on     management to consider requiring direct supervisory oversight or heavy load movements,       engineered mitigation methods. Also, the DEK "Decommissioning Safety Assessment Checklist" provides guidance relative to heavy loads at the station. Finally, implementing procedures related to the use of cranes at KPS provide the needed restrictions upon the handling of heavy loads, with specific requirements related to the handling of such loads in the vicinity of the SFP.
: 9. Procedures will be in DEK practices align with this description. The KPS SFP design precludes an place to control SFP operationally induced rapid decrease in SFP inventory. DEK employs a operations that have the formal procedure to allow for pumping down specified volumes within the potential to rapidly SFP. Similarly, the KPS ISFSI equipment design is such that there are no decrease SFP inventory.
: 10. Routine testing of the     DEK practices align with this description. The KPS design basis credits the alternative fuel pool       station's seismically designed service water (SW) system for meeting the makeup system               NRC Safety Guide 13 requirements for SFP makeup. The SW system has components will be         redundant pumping capability and redundant power supplies adequate to performed and               support the SFP makeup function. The station SW system is continuously administrative controls for operating, allowing for continuous monitoring for proper operation, and equipment out of service   provides pressurized water from Lake Michigan to the inlet of a single will be implemented to     manual isolation valve that can be repositioned locally to supply makeup provide added assurance     water to the SFP. For defense-in-depth, the station also has available, with that the components         supporting procedures for its use, an engine driven emergency makeup would be Available, if     pump capable of delivering Lake Michigan water to the SFP. By procedure, needed.                     both the design basis and defense-in-depth capabilities are routinely tested to ensure their ongoing availability. The process for testing the design basis makeup water supply isolation valve includes an explicit step requiring that a test failure be addressed via the corrective action program. This complies with the corrective action program's general requirement that plant System, Structure, or Component (SSC) failures be entered into the program.
ISFSI-related SFP operations that have the potential to cause a rapid drain These administrative down event. The DEK process governing "procedure use and adherence" controls may require requires procedure users to conform to process requirements established to additional operations or control specific activities. Moreover, work activities, whether performed management review, under a specific procedure designed to control that activity, or under the management physical general process controls of the work control process, are subject to the DEK presence for designated integrated risk management procedure wherein appropriate measures are operations or employed to assess and manage the risk associated with such activities. For administrative limitations example, the integrated risk management procedure requires DEK such as restrictions on management to consider requiring direct supervisory oversight or heavy load movements, engineered mitigation methods. Also, the DEK "Decommissioning Safety Assessment Checklist" provides guidance relative to heavy loads at the station. Finally, implementing procedures related to the use of cranes at KPS provide the needed restrictions upon the handling of heavy loads, with specific requirements related to the handling of such loads in the vicinity of the SFP.
: 10. Routine testing of the DEK practices align with this description. The KPS design basis credits the alternative fuel pool station's seismically designed service water (SW) system for meeting the makeup system NRC Safety Guide 13 requirements for SFP makeup. The SW system has components will be redundant pumping capability and redundant power supplies adequate to performed and support the SFP makeup function. The station SW system is continuously administrative controls for operating, allowing for continuous monitoring for proper operation, and equipment out of service provides pressurized water from Lake Michigan to the inlet of a single will be implemented to manual isolation valve that can be repositioned locally to supply makeup provide added assurance water to the SFP. For defense-in-depth, the station also has available, with that the components supporting procedures for its use, an engine driven emergency makeup would be Available, if pump capable of delivering Lake Michigan water to the SFP. By procedure, needed.
both the design basis and defense-in-depth capabilities are routinely tested to ensure their ongoing availability. The process for testing the design basis makeup water supply isolation valve includes an explicit step requiring that a test failure be addressed via the corrective action program. This complies with the corrective action program's general requirement that plant System, Structure, or Component (SSC) failures be entered into the program.


Serial No. 13-390A Attachment 1 Page 17 of 22 TABLE 2 Staff Decommissioning Assumptions (SDAs) Comparison SDA       SDA Description                         KPS/DEK Alignment to SDA Description No.
Serial No. 13-390A Page 17 of 22 TABLE 2 Staff Decommissioning Assumptions (SDAs) Comparison SDA SDA Description KPS/DEK Alignment to SDA Description No.
I. Licensee's SFP cooling     KPS design aligns with the intent of this description. The KPS SFP design will be at least as Cooling System design is based, in part, on NRC Safety Guide 13. Safety capable as that assumed     Guide 13 requires a seismic category 1 system for providing makeup in the risk assessment,     water to the SFP. This design basis requirement for SFP cooling is including instrumentation. provided by the SW system, which is a Nuclear Safety Design Class 1*
I.
Licensees will have at     system (i.e., it is designed to withstand design basis earthquake least one motor-driven and seismically induced load) protected by a Nuclear Safety Design Class I one diesel-driven fire     structure. The SW system has redundant pumping capability and is pump capable of delivering provided with redundant power supplies adequate to provide SFP makeup inventory to the SFP.       at the required capacity.
Licensee's SFP cooling KPS design aligns with the intent of this description. The KPS SFP design will be at least as Cooling System design is based, in part, on NRC Safety Guide 13. Safety capable as that assumed Guide 13 requires a seismic category 1 system for providing makeup in the risk assessment, water to the SFP. This design basis requirement for SFP cooling is including instrumentation.
provided by the SW system, which is a Nuclear Safety Design Class 1*
Licensees will have at system (i.e., it is designed to withstand design basis earthquake least one motor-driven and seismically induced load) protected by a Nuclear Safety Design Class I one diesel-driven fire structure. The SW system has redundant pumping capability and is pump capable of delivering provided with redundant power supplies adequate to provide SFP makeup inventory to the SFP.
at the required capacity.
The SW pumps are normally powered from offsite power, but can be supplied with backup power from the emergency diesel generators (EDGs), which are also Class 1 components housed in a Class 1 structure.
The SW pumps are normally powered from offsite power, but can be supplied with backup power from the emergency diesel generators (EDGs), which are also Class 1 components housed in a Class 1 structure.
The station design also includes two motor driven fire pumps, each with the ability to be powered from either off site power or from either of two EDGs. The fire pumps have the capability to deliver water to the SFP for makeup.
The station design also includes two motor driven fire pumps, each with the ability to be powered from either off site power or from either of two EDGs. The fire pumps have the capability to deliver water to the SFP for makeup.
Finally, the station also maintains available, with supporting procedures for its use, a diesel engine powered emergency makeup pump capable of supplying water from Lake Michigan to the SFP.
Finally, the station also maintains available, with supporting procedures for its use, a diesel engine powered emergency makeup pump capable of supplying water from Lake Michigan to the SFP.
: 2. Walk-downs of SFP           DEK practices align with the intent of this description. Station procedures systems will be performed   require a member of the staff to tour the SFP area each shift. Proper at least once per shift by system operation is verified once per shift by verifying and recording the operators.             normal SFP level and temperature. Additional verifications are performed daily by direct observation of the proper operation and status of SFP Procedures will be         pumps. The normal and alternate SFP makeup water sources (including developed for and           related tank levels) are also verified daily to ensure that they remain employed by the operators available.
: 2.
to provide guidance on the capability and availability Station procedures require continual verification that the SW system of onsite and offsite       (design basis makeup source) is in operation; thereby ensuring that SFP inventory makeup sources emergency makeup is available.
Walk-downs of SFP DEK practices align with the intent of this description. Station procedures systems will be performed require a member of the staff to tour the SFP area each shift. Proper at least once per shift by system operation is verified once per shift by verifying and recording the operators.
and time available to initiate these sources for Moreover, the KPS guidance in place for "Recovery Plan for Catastrophic various loss of cooling or Event" includes instructions for the following specific actions/events:
normal SFP level and temperature. Additional verifications are performed daily by direct observation of the proper operation and status of SFP Procedures will be pumps. The normal and alternate SFP makeup water sources (including developed for and related tank levels) are also verified daily to ensure that they remain employed by the operators available.
to provide guidance on the capability and availability Station procedures require continual verification that the SW system of onsite and offsite (design basis makeup source) is in operation; thereby ensuring that SFP inventory makeup sources emergency makeup is available.
and time available to initiate these sources for Moreover, the KPS guidance in place for "Recovery Plan for Catastrophic various loss of cooling or Event" includes instructions for the following specific actions/events:
inventory events.
inventory events.
* SEP makeup
SEP makeup Alternate method to ventilate the SFP area SFP leakage control Additional resources Finally, KPS procedure "Validation of Time Sensitive Operator Actions,"
* Alternate method to ventilate the SFP area
* SFP leakage control
                                  " Additional resources Finally, KPS procedure "Validation of Time Sensitive Operator Actions,"
provides the guidance necessary to ensure that operators have sufficient
provides the guidance necessary to ensure that operators have sufficient


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Serial No. 13-390A Page 18 of 22 SDA SDA Description KPS/DEK Alignment to SDA Description No.
time available to initiate makeup water sources for various loss of cooling or inventory events.
time available to initiate makeup water sources for various loss of cooling or inventory events.
: 3. Control room                   KPS design aligns with the intent of this description. Item 7 in KPS instrumentation that            Updated Safety Analysis Report (USAR), Table 9.5-2, "Design monitors SFP temperature        Conformance with Safety Guide 13, states that "Level measuring and water level will directly  instrumentation and radiation monitoring equipment are provided which measure the parameters          alarm both locally and in the Control Room." Additionally, DEK has involved.                      initiated actions to enhance SFP level monitoring instrumentation. These Level instrumentation will      enhancements are intended to provide an additional level monitoring provide alarms at levels        capability (radar). The planned level instrumentation includes two associated with calling in      channels that provide local pool level indication (in the SFP heat offsite resources and with      exchanger room, from which mitigating action would be taken). Each level declaring a general            channel also provides input to a low level alarm in the Control Room. DEK emergency.                      processes in place to respond to an abnormally low level in the SFP direct the plant staff to take appropriate actions to provide SFP makeup, first through normal means, then by utilizing available onsite resources, including both design basis and defense-in-depth capabilities. Ultimately, if the use of onsite means fails to restore SFP inventory to an acceptable level, processes would direct the plant staff to access offsite resources.
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.
KPS design aligns with the intent of this description. Item 7 in KPS Updated Safety Analysis Report (USAR), Table 9.5-2, "Design Conformance with Safety Guide 13, states that "Level measuring instrumentation and radiation monitoring equipment are provided which alarm both locally and in the Control Room." Additionally, DEK has initiated actions to enhance SFP level monitoring instrumentation. These enhancements are intended to provide an additional level monitoring capability (radar). The planned level instrumentation includes two channels that provide local pool level indication (in the SFP heat exchanger room, from which mitigating action would be taken). Each level channel also provides input to a low level alarm in the Control Room. DEK processes in place to respond to an abnormally low level in the SFP direct the plant staff to take appropriate actions to provide SFP makeup, first through normal means, then by utilizing available onsite resources, including both design basis and defense-in-depth capabilities. Ultimately, if the use of onsite means fails to restore SFP inventory to an acceptable level, processes would direct the plant staff to access offsite resources.
To facilitate accessing offsite resources, DEK maintains an Emergency Telephone Directory (ETD). The ETD provides the information necessary to access necessary offsite resources in a timely manner. Appropriate station personnel are trained to use the ETD to obtain offsite resources when needed to support onsite resources. ETD subsection ETD 02 lists contacts for government agencies, emergency equipment contacts (e.g.,
To facilitate accessing offsite resources, DEK maintains an Emergency Telephone Directory (ETD). The ETD provides the information necessary to access necessary offsite resources in a timely manner. Appropriate station personnel are trained to use the ETD to obtain offsite resources when needed to support onsite resources. ETD subsection ETD 02 lists contacts for government agencies, emergency equipment contacts (e.g.,
for fuel, electrical power, makeup water, firefighting equipment). It also identifies private agencies that would be capable of bringing resources when needed such as INPO, NEI, Bartlett Nuclear, and Point Beach Nuclear Plant. ETD subsection ETD 03 has a section specifically for emergency contacts. It lists items such as a portable diesel driven pump, diesel fuel, construction and lifting equipment, firefighting equipment, electrical power equipment, compressed gas and air, and it identifies the Pooled Equipment Inventory Co., to which DEK subscribes.
for fuel, electrical power, makeup water, firefighting equipment). It also identifies private agencies that would be capable of bringing resources when needed such as INPO, NEI, Bartlett Nuclear, and Point Beach Nuclear Plant. ETD subsection ETD 03 has a section specifically for emergency contacts. It lists items such as a portable diesel driven pump, diesel fuel, construction and lifting equipment, firefighting equipment, electrical power equipment, compressed gas and air, and it identifies the Pooled Equipment Inventory Co., to which DEK subscribes.
Regarding the declaration of a general emergency, KPS will be employing Shutdown EALs using an approved NRC EAL Scheme. Consistent with that scheme, there are no conditions that have the capacity to reach any threshold requiring the declaration of a general emergency.
Regarding the declaration of a general emergency, KPS will be employing Shutdown EALs using an approved NRC EAL Scheme. Consistent with that scheme, there are no conditions that have the capacity to reach any threshold requiring the declaration of a general emergency.
    -I.                               I
-I.
: 4.     Licensee determines that       KPS design aligns with this description. Nominal SFP water level is there are no drain paths in    approximately 27 feet above the stored fuel. As stated in KPS USAR, the SFP that could lower        Section 9.3.2.3, "The SFP pump suction lines are located well above the the pool level (by draining,    fuel assemblies and a system failure cannot result in loss of pool water.
I 4.
suction, or pumping) more      The return lines enter the pool above the top of the fuel assemblies and than 15 feet below the          the lines contain check valves at the point of entry into the pool shielding normal pool operating level    concrete. Thus, line failure outside of the SFP cannot cause a loss of pool and that licensee must          water due to siphon action." The SSCs relied upon to preserve the SFP initiate recovery using        inventory are seismically designed SSCs that preclude any significant loss offsite sources.                in fuel pool inventory, via the SFP cooling water suction lines, under design basis conditions. Additionally, the SFP cooling water return lines
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.
KPS design aligns with this description. Nominal SFP water level is approximately 27 feet above the stored fuel. As stated in KPS USAR, Section 9.3.2.3, "The SFP pump suction lines are located well above the fuel assemblies and a system failure cannot result in loss of pool water.
The return lines enter the pool above the top of the fuel assemblies and the lines contain check valves at the point of entry into the pool shielding concrete. Thus, line failure outside of the SFP cannot cause a loss of pool water due to siphon action." The SSCs relied upon to preserve the SFP inventory are seismically designed SSCs that preclude any significant loss in fuel pool inventory, via the SFP cooling water suction lines, under design basis conditions. Additionally, the SFP cooling water return lines


Serial No. 13-390A Attachment 1 Page 19 of 22 SDA       SDA Description                             KPS/DEK Alignment to SDA Description No.
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(also seismically qualified) terminate approximately 15 feet above the stored fuel. This limits any reduction in SFP level via the return lines to approximately 12 feet.
(also seismically qualified) terminate approximately 15 feet above the stored fuel. This limits any reduction in SFP level via the return lines to approximately 12 feet.
: 5. Load Drop consequence           KPS design aligns with this description. The auxiliary building crane, which analyses will be performed     is employed for the lifting and handling of heavy loads in the vicinity of the for facilities with non single spent fuel storage pool, is of a single failure proof design. The NRC failure-proof systems. The     Safety Evaluation Report (SER) for KPS License Amendment 200 analyses and any               documents the single failure proof design of the auxiliary building crane mitigative actions             (SER dated November 20, 2008 (ML082971079)). Additionally, DEK necessary to preclude           controls the handling of heavy loads within the protected area by an catastrophic damage to         administrative procedure designed to meet the guidance provided in the SFP that would lead to     NUREG-0612.
: 5.
a rapid pool draining would be sufficient to               Control of heavy loads is governed by the Technical Requirements Manual demonstrate that there is       (TRM), specifically, TRM 8.9.1, "Spent Fuel Pool - Control of Heavy high confidence in the         Loads," which is subject to the requirements of 10 CFR 50.59.
Load Drop consequence KPS design aligns with this description. The auxiliary building crane, which analyses will be performed is employed for the lifting and handling of heavy loads in the vicinity of the for facilities with non single spent fuel storage pool, is of a single failure proof design. The NRC failure-proof systems. The Safety Evaluation Report (SER) for KPS License Amendment 200 analyses and any documents the single failure proof design of the auxiliary building crane mitigative actions (SER dated November 20, 2008 (ML082971079)). Additionally, DEK necessary to preclude controls the handling of heavy loads within the protected area by an catastrophic damage to administrative procedure designed to meet the guidance provided in the SFP that would lead to NUREG-0612.
a rapid pool draining would be sufficient to Control of heavy loads is governed by the Technical Requirements Manual demonstrate that there is (TRM), specifically, TRM 8.9.1, "Spent Fuel Pool - Control of Heavy high confidence in the Loads," which is subject to the requirements of 10 CFR 50.59.
facilities ability to withstand a heavy load drop.
facilities ability to withstand a heavy load drop.
: 6. Each decommissioning           Item 10 of the seismic checklist provides an alternative wherein the plant will successfully         licensee delays request for a licensing "waiver" (i.e. License Amendment complete the seismic           Request) for Emergency Planning until the plant specific zirconium fire is checklist provided in           no longer a credible concern. DEK has performed an analysis (Calculation Appendix 2B to this study.     2013-11284, "Maximum Cladding and Fuel Temperature Analysis for If the checklist cannot be     Uncovered Spent Fuel Pool"), which concludes that, about 17 months after successfully completed,         reactor shutdown, decay heat cannot raise the spent fuel cladding the decommissioning plant       temperature sufficiently to cause clad failure (565°C) if all water is drained will perform a plant specific   from the SFP. Therefore, as of October 2014, when the requested seismic risk assessment of     changes will be implemented, the plant specific zirconium fire will no the SFP and demonstrate         longer be a credible concern.
: 6.
that SFP seismically induced structural failure     Additionally, KPS is located in a geologically stable region whose seismic and rapid loss of inventory     hazard risk is very low as documented in recent seismic hazard estimates is less than the generic       (based on U.S. Geological Survey (USGS) of 2008). Geologic bounding estimates             investigations throughout the Lake Michigan basin have not found any provided in this study (<1     indication of fault movement in the recent geologic past. As shown in x10 5 per year including       Figure 2 of Generic Issue 199 (GI-1 99, August 2010), the peak horizontal non-seismic events),           acceleration (%g) for 2-percent probability of exceedance in 50 years, for the geographic region where KPS is located, is in the second lowest region of the conterminous United States (between 0.02 and 0.03 g).
Each decommissioning Item 10 of the seismic checklist provides an alternative wherein the plant will successfully licensee delays request for a licensing "waiver" (i.e. License Amendment complete the seismic Request) for Emergency Planning until the plant specific zirconium fire is checklist provided in no longer a credible concern. DEK has performed an analysis (Calculation Appendix 2B to this study.
2013-11284, "Maximum Cladding and Fuel Temperature Analysis for If the checklist cannot be Uncovered Spent Fuel Pool"), which concludes that, about 17 months after successfully completed, reactor shutdown, decay heat cannot raise the spent fuel cladding the decommissioning plant temperature sufficiently to cause clad failure (565&deg;C) if all water is drained will perform a plant specific from the SFP. Therefore, as of October 2014, when the requested seismic risk assessment of changes will be implemented, the plant specific zirconium fire will no the SFP and demonstrate longer be a credible concern.
that SFP seismically induced structural failure Additionally, KPS is located in a geologically stable region whose seismic and rapid loss of inventory hazard risk is very low as documented in recent seismic hazard estimates is less than the generic (based on U.S. Geological Survey (USGS) of 2008). Geologic bounding estimates investigations throughout the Lake Michigan basin have not found any provided in this study (<1 indication of fault movement in the recent geologic past. As shown in x105 per year including Figure 2 of Generic Issue 199 (GI-1 99, August 2010), the peak horizontal non-seismic events),
acceleration (%g) for 2-percent probability of exceedance in 50 years, for the geographic region where KPS is located, is in the second lowest region of the conterminous United States (between 0.02 and 0.03 g).
Based on existing knowledge of the SFP structural capabilities, the SFP seismically induced structural failure and rapid loss of SFP inventory is very unlikely.
Based on existing knowledge of the SFP structural capabilities, the SFP seismically induced structural failure and rapid loss of SFP inventory is very unlikely.
Finally, KPS has procedures in place to ensure successful implementation of mitigation measures to supply alternate cooling water using portable equipment. As a result, no radiological releases with offsite consequence is expected or should occur following a severe earthquake because KPS has been permanently shutdown since May 2013, and mitigation measures for cooling water are in place.
Finally, KPS has procedures in place to ensure successful implementation of mitigation measures to supply alternate cooling water using portable equipment. As a result, no radiological releases with offsite consequence is expected or should occur following a severe earthquake because KPS has been permanently shutdown since May 2013, and mitigation measures for cooling water are in place.


Serial No. 13-390A Attachment 1 Page 20 of 22 SDA         SDA Description                       KPS/DEK Alignment to SDA Description No.
Serial No. 13-390A Page 20 of 22 SDA SDA Description KPS/DEK Alignment to SDA Description No.
: 7. Licensees will maintain a DEK procedures align with this description. KPS does not utilize Boraflex program to provide       in any of its spent fuel storage racks. Rather, the KPS design employs surveillance and         boron carbide (B4C) in the spent fuel storage racks in the north and south monitoring of Boraflex in pools and Boral neutron absorber material in the north canal spent fuel high-density spent fuel   storage racks.
: 7.
Licensees will maintain a DEK procedures align with this description. KPS does not utilize Boraflex program to provide in any of its spent fuel storage racks. Rather, the KPS design employs surveillance and boron carbide (B4C) in the spent fuel storage racks in the north and south monitoring of Boraflex in pools and Boral neutron absorber material in the north canal spent fuel high-density spent fuel storage racks.
racks until such time as spent fuel is no longer stored in these high-density racks.
racks until such time as spent fuel is no longer stored in these high-density racks.
NRC Question MF2567-RAII-ORLOB-Norris-013 Page 43 of 55 references a site-specific adiabatic heat up to address a partial drain down of the SFP (identified as Reference 12), assuming no air-cooling it states that the time necessary for the hottest fuel assembly to reach the critical temperature of 5650C is six hours after the fuel rods have become uncovered. Some previous exemptions were granted based [on] time to reach the cladding auto-ignition temperature of 900 0 C.
NRC Question MF2567-RAII-ORLOB-Norris-013 Page 43 of 55 references a site-specific adiabatic heat up to address a partial drain down of the SFP (identified as Reference 12), assuming no air-cooling it states that the time necessary for the hottest fuel assembly to reach the critical temperature of 5650C is six hours after the fuel rods have become uncovered. Some previous exemptions were 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.
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.


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A copy of this site-specific adiabatic heat up analysis to address a partial drain down of the SFP, assuming no air-cooling, is provided in Enclosure 1 to this submittal (Calculation 2013-07050, Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with No Air Cooling). This site-specific quantitative heat up analysis includes both the time to reach 5650C and the time to reach 9000C (the previous analysis for the heat up to 5650C, identified as Reference 12 in the exemption request, was a qualitative analysis whose results have been subsumed into the current analysis).
A copy of this site-specific adiabatic heat up analysis to address a partial drain down of the SFP, assuming no air-cooling, is provided in Enclosure 1 to this submittal (Calculation 2013-07050, Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with No Air Cooling). This site-specific quantitative heat up analysis includes both the time to reach 5650C and the time to reach 9000C (the previous analysis for the heat up to 5650C, identified as Reference 12 in the exemption request, was a qualitative analysis whose results have been subsumed into the current analysis).


Serial No. 13-390A Attachment 1 Page 21 of 22 NRC Question MF2567-RAII-ORLOB-Norris-014 Page 43 of 55 references spent fuel pool inventory makeup strategies. Please provide additional information related to:
Serial No. 13-390A Page 21 of 22 NRC Question MF2567-RAII-ORLOB-Norris-014 Page 43 of 55 references spent fuel pool inventory makeup strategies. Please provide additional information related to:
: a. What is the availability of trained personnel to perform the required actions?
: a. What is the availability of trained personnel to perform the required actions?
: b. How is the referenced equipment maintained and tested?
: b. How is the referenced equipment maintained and tested?
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===Response===
===Response===
: a. Availability of trained personnel to perform the required actions The on-shift Plant Operators and Fire Brigade members are appropriately trained on the various actions needed to provide makeup to the spent fuel pool (SFP) based on a systematic approach to training. Because KPS is no longer operating, maintaining SFP cooling and inventory would be the highest priority activity; therefore, the personnel needed to perform these actions are available at all times.
: a. Availability of trained personnel to perform the required actions The on-shift Plant Operators and Fire Brigade members are appropriately trained on the various actions needed to provide makeup to the spent fuel pool (SFP) based on a systematic approach to training. Because KPS is no longer operating, maintaining SFP cooling and inventory would be the highest priority activity; therefore, the personnel needed to perform these actions are available at all times.
: b. Referenced equipment maintained and tested Existing plant systems used for SFP makeup are maintained and tested using plant procedures in accordance with the KPS preventive maintenance program. This includes testing the capability to align emergency makeup via the service water system. The diesel-driven portable pump is maintained and tested using plant procedures in accordance with the KPS preventative maintenance program. In addition, flow testing of external makeup capacity is performed periodically to validate that the specified actions can be completed in a timely manner.
: b. Referenced equipment maintained and tested Existing plant systems used for SFP makeup are maintained and tested using plant procedures in accordance with the KPS preventive maintenance program.
This includes testing the capability to align emergency makeup via the service water system. The diesel-driven portable pump is maintained and tested using plant procedures in accordance with the KPS preventative maintenance program.
In addition, flow testing of external makeup capacity is performed periodically to validate that the specified actions can be completed in a timely manner.
: c. Procedures developed to perform this task and how they are controlled Operating procedures (NOP-SFP-001, "Spent Fuel Pool Cooling and Cleanup System" and AOP-SFP-001, "Abnormal Spent Fuel Pool Cooling and Cleanup System Operation") provide direction for supplying makeup water to the spent fuel pool using existing plant systems in the event of a loss of level. If these procedurally directed strategies do not result in restoration of level, then response plans (which are in place to address large area fires) would be implemented which direct personnel to provide external makeup water via a portable diesel-driven pump.
: c. Procedures developed to perform this task and how they are controlled Operating procedures (NOP-SFP-001, "Spent Fuel Pool Cooling and Cleanup System" and AOP-SFP-001, "Abnormal Spent Fuel Pool Cooling and Cleanup System Operation") provide direction for supplying makeup water to the spent fuel pool using existing plant systems in the event of a loss of level. If these procedurally directed strategies do not result in restoration of level, then response plans (which are in place to address large area fires) would be implemented which direct personnel to provide external makeup water via a portable diesel-driven pump.
Administrative controls (GNP-03.01.01, "Directive, Implementing Document, and Procedure Administrative Controls") are in place to ensure that procedures are
Administrative controls (GNP-03.01.01, "Directive, Implementing Document, and Procedure Administrative Controls") are in place to ensure that procedures are


Serial No. 13-390A Attachment 1 Page 22 of 22 maintained and implemented, and that any changes to them are appropriately reviewed and approved (including any applicable requirements of 10 CFR 50.59).
Serial No. 13-390A Page 22 of 22 maintained and implemented, and that any changes to them are appropriately reviewed and approved (including any applicable requirements of 10 CFR 50.59).
: d. Referencing these procedures and equipment in the emergency plan These procedures and equipment are not specifically referenced in the existing KPS Emergency Plan and are not included in the planned Permanently Defueled Emergency Plan (to be submitted for NRC approval). These procedures are required by TS 5.4.1.a, which directs establishing, implementing, and maintaining applicable procedures recommended in RG 1.33, Revision 2, Appendix A.
: d. Referencing these procedures and equipment in the emergency plan These procedures and equipment are not specifically referenced in the existing KPS Emergency Plan and are not included in the planned Permanently Defueled Emergency Plan (to be submitted for NRC approval).
These procedures are required by TS 5.4.1.a, which directs establishing, implementing, and maintaining applicable procedures recommended in RG 1.33, Revision 2, Appendix A.
Therefore, it is not necessary for them to be specifically referenced in the Emergency Plan. Equipment requirements are specified in the pertinent procedures.
Therefore, it is not necessary for them to be specifically referenced in the Emergency Plan. Equipment requirements are specified in the pertinent procedures.
References
References
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10 CFR 50.47(c)(2), AND 10 CFR 50, APPENDIX E, SECTION IV KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.
10 CFR 50.47(c)(2), AND 10 CFR 50, APPENDIX E, SECTION IV KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.


Serial No. 13-390A Attachment 2 Page 1 of 10 Supplement 1:
Serial No. 13-390A Page 1 of 10 Supplement 1:
Request for Exemptions from Portions of 10 CFR 50.47(b),
Request for Exemptions from Portions of 10 CFR 50.47(b),
10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV I. DESCRIPTION By application dated July 31, 2013 (Reference 1), Dominion Energy Kewaunee, Inc.
10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV I.
DESCRIPTION By application dated July 31, 2013 (Reference 1), Dominion Energy Kewaunee, Inc.
(DEK) requested exemptions, pursuant to 10 CFR 50.12 "Specific exemptions," from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV, for Kewaunee Power Station (KPS). The requested exemptions would allow DEK to reduce emergency planning requirements and subsequently revise the KPS Emergency Plan consistent with the permanently defueled condition of the station.
(DEK) requested exemptions, pursuant to 10 CFR 50.12 "Specific exemptions," from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV, for Kewaunee Power Station (KPS). The requested exemptions would allow DEK to reduce emergency planning requirements and subsequently revise the KPS Emergency Plan consistent with the permanently defueled condition of the station.
In response to the staff's comments, DEK is revising the originally proposed exemption request. Attachment 2 to this letter provides a supplement to the proposed exemption request. The analyses provided in Reference 1 remain applicable and bounding to this revised request. The conclusions of the no significant hazards consideration and the environmental considerations contained in Reference 1 are not affected by, and remain applicable to, this revised request.
In response to the staff's comments, DEK is revising the originally proposed exemption request. Attachment 2 to this letter provides a supplement to the proposed exemption request. The analyses provided in Reference 1 remain applicable and bounding to this revised request. The conclusions of the no significant hazards consideration and the environmental considerations contained in Reference 1 are not affected by, and remain applicable to, this revised request.
A. Revised Exemptions Requested from 10 CFR 50.47 Table 1 (Revised) below lists the pertinent portions of 10 CFR 50.47(b) and 10 CFR 50.47(c)(2) in the left column. The specific portion of the requirement within the regulation from which exemption is being requested is emphasized (bold/underlined).
A. Revised Exemptions Requested from 10 CFR 50.47 Table 1 (Revised) below lists the pertinent portions of 10 CFR 50.47(b) and 10 CFR 50.47(c)(2) in the left column.
The specific portion of the requirement within the regulation from which exemption is being requested is emphasized (bold/underlined).
The basis for the exemption from the specific portion of each requirement is provided in the corresponding row of the column on the right.
The basis for the exemption from the specific portion of each requirement is provided in the corresponding row of the column on the right.
The table below shows only the regulations for which a revision to the originally requested exemption (Reference 1) is being proposed. The rows shown in Table 1 (Revised), below, replace the corresponding rows listed in Table 1 of Reference 1 in their entirety. The requested exemptions from all other regulations shown in Table 1 of Reference 1 remains in effect.
The table below shows only the regulations for which a revision to the originally requested exemption (Reference 1) is being proposed. The rows shown in Table 1 (Revised), below, replace the corresponding rows listed in Table 1 of Reference 1 in their entirety. The requested exemptions from all other regulations shown in Table 1 of Reference 1 remains in effect.


Serial No. 13-390A Attachment 2 Page 2 of 10 TABLE 1 (Revised)
Serial No. 13-390A Page 2 of 10 TABLE 1 (Revised)
Exemptions Requested from 10 CFR 50.47 Regulation                           Basis for Requested Exemption (portion being exempted shown emphasized) 10 CFR 50.47(b)(1) - Primary responsibilities for   Revised       radiological   analyses have been emergency response by the nuclear facility licensee developed that show that, 90 days after shutdown, and by State and local organizations within the     the radiological consequences of design basis Emergency Planning Zones have been assigned,         accidents will not exceed the limits of the EPA the emergency responsibilities of the various       Protective Action Guides at the EAB. In addition, supporting organizations have been specifically     analyses have been developed for beyond design established,       and   each   principal response basis events related to the spent fuel pool which organization has staff to respond and to augment its show that, within 17 months after shutdown, the initial response on a continuous basis.             analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.
Exemptions Requested from 10 CFR 50.47 Regulation Basis for Requested Exemption (portion being exempted shown emphasized) 10 CFR 50.47(b)(1) - Primary responsibilities for Revised radiological analyses have been emergency response by the nuclear facility licensee developed that show that, 90 days after shutdown, and by State and local organizations within the the radiological consequences of design basis Emergency Planning Zones have been assigned, accidents will not exceed the limits of the EPA the emergency responsibilities of the various Protective Action Guides at the EAB. In addition, supporting organizations have been specifically analyses have been developed for beyond design established, and each principal response basis events related to the spent fuel pool which organization has staff to respond and to augment its show that, within 17 months after shutdown, the initial response on a continuous basis.
Therefore, there will no longer be a need for Emergency Planning Zones.           State and local government agency response will be in accordance with each agency's plans and procedures, and commensurate with the hazard posed by the emergency.
analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.
10 CFR 50.47(b)(7) - Information is made             Revised       radiological   analyses have been available to the public on a periodic basis on       developed that show that, 90 days after shutdown, how they will be notified and what their initial     the radiological consequences of design basis actions should be in an emergency (e.g.,             accidents will not exceed the limits of the EPA listening to a local broadcast station and           Protective Action Guides at the EAB. In addition, remaining indoors), the principal points of contact analyses have been developed for beyond design with the news media for dissemination of information basis events related to the spent fuel pool which during an emergency (including the Physical         show that, within 17 months after shutdown, the location or locations) are established in advance,   analyzed event is either not credible, is capable of and procedures for coordinated dissemination of     being mitigated, or the event's radiological information to the public are established,           consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. There will be no need for the public to take any protective actions in the event of an emergency at KPS. Therefore, there will no longer be any need for information to be made available to the public about how they will be notified and what their initial protective actions should be.
Therefore, there will no longer be a need for Emergency Planning Zones.
State and local government agency response will be in accordance with each agency's plans and procedures, and commensurate with the hazard posed by the emergency.
10 CFR 50.47(b)(7)
Information is made Revised radiological analyses have been available to the public on a periodic basis on developed that show that, 90 days after shutdown, how they will be notified and what their initial the radiological consequences of design basis actions should be in an emergency (e.g.,
accidents will not exceed the limits of the EPA listening to a local broadcast station and Protective Action Guides at the EAB. In addition, remaining indoors), the principal points of contact analyses have been developed for beyond design with the news media for dissemination of information basis events related to the spent fuel pool which during an emergency (including the Physical show that, within 17 months after shutdown, the location or locations) are established in advance, analyzed event is either not credible, is capable of and procedures for coordinated dissemination of being mitigated, or the event's radiological information to the public are established, consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. There will be no need for the public to take any protective actions in the event of an emergency at KPS. Therefore, there will no longer be any need for information to be made available to the public about how they will be notified and what their initial protective actions should be.


Serial No. 13-390A Attachment 2 Page 3 of 10 Regulation                           Basis for Requested Exemption (portion being exempted shown emphasized) 10 CFR 50.47(b)(9) - Adequate methods, systems,     Revised    radiological  analyses    have been and equipment for assessing and monitoring actual   developed that show that, 90 days after shutdown, or potential offsite consequences of a radiological  the radiological consequences of design basis emergency condition are in use.                      accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.
Serial No. 13-390A Page 3 of 10 Regulation (portion being exempted shown emphasized)
Therefore, assessing and monitoring of offsite consequences       of   radiological   emergency conditions will no longer be required.
Basis for Requested Exemption 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.
Revised radiological analyses have been developed that show that, 90 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.
Therefore, assessing and monitoring of offsite consequences of radiological emergency conditions will no longer be required.
Since a need for monitoring and assessing will no longer exist, DEK no longer intends to maintain the capability to deploy field teams for assessing and monitoring offsite radiological conditions.
Since a need for monitoring and assessing will no longer exist, DEK no longer intends to maintain the capability to deploy field teams for assessing and monitoring offsite radiological conditions.
10 CFR 50.47(b)(14) - Periodic exercises are (will   No exemption from 10 CFR 50.47(b)(14) is being be) conducted to evaluate major portions of         proposed.
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.
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 from 10 CFR 50.47(b)(14) is being proposed.
B. Exemptions Requested from 10 CFR 50, Appendix E Table 2 (Revised) below lists the pertinent portions of 10 CFR 50, Appendix E, Section IV, in the left column. The specific portion of the requirement within the regulation from which exemption is being requested is emphasized (bold/underlined). The basis for the exemption from the specific portion of each requirement is provided in the corresponding row of the column on the right.
B. Exemptions Requested from 10 CFR 50, Appendix E Table 2 (Revised) below lists the pertinent portions of 10 CFR 50, Appendix E, Section IV, in the left column. The specific portion of the requirement within the regulation from which exemption is being requested is emphasized (bold/underlined). The basis for the exemption from the specific portion of each requirement is provided in the corresponding row of the column on the right.
The table below shows only the regulations for which a revision to the originally requested exemption (Reference 1) is being proposed. The rows shown in Table 2 (Revised), below, replace the corresponding rows listed in Table 2 of Reference 1 in their entirety. The requested exemptions from all other regulations shown in Table 2 of Reference 1 remains in effect.
The table below shows only the regulations for which a revision to the originally requested exemption (Reference 1) is being proposed. The rows shown in Table 2 (Revised), below, replace the corresponding rows listed in Table 2 of Reference 1 in their entirety. The requested exemptions from all other regulations shown in Table 2 of Reference 1 remains in effect.


Serial No. 13-390A Attachment 2 Page 4 of 10 TABLE 2 (Revised)
Serial No. 13-390A Page 4 of 10 TABLE 2 (Revised)
Exemptions Requested from 10 CFR 50, Appendix E Regulation (10 CFR 50, Appendix E)                         Basis for Requested Exemption (portion being exempted shown emphasized) i
Exemptions Requested from 10 CFR 50, Appendix E Regulation (10 CFR 50, Appendix E)
&sect; IV.B.1 - The means to be used for determining         Revised      radiological  analyses    have    been the magnitude of, and for continually assessing the     developed that show that, 90 days after shutdown, impact of, the release of radioactive materials shall   the radiological consequences of design basis be described, including emergency action levels         accidents will not exceed the limits of the EPA that are to be used as criteria for determining the     Protective Action Guides at the EAB. In addition, need for notification and participation of local and     analyses have been developed for beyond design State agencies, the Commission, and other Federal       basis events related to the spent fuel pool which agencies, and the emergency action levels that are       show that, within 17 months after shutdown, the to be used for determining when and what type of        analyzed event is either not credible, is capable of protective measures should be considered within         being mitigated, or the event's radiological and outside the site boundary to protect health         consequences will not exceed the limits of the EPA and safety. The emergency action levels shall be         Protective Action Guides at the EAB. Therefore, based on in-plant conditions and instrumentation in     offsite emergency response plans for local addition to onsite and offsite monitoring. By June       government authorities will          no longer be 20, 2012, for nuclear power reactor licensees,           necessary.
(portion being exempted shown emphasized)
these action levels must include hostile action that may adversely affect the nuclear power             Since offsite emergency plans will no longer be plant. The initial emergency action levels shall be     necessary, and based on the significantly reduced discussed and agreed on by the applicant or              scope of EALs for the permanently defueled licensee and state and local governmental                facility, the scope of the annual review of EALs with authorities, and approved by the NRC. Thereafter,        State and local governmental authorities is emergency action levels shall be reviewed with the      expected to be commensurately reduced (e.g.,
Basis for Requested Exemption i
State and local governmental authorities on an          informational mailings, etc.).
&sect; IV.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 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.
annual basis.
Revised radiological analyses have been developed that show that, 90 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. Therefore, offsite emergency response plans for local government authorities will no longer be necessary.
Since offsite emergency plans will no longer be necessary, and based on the significantly reduced scope of EALs for the permanently defueled facility, the scope of the annual review of EALs with State and local governmental authorities is expected to be commensurately reduced (e.g.,
informational mailings, etc.).
Justification from the requirements in Appendix E related to a "hostile action" is provided in the Basis for the requested exemption from &sect; IV. 1 above.
Justification from the requirements in Appendix E related to a "hostile action" is provided in the Basis for the requested exemption from &sect; IV. 1 above.
&sect; IV.F.1 - In addition, a radiological orientation       Revised      radiological  analyses    have    been training program shall be made available to local       developed that show that, 90 days. after shutdown, services personnel; e.g.,        local emergency        the radiological consequences of design basis serviceslCivil Defense, local law enforcement            accidents will not exceed the limits of the EPA personnel, local news media persons.                    Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. Therefore, offsite emergency response plans will no longer be necessary.
&sect; IV.F.1 - In addition, a radiological orientation training program shall be made available to local services personnel; e.g.,
local emergency serviceslCivil Defense, local law enforcement personnel, local news media persons.
Revised radiological analyses have been developed that show that, 90 days. after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. Therefore, offsite emergency response plans will no longer be necessary.
Local news media persons no longer need radiological orientation training since they will not be called upon to respond to a radiological event.
Local news media persons no longer need radiological orientation training since they will not be called upon to respond to a radiological event.
The term "Civil Defense" is no longer commonly used; therefore, reference to this term in the examples provided in the regulation is not needed.
The term "Civil Defense" is no longer commonly used; therefore, reference to this term in the examples provided in the regulation is not needed.


Serial No. 13-390A Attachment 2 Page 5 of 10 Regulation (10 CFR 50, Appendix E)                       Basis for Requested Exemption (portion being exempted shown emphasized)
Serial No. 13-390A Page 5 of 10 Regulation (10 CFR 50, Appendix E)
&sect; IV.F.2 - The plan shall describe provisions for the Revised      radiological  analyses    have    been conduct of emergency preparedness exercises as       developed that show that, 90 days after shutdown, follows: Exercises shall test the adequacy of timing  the radiological consequences of design basis and content of implementing procedures and            accidents will not exceed the limits of the EPA methods,      test  emergency    equipment    and  Protective Action Guides at the EAB. In addition, communications networks, test the public alert        analyses have been developed for beyond design and notification system, and ensure that              basis events related to the spent fuel pool which emergency organization personnel are familiar with    show that, within 17 months after shutdown, the their duties.                                        analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.
(portion being exempted shown emphasized)
Basis for Requested Exemption
&sect; IV.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.
Revised radiological analyses have been developed that show that, 90 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.
There will be no need for the public to take any protective actions in the event of an emergency at KPS. Therefore, participation by offsite entities will no longer be necessary, public alert and notification system will no longer be required.
There will be no need for the public to take any protective actions in the event of an emergency at KPS. Therefore, participation by offsite entities will no longer be necessary, public alert and notification system will no longer be required.
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility.
Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.


Serial No. 13-390A Attachment 2 Page 6 of 10 Regulation (10 CFR 50, Appendix E)                       Basis for Requested Exemption (portion being exempted shown emphasized)
Serial No. 13-390A Page 6 of 10 Regulation (10 CFR 50, Appendix E)
&sect; IV.F.2.b - Each licensee at each site shall         Revised      radiological  analyses    have    been conduct a subsequent exercise of its onsite           developed that show that, 90 days after shutdown, emergency plan every 2 years. Nuclear power           the radiological consequences of design basis reactor licensees shall submit exercise               accidents will not exceed the limits of the EPA scenarios under - 50.4 at least 60 days before         Protective Action Guides at the EAB. In addition, use in an exercise required by this paragraph         analyses have been developed for beyond design 2.b. The exercise may be included in the full         basis events related to the spent fuel pool which participation biennial exercise required by           show that, within 17 months after shutdown, the paragraph 2.c. of this section. In addition, the     analyzed event is either not credible, is capable of licensee shall take actions necessary to ensure       being mitigated, or the event's radiological that adequate emergency response capabilities are     consequences will not exceed the limits of the EPA maintained during the interval between biennial       Protective Action Guides at the EAB. There will be exercises by conducting drills, including at least     no need for the public to take any protective one drill involving a combination of some of the       actions in the event of an emergency at KPS.
Basis for Requested Exemption (portion being exempted shown emphasized)
principal functional areas of the licensee's onsite   Therefore, participation by offsite entities will no emergency response capabilities. The principal         longer be necessary and associated exercises will functional areas of emergency response include         no longer need to be conducted.
&sect; IV.F.2.b -
activities such as management and coordination of emergency response, accident assessment, event         Although NUREG-0654 states that "an exercise classification, notification of offsite authorities,   shall include mobilization of State and local assessment of the onsite and offsite impact of         personnel and resources adequate to verify the radiological     releases,     protective     action capability to respond to an accident scenario recommendation           development.     protective requiring response," such an exercise scenario action decision making, plant system repair and       scope is not necessary for a permanently defueled mitigative action implementation.       During these facility. Performance of reduced scope exercises drills, activation of all of the licensee's emergency is sufficient to maintain and assess the capability of response facilities (Technical Support Center         the emergency response organization to properly (TSC). Operations Support Center (OSC), and           perform activities.
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
the Emergency Operations Facility (EOF)) would not be necessary, licensees would have the             Offsite emergency response plans will no longer be opportunity to consider accident management           necessary and there will be no required response strategies, supervised instruction would           be by offsite agencies to the EOF. An EOF will no permitted, operating staff in all participating       longer be maintained.
: releases, protective action recommendation development.
facilities would have the opportunity to resolve problems (success paths) rather than have             An onsite facility (whether the control room or a controllers intervene, and the drills may focus on    facility similar to the technical support center) the onsite exercise training objectives.              would continue to be maintained, from which effective control can be exercised during an emergency.
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.
Revised radiological analyses have been developed that show that, 90 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. There will be no need for the public to take any protective actions in the event of an emergency at KPS.
Therefore, participation by offsite entities will no longer be necessary and associated exercises will no longer need to be conducted.
Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility.
Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.
Offsite emergency response plans will no longer be necessary and there will be no required response by offsite agencies to the EOF. An EOF will no longer be maintained.
An onsite facility (whether the control room or a facility similar to the technical support center) would continue to be maintained, from which effective control can be exercised during an emergency.


Serial No. 13-390A Attachment 2 Page 7 of 10 Regulation (10 CFR 50, Appendix E)                     Basis for Requested Exemption (portion being exempted shown emphasized)
Serial No. 13-390A Page 7 of 10 Regulation (10 CFR 50, Appendix E)
&sect; IV.F.2.e - Licensees shall enable any State or   Revised radiological analyses have been local government located within the plume         developed that show that, 90 days after shutdown, exposure Pathway EPZ to participate in the         the radiological consequences of design basis licensee's drills when requested by such State or   accidents will not exceed the limits of the EPA local government.                                   Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. Therefore, the plume exposure pathway emergency planning zone, and offsite plans and drills will no longer be necessary. In the context of this paragraph of the regulation, "any State" means Wisconsin and "local government" means the organizations that provide emergency support services (i.e. ambulance, fire, police) to KPS upon request.
Basis for Requested Exemption (portion being exempted shown emphasized)
&sect; IV.F.2.f - Remedial exercises will be required if Revised radiological     analyses have been the emergency plan is not satisfactorily tested     developed that show that, 90 days after shutdown, during the biennial exercise, such that NRC, in     the radiological consequences of design basis consultation with FEMA. cannot (1) find             accidents will not exceed the limits of the EPA reasonable assurance that adequate protective       Protective Action Guides at the EAB. In addition, measures can and will be taken in the event of a   analyses have been developed for beyond design radiological emergency or (2) determine that the   basis events related to the spent fuel pool which Emergency Response Organization (ERO) has           show that, within 17 months after shutdown, the maintained key skills specific to emergency         analyzed event is either not credible, is capable of response. The extent of State and local             being mitigated, or the event's radiological participation in remedial exercises must be         consequences will not exceed the limits of the EPA sufficient to show that appropriate corrective     Protective Action Guides at the EAB. Therefore, measures have been taken regqarding the             offsite emergency response plans will no longer be elements of the plan not Properly tested in the     necessary and the scope of exercises can be previous exercises.                                 commensurately reduced.
&sect; IV.F.2.e - Licensees shall enable any State or Revised radiological analyses have been local government located within the plume developed that show that, 90 days after shutdown, exposure Pathway EPZ to participate in the the radiological consequences of design basis licensee's drills when requested by such State or accidents will not exceed the limits of the EPA local government.
Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.
Therefore, the plume exposure pathway emergency planning zone, and offsite plans and drills will no longer be necessary. In the context of this paragraph of the regulation, "any State" means Wisconsin and "local government" means the organizations that provide emergency support services (i.e. ambulance, fire, police) to KPS upon request.
&sect; IV.F.2.f - Remedial exercises will be required if Revised radiological analyses have been the emergency plan is not satisfactorily tested developed that show that, 90 days after shutdown, during the biennial exercise, such that NRC, in the radiological consequences of design basis consultation with FEMA.
cannot (1) find accidents will not exceed the limits of the EPA reasonable assurance that adequate protective Protective Action Guides at the EAB. In addition, measures can and will be taken in the event of a analyses have been developed for beyond design radiological emergency or (2) determine that the basis events related to the spent fuel pool which Emergency Response Organization (ERO) has show that, within 17 months after shutdown, the maintained key skills specific to emergency analyzed event is either not credible, is capable of response. The extent of State and local being mitigated, or the event's radiological participation in remedial exercises must be consequences will not exceed the limits of the EPA sufficient to show that appropriate corrective Protective Action Guides at the EAB. Therefore, measures have been taken regqarding the offsite emergency response plans will no longer be elements of the plan not Properly tested in the necessary and the scope of exercises can be previous exercises.
commensurately reduced.


Serial No. 13-390A Attachment 2 Page 8 of 10 Regulation (10 CFR 50, Appendix E)                     Basis for Requested Exemption (portion being exempted shown emphasized)
Serial No. 13-390A Page 8 of 10 Regulation (10 CFR 50, Appendix E)
&sect; IV.F.2.i - Licensees shall use drill and exercise Revised    radiological    analyses    have    been scenarios that provide reasonable assurance that   developed that show that, 90 days after shutdown, anticipatory responses will not result from        the radiological consequences of design basis preconditioning of participants. Such scenarios    accidents will not exceed the limits of the EPA for nuclear power reactor licensees must            Protective Action Guides at the EAB. In addition, include a wide spectrum of radiological            analyses have been developed for beyond design releases and events. including hostile action.      basis events related to the spent fuel pool which Exercise and drill scenarios as appropriate must    show that, within 17 months after shutdown, the emphasize coordination among onsite and offsite    analyzed event is either not credible, is capable of response organizations.                              being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective   Action   Guides     at   the   EAB.
Basis for Requested Exemption (portion being exempted shown emphasized)
&sect; IV.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.
Revised radiological analyses have been developed that show that, 90 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.
Requirements for offsite planning will no longer be necessary. Therefore, the scope of exercises can be commensurately reduced.
Requirements for offsite planning will no longer be necessary. Therefore, the scope of exercises can be commensurately reduced.
Following docketing of its "Certification of Permanent Removal of Fuel from the Reactor Vessel," dated May 14, 2013, KPS is a permanently shutdown facility with spent fuel stored in the spent fuel pool and ISFSI. In the EP Final Rule (76 FR 72596, Nov. 23, 2011), the Commission defined "hostile action" as, in part, an act directed toward a nuclear power plant or its personnel. The NRC excluded non-power reactors (NPR) 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.
Following docketing of its "Certification of Permanent Removal of Fuel from the Reactor Vessel,"
dated May 14,
: 2013, KPS is a
permanently shutdown facility with spent fuel stored in the spent fuel pool and ISFSI. In the EP Final Rule (76 FR 72596, Nov. 23, 2011), the Commission defined "hostile action" as, in part, an act directed toward a nuclear power plant or its personnel. The NRC excluded non-power reactors (NPR) 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.
Likewise, spent fuel pools and ISFSIs are not a nuclear power plant.
Likewise, spent fuel pools and ISFSIs are not a nuclear power plant.
The following similarities between the KPS facility and NPRs show that the KPS facility should be treated similarly to NPRs. Similar to NPRs, KPS poses lower radiological risks to the public from accidents than do power reactors because: (1)
The following similarities between the KPS facility and NPRs show that the KPS facility should be treated similarly to NPRs. Similar to NPRs, KPS poses lower radiological risks to the public from accidents than do power reactors because: (1)
KPS is a permanently shutdown facility (with fuel stored in the spent fuel pool and ISFSI) and no longer generates fission products; 2) Fuel stored in the KPS SFP has lower decay heat, resulting in lower risk of fission product release in the event of a non-credible boil off or draindown event; and 3) no credible accident at KPS will result in radiological releases requiring offsite protective actions. NPRs have lower decay heat associated with a lower risk of core melt and fission product release in a loss-of-coolant accident. Likewise, KPS has a low likelihood of a credible accident resulting in radiological releases requiring offsite protective actions.
KPS is a permanently shutdown facility (with fuel stored in the spent fuel pool and ISFSI) and no longer generates fission products; 2) Fuel stored in the KPS SFP has lower decay heat, resulting in lower risk of fission product release in the event of a non-credible boil off or draindown event; and 3) no credible accident at KPS will result in radiological releases requiring offsite protective actions. NPRs have lower decay heat associated with a lower risk of core melt and fission product release in a loss-of-coolant accident.
: Likewise, KPS has a low likelihood of a credible accident resulting in radiological releases requiring offsite protective actions.


Serial No. 13-390A Attachment 2 Page 9 of 10 The portions of 10 CFR 50.47 and 10 CFR 50, Appendix E that are not identified in Tables 1 and 2 of Reference 1, as modified in the two tables above (i.e., those portions for which exemption is not being requested), will remain applicable to KPS.
Serial No. 13-390A Page 9 of 10 The portions of 10 CFR 50.47 and 10 CFR 50, Appendix E that are not identified in Tables 1 and 2 of Reference 1, as modified in the two tables above (i.e., those portions for which exemption is not being requested), will remain applicable to KPS.
II. BACKGROUND The background information contained in Reference 1 remains applicable to this supplement. As discussed in Reference 1, an analysis of the potential radiological impact of a design basis accident at KPS in a permanently defueled condition indicates that any potential radiological releases beyond the site boundary would be below the EPA PAG exposure levels, as detailed in the EPA's "Protective Action Guide and Planning Guidance for Radiological Incidents," Draft for Interim Use and Public Comment dated March 2013 (PAG Manual).
II.
BACKGROUND The background information contained in Reference 1 remains applicable to this supplement.
As discussed in Reference 1, an analysis of the potential radiological impact of a design basis accident at KPS in a permanently defueled condition indicates that any potential radiological releases beyond the site boundary would be below the EPA PAG exposure levels, as detailed in the EPA's "Protective Action Guide and Planning Guidance for Radiological Incidents," Draft for Interim Use and Public Comment dated March 2013 (PAG Manual).
As stated in Reference 1, the KPS USAR contains the following description regarding spent fuel pool indication available to operators for responding to a postulated loss of heat removal capability for the spent fuel pool.
As stated in Reference 1, the KPS USAR contains the following description regarding spent fuel pool indication available to operators for responding to a postulated loss of heat removal capability for the spent fuel pool.
Both temperature and level indicators in the pool would alert operators to a loss of cooling. Local and remote alarms are provided. This allows the operator to take corrective measures in a timely manner to restore cooling capability to the spent fuel pool cooling loop.
Both temperature and level indicators in the pool would alert operators to a loss of cooling. Local and remote alarms are provided. This allows the operator to take corrective measures in a timely manner to restore cooling capability to the spent fuel pool cooling loop.
A recent review of this USAR description revealed that the sentence regarding local and remote alarms is ambiguous and could imply that local and remote alarms are provided both for pool temperature and for pool level. Although installation of a new spent fuel pool level indication system will include both a local and remote alarm, only a remote alarm is provided for spent fuel pool temperature. These indications and alarms are considered adequate to allow timely operator response to an abnormal spent fuel pool condition.
A recent review of this USAR description revealed that the sentence regarding local and remote alarms is ambiguous and could imply that local and remote alarms are provided both for pool temperature and for pool level. Although installation of a new spent fuel pool level indication system will include both a local and remote alarm, only a remote alarm is provided for spent fuel pool temperature. These indications and alarms are considered adequate to allow timely operator response to an abnormal spent fuel pool condition.
II1. JUSTIFICATION FOR EXEMPTIONS AND SPECIAL CIRCUMSTANCES 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 Part 50 which are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the defense and security. 10 CFR 50.12 also states that the Commission will not consider granting an exemption unless special circumstances are present. The justification for exemptions and special circumstances contained in Reference 1 are not affected by, and remain applicable to, this supplement. Therefore, this exemption request satisfies the provisions of Section 50.12.
II1. JUSTIFICATION FOR EXEMPTIONS AND SPECIAL CIRCUMSTANCES 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 Part 50 which are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the defense and security.
10 CFR 50.12 also states that the Commission will not consider granting an exemption unless special circumstances are present.
The justification for exemptions and special circumstances contained in Reference 1 are not affected by, and remain applicable to, this supplement. Therefore, this exemption request satisfies the provisions of Section 50.12.


Serial No. 13-390A Attachment 2 Page 10 of 10 IV. ENVIRONMENTAL CONSIDERATION The conclusions of the environmental considerations contained in Reference 1 are not affected by, and remain applicable to, this supplement.
Serial No. 13-390A Page 10 of 10 IV. ENVIRONMENTAL CONSIDERATION The conclusions of the environmental considerations contained in Reference 1 are not affected by, and remain applicable to, this supplement.
V. CONCLUSION The conclusions contained in Reference 1 are not affected by, and remain applicable to, this supplement. Therefore, the requested exemptions, as supplemented herein, 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, and special circumstances are present as set forth in 10 CFR 50.12(a)(2).
V.
CONCLUSION The conclusions contained in Reference 1 are not affected by, and remain applicable to, this supplement. Therefore, the requested exemptions, as supplemented herein, 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, and special circumstances are present as set forth in 10 CFR 50.12(a)(2).
REFERENCES
REFERENCES
: 1. Letter from A. J. Jordan (DEK) to NRC Document Control Desk, "Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E," dated July 31, 2013
: 1. Letter from A. J. Jordan (DEK) to NRC Document Control Desk, "Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E," dated July 31, 2013
Line 362: Line 473:
: 1. CALCULATION 2013-07050, MAXIMUM CLADDING TEMPERATURE ANALYSIS FOR AN UNCOVERED SPENT FUEL POOL WITH NO AIR COOLING KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.
: 1. CALCULATION 2013-07050, MAXIMUM CLADDING TEMPERATURE ANALYSIS FOR AN UNCOVERED SPENT FUEL POOL WITH NO AIR COOLING KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.


Calculation Cover Sheet SFDominion                                                            I     CMAACC-0             ATAHET26ae1o Note: This form is only applicable to Revision 6 of this procedure.
SFDominion Calculation Cover Sheet I
CMAACC-0 ATAHET26ae1o Note: This form is only applicable to Revision 6 of this procedure.
Complete the fields with text or an X as required.
Complete the fields with text or an X as required.
Calculation Number:
Calculation Number:
2013-07050 Vendor (If not Dominion):                                                 Calculation Preparation Risk: OLow       E]Medium         -]High Sargent & Lundy                                                              HU-RISK-20130082 Vendor Proprietary: E]Yes               SNo                             Pre-Job Brief Completed: ElYes EINo ENA Calculation Quality Class:   E-Safety Related             ONSQ             E]Non-Safety Related Subject (Calculation Title):
2013-07050 Vendor (If not Dominion):
Sargent & Lundy Calculation Preparation Risk: OLow E]Medium  
-]High HU-RISK-20130082 Vendor Proprietary: E]Yes SNo Pre-Job Brief Completed: ElYes EINo ENA Calculation Quality Class:
E-Safety Related ONSQ E]Non-Safety Related Subject (Calculation Title):
Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with no Air Cooling Addendum
Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with no Air Cooling Addendum


==Title:==
==Title:==
N/A Station(s) and Unit(s): NA   Eli   E12   1773                 Affected System(s), Structure(s), or Component(s):
N/A Station(s) and Unit(s): NA Eli E12 1773 Affected System(s), Structure(s), or Component(s):
SU [-1l     E-2     KW 0       MP   Ell 1[2 Ei3                 21-SFP- Spent Fuel Cooling EICO (Note: If both SU and NA then only check CO)
SU [-1l E-2 KW 0 MP Ell 1[2 Ei3 21-SFP-Spent Fuel Cooling EICO (Note: If both SU and NA then only check CO)
Purpose (Executive Summary):
Purpose (Executive Summary):
This vendor calculation was performed by Sargent & Lundy to their Quality Program. It evaluates the length of time it takes for uncovered spent fuel assemblies to reach the temperature where the zircaloy might ignite for the unlikely postulated event of a loss of spent fuel pool inventory. The analysis assumes no air cooling of the fuel assemblies.
This vendor calculation was performed by Sargent & Lundy to their Quality Program. It evaluates the length of time it takes for uncovered spent fuel assemblies to reach the temperature where the zircaloy might ignite for the unlikely postulated event of a loss of spent fuel pool inventory. The analysis assumes no air cooling of the fuel assemblies.
Originator (Qual. Required): Printed Name (1) (3)                           Signature: (1) (3)                   Date: (1) (3)
Originator (Qual. Required): Printed Name (1) (3)
N/A N/A                                                                         N/A Reviewer (Qual. Required): Printed Name (1)         Type of Review: (2)   Signature:                           Date:
Signature: (1) (3)
M. S. Lico                                           OwnerNm               Sgare                                               g /u Approveair: Printed Namew.J                                                 Sign. -*/b*at ure-/                         Date: *:*t**//_*
Date: (1) (3)
N/A N/A N/A Reviewer (Qual. Required): Printed Name (1)
Type of Review: (2)
Signature:
Date:
M. S. Lico OwnerNm Sgare g /u Approve r: Printed Namew.J ai Sign. -*/b*at ure-/
Date: *:*t**//_*
(i Note: Physical or electronic signatures are acceptable.
(i Note: Physical or electronic signatures are acceptable.
Note: At the discretion of the originator, a facsimile of this cover sheet that does not contain the "CM-AA-CLC-301"or "Attachment 1" headers may be used. Facsimiles must contain all of the elements of the cover sheet in the current revision of CM-AA-CLC-301. (1) Add lines for additional originators or reviewers as necessary. (2) Note if reviews are "Independent," "Peer", "Subject Matter Expert", "Supervisor", or "Owner's". (3) Enter N/A for Owner's Review of Vendor Calculation.
Note: At the discretion of the originator, a facsimile of this cover sheet that does not contain the "CM-AA-CLC-301"or "Attachment 1" headers may be used. Facsimiles must contain all of the elements of the cover sheet in the current revision of CM-AA-CLC-301. (1) Add lines for additional originators or reviewers as necessary. (2) Note if reviews are "Independent," "Peer", "Subject Matter Expert", "Supervisor", or "Owner's". (3) Enter N/A for Owner's Review of Vendor Calculation.
731189 (Mar 2012)
731189 (Mar 2012)


CalculationReview Checklist JFDominion                                            I   CMAACC-0           ATAHET4                   ae1o Calculation# 2013-07050           Rev. 2 Add. N/A NOTE: If "Yes" is not answered, an explanation may be provided below. Reference may be made to explanations contained in the calculation or addendum.
JFDominion Calculation Review Checklist I
: 1. Have the sources of design inputs been correctly selected and referenced in the calculation?                                                                                           I     1 ]
CMAACC-0 ATAHET4 ae1o Calculation # 2013-07050 Rev. 2 Add. N/A NOTE: If "Yes" is not answered, an explanation may be provided below. Reference may be made to explanations contained in the calculation or addendum.
: 2. Are the sources of design inputs up-to-date and retrievable/attached to the calculation?             [X]       [ ]
: 1. Have the sources of design inputs been correctly selected and referenced in the calculation?
: 3. Where appropriate, have the other disciplines reviewed or provided the design inputs for which they are responsible?                                                                         [X       1[
I 1 ]
: 4. Have design inputs been confirmed by analysis, test, measurement, field walkdown, or other pertinent means as appropriate for the configuration analyzed?                                 I I       V]
: 2. Are the sources of design inputs up-to-date and retrievable/attached to the calculation?
: 5. Have the bases for assumptions been adequately and clearly presented and are they bounded by the Station Design Basis?                                                                 [X       I I
[X]
: 6. Were appropriate calculation/analytic methods used and are outputs reasonable when compared to inputs?                                                                                   [X 1     [
[ ]
: 7. Are computations technically accurate?                                                                 [X I     [
: 3. Where appropriate, have the other disciplines reviewed or provided the design inputs for which they are responsible?
: 8. Has the calculation made appropriate allowances for instrument errors and calibration equipment errors?                                                                                   [X]       [
[X 1 [
: 9. Have those computer codes used in the analysis been referenced in the calculation?                   [X ]     [ I
: 4. Have design inputs been confirmed by analysis, test, measurement, field walkdown, or other pertinent means as appropriate for the configuration analyzed?
: 10. Have all exceptions to station design basis criteria and regulatory requirements been identified and justified in accordance with NQA-1-1994?                                               I I     [X]
I I V ]
: 11. Has the design authority/original preparer for this calculation been informed of its revision or addendum, if required?                                                                             I I     [X]
: 5. Have the bases for assumptions been adequately and clearly presented and are they bounded by the Station Design Basis?
[X I I
: 6. Were appropriate calculation/analytic methods used and are outputs reasonable when compared to inputs?
[X 1
[
: 7.
Are computations technically accurate?
[X I
[
: 8. Has the calculation made appropriate allowances for instrument errors and calibration equipment errors?
[X]
[
: 9.
Have those computer codes used in the analysis been referenced in the calculation?
[X ]
[ I
: 10. Have all exceptions to station design basis criteria and regulatory requirements been identified and justified in accordance with NQA-1-1994?
I I
[X]
: 11. Has the design authority/original preparer for this calculation been informed of its revision or addendum, if required?
I I
[X]
Comments provided to S & L resolved in the final draft of the calculation to Dominion's satisfaction.
Comments provided to S & L resolved in the final draft of the calculation to Dominion's satisfaction.
Item # 4 - Input data is from the literature or provided in ETE-NAF-2013-0077, Rev. 0 (documented in S & Lcalc).
Item # 4 - Input data is from the literature or provided in ETE-NAF-2013-0077, Rev. 0 (documented in S & L calc).
#10 - Calculation provides information and is not used for design basis criteria.
#10 - Calculation provides information and is not used for design basis criteria.
# 11 - The original preparer of Rev. 2 is the same as Rev. 0 and 1.
# 11 - The original preparer of Rev. 2 is the same as Rev. 0 and 1.
Signature:   N/A                                                             Date:     N/A (Preparer)
Signature:
Signature:   M. S. Lico   ,        -                                          Date:   9"16 F/f3 (Reviewer)
N/A Date:
N/A (Preparer)
Signature:
M. S. Lico Date:
9"16 F/f3 (Reviewer)
Note: Physical or electronic signatures are acceptable.
Note: Physical or electronic signatures are acceptable.
731190 (Mar 2012)
731190 (Mar 2012)
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==SUMMARY==
==SUMMARY==
CLIENT:
Dominion UNIT:
1 PAGE NO.:
1 PROJECT NAME:
Kewaunee Power Station S&L NUCLEAR OA PROGRAM PROJECT NO.:
11862-198 APPLICABLE El YES [l NO CALC. NO.:
2013-07050 SAFETY RELATED E: YES Z NO TITLE:
Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with no Air Cooling EQUIPMENT NO.:
IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDNOIDED & REVIEW METHOD Initial Issue. The main body is pages 1 through 10. The final page is INPUTS/ASSUMPTIONS Attachment C, page C2.
M VERIFIED El UNVERIFIED REVIEW METHOD:
Detailed Review REV.:
0 STATUS:
Z APPROVED El SUPERSEDED BY CALCULATION NO.
Ql VOID DATE FOR REV.:
7-22-13 PREPARER:
Matthew M. Ross Signature on file DATE:
7-22-13 REVIEWER:
Joseph J. Pawasarat Signature on file DATE:
7-22-13 APPROVER:
Robert J. Peterson Signature on file DATE:
7-22-13 IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDED/VOIDED & REVIEW METHOD Revision 1 corrects a typo in Figure 7-1 and incorporates minor editorial changes. Pages 1 through 10.are revised and the changes are tracked with INPUTS/ASSUMPTIONS revision bars. None of the attachments are revised. The main body is pages 1 2 VERIFIED through 10. The final page is Attachment C, page C2.
El UNVERIFIED REVIEW METHOD:
Detailed Review REV.:
1 STATUS:
E APPROVED
[E SUPERSEDED BY CALCULATION NO.
El VOID DATE FOR REV.:
7-23-13 PREPARER:
Matthew M. Ross Signature on file DATE:
7-23-13 REVIEWER:
Joseph J. Pawasarat Signature on file DATE:
7-23-13 APPROVER:
Robert J. Peterson Signature on file DATE:
7-23-13 IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDIVOIDED & REVIEW METHOD Revision 2 calculates the date when it will take 2, 4, 6, or 10 hours for the spent fuel to heat up from 32 0C to 900 0. Changes to the main body are tracked with revision bars. Changes to Attachment B are tracked with revision bars.
INPUTS/ASSUMPTIONS There are no changes to Attachments A and C. The main body is pages 1 Z VERIFIED through 10. The final page of the calculation is Attachment C, page C2.
El UNVERIFIED REVIEW METHOD:
Detailed Review REV.:
2 STATUS:
[0 APPROVED El SUPERSEDED BY ALCULATION NO.
El VOID DATE FOR REV.:
7-,
&#xfd;-
PREPARER:
Matthew M. Ross DATE:
Z,9/tg/-
0 I REVIEWER:
Joseph J. Pawasarat DATE:,-Z-
_j APPROVER:
Robert J. Peterson DATE:


CLIENT:              Dominion                                                  UNIT:    1        PAGE NO.:  1 PROJECT NAME:        Kewaunee Power Station                                         S&L NUCLEAR OA PROGRAM PROJECT NO.:          11862-198                                                      APPLICABLE El YES [l NO CALC. NO.:            2013-07050                                                      SAFETY RELATED E: YES Z NO TITLE:                Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with no Air Cooling EQUIPMENT NO.:
Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 2 of 10i S. Purpose and Scope.............................................................................................................................
IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDNOIDED &REVIEW METHOD Initial Issue. The main body is pages 1 through 10. The final page is                          INPUTS/ASSUMPTIONS Attachment C, page C2.                                                                           M VERIFIED El  UNVERIFIED REVIEW METHOD:      Detailed Review                                                                    REV.:        0 STATUS:      Z APPROVED      El SUPERSEDED  BY CALCULATION NO.             Ql VOID      DATE FOR REV.:      7-22-13 PREPARER:      Matthew M. Ross                    Signature on file                                  DATE:    7-22-13 REVIEWER:      Joseph J. Pawasarat                Signature on file                                  DATE:    7-22-13 APPROVER:      Robert J. Peterson                  Signature on file                                  DATE:    7-22-13 IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDED/VOIDED &REVIEW METHOD Revision 1 corrects a typo in Figure 7-1 and incorporates minor editorial changes. Pages 1 through 10.are revised and the changes are tracked with                        INPUTS/ASSUMPTIONS revision bars. None of the attachments are revised. The main body is pages 1                      2 VERIFIED through 10. The final page is Attachment C, page C2.                                             El UNVERIFIED REVIEW METHOD:      Detailed Review                                                                    REV.:        1 STATUS:      E APPROVED      [E SUPERSEDED BY CALCULATION NO.                 El VOID      DATE FOR REV.:      7-23-13 PREPARER:      Matthew M. Ross                      Signature on  file                              DATE:      7-23-13 REVIEWER:      Joseph J. Pawasarat                Signature on file                                DATE:      7-23-13 APPROVER:      Robert J. Peterson                  Signature on file                                DATE:      7-23-13 IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDIVOIDED &REVIEW METHOD Revision 2 calculates the date when it will take 2, 4, 6, or 10 hours for the spent fuel to heat up from 32 0C to 900 0. Changes to the main body are tracked with revision bars. Changes to Attachment B are tracked with revision bars.                     INPUTS/ASSUMPTIONS There are no changes to Attachments A and C. The main body is pages 1                            Z VERIFIED through 10. The final page of the calculation is Attachment C, page C2.                         El UNVERIFIED REVIEW METHOD:      Detailed Review                                                                    REV.:        2 STATUS:      [0 APPROVED    El SUPERSEDED BY ALCULATION NO.                 El VOID      DATE FOR REV.: 7-,          &#xfd;-
3
PREPARER:      Matthew  M. Ross                                                                      DATE:      Z,9/tg/-
: 2.
0 I REVIEWER:      Joseph J. Pawasarat                                                                    DATE:  ,-Z-        _j APPROVER:      Robert J. Peterson                                                                    DATE:
References...........................................................................................................................................
 
4
Document No. 2013-07050                                                                                                              Kewaunee Power Station Revision 2                                                                                                                                                    Page 2 of 10i S. Purpose and Scope .............................................................................................................................                     3
: 3.
: 2. References ...........................................................................................................................................               4
Definitions...........................................................................................................................................
: 3. Definitions ...........................................................................................................................................             4
4
: 4. Input Data ...........................................................................................................................................              5
: 4.
: 5. Assum ptions ........................................................................................................................................               6
Input Data...........................................................................................................................................
: 6. M ethodology .......................................................................................................................................                 7
5
: 7. R e su lts .................................................................................................................................................        9
: 5.
: 8. Conclusions and Recom m endations ...........................................................................................                                     10 Attachments:                                                                                                                                                  No. of Pages:
Assum ptions........................................................................................................................................
Attachm ent A : Generation Rate vs. Decay Tim e (Reference 2.5) ............................................................                                               6 Attachm ent B : Analysis ................................................................................................................................                 2 Attachm ent C: Current SFP Tem perature .................................................................................................                                  2
6
                                                                                    &~viLS I--jfr~   d     -
: 6.
M ethodology.......................................................................................................................................
7 7.
R e su lts.................................................................................................................................................
9
: 8.
Conclusions and R ecom m endations...........................................................................................
10 Attachments:
No. of Pages:
Attachm ent A : Generation Rate vs. Decay Tim e (Reference 2.5)............................................................
6 Attachm ent B : Analysis................................................................................................................................
2 Attachm ent C: Current SFP Tem perature.................................................................................................
2 I--jfr~  
&~vi LS d


Document No. 2013-07050                                                   Kewaunee Power Station Revision 2                                                                             Page 3 of 10
Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 3 of 10
: 1. Purpose and Scope 1.1. Purpose The purpose of this calculation is to conservatively evaluate the length of time (number of hours) it takes for uncovered spent fuel assemblies to reach the temperature where the zirconium cladding would fail. This analysis conservatively assumes that there is no air cooling of the assemblies: the flow paths that would provide natural circulation cooling are assumed to be blocked.
: 1. Purpose and Scope 1.1.
Purpose The purpose of this calculation is to conservatively evaluate the length of time (number of hours) it takes for uncovered spent fuel assemblies to reach the temperature where the zirconium cladding would fail. This analysis conservatively assumes that there is no air cooling of the assemblies: the flow paths that would provide natural circulation cooling are assumed to be blocked.
1.2. Scope The length of time for the fuel to heat up (the heat-up time) is determined as a function of the day that the analysis is performed (the decay time). The heat load from Westinghouse 422V+ fuel is used in this analysis (Reference 2.5 and Assumption 5.1).
1.2. Scope The length of time for the fuel to heat up (the heat-up time) is determined as a function of the day that the analysis is performed (the decay time). The heat load from Westinghouse 422V+ fuel is used in this analysis (Reference 2.5 and Assumption 5.1).
The zirconium cladding must remain below the temperature where it will fail. Per NUREG/CR-6451 (Ref. 2.1, see Design Input 4.1), 565 'C (1049 'F) is the lowest temperature where incipient cladding failure might occur. NUREG- 1738 (Ref. 2.7, pg.
The zirconium cladding must remain below the temperature where it will fail. Per NUREG/CR-6451 (Ref. 2.1, see Design Input 4.1), 565 'C (1049 'F) is the lowest temperature where incipient cladding failure might occur. NUREG-1738 (Ref. 2.7, pg.
3-7) states that runaway oxidation of zirconium occurs at 900 'C. For this analysis, the NUREG/CR-6451 temperature (565 'C, 1049 'F) and the NUREG-1738 temperature (900 'C, 1652 'F) are the temperatures of interest for the zirconium cladding.
3-7) states that runaway oxidation of zirconium occurs at 900 'C. For this analysis, the NUREG/CR-6451 temperature (565 'C, 1049 'F) and the NUREG-1738 temperature (900 'C, 1652 'F) are the temperatures of interest for the zirconium cladding.
There are no specific acceptance criteria for this analysis, however, SECY-99-168 (Ref.
There are no specific acceptance criteria for this analysis, however, SECY-99-168 (Ref.
2.4) suggests that "10 hours (is) sufficient time to take mitigative action" and that for PWRs, 2.5 years is expected to be the decay time needed to reach a 10 hour heat-up time from 30 'C to 900 'C. NUREG-1738 shows that a 10 hour heat up time to 900 'C for a PWR would occur at less than 2 years (Ref. 2.7, Fig. 2-2).
2.4) suggests that "10 hours (is) sufficient time to take mitigative action" and that for PWRs, 2.5 years is expected to be the decay time needed to reach a 10 hour heat-up time from 30 'C to 900 'C. NUREG-1738 shows that a 10 hour heat up time to 900 'C for a PWR would occur at less than 2 years (Ref. 2.7, Fig. 2-2).
                                        -,mrar tntC. L.In l'y
-,mrar tnt C. L.In l'y


Document No. 2013-07050                                                 Kewaunee Power Station Revision 2                                                                               Page 4 of 10
Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 4 of 10
: 2. References 2.1. NUREG/CR-6451, "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants," August 1997.
: 2. References 2.1.
NUREG/CR-6451, "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants," August 1997.
2.2. Incropera, Frank P., and David P. DeWitt, Introduction to Heat Transfer, Fourth Edition, John Wiley & Sons.
2.2. Incropera, Frank P., and David P. DeWitt, Introduction to Heat Transfer, Fourth Edition, John Wiley & Sons.
2.3. Kewaunee USAR, Chapter 3: Reactor, Revision 24.02 - Updated Online 04/15/13.
2.3. Kewaunee USAR, Chapter 3: Reactor, Revision 24.02 - Updated Online 04/15/13.
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July 2 2nd, 2013. Included as Attachment C.
July 2 2nd, 2013. Included as Attachment C.
2.7. NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," February 2001.
2.7. NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," February 2001.
: 3. Definitions 3.1. Decay Time The decay time is the time since the reactor was shut down (May   7 th, 2013).
: 3. Definitions 3.1. Decay Time The decay time is the time since the reactor was shut down (May 7 th, 2013).
3.2. Heat-up Time The heat-up time is the amount of time between when the fuel becomes uncovered and when the zirconium cladding reaches the failure temperatures of interest, 565 'C (1049
3.2. Heat-up Time The heat-up time is the amount of time between when the fuel becomes uncovered and when the zirconium cladding reaches the failure temperatures of interest, 565 'C (1049
        'F) and 900 'C (1652 'F).
'F) and 900 'C (1652 'F).
sAr,0orn rStucl
sAr,0orn rStucl


Document No. 2013-07050                                                   Kewaunee Power Station Revision 2                                                                             Page 5 of 10
Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 5 of 10
: 4. Input Data 4.1. Maximum Zirconium Temperature Several studies are presented in NUREG/CR-6451 (Ref. 2.1) discussing the maximum allowable temperature of zirconium cladding that will ensure that failure of the zirconium cladding will not occur. Per NUREG/CR-6451 (Ref. 2. 1, see Design Input 4.1), 565 -C (1049 OF) is the lowest temperature where incipient cladding failure might occur. NUREG-1738 uses 900 'C (1652 &deg;F) as the temperature where "runaway oxidation" is expected to occur (Ref. 2.7, pg. 3-7). These two temperatures are the failure temperatures of interest for this calculation 4.2. Zirconium Properties The specific heat of zirconium at 600 K (620 OF) is 322 J/kg-K and the density of zirconium is 6570 kg/m 3 (Ref. 2.2, pg. 822). A temperature of 620 &deg;F is in the temperature range (less than the midpoint for both ranges) of this analysis. From Reference 2.2, the specific heat slightly increases with an increase in temperature. At higher temperatures, the zirconium would heat up more slowly. This temperature is representative of the full temperature range for this analysis.
: 4. Input Data 4.1.
Maximum Zirconium Temperature Several studies are presented in NUREG/CR-6451 (Ref. 2.1) discussing the maximum allowable temperature of zirconium cladding that will ensure that failure of the zirconium cladding will not occur. Per NUREG/CR-6451 (Ref. 2. 1, see Design Input 4.1), 565 -C (1049 OF) is the lowest temperature where incipient cladding failure might occur. NUREG-1738 uses 900 'C (1652 &deg;F) as the temperature where "runaway oxidation" is expected to occur (Ref. 2.7, pg. 3-7). These two temperatures are the failure temperatures of interest for this calculation 4.2. Zirconium Properties The specific heat of zirconium at 600 K (620 OF) is 322 J/kg-K and the density of zirconium is 6570 kg/m3 (Ref. 2.2, pg. 822). A temperature of 620 &deg;F is in the temperature range (less than the midpoint for both ranges) of this analysis. From Reference 2.2, the specific heat slightly increases with an increase in temperature. At higher temperatures, the zirconium would heat up more slowly. This temperature is representative of the full temperature range for this analysis.
4.3. Uranium Properties The specific heat of uranium at 600 K (620 &deg;F) is 146 J/kg-K and the density of uranium is 19070 kg/m 3 (Ref. 2.2, pg. 822). A temperature of 620 OF is in the temperature range (less than the midpoint for both ranges) of this analysis. From Reference 2.2, the specific heat slightly increases with an increase in temperature. At higher temperatures, the uranium would heat up more slowly. This temperature is representative of the full temperature range for this analysis.
4.3. Uranium Properties The specific heat of uranium at 600 K (620 &deg;F) is 146 J/kg-K and the density of uranium is 19070 kg/m 3 (Ref. 2.2, pg. 822). A temperature of 620 OF is in the temperature range (less than the midpoint for both ranges) of this analysis. From Reference 2.2, the specific heat slightly increases with an increase in temperature. At higher temperatures, the uranium would heat up more slowly. This temperature is representative of the full temperature range for this analysis.
4.4. Geometry for Westinghouse 422V+ Assemblies The table below shows the geometry inputs for the fuel assemblies used in this analysis.
4.4. Geometry for Westinghouse 422V+ Assemblies The table below shows the geometry inputs for the fuel assemblies used in this analysis.
Table 4-1: Fuel Assembly Inputs (from USAR Table 3.2-8, Ref. 2.3)
Table 4-1: Fuel Assembly Inputs (from USAR Table 3.2-8, Ref. 2.3)
Uranium Pellet Diameter                               0.3659 inches Inner Diameter of Cladding                             0.3734 inches Outer Diameter of Cladding                             0.422 inches Rod Configuration and Total Rods                       14 x 14, 196 total spaces Number of Guide Tubes, Instrument Tubes               16 guide, I instrument Total Number of Heated Rods                           179 rods Inner Diameter of Guide Tubes (Above Dashpot) 0.492 inches Outer Diameter of Guide Tubes (Above Dashpot) 0.526 inches snrvg~nlt C&#xfd; L..ndjv,
Uranium Pellet Diameter 0.3659 inches Inner Diameter of Cladding 0.3734 inches Outer Diameter of Cladding 0.422 inches Rod Configuration and Total Rods 14 x 14, 196 total spaces Number of Guide Tubes, Instrument Tubes 16 guide, I instrument Total Number of Heated Rods 179 rods Inner Diameter of Guide Tubes (Above Dashpot) 0.492 inches Outer Diameter of Guide Tubes (Above Dashpot) 0.526 inches snrvg~nlt C&#xfd; L..ndjv,


Document No. 2013-07050                                                     Kewaunee Power Station Revision 2                                                                                 Page 6 of 10 Table 4-1 Continued Heated Height of Rods                                   143.25 inches Cladding and Guide Tube Material                       ZIRLO Zirconium Theoretical Uranium Density Percentage                 96.56%
Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 6 of 10 Table 4-1 Continued Heated Height of Rods 143.25 inches Cladding and Guide Tube Material ZIRLO Zirconium Theoretical Uranium Density Percentage 96.56%
4.5. Heat Load Reference 2.5 determines the maximum heat load from a single assembly. The assembly with the highest heat load will have the shortest heat-up time. The table showing the maximum fuel assembly heat generation rate for several years is located in Attachment A. The heat generation rates were calculated using the computer program HEATUP. Per Reference 2.5, the results in HEATUP are conservative compared to ORIGEN models.
4.5. Heat Load Reference 2.5 determines the maximum heat load from a single assembly. The assembly with the highest heat load will have the shortest heat-up time. The table showing the maximum fuel assembly heat generation rate for several years is located in Attachment A. The heat generation rates were calculated using the computer program HEATUP. Per Reference 2.5, the results in HEATUP are conservative compared to ORIGEN models.
: 5. Assumptions 5.1. All of the fuel assemblies are assumed to be Westinghouse 422V+ fuel. This is appropriate because the most recent design consisted of a full core of 422V+ assemblies (Ref. 2.3, pg. 3.2-22). The most recently offloaded assemblies are limiting in terns of heat generation.
: 5. Assumptions 5.1.
All of the fuel assemblies are assumed to be Westinghouse 422V+ fuel. This is appropriate because the most recent design consisted of a full core of 422V+ assemblies (Ref. 2.3, pg. 3.2-22). The most recently offloaded assemblies are limiting in terns of heat generation.
5.2. The properties of pure zirconium are used for the specific heat and deilsity of the zirconium alloy cladding. Based on an examination of alloys of some metals (e.g.
5.2. The properties of pure zirconium are used for the specific heat and deilsity of the zirconium alloy cladding. Based on an examination of alloys of some metals (e.g.
aluminum, nickel, or steel) in Table A. 1 of Reference 2.2, the density and specific heat are not significantly impacted by alloying.
aluminum, nickel, or steel) in Table A. 1 of Reference 2.2, the density and specific heat are not significantly impacted by alloying.
Line 460: Line 672:
5.4. The starting temperature for the heat-up analysis is assumed to be uniform and 90 'F (32 'C). A temperature of 90 'F is selected as representative of the current pool conditions (see Attachment C). The water temperature in the pool will continue to decrease over time due to a reduction in the heat load. It is appropriate to use a realistic value for the initial temperature due to the inherently conservative methodology (i.e. no heat transfer to the environment). In addition, this temperature is consistent with the sample analysis performed in SECY-99-168, where the starting temperature was 30 'C (86 'F).
5.4. The starting temperature for the heat-up analysis is assumed to be uniform and 90 'F (32 'C). A temperature of 90 'F is selected as representative of the current pool conditions (see Attachment C). The water temperature in the pool will continue to decrease over time due to a reduction in the heat load. It is appropriate to use a realistic value for the initial temperature due to the inherently conservative methodology (i.e. no heat transfer to the environment). In addition, this temperature is consistent with the sample analysis performed in SECY-99-168, where the starting temperature was 30 'C (86 'F).
5.5. The heat-up time is assumed to start when the spent fuel pool has been completely drained. This is conservative. It is likely that site personnel will start to respond to an incident when draindown starts.
5.5. The heat-up time is assumed to start when the spent fuel pool has been completely drained. This is conservative. It is likely that site personnel will start to respond to an incident when draindown starts.
                                                  &.'V'J~ndy' V.
&.'V' V.
J~ndy'


Document No. 2013-07050                                                 Kewaunee Power Station Revision 2                                                                            Paqe7 ofl1 I
Document No. 2013-07050 Revision 2 Kewaunee Power Station Paqe7 ofl1 I
: 6. Methodology This analysis determines the heat-up time of the fuel assembly using the thermal capacity of materials (Based on Section 2.3 of.Ref. 2.2).
: 6. Methodology This analysis determines the heat-up time of the fuel assembly using the thermal capacity of materials (Based on Section 2.3 of.Ref. 2.2).
4=P x V x Cp x AT                                             Equation 6-1 Where:
4=P x V x Cp x AT Equation 6-1 Where:
4 is the heat generation rate in BTU/hr 3
4 is the heat generation rate in BTU/hr p is the density of the material in lb/ft3 Vis the volume of the material in ft3 cp is the specific heat in BTU/lb-&deg;F AT is the temperature increase in 'F t is the heat-up time in hr For this analysis, there are two materials being heated: the uranium fuel pellets and the ZIRLO zirconium alloy cladding. The zirconium is in the cladding and the instrument tubes, which are also being heated. The zirconium and the uranium are modeled as heating up at the same rate, so the AT/t will be the same for both materials.
p is the density of the material in lb/ft 3
q=AT x &#xfd;px &#xfd;x cP'+
Vis the volume of the material in ft cp is the specific heat in BTU/lb-&deg;F AT is the temperature increase in 'F t is the heat-up time in hr For this analysis, there are two materials being heated: the uranium fuel pellets and the ZIRLO zirconium alloy cladding. The zirconium is in the cladding and the instrument tubes, which are also being heated. The zirconium and the uranium are modeled as heating up at the same rate, so the AT/t will be the same for both materials.
P:x Vx Cp)
q=AT x &#xfd;px &#xfd;x cP'+           P:x Vx Cp)                     Equation 6-2 I
I Where:
Where:
X,, signifies the property is for uranium XK signifies the property is for zirconium Equation 6-2 This calculation seeks the heat-up time, so Equation 6-2 is solved for t.
X,, signifies the property is for uranium XK signifies the property is for zirconium This calculation seeks the heat-up time, so Equation 6-2 is solved for t.
AT( x 1  
AT( x     1   ,x     +p       xV Equation 6-3 The volume of uranium is given below.
, x  
1x r= DD2jNhr, xL                                             Equation 6-4 Where:
+ p xV Equation 6-3 The volume of uranium is given below.
D, is the diameter of the uranium pellet Ni,. is the number of heated rods L is the heated length of the rods S*rrj"   C6 L, Indy z
r=
1x DD2jNhr, xL Where:
D, is the diameter of the uranium pellet Ni,. is the number of heated rods L is the heated length of the rods Equation 6-4 S*rrj" C6 L, Indy z


Document No. 2013-07050                                                     Kewaunee Power Station Revision 2                                                                               Page 8 of 10 The volumes of zirconium in the heated rods and in the guide tubes are given below. The length of the cladding and guide tubes that are heated is conservatively modeled as being the same as the heated length of uranium. The guide tubes and cladding are longer than the length of the uranium pellets.
Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 8 of 10 The volumes of zirconium in the heated rods and in the guide tubes are given below. The length of the cladding and guide tubes that are heated is conservatively modeled as being the same as the heated length of uranium. The guide tubes and cladding are longer than the length of the uranium pellets.
D.,&deg;2 _D j   "i V=
D.,&deg;2 _D j "i V=  
                    =7 x         c     -      Nh,. x L                         Equation 6-5 V    =
=7 x c
K [gx D* o" - Dgj 4
Nh,. x L Equation 6-5 D* o" - Dgj '"
Ng, x L                         Equation 6-6 V&#xfd; = Vg + V_."                                                 Equation 6-7 Where:
V
[g
= K x Ng, x L Equation 6-6 4
V&#xfd; = Vg + V_."
Equation 6-7 Where:
J'%, is the volume of zirconium in the cladding of heated tubes VP.g is the volume of zirconium in the guide tubes D0,o is the outer diameter of the cladding Dcj is the inner diameter of the cladding Dgo is the outer diameter of the guide tubes Dgji is the inner diameter of the guide tubes Ng, is the number of guide tubes The temperature increase (AT) for this analysis is taken to be from the initial temperature of the pool, 90 'F (Assumption 5.4), to the zirconium cladding failure temperatures of interest, 1049 'F and 1652 'F (Input 4.1).
J'%, is the volume of zirconium in the cladding of heated tubes VP.g is the volume of zirconium in the guide tubes D0,o is the outer diameter of the cladding Dcj is the inner diameter of the cladding Dgo is the outer diameter of the guide tubes Dgji is the inner diameter of the guide tubes Ng, is the number of guide tubes The temperature increase (AT) for this analysis is taken to be from the initial temperature of the pool, 90 'F (Assumption 5.4), to the zirconium cladding failure temperatures of interest, 1049 'F and 1652 'F (Input 4.1).
The heat-up time is calculated as a function of the decay time.
The heat-up time is calculated as a function of the decay time.
To avoid rounding, the Hottest Assembly column is recalculated in Attachment A based on the equations presented in Reference 2.5. Per Reference 2.5, the hottest assembly is calculated as:
To avoid rounding, the Hottest Assembly column is recalculated in Attachment A based on the equations presented in Reference 2.5. Per Reference 2.5, the hottest assembly is calculated as:
Hottest Assembly (Heat Load from Cycle 32 Discharge Assemblies x 1.449 HtsA b                                   121                       .
Hottest Assembly (Heat Load from Cycle 32 Discharge Assemblies x 1.449 HtsA b 121 f_--g~n L&c L..ndy
L..ndy f_--g~n L&c


Document No. 2013-07050                                                   Kewaunee Power Station Revision 2                                                                          Paqe9of 10
Document No. 2013-07050 Revision 2 Kewaunee Power Station Paqe9of 10
: 7. Results The results are shown in Table 7-1 below (from Attachment B).
: 7. Results The results are shown in Table 7-1 below (from Attachment B).
Table 7-1: Results Date           End Temperature       Decay Time Heat-Up Time (0 C, 'F)           (months)       (hours)
Table 7-1: Results Date End Temperature Decay Time Heat-Up Time (0C, 'F)
October 4th, 2013         565, 1049                 -5           2.0 April 8th, 2014         565, 1049               - 11         4.0 July 7tt, 2014         565, 1049                 14         4.9 October 7t, 2014         565, 1049                 17         6.0 August 21s t , 2015       565, 1049               -28         10.0 July 18th, 2013         900, 1652               -2           2.0 November 16t', 2013         900, 1.652               -6           4.0 March 11"h, 2014         900, 1652               -10           6.0 October 21 st , 2014       900, 1652               -17         10.0 The 10 hour heat-up time to a temperature of 565 'C (1049 'F) occurs at a decay time of under 2.5 years, which is the expected decay time to a temperature of 900 'C (1652 'F) stated in SECY-99-168 (Ref. 2.4). The 10 hour heat-up time to a temperature of 900 'C (I 652 'F) occurs at a decay time of roughly 1.5 years, which is less than the expected decay time calculated in NUREG-1738 (Ref. 2.7, pg. 2-3).
(months)
(hours)
October 4th, 2013 565, 1049  
-5 2.0 April 8th, 2014 565, 1049  
- 11 4.0 July 7tt, 2014 565, 1049 14 4.9 October 7t, 2014 565, 1049 17 6.0 August 21s t, 2015 565, 1049  
-28 10.0 July 18th, 2013 900, 1652  
-2 2.0 November 16t', 2013 900, 1.652  
-6 4.0 March 11"h, 2014 900, 1652  
-10 6.0 October 21 st, 2014 900, 1652  
-17 10.0 The 10 hour heat-up time to a temperature of 565 'C (1049 'F) occurs at a decay time of under 2.5 years, which is the expected decay time to a temperature of 900 'C (1652 'F) stated in SECY-99-168 (Ref. 2.4). The 10 hour heat-up time to a temperature of 900 'C (I 652 'F) occurs at a decay time of roughly 1.5 years, which is less than the expected decay time calculated in NUREG-1738 (Ref. 2.7, pg. 2-3).
A plot showing the heat-up time to the temperatures of interest as a function of decay time is Figure 7-1.
A plot showing the heat-up time to the temperatures of interest as a function of decay time is Figure 7-1.
S;   Luncldy
S;  
                                              &&#xfd;M;r
&&#xfd;M;r Luncldy


Document No. 2013-07050                                                         Kewaunee Power Station Revision 2                                                                                  Paae 10 of 10 Figure 7-1: Heat-Up Time vs. Decay Time 12 0) 10 16 8
Document No. 2013-07050 Revision 2 Kewaunee Power Station Paae 10 of 10 Figure 7-1: Heat-Up Time vs. Decay Time 12 0) 10 16 8
E 0
E 0
D 0 U
D 0 U
C4, E
C4, E
2     4 U.
4 2
U.
0E CL R0 0'
0E CL R0 0'
1/1/2013   7/2/2013   1/1/2014       7/2/2014     1/1/2015 7/3/2015   1/1/2016   7/2/2016 Day (Shutdown was May 7, 2013)
1/1/2013 7/2/2013 1/1/2014 7/2/2014 1/1/2015 7/3/2015 1/1/2016 7/2/2016 Day (Shutdown was May 7, 2013)
: 8. Conclusions and Recommendations The Kewaunee results are more favorable than the analyses performed for SECY-99-168 (Ref 2.4) and NUREG-1738 (Ref 2.7). There are no acceptance criteria for this analysis.
: 8. Conclusions and Recommendations The Kewaunee results are more favorable than the analyses performed for SECY-99-168 (Ref 2.4) and NUREG-1738 (Ref 2.7). There are no acceptance criteria for this analysis.
There are no specific recommendations for this analysis.
There are no specific recommendations for this analysis.
The primary input to this analysis is the heat generation rate, which is conservative. The heat generation rates were calculated using the computer program HEATUP. Per Reference 2.5, the results in HEATUP are conservative compared to ORIGEN models.
The primary input to this analysis is the heat generation rate, which is conservative. The heat generation rates were calculated using the computer program HEATUP. Per Reference 2.5, the results in HEATUP are conservative compared to ORIGEN models.
                                            .nr'tont   &. LArndy',
.nr'tont  
&. LArndy',


Calculation 2013-07050 Rev. 0 Kewaunee Power Station                                                           Pg Pg       lo lo Attachment A: Heat Generation Rate vs. DcyTime (from Ref. 2.5 Heat Load from   Hottest Fuel                               Recalculated Cycle 32 Discharge Assembly                                     Hottest Assemblies Only   Estimate                 Date Days since Assembly Date   Time     (MBTU/hr)     (MBTU/hr)   __(Reprinted)     May 8, 2013 (MBTU/hr) 5/8/2013 0:00         32.25         0.386     _        5/8/2013       0-     0.3862 5/8/2013 18:00       28.23           0.338           1_ 5/8/2013       0.33     0.3381 5/8/2013 16:00       26.62           0.319     _        5/8/2013     0.67     0.3188 5/9/2013   0:00       25.39           0.304               5/9/2013       1       0.3041 5/9/2013   8:00       24.36           0.292     _        5/9/2013     1.33     0.2917 5/9/2013 16:00       23.45           0.281               5/9/2013     1.67     0.2808 5/10/2013 0:00         22.62           0.271             5/10/2013       2       0.2709 5/10/2013 8:00         21.86           0.262           1_5/10/2013       2.33     0.2618 5/10/2013 16:00       21.15           0.253             5/10/2013     2.67     0.2533 5/11/2013 0.00         20.5           0.246             5/11/2013       3       0.2455 5/11/2013 8:00           19.9         0.238     _        5/11/2013     3.33     0.2383 5/11/2013 16:00         19.33         0.232             5/11/2013     3.67     0.2315 5/12/2013 0:00         18.81         0.225     __5/12/2013               4       0.2253 5/13/2013 0:00         17.44         0.209           15/13/2013         5       0.2088 5/14/2013 0:00           16.3         0.195             5/14/2013       6       0.1952 5/15/2013 0:00         15.34         0.184             5/15/2013       7       0.1837 5/16/2013 0:00         14.52         0.174             5/16/2013       8       0.1739 5/17/2013 0:00         13.81         0.165             5/17/2013 ___9           0.1654 5/18/2013 0:00         13.19         0.158             5/18/2013       10       0.1580 5/19/2013 10:00         12.65         0.151         15/19/2013         11       0.1515 5/20/2013 0:00         12.16         0.146             5/20/2013       12       0.1456 5/21/2013 0:00         11.73         0.140             5/21/2013       13       0.1405 5/22/2013 0:00           11.34         0.136             5/22/2013       14       0.1358 5/23/2013 0:00           10.99         0.132       -5/23/2013             15       0.1316 5/24/2013 0:00           10.67         0.128             5/24/2013       16       0.1278 5/25/2013 10:00         10.38         0.124         1 5/25/2013         17       0.1243 5/26/2013 0:00           10.11         0.121             5/26/2013       18       0.1211 5/27/2013 0:00           9.87           0.118             5/27/2013       19       0.1182 5/28/2013 0:00           9.64           0.115             5/28/2013       20       0.1154 5/29/2013 0:00           9.43           0.113             5/29/2013       21       0.1129 5/30/2013 0:00           9.24           0.111             5/30/2013       22       0.1107 5/31/2013 10:00         9.06           0.108         15/31/2013         23       0.1085 6/1/2013   0:00         8.88           0.106               6/1/2013       24       0.1063 6/3/2013   0:00         8.57           0.103             6/3/2013       26       0.1026 6/5/2013   0:00         8.29           0.099             6/5/2013       28       0.0993 6/7/2013   0:00         8.04           0.096             6/7/2013       30       0.0963 6/9/2013   0:00         7.8           0.093             6/9/2013       32       0.0934 6/11/2013 0:00           7.59           0.091             6/11/2013       34       0.0909 6/13/2013 10:00         7.38           0.088         16/13/2013         36       0.0884 6/15/2013 0:00           7.19           0.086             6/15/2013       38       0.0861 6/17/2013 0:00           7.01           0.084             6/17/2013       40       0.0839 6/19/2013 0:00           6.84           0.082             6/19/2013       42       0.0819 6/21/2013 0:00           6.68         0.080             6/21/2013       44       0.0800 6./25/2013 0:00         6.38         0.076             6/25/2013       48       0.0764 6/29/2013 0:00           6.11         0.073             6/29/2013       52       0.0732 7/3/2013   0:00         5.86         0.070               7/3/2013       56       0.0702
Calculation 2013-07050 Rev. 0 Pg lo Kewaunee Power Station Pg lo Attachment A: Heat Generation Rate vs. DcyTime (from Ref. 2.5 Heat Load from Hottest Fuel Recalculated Cycle 32 Discharge Assembly Hottest Assemblies Only Estimate Date Days since Assembly Date Time (MBTU/hr)
(MBTU/hr)
__(Reprinted)
May 8, 2013 (MBTU/hr) 5/8/2013 0:00 32.25 0.386 5/8/2013 0-0.3862 5/8/2013 18:00 28.23 0.338 1_ 5/8/2013 0.33 0.3381 5/8/2013 16:00 26.62 0.319 5/8/2013 0.67 0.3188 5/9/2013 0:00 25.39 0.304 5/9/2013 1
0.3041 5/9/2013 8:00 24.36 0.292 5/9/2013 1.33 0.2917 5/9/2013 16:00 23.45 0.281 5/9/2013 1.67 0.2808 5/10/2013 0:00 22.62 0.271 5/10/2013 2
0.2709 5/10/2013 8:00 21.86 0.262 1_ 5/10/2013 2.33 0.2618 5/10/2013 16:00 21.15 0.253 5/10/2013 2.67 0.2533 5/11/2013 0.00 20.5 0.246 5/11/2013 3
0.2455 5/11/2013 8:00 19.9 0.238 5/11/2013 3.33 0.2383 5/11/2013 16:00 19.33 0.232 5/11/2013 3.67 0.2315 5/12/2013 0:00 18.81 0.225
__5/12/2013 4
0.2253 5/13/2013 0:00 17.44 0.209 15/13/2013 5
0.2088 5/14/2013 0:00 16.3 0.195 5/14/2013 6
0.1952 5/15/2013 0:00 15.34 0.184 5/15/2013 7
0.1837 5/16/2013 0:00 14.52 0.174 5/16/2013 8
0.1739 5/17/2013 0:00 13.81 0.165 5/17/2013
___9 0.1654 5/18/2013 0:00 13.19 0.158 5/18/2013 10 0.1580 5/19/2013 10:00 12.65 0.151 15/19/2013 11 0.1515 5/20/2013 0:00 12.16 0.146 5/20/2013 12 0.1456 5/21/2013 0:00 11.73 0.140 5/21/2013 13 0.1405 5/22/2013 0:00 11.34 0.136 5/22/2013 14 0.1358 5/23/2013 0:00 10.99 0.132  
-5/23/2013 15 0.1316 5/24/2013 0:00 10.67 0.128 5/24/2013 16 0.1278 5/25/2013 10:00 10.38 0.124 1 5/25/2013 17 0.1243 5/26/2013 0:00 10.11 0.121 5/26/2013 18 0.1211 5/27/2013 0:00 9.87 0.118 5/27/2013 19 0.1182 5/28/2013 0:00 9.64 0.115 5/28/2013 20 0.1154 5/29/2013 0:00 9.43 0.113 5/29/2013 21 0.1129 5/30/2013 0:00 9.24 0.111 5/30/2013 22 0.1107 5/31/2013 10:00 9.06 0.108 15/31/2013 23 0.1085 6/1/2013 0:00 8.88 0.106 6/1/2013 24 0.1063 6/3/2013 0:00 8.57 0.103 6/3/2013 26 0.1026 6/5/2013 0:00 8.29 0.099 6/5/2013 28 0.0993 6/7/2013 0:00 8.04 0.096 6/7/2013 30 0.0963 6/9/2013 0:00 7.8 0.093 6/9/2013 32 0.0934 6/11/2013 0:00 7.59 0.091 6/11/2013 34 0.0909 6/13/2013 10:00 7.38 0.088 16/13/2013 36 0.0884 6/15/2013 0:00 7.19 0.086 6/15/2013 38 0.0861 6/17/2013 0:00 7.01 0.084 6/17/2013 40 0.0839 6/19/2013 0:00 6.84 0.082 6/19/2013 42 0.0819 6/21/2013 0:00 6.68 0.080 6/21/2013 44 0.0800 6  
./25/2013 0:00 6.38 0.076 6/25/2013 48 0.0764 6/29/2013 0:00 6.11 0.073 6/29/2013 52 0.0732 7/3/2013 0:00 5.86 0.070 7/3/2013 56 0.0702


Calculation 2013-07050 Rev. 0 Kewaunee Power Station                                                           Page A2 of 6 Attachment A: Heat Generation Rate vs. DcyTime (from Ref.IL5 Heat Load from Hottest Fuel                                 Recalculated Cycle 32 Discharge Assembly                                       Hottest Assemblies Only   Estimate                   Date Days since   Assembly Date   Time         (MBTU/hr) (MBTU/hr)     __(Reprinted)     May 8, 2013 (MBTU/hr) 7/7/2013   0:00           5.64       0.068           17/7/2013         60         0.0675 7/11/2013 10:00           5.43       0.065               7/11/2013     64         0.0650 7/15/2013 0:00             5.24       0.063               7/15/2013     68         0.0628 7/19/2013 0:00             5.07       0.061               7/19/2013     72         0.0607 7/23/2013 0:00             4.91       0.059               7/23/2013     76         0.0588 7/27/2013 0:00             4.76       0.057               7/27/2013     80         0.0570 8/6/2013   0:00           4.42       0.053           18/6/2013         90         0.0529 8/16/2013 10:00           4.13       0.049             8/16/2013     100         0.0495 8/26/2013 0:00             3.88       0.046             8/26/2013     110         0.0465 9/5/2013   0:00           3.66       0.044             .9/5/2013     120         0.0438 9/15/2013 0:00             3.46       0.041             9/15/2013     130         0.0414 9/25/2013 0:00             3.27       0.039             9/25/2013     140         0.0392 10/5/2013 0:00             3.11       0.037               10/5/2013   150         0.0372 10/15/20131 0:00           2.96       0.035           110/15/2013     160         0.0354 10/25/2013 0:00             2.82       0.034             10/25/2013     170         0.0338 11/4/2013 0:00             2.69       0.032               11/4/2013   180         0.0322 11/24/2013 0:00             2.46       0.030             11/24/2013     200         0.0295 12/14/2013 0:00             2.27       0.027             12/14/2013     220         0.0272 1/3/2014 0:00 ___        2.1       0.025               1/3/2014   240         0.0251 1/23/2014 10:00 ____1.96               0.023           11/23/2014     260         0.0235 2/12/2014 0:00   ____1.84               0.022             2/12/2014     280         0.0220 3/4/2014   0:00 ____1.73               0.021               3/4/2014   300         0.0207 3/24/2014 0:00   ____1.63               0.020             3/24/2014     320         0.0195 4/13/2014 0:00   ____1.54               0.018             4/13/2014     340         0.0184 5/3/2014   0:00 ____1.47               0.018               5/3/2014   360         0.0176 5/23/2014 0:00   ___        1.4       0.017           15/23/2014       380         0.0168 6/12/2014 0:00   ___        1.33       0.016             6/12/2014     400         0.0159 7/2/2014   0:00 ___        1.28       0.015               7/2/2014   420         0.0153 7/22/2014 0:00   ____1.22               0.015             7/22/2014     440         0.0146 8/11/2014 0:00   ____1.17               0.014             8/11/2014     460         0.0140 8/31/2014 10:00             1.13       0.013             8/31/2014     480         0.0135 9/20/2014 0:00             1.08       0.013           19/20/2014       500         0.0129 10/10/2014 0:00             1.04       0.012           110/10/2014     520         0.0125 10/30/2014 0:00               1         0.012             10/30/2014     540         0.0120 11/19/2014 0:00             0.97       0.012             11/19/2014     560         0.0116 12/9/2014 0:00             0.93       0.011               12/9/2014   580         0.0111 12/29/2014 10:00             0.9       0.011             12/29/2014     600         0.0108 1/18/2015 0:00             0.87       0.010               1/18/2015   620         0.0104 2/7/2015   0:00           0.84       0.010           12/7/2015       640         0.0101 2/27/2015 0:00             0.81       0.010               2/27/2015     660         0.0097 3/19/2015 0:00             0.79       0.009               3/19/2015     680         0.0095 4/8/2015   0:00           0.76       0.009               4/8/2015     700         0.0091 4/28/2015 10:00             0.74       0.009               4/28/2015     720         0.0089 5/18/2015 0:00             0.72       0.009               5/18/2015     740         0.0086 6/7/2015   0:00             0.7       0.008     116/7/2015             760         0.0084 6/27/2015 0:00             0.68       0.008     116/27/2015           780         0.0081
Calculation 2013-07050 Rev. 0 Kewaunee Power Station Page A2 of 6 Attachment A: Heat Generation Rate vs. DcyTime (from Ref.IL5 Heat Load from Hottest Fuel Recalculated Cycle 32 Discharge Assembly Hottest Assemblies Only Estimate Date Days since Assembly Date Time (MBTU/hr)
(MBTU/hr)
__(Reprinted)
May 8, 2013 (MBTU/hr) 7/7/2013 0:00 5.64 0.068 17/7/2013 60 0.0675 7/11/2013 10:00 5.43 0.065 7/11/2013 64 0.0650 7/15/2013 0:00 5.24 0.063 7/15/2013 68 0.0628 7/19/2013 0:00 5.07 0.061 7/19/2013 72 0.0607 7/23/2013 0:00 4.91 0.059 7/23/2013 76 0.0588 7/27/2013 0:00 4.76 0.057 7/27/2013 80 0.0570 8/6/2013 0:00 4.42 0.053 18/6/2013 90 0.0529 8/16/2013 10:00 4.13 0.049 8/16/2013 100 0.0495 8/26/2013 0:00 3.88 0.046 8/26/2013 110 0.0465 9/5/2013 0:00 3.66 0.044  
.9/5/2013 120 0.0438 9/15/2013 0:00 3.46 0.041 9/15/2013 130 0.0414 9/25/2013 0:00 3.27 0.039 9/25/2013 140 0.0392 10/5/2013 0:00 3.11 0.037 10/5/2013 150 0.0372 10/15/20131 0:00 2.96 0.035 110/15/2013 160 0.0354 10/25/2013 0:00 2.82 0.034 10/25/2013 170 0.0338 11/4/2013 0:00 2.69 0.032 11/4/2013 180 0.0322 11/24/2013 0:00 2.46 0.030 11/24/2013 200 0.0295 12/14/2013 0:00 2.27 0.027 12/14/2013 220 0.0272 1/3/2014 0:00 2.1 0.025 1/3/2014 240 0.0251 1/23/2014 10:00
____1.96 0.023 11/23/2014 260 0.0235 2/12/2014 0:00
____1.84 0.022 2/12/2014 280 0.0220 3/4/2014 0:00
____1.73 0.021 3/4/2014 300 0.0207 3/24/2014 0:00
____1.63 0.020 3/24/2014 320 0.0195 4/13/2014 0:00
____1.54 0.018 4/13/2014 340 0.0184 5/3/2014 0:00
____1.47 0.018 5/3/2014 360 0.0176 5/23/2014 0:00 1.4 0.017 15/23/2014 380 0.0168 6/12/2014 0:00 1.33 0.016 6/12/2014 400 0.0159 7/2/2014 0:00 1.28 0.015 7/2/2014 420 0.0153 7/22/2014 0:00
____1.22 0.015 7/22/2014 440 0.0146 8/11/2014 0:00
____1.17 0.014 8/11/2014 460 0.0140 8/31/2014 10:00 1.13 0.013 8/31/2014 480 0.0135 9/20/2014 0:00 1.08 0.013 19/20/2014 500 0.0129 10/10/2014 0:00 1.04 0.012 110/10/2014 520 0.0125 10/30/2014 0:00 1
0.012 10/30/2014 540 0.0120 11/19/2014 0:00 0.97 0.012 11/19/2014 560 0.0116 12/9/2014 0:00 0.93 0.011 12/9/2014 580 0.0111 12/29/2014 10:00 0.9 0.011 12/29/2014 600 0.0108 1/18/2015 0:00 0.87 0.010 1/18/2015 620 0.0104 2/7/2015 0:00 0.84 0.010 12/7/2015 640 0.0101 2/27/2015 0:00 0.81 0.010 2/27/2015 660 0.0097 3/19/2015 0:00 0.79 0.009 3/19/2015 680 0.0095 4/8/2015 0:00 0.76 0.009 4/8/2015 700 0.0091 4/28/2015 10:00 0.74 0.009 4/28/2015 720 0.0089 5/18/2015 0:00 0.72 0.009 5/18/2015 740 0.0086 6/7/2015 0:00 0.7 0.008 116/7/2015 760 0.0084 6/27/2015 0:00 0.68 0.008 116/27/2015 780 0.0081


Calculation 2013-07050 Rev. 0 Kewaunee Power Station                                                   Page A3 of 6 Attachment A: Heat Generation Rate vs. Decay Time (from Ref. 2.5)
Calculation 2013-07050 Rev. 0 Kewaunee Power Station Page A3 of 6 Attachment A: Heat Generation Rate vs. Decay Time (from Ref. 2.5)
Heat Load from   Hottest Fuel                         Recalculated Cycle 32 Discharge Assembly                               Hottest Assemblies Only   Estimate         Date   Days since   Assembly Date   Time     (MBTU/hr)     (MBTU/hr)     (Reprinted) May 8, 2013 (MBTU/hr) 7/17/2015 0:00       0.66           0.008     1 7/17/2015     800       0.0079 8/6/2015 0:00       0.64           0.008       8/6/2015     820       0.0077 8/26/2015 0:00       0.62           0.007       8/26/2015     840       0.0074 9/15/2015 0:00         0.6           0.007       9/15/2015     860       0.0072 10/5/2015 0:00       0.59           0.007       10/5/2015     880       0.0071 10/25/2015 0:00       0.57         0.007       10/25/2015     900       0.0068 11/14/2015 0:00       0.56           0.007     111/14/2015     920       0.0067 12/4/2015 0:00       0.54         0.007       12/4/2015     940       0.0065 12/24/2015 0:00       0.53         0.006       12/24/2015     960       0.0063 1/13/2016 0:00       0.52         0.006       1/13/2016     980       0.0062 2/2/2016   0:00         0.5         0.006         2/2/2016     1000       0.0060 2/22/2016 0:00       0.49         0.006       2/22/2016     1020       0.0059 3/13/2016 0:00       0.48         0.006       3/13/2016     1040       0.0057 4/2/2016   0:00       0.47         0.006     1 4/2/2016     1060       0.0056 4/22/2016   0:00       0.46         0.005       4/22/2016     1080       0.0055
Heat Load from Hottest Fuel Recalculated Cycle 32 Discharge Assembly Hottest Assemblies Only Estimate Date Days since Assembly Date Time (MBTU/hr)
(MBTU/hr)
(Reprinted) May 8, 2013 (MBTU/hr) 7/17/2015 0:00 0.66 0.008 1 7/17/2015 800 0.0079 8/6/2015 0:00 0.64 0.008 8/6/2015 820 0.0077 8/26/2015 0:00 0.62 0.007 8/26/2015 840 0.0074 9/15/2015 0:00 0.6 0.007 9/15/2015 860 0.0072 10/5/2015 0:00 0.59 0.007 10/5/2015 880 0.0071 10/25/2015 0:00 0.57 0.007 10/25/2015 900 0.0068 11/14/2015 0:00 0.56 0.007 111/14/2015 920 0.0067 12/4/2015 0:00 0.54 0.007 12/4/2015 940 0.0065 12/24/2015 0:00 0.53 0.006 12/24/2015 960 0.0063 1/13/2016 0:00 0.52 0.006 1/13/2016 980 0.0062 2/2/2016 0:00 0.5 0.006 2/2/2016 1000 0.0060 2/22/2016 0:00 0.49 0.006 2/22/2016 1020 0.0059 3/13/2016 0:00 0.48 0.006 3/13/2016 1040 0.0057 4/2/2016 0:00 0.47 0.006 1 4/2/2016 1060 0.0056 4/22/2016 0:00 0.46 0.005 4/22/2016 1080 0.0055


Calculation 2013-07050 Rev. 0 Kewaunee Power Station                                                               Page A4 of 6 Attachment A: Heat Generation Rate vs. Deca' Time (from Ref. 2.5 Heat Load from Cycle   Hottest Fuel 32 Discharge       Assembly Assemblies Only       Estimate       Date     Days since Recalculated Hottest Date     Time           (MBTU/hr)         (MBTU/hr)     (Reprinted) May 8, 2013 Assembly (MBTU/hr) 41402   0         32.25                 0.386           =A3+B3       =0         =(C3/121)*1.449 41402   0.3333333, 28.23                 0.338           =A4+64       =F4-F$3     =(C4/121)*1.449 41402   0.6666666E 26.62                 0.319           =A5+B5       =F5-F$3     =(C5/121)*1.449 41403   0         25.39                 0.304           =A6+B6       =F6-F$3     =(C6/121)*1.449 41403   0.3333333" 24.36                 0.292           =A7+B7       =F7-F$3     =(C7/121)'1.449 41403   0.6666666E 23.45                 0.281           =A8+B8       =F8-F$3     =(C8/121)*1.449 41404   0         22.62                 0.271           =A9+B9       =F9-F$3     =(C9/121)*1.449 41404   0.33333333 21.86                 0.262           =A10+B10     =F10-F$3   =(C1O/121)*1.449 41404   0.6666666E 21.15                 0.253           =A11+B11     =F11-F$3   =(C11/121)*1.449 41405   0         20.5                 0.246           =A12+B12     =F12-F$3   =(C12/121)*1.449 41405   0.33333332 19.9                 0.238           =A13+B13     =F13-F$3   =(C13/121)*1.449 41405   0.6666666E 19.33                 0.232           =A14+B14     =F14-F$3   =(C14/121)*1.449 41406   0         18.81                 0.225           =A15+B15     =F15-F$3   =(C15/121)*1.449 41407   0         17.44                 0.209           =A16+B16     =F16-F$3   =(C16/121)*1.449 41408   0         16.3                 0.195           =A17+B17     =F17-F$3   =(C17/121)*1.449 41409   0         15.34                 0.184           =A18+B18     =F18-F$3   =(C18/121)*1.449 41410   0         14.52                 0.174           =A19+B19     =F19-F$3   =(C19/121)*1.449 41411   0         13.81                 0.165           =A20+B20     =F20-F$3   =(C20/121)'1.449 41412   0         13.19                 0.158         1=A21+B21     =F21-F$3   =(C21/121)*1.449 41413   0         12.65                 0.151           =A22+B22     =F22-F$3   =(C22/121)*1.449 41414   0         12.16                 0.146           =A23+B23     =F23-F$3   =(C23/121)*1.449 41415   0         11.73                 0.14           =A24+B24     =F24-F$3   =(C24/121)*1.449 41416   0         11.34                 0.136           =A25+B25     =F25-F$3   =(C25/121)*1.449 41417   0         10.99                 0.132           =A26+B26     =F26-F$3   =(C26/121)*1.449 41418
Calculation 2013-07050 Rev. 0 Kewaunee Power Station Page A4 of 6 Attachment A: Heat Generation Rate vs. Deca' Time (from Ref. 2.5 Heat Load from Cycle Hottest Fuel 32 Discharge Assembly Assemblies Only Estimate Date Days since Recalculated Hottest Date Time (MBTU/hr)
_        0          10.67                 0.128           =A27+B27     =F27-F$3   =(C27/121)*1.449 41419   0         10.38                 0.124           =A28+B28     =F28-F$3   =(C28/121)*1.449 41420   0         10.11                 0.121           =A29+B29     =F29-F$3   =(C29/121)'1.449 41421   0         9.87                 0.118           =A30+B30     =F30-F$3   =(C30/121)*1.449 41422   0         9.64                 0.115           =A31+B31     =F31-F$3   =(C31/121)*1.449 41423   0         9.43                 0.113           =A32+B32     =F32-F$3   =(C32/121)*1.449 41424   0         9.24                 0.111           =A33+B33     =F33-F$3   =(C33/121)*1.449 41425   0         9.06                 0.108           =A34+B34     =F34-F$3   =(C34/121)*1.449 41426   0         8.88                 0.106           =A35+B35     =F35-F$3   =(C35/121)*1.449 41428   0         8.57                 0.103           =A36+B36     =F36-F$3   =(C36/121)*1.449 41430   0         8.29                 0.099           =A37+B37     =F37-F$3   =(C37/121)*1.449 41432   0         8.04                 0.096           =A38+B38     =F38-F$3   =(C38/121)*1.449 41434   0         7.8                   0.093           =A39+B39     =F39-F$3   =(C39/121)*1.449 41436   0         7.59                 0.091           =A40+1B40   =F40-F$3   =(C40/121)*1.449 41438   0         7.38                 0.088           =A41+B41     =F41-F$3   =(C41/121)*1.449 41440   0         7.19                 0.086           =A42+B42     =F42-F$3   =(C42/121)*1.449 41442   0         7.01                 0.084           =A43+B43     =F43-F$3   =(C43/121)*1.449 41444   0         6.84                 0.082           =A44+B44     =F44-F$3   =(C44/121)*1.449 41446   0         6.68                 0.08           =A45+B45     =F45-F$3   =(C45/121)*1.449 41450   0         6.38                 0.076           =A46+B46     =F46-F$3   =(C46/121)*1.449 41454   0         6.11                 0.073           =A47+B47     =F47-F$3   =(C47/121)*1.449 41458   0         5.86                 0.07           =A48+B48     =F48-F$3   =(C48/121)*1.449 41462   0         5.64                 0.068           =A49+B49     =F49-F$3   =(C49/121)*1.449 41466   0         5.43                 0.065           =A50+B50     =F50-F$3   =(C50/121)*1.449
(MBTU/hr)
(Reprinted)
May 8, 2013 Assembly (MBTU/hr) 41402 0
32.25 0.386  
=A3+B3  
=0  
=(C3/121)*1.449 41402 0.3333333, 28.23 0.338  
=A4+64  
=F4-F$3  
=(C4/121)*1.449 41402 0.6666666E 26.62 0.319  
=A5+B5  
=F5-F$3  
=(C5/121)*1.449 41403 0
25.39 0.304  
=A6+B6  
=F6-F$3  
=(C6/121)*1.449 41403 0.3333333" 24.36 0.292  
=A7+B7  
=F7-F$3  
=(C7/121)'1.449 41403 0.6666666E 23.45 0.281  
=A8+B8  
=F8-F$3  
=(C8/121)*1.449 41404 0
22.62 0.271  
=A9+B9  
=F9-F$3  
=(C9/121)*1.449 41404 0.33333333 21.86 0.262  
=A10+B10  
=F10-F$3  
=(C1O/121)*1.449 41404 0.6666666E 21.15 0.253  
=A11+B11  
=F11-F$3  
=(C11/121)*1.449 41405 0
20.5 0.246  
=A12+B12  
=F12-F$3  
=(C12/121)*1.449 41405 0.33333332 19.9 0.238  
=A13+B13  
=F13-F$3  
=(C13/121)*1.449 41405 0.6666666E 19.33 0.232  
=A14+B14  
=F14-F$3  
=(C14/121)*1.449 41406 0
18.81 0.225  
=A15+B15  
=F15-F$3  
=(C15/121)*1.449 41407 0
17.44 0.209  
=A16+B16  
=F16-F$3  
=(C16/121)*1.449 41408 0
16.3 0.195  
=A17+B17  
=F17-F$3  
=(C17/121)*1.449 41409 0
15.34 0.184  
=A18+B18  
=F18-F$3  
=(C18/121)*1.449 41410 0
14.52 0.174  
=A19+B19  
=F19-F$3  
=(C19/121)*1.449 41411 0
13.81 0.165  
=A20+B20  
=F20-F$3  
=(C20/121)'1.449 41412 0
13.19 0.158 1 =A21+B21  
=F21-F$3  
=(C21/121)*1.449 41413 0
12.65 0.151  
=A22+B22  
=F22-F$3  
=(C22/121)*1.449 41414 0
12.16 0.146  
=A23+B23  
=F23-F$3  
=(C23/121)*1.449 41415 0
11.73 0.14  
=A24+B24  
=F24-F$3  
=(C24/121)*1.449 41416 0
11.34 0.136  
=A25+B25  
=F25-F$3  
=(C25/121)*1.449 41417 0
10.99 0.132  
=A26+B26  
=F26-F$3  
=(C26/121)*1.449 41418 0
10.67 0.128  
=A27+B27  
=F27-F$3  
=(C27/121)*1.449 41419 0
10.38 0.124  
=A28+B28  
=F28-F$3  
=(C28/121)*1.449 41420 0
10.11 0.121  
=A29+B29  
=F29-F$3  
=(C29/121)'1.449 41421 0
9.87 0.118  
=A30+B30  
=F30-F$3  
=(C30/121)*1.449 41422 0
9.64 0.115  
=A31+B31  
=F31-F$3  
=(C31/121)*1.449 41423 0
9.43 0.113  
=A32+B32  
=F32-F$3  
=(C32/121)*1.449 41424 0
9.24 0.111  
=A33+B33  
=F33-F$3  
=(C33/121)*1.449 41425 0
9.06 0.108  
=A34+B34  
=F34-F$3  
=(C34/121)*1.449 41426 0
8.88 0.106  
=A35+B35  
=F35-F$3  
=(C35/121)*1.449 41428 0
8.57 0.103  
=A36+B36  
=F36-F$3  
=(C36/121)*1.449 41430 0
8.29 0.099  
=A37+B37  
=F37-F$3  
=(C37/121)*1.449 41432 0
8.04 0.096  
=A38+B38  
=F38-F$3  
=(C38/121)*1.449 41434 0
7.8 0.093  
=A39+B39  
=F39-F$3  
=(C39/121)*1.449 41436 0
7.59 0.091  
=A40+1B40  
=F40-F$3  
=(C40/121)*1.449 41438 0
7.38 0.088  
=A41+B41  
=F41-F$3  
=(C41/121)*1.449 41440 0
7.19 0.086  
=A42+B42  
=F42-F$3  
=(C42/121)*1.449 41442 0
7.01 0.084  
=A43+B43  
=F43-F$3  
=(C43/121)*1.449 41444 0
6.84 0.082  
=A44+B44  
=F44-F$3  
=(C44/121)*1.449 41446 0
6.68 0.08  
=A45+B45  
=F45-F$3  
=(C45/121)*1.449 41450 0
6.38 0.076  
=A46+B46  
=F46-F$3  
=(C46/121)*1.449 41454 0
6.11 0.073  
=A47+B47  
=F47-F$3  
=(C47/121)*1.449 41458 0
5.86 0.07  
=A48+B48  
=F48-F$3  
=(C48/121)*1.449 41462 0
5.64 0.068  
=A49+B49  
=F49-F$3  
=(C49/121)*1.449 41466 0
5.43 0.065  
=A50+B50  
=F50-F$3  
=(C50/121)*1.449


Calculation 2013-07050 Rev. 0 Kewaunee Power Station                                                                   Page A5 of 6 Attachment A: Heat Generation Rate vs. Deca' Time (from Ref. 2.5)
Calculation 2013-07050 Rev. 0 Kewaunee Power Station Page A5 of 6 Attachment A: Heat Generation Rate vs. Deca' Time (from Ref. 2.5)
Heat Load from Cycle   Hottest Fuel 32 Discharge       Assembly Assemblies Only       Estimate               Date     Days since Recalculated Hottest Date     Time       (MBTU/hr)         (MBTU/hr)           (Reprinted) May 8, 2013 Assembly (MBTU/hr) 41470     0     5.24                 0.063                 =A51+B51     =F51-F$3   =(C51/121)*1.449 41474     0     5.07                 0.061                 =A52+B52     =F52-F$3   =(C52/121)*1.449 41478     0     4.91                 0.059           _=A53+B53         =F53-F$3   =(C53/121)'1.449 41482     0     4.76                 0.057                 =A54+B54     =F54-F$3   =(C54/121)*1.449 41492     0     4.42                 0.053                 =A55+B55     =F55-F$3   =(C55/121)*1.449 41502     0     4.13                 0.049                 =A56+B56     =F56-F$3   =(C56/121)*1.449 41512     0     3.88                 0.046                 =A57+B57     =F57-F$3   =(C57/121)*1.449 41522     0     3.66                 0.044                 =A58+B58     =F58-F$3   =(C58/121)*1.449 41532     0     3.46                 0.041                 =A59+B59     =F59-F$3   =(C59/121)*1.449 41542     0     3.27                 0.039                 =A60+B60     =F60-F$3   =(C60/121)*1.449 41552     0     3.11                 0.037         _      =A61+B61     =F61-F$3   =(C61/121)*1.449 41562     0     2.96                 0.035                 =A62+B62     =F62-F$3   =(C62/121)*1.449 41572     0     2.82                 0.034                 =A63+B63     =F63-F$3   =(C63/121)*1.449 41582     0     2.69                 0.032                 =A64+B64     =F64-F$3   =(C64/121)*1.449 41602     0     2.46                 0.03                 =A65+B65     =F65-F$3   =(C65/121)*1.449 41622     0     2.27                 0.027                 =A66+B66     =F66-F$3   =(C66/121)'1.449 41642     0     2.1                   0.025                 =A67+B67     =F67-F$3   =(C67/121)*1.449 41662     0     1.96                 0.023                 =A68+B68     =F68-F$3   =(C68/121)*1.449 41682     0     1.84                 0.022                 =A69+B69     =F69-F$3   =(C69/121)*1.449 41702     0     1.73                 0.021                 =A70+B70     =F70-F$3   =(C70/121)*1.449 41722     0     1.63                 0.02                 =A71+B71     =F71-F$3   =(C71/121)*1.449 41742     0     1.54                 0.018                 =A72+B72     =F72-F$3   =(C72/121)*1.449 41762   .0     1.47                 0.018                 =A73+B73     =F73-F$3   =(C73/121)*1.449 41782     0     1.4                   0.017                 =A74+B74     =F74-F$3   =(C74/121)*1.449 41802     0     1.33                 0.016                 =A75+B75     =F75-F$3   =(C75/121)*1.449 41822     0     1.28                 0.015                 =A76+B76     =F76-F$3   =(C76/121)*1.449 41842     0     1.22                 0.015                 =A77+B77     =F77-F$3   =(C77/121)'1.449 41862     0     1.17                 0.014         _      =A78+B78     =F78-F$3   =(C78/121)*1.449 41882     0     1.13                 0.013                 =A79+B79     =F79-F$3   =(C79/121)*1.449 41902     0     1.08                 0.013                 =A80+B80     =F80-F$3   =(C80/121)*1.449 41922     0     1.04                 0.012                 =A81+B81     =F81-F$3   =(C81/121)*1.449 41942     0     1                     0.012                 =A82+B82     =F82-F$3   =(C82/121)*1.449 41962     0     0.97                 0.012                 =A83+B83     =F83-F$3   =(C83/121)'1.449 41982     0     0.93                 0.011                 =A84+B84     =F84-F$3   =(C84/121)*1.449 42002     0     0.9                   0.011                 =A85+B85     =F85-F$3   =(C85/121)*1.449 42022     0     0.87                 0.01                 =A86+886     =F86-F$3   =(C86/121)*1.449 42042     0     0.84                 0.01                 =A87+B87     =F87-F$3   =(C87/121)*1.449 42062     0     0.81                 0.01                 =A88+B88     =F88-F$3   =(C88/121)*1.449 42082     0     0.79                 0.009                 =A89+B89     =F89-F$3   =(C89/121)*1.449 42102     0     0.76                 0.009               =A90+B90       =F90-F$3   =(C90/121)*1.449 42122     0     0.74                 0.009                 =A91+B91     =F91-F$3   =(C91/121)*1.449 42142     0     0.72                 0.009                 =A92+B92     =F92-F$3   =(C92/121)*1.449 42162     0     0.7                   0.008                 =A93+B93     =F93-F$3   =(C93/121)*1.449 42182     0     0.68                 0.008                 =A94+B94     =F94-F$3   =(C94/121)*1.449 42202     0     0.66                 0.008                 =A95+B95     =F95-F$3   =(C95/121)*1.449 42222     0     0.64                 0.008                 =A96+B96     =F96-F$3   =(C96/121)'1.449 42242     0     0.62                 0.007                 =A97+B97     =F97-F$3   =(C97/121)*1.449 42262     0     0.6                   0.007               =A98+B98       =F98-F$3   =(C98/1211)*1.449
Heat Load from Cycle Hottest Fuel 32 Discharge Assembly Assemblies Only Estimate Date Days since Recalculated Hottest Date Time (MBTU/hr)
(MBTU/hr)
(Reprinted)
May 8, 2013 Assembly (MBTU/hr) 41470 0
5.24 0.063  
=A51+B51  
=F51-F$3  
=(C51/121)*1.449 41474 0
5.07 0.061  
=A52+B52  
=F52-F$3  
=(C52/121)*1.449 41478 0
4.91 0.059
_=A53+B53  
=F53-F$3  
=(C53/121)'1.449 41482 0
4.76 0.057  
=A54+B54  
=F54-F$3  
=(C54/121)*1.449 41492 0
4.42 0.053  
=A55+B55  
=F55-F$3  
=(C55/121)*1.449 41502 0
4.13 0.049  
=A56+B56  
=F56-F$3  
=(C56/121)*1.449 41512 0
3.88 0.046  
=A57+B57  
=F57-F$3  
=(C57/121)*1.449 41522 0
3.66 0.044  
=A58+B58  
=F58-F$3  
=(C58/121)*1.449 41532 0
3.46 0.041  
=A59+B59  
=F59-F$3  
=(C59/121)*1.449 41542 0
3.27 0.039  
=A60+B60  
=F60-F$3  
=(C60/121)*1.449 41552 0
3.11 0.037  
=A61+B61  
=F61-F$3  
=(C61/121)*1.449 41562 0
2.96 0.035  
=A62+B62  
=F62-F$3  
=(C62/121)*1.449 41572 0
2.82 0.034  
=A63+B63  
=F63-F$3  
=(C63/121)*1.449 41582 0
2.69 0.032  
=A64+B64  
=F64-F$3  
=(C64/121)*1.449 41602 0
2.46 0.03  
=A65+B65  
=F65-F$3  
=(C65/121)*1.449 41622 0
2.27 0.027  
=A66+B66  
=F66-F$3  
=(C66/121)'1.449 41642 0
2.1 0.025  
=A67+B67  
=F67-F$3  
=(C67/121)*1.449 41662 0
1.96 0.023  
=A68+B68  
=F68-F$3  
=(C68/121)*1.449 41682 0
1.84 0.022  
=A69+B69  
=F69-F$3  
=(C69/121)*1.449 41702 0
1.73 0.021  
=A70+B70  
=F70-F$3  
=(C70/121)*1.449 41722 0
1.63 0.02  
=A71+B71  
=F71-F$3  
=(C71/121)*1.449 41742 0
1.54 0.018  
=A72+B72  
=F72-F$3  
=(C72/121)*1.449 41762  
.0 1.47 0.018  
=A73+B73  
=F73-F$3  
=(C73/121)*1.449 41782 0
1.4 0.017  
=A74+B74  
=F74-F$3  
=(C74/121)*1.449 41802 0
1.33 0.016  
=A75+B75  
=F75-F$3  
=(C75/121)*1.449 41822 0
1.28 0.015  
=A76+B76  
=F76-F$3  
=(C76/121)*1.449 41842 0
1.22 0.015  
=A77+B77  
=F77-F$3  
=(C77/121)'1.449 41862 0
1.17 0.014  
=A78+B78  
=F78-F$3  
=(C78/121)*1.449 41882 0
1.13 0.013  
=A79+B79  
=F79-F$3  
=(C79/121)*1.449 41902 0
1.08 0.013  
=A80+B80  
=F80-F$3  
=(C80/121)*1.449 41922 0
1.04 0.012  
=A81+B81  
=F81-F$3  
=(C81/121)*1.449 41942 0
1 0.012  
=A82+B82  
=F82-F$3  
=(C82/121)*1.449 41962 0
0.97 0.012  
=A83+B83  
=F83-F$3  
=(C83/121)'1.449 41982 0
0.93 0.011  
=A84+B84  
=F84-F$3  
=(C84/121)*1.449 42002 0
0.9 0.011  
=A85+B85  
=F85-F$3  
=(C85/121)*1.449 42022 0
0.87 0.01  
=A86+886  
=F86-F$3  
=(C86/121)*1.449 42042 0
0.84 0.01  
=A87+B87  
=F87-F$3  
=(C87/121)*1.449 42062 0
0.81 0.01  
=A88+B88  
=F88-F$3  
=(C88/121)*1.449 42082 0
0.79 0.009  
=A89+B89  
=F89-F$3  
=(C89/121)*1.449 42102 0
0.76 0.009  
=A90+B90  
=F90-F$3  
=(C90/121)*1.449 42122 0
0.74 0.009  
=A91+B91  
=F91-F$3  
=(C91/121)*1.449 42142 0
0.72 0.009  
=A92+B92  
=F92-F$3  
=(C92/121)*1.449 42162 0
0.7 0.008  
=A93+B93  
=F93-F$3  
=(C93/121)*1.449 42182 0
0.68 0.008  
=A94+B94  
=F94-F$3  
=(C94/121)*1.449 42202 0
0.66 0.008  
=A95+B95  
=F95-F$3  
=(C95/121)*1.449 42222 0
0.64 0.008  
=A96+B96  
=F96-F$3  
=(C96/121)'1.449 42242 0
0.62 0.007  
=A97+B97  
=F97-F$3  
=(C97/121)*1.449 42262 0
0.6 0.007  
=A98+B98  
=F98-F$3  
=(C98/1211)*1.449


Calculation 2013-07050 Rev. 0 Kewaunee Power Station                                                           Page A6 of 6 Attachment A: Heat Generation Rate vs. Decay Time (from Ref. 2.5 Heat Load from Cycle   Hottest Fuel 32 Discharge       Assembly Assemblies Only       Estimate       Date     Days since Recalculated Hottest Date   Time         (MBTU/hr)         (MBTU/hr)   (Reprinted) May 8, 2013 Assembly (MBTU/hr) 42282   0       0.59                 0.007         =A99+B99     =F99-F$3   =(C99/121)*1.449 42302   0       0.57                 0.007         =A100+B100   =F100-F$3   =(C100/121)*1.449 42322   0       0.56                 0.007         =A101+B101   =F101-F$3   =(C101/121)*1.449 42342   0       0.54                 0.007         =A102+B102   =F102-F$3   =(C102/121)*1.449 42362   0       0.53                 0.006         =A103+B103   =F103-F$3   =(C103/121)'1.449 42382   0       0.52                 0.006         =A104+B104   =F104-F$3   =(C104/121)'1.449 42402   0       0.5                   0.006         =A105+B105   =F105-F$3   =(C105/121)*1.449 42422   0       0.49                 0.006         =A106+B106   =F106-F$3   =(C106/121)*1.449 42442   0       0.48                 0.006         =A107+B107   =F107-F$3   =(C107/121)*1.449 42462   0       0.47                 0.006         =A108+B108   =F108-F$3   =(C108/121)'1.449 42482   0       0.46                 0.005         =A109+B109   =F109-F$3   =(C109/121)*1.449
Calculation 2013-07050 Rev. 0 Kewaunee Power Station Page A6 of 6 Attachment A: Heat Generation Rate vs. Decay Time (from Ref. 2.5 Heat Load from Cycle Hottest Fuel 32 Discharge Assembly Assemblies Only Estimate Date Days since Recalculated Hottest Date Time (MBTU/hr)
(MBTU/hr)
(Reprinted)
May 8, 2013 Assembly (MBTU/hr) 42282 0
0.59 0.007  
=A99+B99  
=F99-F$3  
=(C99/121)*1.449 42302 0
0.57 0.007  
=A100+B100  
=F100-F$3  
=(C100/121)*1.449 42322 0
0.56 0.007  
=A101+B101  
=F101-F$3  
=(C101/121)*1.449 42342 0
0.54 0.007  
=A102+B102  
=F102-F$3  
=(C102/121)*1.449 42362 0
0.53 0.006  
=A103+B103  
=F103-F$3  
=(C103/121)'1.449 42382 0
0.52 0.006  
=A104+B104  
=F104-F$3  
=(C104/121)'1.449 42402 0
0.5 0.006  
=A105+B105  
=F105-F$3  
=(C105/121)*1.449 42422 0
0.49 0.006  
=A106+B106  
=F106-F$3  
=(C106/121)*1.449 42442 0
0.48 0.006  
=A107+B107  
=F107-F$3  
=(C107/121)*1.449 42462 0
0.47 0.006  
=A108+B108  
=F108-F$3  
=(C108/121)'1.449 42482 0
0.46 0.005  
=A109+B109  
=F109-F$3  
=(C109/121)*1.449


"' Calculation 2013-07050 Rev. 2 Kewaunee Power Station                                                                         Page B1 of 2 Attachment B: Analysis Specific Heat of Uranium         146  J/kg-K    Input 4.3 Specific Heat of Uranium       0.035  BTU/Ib-F  Conversion Specific Heat of Zirconium       322  J/kg-K    Input 4.2 Specific Heat of Zirconium     0.077  BTU/Ib-F  Conversion Diameter of Fuel Uranium       0.3659  inches    Input 4.4 Inner Diameter of Zirconium     0.3734  inches    Input 4.4 Outer Diameter of Zirconium       0.422  inches    Input 4.4 Heated Rods per Assem           179  Rods      Input 4.4 Unheated Rods (Guide or Instrument Tubes)             17  Tubes    Input 4.4 ID of Guide Tubes     0.492  inches    Input 4.4 OD of Guide Tubes       0.526   inches    Input 4.4 Density of Uranium      19,070 kg/m3    Input 4.3 Theoretical Density    96.56%             Input 4.4 3
Calculation 2013-07050 Rev. 2 Kewaunee Power Station Page B1 of 2 Attachment B: Analysis Specific Heat of Uranium Specific Heat of Uranium Specific Heat of Zirconium Specific Heat of Zirconium Diameter of Fuel Uranium Inner Diameter of Zirconium Outer Diameter of Zirconium Heated Rods per Assem Unheated Rods (Guide or Instrument Tubes)
Density of Uranium      1149.5 lb/ft   3 Conversion Density of Zirconium      6570  kg/M      Input 4.2 3
ID of Guide Tubes OD of Guide Tubes Density of Uranium Theoretical Density Density of Uranium Density of Zirconium Density of Zirconium Heated Length of Uranium Initial Temperature Final Temperature Total temperature Increase Volume of Uranium Volume of Zirconium in a Heated Rod Volume of Zirconium in a Guide Tube Total Volume of Zirconium 146 0.035 322 0.077 0.3659 0.3734 0.422 179 17 0.492 0.526 19,070 96.56%
Density of Zirconium      410.2   lb/ft    Conversion Heated Length of Uranium        11.9375  feet      Input 4.4 Initial Temperature        90  F        Assumption 5.4 Final Temperature        1049  F        Input 4.1 Total temperature Increase        959  F        Initial Minus Final 3
1149.5 6570 410.2 11.9375 90 1049 959 1.560 0.451 0.038 0.489 J/kg-K BTU/Ib-F J/kg-K BTU/Ib-F inches inches inches Rods Tubes inches inches kg/m3 lb/ft3 kg/M 3
ft Volume of Uranium        1.560    3      Equation   6-4 ft Volume of Zirconium in a Heated Rod        0.451    3      Equation   6-5 ift Volume of Zirconium in a Guide Tube        0.038    3      Equation   6-6 ft.
lb/ft3 feet F
Total Volume of Zirconium        0.489            Equation   6-7 Assem Heat Generation at 14 Months         0.01515   MBTU/hr Interpolated from Att. A Time to Failure     4.94   hrs       Equation 6-3 Assem Heat Generation at 17 Months         0.01253   MBTU/hr Interpolated from Att. A Time to Failure     5.97   hrs       Equation 6-3 Heat Generation that Gives 2 Hour Heat-Up       0.03739   MBTU/hr Iterated Time to Failure     2.00 hrs       Equation 6-3 Date of Associated Heat Generationr 10/4/2013-1             Interpolated from Att. A Heat Generation that Gives 4 Hour Heat-Up       0.01869 MBTU/hr Iterated Time to Failure     4.00 hrs       Equation 6-3 Date of Associated Heat Generationr 4/8/2014 -               Interpolated from Att. A Heat Generation that Gives 10 Hour Heat-Up       0.00748 MBTU/hr Iterated Time to Failure   10.00 hrs       Equation 6-3 Date of Associated Heat Generationr 8/21/2015-1             Interpolated from Att. A NUREG-1783 Maximum Temperature (900 C)               1652 F         Input 4.1 Temperature Increase         1562 F         Initial Minus Final Heat Generation that Gives 10 Hour Heat-Up       0.01218 MBTU/hr   Iterated Time to Failure   10.00 hrs       Equation 6-3 Date of Associated Heat Generation 10/21/2014               Interpolated from Att. A Heat Generation that Gives 6 Hour Heat-Up       0.02030   ABTU/hr Iterated Time to Failure     6.00 hirs      Equation 6-3 Date of Associated Heat Generationr 3/11/2014-1             Interpolated from Att. A Heat Generation that Gives 4 Hour Heat-Up       0.03045   MBTU/hr Iterated Time to Failure     4.00 h"irs      Equation 6-3 Date of Associated Heat Generationl 11/16/2013               Interpolated from Att. A Heat Generation that Gives 2 Hour Heat-Up       0.06089   ABTU/hr Iterated Time to Failure     2.00 hirs      Equation 6-3 Date of Associated Heat Generation     7/18/2013           Interpolated from Att. A
F F
ft3 ft3 ift3 ft.3 Input 4.3 Conversion Input 4.2 Conversion Input 4.4 Input 4.4 Input 4.4 Input 4.4 Input 4.4 Input 4.4 Input 4.4 Input 4.3 Input 4.4 Conversion Input 4.2 Conversion Input 4.4 Assumption 5.4 Input 4.1 Initial Minus Final Equation 6-4 Equation 6-5 Equation 6-6 Equation 6-7 Assem Heat Generation at 14 Months 0.01515 MBTU/hr Interpolated from Att. A Time to Failure 4.94 hrs Equation 6-3 Assem Heat Generation at 17 Months 0.01253 MBTU/hr Interpolated from Att. A Time to Failure 5.97 hrs Equation 6-3 Heat Generation that Gives 2 Hour Heat-Up 0.03739 MBTU/hr Iterated Time to Failure 2.00 hrs Equation 6-3 Date of Associated Heat Generationr 10/4/2013-1 Interpolated from Att. A Heat Generation that Gives 4 Hour Heat-Up 0.01869 MBTU/hr Iterated Time to Failure 4.00 hrs Equation 6-3 Date of Associated Heat Generationr 4/8/2014 -
Interpolated from Att. A Heat Generation that Gives 10 Hour Heat-Up 0.00748 MBTU/hr Iterated Time to Failure 10.00 hrs Equation 6-3 Date of Associated Heat Generationr 8/21/2015-1 Interpolated from Att. A NUREG-1783 Maximum Temperature (900 C) 1652 F
Input 4.1 Temperature Increase 1562 F
Initial Minus Final Heat Generation that Gives 10 Hour Heat-Up 0.01218 MBTU/hr Iterated Time to Failure 10.00 hrs Equation 6-3 Date of Associated Heat Generation 10/21/2014 Interpolated from Att. A Heat Generation that Gives 6 Hour Heat-Up 0.02030 Time to Failure 6.00 h
Date of Associated Heat Generationr 3/11/2014-1 Heat Generation that Gives 4 Hour Heat-Up 0.03045 Time to Failure 4.00 h"
Date of Associated Heat Generationl 11/16/2013 Heat Generation that Gives 2 Hour Heat-Up 0.06089 Time to Failure 2.00 h
Date of Associated Heat Generation 7/18/2013 ABTU/hr Iterated irs Equation 6-3 Interpolated from Att. A MBTU/hr Iterated irs Equation 6-3 Interpolated from Att. A ABTU/hr Iterated irs Equation 6-3 Interpolated from Att. A


Calculation 2013-07050 Rev. 2 Kewaunee Power Station                                                                                                                                                                     Page B2 of 2 A                                                                           BC                                                   J               D               E F 1 Attachment B: Analysis                             ____
Calculation 2013-07050 Rev. 2 Kewaunee Power Station Page B2 of 2 A
2 3                       Specific Heat of Uranium     146                                                                                     J/kg-K       input4.3 4                       Specific Heat of Uranium     =B3"0.0009478/2.20462/(9/5)                                                             BTU/lb-F Conversion 5                     Specific Heat of Zirconium   322                                                                                     J/kg-K _jInput 4.2 6                     Specific Heat of Zirconium   =B5*0.000947812.20462/(915)                                                             BTU/lb-F 'Conversion                   i 7                     Diameter of Fuel Uranium     0.3659                                                                                   inches       IInput 4.4                 1 8                   Inner Diameter ot Zirconium     0.3734                                                                                   inches       lnnput 4.4 9                   Outer Diameter of Zirconium     0.422                                                                                   inches       tInput 4.4 10                       Heated Rods erAssem         179                                                                                     Rods         tnaput4.4 11   Unheated Rods (Guide or Instrument Tubes)     17                                                                                       Tubes         Input 4.4 12                             ID of Guide Tubes   0.492                                                                                   inches         Input 4.4 13                           OD of Guide Tubes     0.526                                                                                   inches         Input 4.4 14                             Density of Uranium   19070                                                                                   kg/im         tInput4.3 15                           Theoretical Density   0.9656                                                                                                 Input 4.4 16                             Density of Uranium   =814*2.20462/3.280841^3*B15                                                             lb/ft'       !Conversion 17                           Density of Zirconium   6570                                                                                     kg/mf         CoInput 4.2 3
BC J
18                           Density of Zirconium   =B17"2.20462/3.28084^3                                                                   lb/hf       !Conversion                 _
D E
19                     Heated Length of Uranium     =143.25/12                                                                               feet         itnput 4.4 20                             Initial Temperature   90                                                                                       F           ]Assumption 5.4 21                             Final Temperature   1049                                                                                     F           !Input 4.1 22                   Total temperature Increase     =B21-B20                                                                                 F           itnitial Minus Final 23                                                                                                                                                         1 3
F 1 Attachment B: Analysis 2
24                             Volume of Uranium =PIt*B7^214*B19/144*B10                                                                     ft             Equation 6-4 25         Volume of Zirconium in a Heated Rod =Pl()*(B9^2-B8^2)/4*B19/144*BlO                                                               ft3           Equation 6-55 26         Volume of Zirconium in a Guide Tube =PlO*(B13^2-B12^2)/4"B19/144*B11                                                             ft3         'Equation 6-6 27 28                    Total Volume of Zirconium =B25+126                                                                                     ft             E           -7io 29         Assent Heat Generation at 14 Months ='Attachment A'!H76-(5/20)*('Attachment A'!H76-'Attachment A1H77)                             MBTU/hr Ilnterpolated from Al. A 1 30                                   Time to Failure =$BS22/(B29&deg;10 *6)*(SBS16"$B$24"$B$4+$B$18*SB$27*SBS6)                                   hrs         'Equation 6-3 31 32         Assent Heat Generation at 17 Months ='Attachment A'!H80-(17/20)&deg;(Attachment A'!H80-'Attachment A'1H81)                           MBTU/hr IInterpolated from Att. A 33                                   Time to Failure =$B$22/(B32*10^6)*($BS16*$BS24"$B$4+$B$18"SB$27'$BS6)                                   hrs         iEquation 6-3 35   Heat Generation that Gives 2 Hour Heat-Up     0.0373850764604915                                                                       MBTU/hr Iterated                         _
3 Specific Heat of Uranium 146 J/kg-K input4.3 4
36                                 Time to Failure =SBS22/(B35*10^6)*($B$16*$BS24&deg;$B$4+$B$18*SB$27"$BS6)                                   hrs         'Equation 6-3 37           Date of Associated Heat Generation     ='Attachment A'F60+('Attachment A'!H60-B35)/('Attachment A'!H60-'Attachment A'!H61)*10               'Interpolated from Att. A 38                                                                                                                                           1 39   Heat Generation that Gives 4 Hour Heat-Up     0.0186925354210505                                                                       MBTU/hr Iterated 40                                   Time to Failure =SBS22/(B39&deg;10^6)*($BS16*$B$24*$B$4+$B$18*SBS27*$BS6)                                   hrs           Equation 6-3 41           Date of Associated Heat Generation     ='Attachment A'IF71+i'Attachment A'tH71-B39)/('Attachment A!H7I-'Attachment A'!H72)*20                 Interpolated from Att. A 1 42 _           __
Specific Heat of Uranium =B3"0.0009478/2.20462/(9/5)
43   Heat Generation that Gives 10 Hour Heat-Up     0.00747701366999636                                                                     MBTU/hr Iterated 44                                 Time to Failure =SB$221(B43"10^6)*($BS16"$B$24*$B$4+$B$18*$B$27&deg;SBS6)                                   hrs         'Equation 6-3 45           Date of Associated Heat Generation     ='Attachment A'!F96+('Attachment Al1H96-B43)/('Attachment Al1H96-'Attachment A:IH97)*20               IInterpolated from Att. A 461i 47   NUREG-1783 Maximum Temperature (900 C)11652                                                                                             F             Input 4.1 48                                                       7-B20                                                                               F             Ilnitial Minus Final 49   Heat Generatic                                                                                                                           MBTU/hr       Iterated hirs         Equation 6-3 Date of Associated Heat C                                                                                                                    .Interpolated from Att. A 53   Heat Generation that Gives 6 Hour Heat-Up 10.0202973514135572                                                                           MBTU/hr Iterated 54                                 Time to Failure =$6S8/(B5310v6)*(B$16$B$24S$B               $18SB$27S$B$6                                 hrs -          neuateron6-3 551         Date of Associated Heat Generation ='Attachment A'!F70v('Attachment A'!H70-853/('Attachment A'l                                                 iterpolated trom Alt. A 561 57   Heat Generation that Gives 4 Hour Heat-Up 0.0304460281462153                                                                           MBTU/hr       Iterated 58                                   Time to Failure =$BS48/[B57*1                         4+$B$18"SB$27*$BS6)                               hrs           Equation 6-3 59           Date of Associated Heat Generation             rent All ttachment A'-H65)'20               !Interpolated from ACt. A 60 61    Heat Generation that Gives 2 Hour Heat-Up                                                                                                      'U/hr Iterated 62                                  Time to Failure =$BS48/(B61"10 6)(*$BS16"$B$24*$B$4 $B$18*SB$27*$B$6)                                                 IEquation 6-3 63          Date of Associated Heat Generation ='Attachment A'1F51+('Attachment A!H51-B61 )/('Attachment A'!H51-'Attachment A'lH52)*4                     Ilnteroolated from Att. A
BTU/lb-F Conversion 5
Specific Heat of Zirconium 322 J/kg-K _jInput 4.2 6
Specific Heat of Zirconium =B5*0.000947812.20462/(915)
BTU/lb-F 'Conversion i
7 Diameter of Fuel Uranium 0.3659 inches I Input 4.4 1
8 Inner Diameter ot Zirconium 0.3734 inches lnnput 4.4 9
Outer Diameter of Zirconium 0.422 inches t Input 4.4 10 Heated Rods erAssem 179 Rods tnaput 4.4 11 Unheated Rods (Guide or Instrument Tubes) 17 Tubes Input 4.4 12 ID of Guide Tubes 0.492 inches Input 4.4 13 OD of Guide Tubes 0.526 inches Input 4.4 14 Density of Uranium 19070 kg/im tInput 4.3 15 Theoretical Density 0.9656 Input 4.4 16 Density of Uranium =814*2.20462/3.280841^3*B15 lb/ft'  
!Conversion 17 Density of Zirconium 6570 kg/mf CoInput 4.2 18 Density of Zirconium =B17"2.20462/3.28084^3 lb/hf 3
!Conversion 19 Heated Length of Uranium =143.25/12 feet i tnput 4.4 20 Initial Temperature 90 F  
]Assumption 5.4 21 Final Temperature 1049 F  
!Input 4.1 22 Total temperature Increase =B21-B20 F
itnitial Minus Final 23 1
24 Volume of Uranium =PIt*B7^214*B19/144*B10 ft 3
Equation 6-4 25 Volume of Zirconium in a Heated Rod =Pl()*(B9^2-B8^2)/4*B19/144*BlO ft3 Equation 6-5 5 26 Volume of Zirconium in a Guide Tube =PlO*(B13^2-B12^2)/4"B19/144*B11 ft3  
'Equation 6-6 27 Total Volume of Zirconium =B25+126 ft E  
-7io 28 29 Assent Heat Generation at 14 Months ='Attachment A'!H76-(5/20)*('Attachment A'!H76-'Attachment A1H77)
MBTU/hr Ilnterpolated from Al. A 1 30 Time to Failure =$BS22/(B29&deg;10 *6)*(SBS16"$B$24"$B$4+$B$18*SB$27*SBS6) hrs  
'Equation 6-3 31 32 Assent Heat Generation at 17 Months ='Attachment A'!H80-(17/20)&deg;(Attachment A'!H80-'Attachment A'1H81)
MBTU/hr IInterpolated from Att. A 33 Time to Failure =$B$22/(B32*10^6)*($BS16*$BS24"$B$4+$B$18"SB$27'$BS6) hrs iEquation 6-3 35 Heat Generation that Gives 2 Hour Heat-Up 0.0373850764604915 MBTU/hr Iterated 36 Time to Failure =SBS22/(B35*10^6)*($B$16*$BS24&deg;$B$4+$B$18*SB$27"$BS6) hrs  
'Equation 6-3 37 Date of Associated Heat Generation ='Attachment A'F60+('Attachment A'!H60-B35)/('Attachment A'!H60-'Attachment A'!H61)*10  
'Interpolated from Att. A 38 1
39 Heat Generation that Gives 4 Hour Heat-Up 0.0186925354210505 MBTU/hr Iterated 40 Time to Failure =SBS22/(B39&deg;10^6)*($BS16*$B$24*$B$4+$B$18*SBS27*$BS6) hrs Equation 6-3 41 Date of Associated Heat Generation ='Attachment A'IF71+i'Attachment A'tH71-B39)/('Attachment A!H7I-'Attachment A'!H72)*20 Interpolated from Att. A 1 42 _
43 Heat Generation that Gives 10 Hour Heat-Up 0.00747701366999636 MBTU/hr Iterated 44 Time to Failure =SB$221(B43"10^6)*($BS16"$B$24*$B$4+$B$18*$B$27&deg;SBS6) hrs  
'Equation 6-3 45 Date of Associated Heat Generation ='Attachment A'!F96+('Attachment Al1H96-B43)/('Attachment Al1H96-'Attachment A:IH97)*20 I Interpolated from Att. A 461i 47 NUREG-1783 Maximum Temperature (900 C) 11652 F
Input 4.1 48 7-B20 F
Ilnitial Minus Final 49 Heat Generatic MBTU/hr Iterated Date of Associated Heat C hirs Equation 6-3
.Interpolated from Att. A 53 Heat Generation that Gives 6 Hour Heat-Up 10.0202973514135572 MBTU/hr Iterated 54 Time to Failure =$6S8/(B5310v6)*(B$16$B$24S$B  
$18SB$27S$B$6 551 Date of Associated Heat Generation ='Attachment A'!F70v('Attachment A'!H70-853/('Attachment A'l 561 hrs neuateron 6-3 iterpolated trom Alt. A 57 Heat Generation that Gives 4 Hour Heat-Up 0.0304460281462153 MBTU/hr Iterated 58 Time to Failure =$BS48/[B57*1 4+$B$18"SB$27*$BS6) hrs Equation 6-3 59 Date of Associated Heat Generation 60 61 Heat Generation that Gives 2 Hour Heat-Up 62 Time to Failure 63 Date of Associated Heat Generation rent All ttachment A'-H65)'20  
! Interpolated from ACt. A
'U/hr Iterated IEquation 6-3 Ilnteroolated from Att. A
=$BS48/(B61"10 6)(*$BS16"$B$24*$B$4 $B$18*SB$27*$B$6)
='Attachment A'1F51 +('Attachment A!H51-B61 )/('Attachment A'!H51-'Attachment A'lH52)*4


4 L
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Calculation   2013-07050 Kewaunee Power Station                                                                       Page C1 of 2 FW: sfp temp today Michael S Lico (Generation - 6) 07/22/2013 11:28 AM To:
Calculation 2013-07050 Kewaunee Power Station Page C1 of 2 FW: sfp temp today Michael S Lico (Generation - 6) 07/22/2013 11:28 AM To:
  'MATTHEW.M.ROSS@Sargentlundy.com' Show Details Max, The attached printout from the KPS Plant Parameters List shows the present SFP temperature to be 80 0 F.
'MATTHEW.M.ROSS@Sargentlundy.com' Show Details
: Max, The attached printout from the KPS Plant Parameters List shows the present SFP temperature to be 800 F.
Clearly, the temperature will only continue to drop as the SFP heat load decreases. Thanks.
Clearly, the temperature will only continue to drop as the SFP heat load decreases. Thanks.
Mike Lico From: John F Helfenberger (Generation - 4)
Mike Lico From: John F Helfenberger (Generation - 4)
Line 545: Line 1,270:
sfp temp today FYI.
sfp temp today FYI.
John F. Helfenberger, Lead Reactor Engineer, Kewaunee Power Station bus. 920.388.8294 Dominion Tie-line - 8.691.8294 pag. 920.704.4471 CONFIDENTIALITY NOTICE: This electronic message contains information which may be legally confidential and/or privileged and does not in any case represent a firm ENERGY COMMODITY bid or offer relating thereto which binds the sender without an additional express written confirmation to that effect. The information is intended solely for the individual or entity named above and access by anyone else is unauthorized. If you are not the intended recipient, any disclosure, copying, distribution, or use of the contents of this information is prohibited and may be unlawful. If you have received this electronic transmission in error, please reply immediately to the sender that you have received the message in error, and delete it. Thank you.
John F. Helfenberger, Lead Reactor Engineer, Kewaunee Power Station bus. 920.388.8294 Dominion Tie-line - 8.691.8294 pag. 920.704.4471 CONFIDENTIALITY NOTICE: This electronic message contains information which may be legally confidential and/or privileged and does not in any case represent a firm ENERGY COMMODITY bid or offer relating thereto which binds the sender without an additional express written confirmation to that effect. The information is intended solely for the individual or entity named above and access by anyone else is unauthorized. If you are not the intended recipient, any disclosure, copying, distribution, or use of the contents of this information is prohibited and may be unlawful. If you have received this electronic transmission in error, please reply immediately to the sender that you have received the message in error, and delete it. Thank you.
file://C:\Documents and Settings\0n7447\Local Settings\Temp\notesFFF692\-web5264.htm               7/22/2013
file://C:\\Documents and Settings\\0n7447\\Local Settings\\Temp\\notesFFF692\\-web5264.htm 7/22/2013


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Latest revision as of 00:16, 11 January 2025

And Response to Request for Additional Information Re Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E
ML13351A040
Person / Time
Site: Kewaunee Dominion icon.png
Issue date: 12/11/2013
From: Mark D. Sartain
Dominion Energy Kewaunee
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
13-390A
Download: ML13351A040 (59)
v • d • e


Text

Dominion Energy Kewaunee, Inc.

5000 Dominion Boulevard, Glen Allen, VA 23060 DoEEE oflI Web Address: www.dom.com December 11, 2013 ATTN: Document Control Desk Serial No.

13-390A U. S. Nuclear Regulatory Commission LIC/JG/RO Washington, DC 20555-0001 Docket No.: 50-305 License No.: DPR-43 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 By application dated July 31, 2013 (Reference 1), Dominion Energy Kewaunee, Inc.

(DEK) requested exemptions from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2),

and 10 CFR 50, Appendix E, Section IV, for Kewaunee Power Station (KPS).

The requested exemptions would allow DEK to reduce emergency planning requirements and subsequently revise the KPS Emergency Plan consistent with the permanently defueled condition of the station.

Subsequently, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) regarding the proposed exemptions (Reference 2).

The RAI questions and associated DEK response are provided in Attachment 1 to this letter.

In response to the staff's comments, DEK is revising the originally proposed exemption request. Attachment 2 to this letter provides a supplement to the proposed exemption request describing the revisions. The analyses and conclusions provided in Reference 1 are not changed by the proposed revisions. The conclusions of the no significant hazards consideration and the environmental considerations contained in Reference 1 are not affected by, and remain applicable to, this revised request.

The July 31, 2014 requested approval date for the submittal remains unchanged.

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

Very truly yours, Mark D. Sartain Vice President - Nuclear Engineering and Development

Serial No. 13-390A Request for Exemptions Page 2 of 2 Attachments:

1. Response to NRC Request for Additional Information
2. Supplement 1 to DEK Request for Exemptions

Enclosure:

1. Supporting Calculation

References:

1. Letter from A. J. Jordan (DEK) to NRC Document Control Desk, "Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E," dated July 31, 2013.
2. Email from Dr. Karl D. Feintuch (NRC) to Margaret Earle, Jack Gadzala, Craig Sly, et al (DEK), "MF2567 Kewaunee Emergency Plan Requests for Exemption MF2567-RAII-ORLT-Norris-001 to -014 8 October 2013," dated October 8, 2013.

Commitments made by this letter: None cc:

Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road Suite 210 Lisle, IL 60532-4352 Dr. Karl D. Feintuch Project Manager U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08-D15 11555 Rockville Pike Rockville, MD 20852-2738 Mr. Theodore Smith Project Manager U.S. Nuclear Regulatory Commission Two White Flint North, Mail Stop 08-F5 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Kewaunee Power Station

Serial No. 13-390A ATTACHMENT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

REQUEST FOR EXEMPTIONS FROM PORTIONS OF 10 CFR 50.47(b),

10 CFR 50.47(c)(2), AND 10 CFR 50, APPENDIX E, SECTION IV KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No. 13-390A Page 1 of 22 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

REQUEST FOR EXEMPTIONS FROM PORTIONS OF 10 CFR 50.47(b),

10 CFR 50.47(c)(2), AND 10 CFR 50, APPENDIX E, SECTION IV By application dated July 31, 2013 (Reference 1), Dominion Energy Kewaunee, Inc.

(DEK) requested exemptions from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2),

and 10 CFR 50, Appendix E, Section IV, for Kewaunee Power Station (KPS). The requested exemptions would allow DEK to reduce emergency planning requirements and subsequently revise the KPS Emergency Plan consistent with the permanently defueled condition of the station.

Subsequently, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) regarding the proposed exemptions (Reference 2).

The RAI questions and associated DEK responses are provided below.

In the NRC RAI questions, the specific portion of the requirement within the regulation from which exemption is being requested is depicted in emphasized (boldlunderlined) font. The table numbers refer to the tables contained in Reference 1. In the tables below, the column titled "Kewaunee Request Wording" indicates DEK's originally requested exemption as contained in Reference 1.

The right column (titled "Past Precedence Wording") indicates exemptions (i.e., exempted wording) as previously granted by NRC for the associated regulation.

NRC Question MF2567-RAII-ORLOB-Norris-001 Table 1 Kewaunee Request Wording Past Precedence Wording 50.47(b)(1)

... State and local organizations within

...State and local organizations within the the Emergency Planning Zones-....

Emergency Planning Zones....

Although formal offsite emergency plans have typically been exempted for decommissioning sites, State and local organizations continue to be relied upon for firefighting, law enforcement, ambulance and medical services. Please explain why this requirement would not be applicable.

Response

The intent of the originally requested exemption was to continue to rely on State and local organizations for firefighting, law enforcement, ambulance and medical services as needed for events at the site, but without an expected need for these organizations for offsite events.

However, the past precedence wording also meets this intent.

Therefore, DEK is revising the originally requested exemption from portions of 10 CFR

Serial No. 13-390A Page 2 of 22 50.47(b)(1) in Reference 1, to read as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.

Memoranda of Understanding (MOU) establish agreements for assistance of the local organizations for firefighting, law enforcement, ambulance and medical services. The MOU continue to be required per 10 CFR 50.47(b)(3) even after approval of the proposed exemption to a portion of that regulation. Applicable details are contained in the Permanently Defueled Emergency Plan.

NRC Question MF2567-RAII-ORLOB-Norris-002 Table 1 Kewaunee Request Wording Past Precedence Wording 50.47(b)(7)

Information is made available to the Information is made available to the public on a periodic basis on how they public on a periodic basis on how they will be notified and what their initial will be notified and what their initial actions should be in an emergency actions should be in an emergency (e.g., listening to a local broadcast (e.g., listening to a local broadcast station and remaining indoors), the station and remaining indoors), [T]he principal points of contact with the principal points of contact with the news news media for dissemination of media for dissemination of information information during an emergency during an emergency (including the (including the physical location or physical location or locations) are locations) are established in advance, established in advance, and procedures for and procedures for coordinated coordinated dissemination of information to dissemination of information to the the public are established.

I public are established.

II The regulations in 10 CFR 72.32(a)(16) states, "Arrangements made for providing information to the public." Although Kewaunee has a general licensed ISFSI, the staff informed the previous exemption granted with the regulations in Part 72 for a specific licensed ISFSI. Please describe how information would be disseminated to the public should an event occur at the KPS site.

Response

Dominion Resources, Inc. (Dominion), the parent company for DEK, maintains a corporate communications organization, which includes a media relations group. News media contacts for the KPS location continue to be maintained. Should an event occur at the KPS site, information would be disseminated to the public and briefings with pertinent media organizations would be conducted per Dominion corporate communication protocols. 10 CFR 50.72(b)(2)(xi) requires that the NRC be notified, via the Emergency Notification System, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of certain events for which a news release is planned.

The process for complying with this requirement is procedurally implemented.

Serial No. 13-390A Page 3 of 22 Since there are no longer any pre-planned actions that the public needs to take as a result of an anticipated emergency at KPS, it is no longer necessary to pre-plan dissemination of emergency information to the public.

The intent of the originally requested exemption was to discontinue specific emergency response organizational requirements for major interactions with news media.

However, the past precedence wording also meets this intent.

Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50.47(b)(7) in Reference 1, to read as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.

Applicable details regarding how information would be disseminated to the public should an event occur at the KPS site are contained in the Permanently Defueled Emergency Plan.

NRC Question MF2567-RAII-ORLOB-Norris-003 Table 1 Kewaunee Request Wording Past Precedence Wording 50.47(b)(9)

Adequate methods, systems, and Adequate methods, systems, and equipment for assessing and equipment for assessing and monitoring monitoring actual or potential offsite actual or potential offsite consequences consequences of a radiological of a radiological emergency condition are emergency condition are in use.

in use.

The regulations in 10 CFR 72.32(a)(4) states, "Detection of accidents. Identification of the means of detecting an accident condition." Previous exemptions were granted for only the "offsite" assessment and monitoring.

Please provide specific justification for exempting this requirement.

Response

The intent of the originally requested exemption was to discontinue only those requirements associated with "offsite" assessment and monitoring. However, the past precedence wording also meets this intent. Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50.47(b)(9) in Reference 1, as shown in of this submittal.

The revised request is consistent with the past precedence wording shown above.

Applicable details regarding methods, systems, and equipment for assessing and monitoring actual or potential onsite consequences of a radiological emergency condition at the KPS site are contained in the Permanently Defueled Emergency Plan.

Serial No. 13-390A Page 4 of 22 NRC Question MF2567-RAII-ORLOB-Norris-004 Table 1 Kewaunee Request Wording Past Precedence Wording 50.47(b)(14)

Periodic exercises are (will be)

Periodic exercises are (will be) conducted conducted to evaluate maior to evaluate major portions of emergency portions of emerciency response response capabilities, periodic drills are (will capabilities, periodic drills are (will be) conducted to develop and maintain key be) conducted to develop and maintain skills, and deficiencies identified as a result key skills, and deficiencies identified as of exercises or drills are (will be) corrected.

a result of exercises or drills are (will be) corrected.

10 CFR 72.32(a)(12) states:

Exercises. (i) Provisions for conducting semiannual communications checks with offsite response organizations and biennial onsite exercises to test response to simulated emergencies. Radiological/Health Physics, Medical, and Fire drills shall be conducted annually. Semiannual communications checks with offsite response organizations must include the check and update of all necessary telephone numbers. The licensee shall invite offsite response organizations to participate in the biennial exercise.

(ii) Participation of offsite response organizations in biennial exercises, although recommended, is not required. Exercises must use scenarios not known to most exercise participants. The licensee shall critique each exercise using individuals not having direct implementation responsibility for conducting the exercise. Critiques of exercises must evaluate the appropriateness of the plan, emergency procedures, facilities, equipment, training of personnel, and overall effectiveness of the response.

Deficiencies found by the critiques must be corrected.

Previous exemptions did not grant an exemption as requested by DEK. Additionally, the regulations in 10 CFR 30, 10 CFR 40 and 10 CFR 70 related to emergency plans require licensees to conduct a biennial exercise within the scope of 10 CFR 72.32(a)(12). Please provide specific justification for exempting this requirement.

Response

DEK is retracting the originally requested exemption from 10 CFR 50.47(b)(14) in Reference 1, as shown in Attachment 2 of this submittal.

Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises

Serial No. 13-390A Page 5 of 22 is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.

KPS plans to conduct biennial exercises to test the adequacy of timing and content of implementing procedures and methods; to test emergency equipment and communication networks; and to ensure that emergency personnel are familiar with their duties.

KPS plans to invite offsite response organizations to participate in the exercises.

For alternating years, a drill would be conducted for the purpose of testing, developing, and maintaining the proficiency of on-site emergency responders.

Exercise and drill scenarios would include, at a minimum, the following:

  • The basic objective(s) of the drill.

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

" A time schedule of real and simulated initiating events.

  • A narrative summary describing conduct of the drill, including simulated casualties, off-site fire assistance, rescue of personnel, and use of protective clothing.

Critiques would evaluate the performance of the organization.

Applicable details regarding specific requirements for exercises and drills at KPS are contained in the Permanently Defueled Emergency Plan.

NRC Question MF2567-RAII-ORLOB-Norris-005 Table 2 Kewaunee Request Wording Past Precedence Wording IOCFR50

... and outside the site boundary to

...... and outside the site boundary to Appendix protect health and safety. The emergency protect health and safety. The emergency E IV.B.1 action levels shall be based on in-plant action levels shall be based on in-plant conditions and instrumentation in addition conditions and instrumentation in addition to onsite and offsite monitoring. By June to onsite and offsite monitoring. By June 20, 2012. for nuclear power reactor 20, 2012, for nuclear power reactor licensees, these action levels must licensees, these action levels must include hostile action that may include hostile action that may adversely affect the nuclear power plant, adversely affect the nuclear power The initial emergency action levels shall be plant. The initial emergency action levels discussed and agreed on by the applicant shall be discussed and agreed on by the or licensee and state and local applicant or licensee and state and local governmental authorities, and approved by governmental authorities, and approved by the NRC. Thereafter, emergency action the NRC. Thereafter, [E]mergency action levels shall be reviewed with the State levels shall be reviewed with the State and and local governmental authorities on local governmental authorities on an an annual basis.

annual basis.

Serial No. 13-390A Page 6 of 22 Maintaining the requirement for the offsite response organizations (OROs) to review the EALs on an annual basis will ensure the proper awareness by OROs of applicable emergency classifications and will also ensure that communications with the proper authorities are maintained.

Please provide specific justification for exempting this requirement.

Response

The intent of the originally requested exemption to review emergency action levels (EALs) with State and local governmental authorities was based on the proposed elimination of the requirement for offsite emergency response plans. However, based on the reviewer's comments, this portion of the exemption request is being retracted.

Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.B.1 in Reference 1, as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.

Based on the significantly reduced scope of EALs for the permanently defueled facility, the scope of the annual review of EALs with State and local governmental authorities (to ensure proper awareness of applicable emergency classifications) is expected to be commensurately reduced (e.g., informational mailings, etc.).

Applicable details regarding review of EALs with State and local governmental authorities are contained in the Permanently Defueled Emergency Plan.

NRC Question MF2567-RAII-ORLOB-Norris-006 Table 2 Kewaunee Request Wording Past Precedence Wording IOCFR50 In addition, a radiological orientation Appendix training program shall be made E IV.F.1 available to local services personnel; e.g., local emergency services/Civil Defense, local law enforcement personnel, local news media persons.

Previous exemption requests did not include an exemption for this requirement.

Firefighting, local law enforcement and ambulance/medical facilities are still expected to play a role in some onsite emergencies; therefore they should have some basic knowledge about radiation. 10 CRF 50.47(b)(15) requires that radiological emergency response training is provided to those who may be called on to assist in an emergency.

10 CFR 72.32(a)(8) requires a commitment to and a brief description of the means to promptly notify offsite response organizations and request offsite assistance, including medical assistance for the treatment of contaminated injured onsite workers when

Serial No. 13-390A Page 7 of 22 appropriate. 10 CFR 72.32(a)(10) requires training including any special instructions or orientation tours the licensee would offer to fire, police medical and other emergency personnel.

Additionally, 10 CFR 72.32(a)(16) requires arrangements made for providing information to the public.

Please provide justification for exempting this requirement.

Response

Based on the reviewer's comments, this portion of the exemption request is being retracted. Therefore, DEK is revising the originally requested exemption from 10 CFR 50, Appendix E, IV.F.1 in Reference 1, as shown in Attachment 2 of this submittal. The revised request proposes only to delete a reference to two of the examples listed in the regulation of local services personnel (Civil Defense and local news media persons).

The intent of the originally requested exemption was premised on local services personnel, such as local law enforcement personnel and local news media persons, no longer needing radiological orientation training since they will not be called upon to respond to a radiological event. However, their response to certain onsite emergencies was expected to be maintained.

10 CFR 50.47(b)(15) requires that radiological emergency response training be provided to those who may be called on to assist in an emergency. This requirement would encompass training offered to offsite response organizations (firefighting, law enforcement, ambulance and medical services).

A discussion of how information would be disseminated to the public is discussed in the response to Question 2 above. Since there are no longer 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.

The phrase "Civil Defense" is no longer a commonly used term and is no longer applicable as an example in the regulation.

Applicable details regarding the extent of the radiological orientation training program available to local services personnel are contained in the Permanently Defueled Emergency Plan.

Serial No. 13-390A Page 8 of 22 NRC Question MF2567-RAII-ORLOB-Norris-007 Table 2 Kewaunee Request Wording Past Precedence Wording 10CFR50 The plan shall describe provisions for the The plan shall describe provisions for the Appendix conduct of emergency preparedness conduct of emergency preparedness

  • E IV.F.2 exercises as follows: Exercises shall exercises as follows: Exercises shall test the test the adequacy of timing and content adequacy of timing and content of of implementing procedures and implementing procedures and methods, test methods, test emergency equipment and emergency equipment and communications communications networks, test the networks, test the public alert and public alert and notification system, notification system, and ensure that and ensure that emergency organization emergency organization personnel are personnel are familiar with their duties.

familiar with their duties.

See RAI 4, above.

Previous exemptions did not grant an exemption as requested.

Additionally, the regulations in 10 CFR 30, 10 CFR 40, 10 CFR 70, and 10 CFR 72 related to emergency plans require licensees to conduct a biennial exercise.

Please provide specific justification for exempting this requirement.

Response

The intent of the originally requested exemption was to continue testing the adequacy of timing and content of implementing procedures and methods, emergency equipment, and communications networks, except to perform these tests during the conduct of drills. However, the past precedence wording also meets this intent. Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.F.2 in Reference 1, as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.

Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.

Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.

Serial No. 13-390A Page 9 of 22 NRC Question MF2567-RAII-ORLOB-Norris-008 Table 2 Kewaunee Request Wording Past Precedence Wording IOCFR50 Appendix E IV.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 F4 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, 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.

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 4 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, 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 Sutport 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.

See RAI 4, above.

Serial No. 13-390A Page 10 of 22 Previous exemptions did not grant an exemption as requested.

Additionally, the regulations in 10 CFR 30, 10 CFR 40, 10 CFR 70, and 10 CFR 72 related to emergency plans require licensees to conduct a biennial exercise.

Please provide specific justification for exempting this requirement.

Response

As discussed in the response to Question 7 above, the intent of the originally requested exemption was to continue testing the adequacy of the emergency response organization, except to perform these tests during the conduct of drills. However, the past precedence wording also meets this intent.

Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.F.2.b in Reference 1, as shown in Attachment 2 of this submittal.

The revised request is consistent with the past precedence wording shown above.

Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.

Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.

NRC Question MF2567-RAII-ORLOB-Norris-009 Table 2 Kewaunee Request Wording Past Precedence Wording IOCFR5O Licensees shall enable any State or Licensees shall enable any State or local Appendix local government located within the government located within the plume E IV.F.2.e plume exposure Pathway EPZ to exposure Pathway EPZ to participate in the participate in the licensee's drills licensee's drills when requested by such State when requested by such State or local or local government.

I government.

Previous exemption requests did not include an exemption for this requirement.

10 CFR 72,32(a)(10) requires training, including any special instructions or orientation tours the licensee would offer to fire, police medical and other emergency personnel.

Additionally, 10 CFR 72.32(a)(12)(i) requires the licensee to invite offsite response organizations to participate in the biennial exercise. Additionally, CFR 72.32(a)(12) (ii) states in part, that participation of offsite response organizations in biennial exercises,

Serial No. 13-390A Page 11 of 22 although recommended, is not required. Please provide justification for exempting this requirement.

Response

Similar to the discussion contained in the response to Question 1 above, the intent of the originally requested exemption was to continue to rely on State and local organizations for firefighting, law enforcement, ambulance and medical services as needed for events at the site, but without an expected need for these organizations for offsite events.

However, the past precedence wording also meets this intent.

Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.F.2.e in Reference 1, as shown in Attachment 2 of this submittal.

The revised request is consistent with the past precedence wording shown above.

Applicable details regarding offsite agency personnel participation in the licensee's drills are contained in the Permanently Defueled Emergency Plan.

NRC Question MF2567-RAII-ORLOB-Norris-010 Table 2 I

Kewaunee Request Wording f

Past Precedence Wording 10CFR50 Appendix E IV.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 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.

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 t~~t~d in the~ nr~vin.i~ ~

teste in~~~

th previous_____exercises___

Previous exemptions did not grant an exemption as requested. Biennial exercises are required and are subject to NRC inspection. 10 CFR 50.47(b)(14) states that 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. A remedial exercise ensures that, in the event that an exercise did not provide reasonable

Serial No. 13-390A Page 12 of 22 assurance to the NRC that the license can and will take adequate protective measures in the event of a radiological emergency, the deficiencies are corrected. Please provide justification for exempting this requirement.

Response

As discussed in the response to Questions 7 and 8 above, the intent of the originally requested exemption was to continue testing the adequacy of the emergency response organization, except to perform these tests during the conduct of drills. However, the past precedence wording also meets this intent.

Therefore, DEK is revising the originally requested exemption from portions of 10 CFR 50, Appendix E, IV.F.2.f in Reference 1, as shown in Attachment 2 of this submittal.

The revised request is consistent with the past precedence wording shown above.

Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.

Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.

NRC Question MF2567-RAII-ORLOB-Norris-O1l Table 2 Kewaunee Request Wording Past Precedence Wording IOCFR50 Licensees shall use drill and exercise Licensees shall use drill and exercise Appendix scenarios that provide reasonable scenarios that provide reasonable assurance E IV.F.2.i assurance that anticipatory responses that anticipatory responses will not result from will not result from preconditioning of preconditioning of participants. Such participants. Such scenarios for scenarios for nuclear power reactor nuclear power reactor licensees must licensees must include a wide spectrum of include a wide spectrum of radiological releases and events, including radiological releases and events, hostile action.

including hostile action. Exercise and Exercise and drill scenarios as appropriate drill scenarios as appropriate must must emphasize coordination among onsite emphasize coordination among onsite and offsite response organizations.

and offsite response organizations.

Previous exemptions did not grant an exemption as requested. Biennial exercises are required and are subject to NRC inspection.

10 CFR 50.47(b)(14) that 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,

Serial No. 13-390A Page 13 of 22 and deficiencies identified as a result of exercises or drills are (will be) corrected.

10 CFR 72.32(a)(10) requires training including any special instructions or orientation tours the licensee would offer to fire, police medical and other emergency personnel.

Additionally, 10 CFR 72.32(a)(12)(i) requires the licensee to invite offsite response organizations to participate in the biennial exercise. Additionally, CFR 72.32(a)(12) (ii) states in part, that participation of offsite response organizations in biennial exercises, although recommended, is not required. Please provide justification for exempting this requirement.

Response

As discussed in the response to Questions 7, 8, and 10 above, the intent of the originally requested exemption was to continue testing the adequacy of the emergency response organization, except to perform these tests during the conduct of drills.

However, the past precedence wording also meets this intent.

Therefore, DEK is revising the exemption originally requested from portions of 10 CFR 50, Appendix E, IV.F.2.i in Reference 1, as shown in Attachment 2 of this submittal. The revised request is consistent with the past precedence wording shown above.

Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility. Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.

Applicable details regarding the extent of drills (training activities) and exercises (evaluated activities) are contained in the Permanently Defueled Emergency Plan.

NRC Question MF2567-RAII-ORLOB-Norris-012 The Executive Summary in NUREG-1738 states, in part, "the staffs 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 industry decommissioning commitments (IDCs) and staff decommissioning assumptions (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.

Please explain if/how KPS meets each of these IDCs and SDAs.

Serial No. 13-390A Page 14 of 22

Response

Results of a comparison of the KPS spent fuel pool against the IDCs and SDAs listed in tables 4.1-1 and 4.1-2, respectively, of NUREG-1738, are shown below.

TABLE 1 Industry Decommissioning Commitments (IDCs) Comparison IDC IDC Description KPSIDEK Alignment to IDC Description No.

1. Cask drop analyses will KPS design aligns with this description. DEK controls the handling of heavy be performed or single loads within the protected area by an administrative procedure designed to failure-proof cranes will meet the guidance provided in NUREG-0612. Additionally, the auxiliary be in use for handling of building crane, which is employed for the lifting and handling of heavy loads heavy loads (i.e., phase II in the vicinity of the spent fuel storage pool, is of a single failure proof of NUREG-0612 will be design. The NRC Safety Evaluation Report (SER) for KPS License implemented).

Amendment 200 documents the single failure proof design of the auxiliary building crane (SER dated November 20, 2008 (ML082971079)).

2. Procedures and training DEK practices align with this description. Consistent with KPS Emergency of personnel will be in Plan requirements, DEK has procedures in place to ensure that onsite and place to ensure that offsite resources are available and personnel are trained on the access and onsite and offsite use thereof during an event. The Permanently Defueled Emergency Plan resources can be brought (PDEP) being submitted for NRC approval addresses these requirements.

to bear during an event.

KPS also maintains Letters of Agreement (LOA) or Memoranda of Understanding (MOU) with offsite agencies to ensure additional resources are available if needed. The discussion in KPS/DEK Alignment to IDC Description #4 contains additional details.

3. Procedures will be in DEK practices align with this description. Should severe weather or seismic place to establish events occur that result in an Emergency Plan entry, procedures are in place communication between that direct personnel to establish the necessary communications and make onsite and offsite the appropriate notifications. For example, the Emergency Director (Shift organizations during Manager) would direct notification of the ERO, applicable State and Counties severe weather and officials, and the NRC. As decommissioning progresses, DEK anticipates seismic events, maintaining procedures in place to require the appropriate communication between onsite and offsite organizations.
4. An offsite resource plan DEK practices align with this description. Consistent with directions provided will be developed which in the Emergency Plan, DEK maintains an Emergency Telephone Directory will include access to (ETD). The ETD provides the information necessary to access necessary portable pumps and offsite resources in a timely manner. Appropriate station personnel are emergency power to trained to use the ETD to obtain offsite resources when needed to support supplement onsite onsite resources.

resources. The plan would principally identify ETD subsection ETD 02 lists contacts for government agencies, emergency organizations or suppliers equipment contacts (e.g., for fuel, electrical power, makeup water, where offsite resources firefighting equipment). It also identifies private agencies that would be could be obtained in a capable of providing resources if requested (such as INPO, NEI, Bartlett timely manner.

Nuclear, and Point Beach Nuclear Plant).

ETD subsection ETD 03 has a section specifically for emergency contacts.

It lists items such as a portable diesel driven pump, diesel fuel, construction

Serial No. 13-390A Page 15 of 22 IOC IDC Description KPSIDEK Alignment to IDC Description No.

and lifting equipment, firefighting equipment, electrical power equipment, compressed gas and air and it identifies the Pooled Equipment Inventory Co., to which DEK subscribes.

As decommissioning progresses, DEK intends to maintain an ETD and necessary subscriptions in place to facilitate the timely acquisition of outside resources if they are needed during Emergency Plan implementation activities.

5.

SFP instrumentation will The KPS design aligns with the intent of this description. Spent fuel pool include readouts and (SFP) instrumentation provides alarms in the control room for: Elevated SFP alarms in the control Area Radiation Level (alert and high level setpoints), SFP Level Hi (two room (or where personnel channels), SFP Level Lo (two channels), and SFP Temperature Hi (two are stationed) for SFP channels). Also, the SFP area radiation level reads out in the control room.

temperature, water level, Additionally, the SFP instrumentation provides local readouts of SFP level and area radiation levels, and temperature. A local alarm to notify personnel of high area radiation levels is also in place. Additionally, DEK intends to improve SFP level monitoring instrumentation. After the planned modifications, KPS will have an additional level monitoring capability (radar). This added feature will provide two channels of pool level readout locally (in the SFP heat exchanger room from which mitigating action would be taken) and low level alarms in the control room.

6. SFP seals that could The KPS design aligns with this description. The design of the SFP gates is cause leakage leading to self limiting to prevent draining to a point where the fuel would be uncovered.

fuel uncovery in the event The relative elevation between the bottom of the SFP gate openings and the of seal failure shall be top of spent fuel assemblies while stored within the spent fuel storage racks self limiting to leakage or ensures that the inadvertent drainage or leakage via spent fuel gate opening otherwise engineered so cannot uncover the fuel - i.e., the bottom of the gate opening is above the that drainage cannot top of stored spent fuel assemblies.

occur.

7. Procedures or DEK practices align with this description. DEK employs a formal procedure administrative controls to to allow for pumping down specified volumes within the SFP. Although the reduce the likelihood of method used does not have the capability to drain down the SFP rapidly, the rapid draindown events procedure includes the key elements called out in IDC 7. Similarly, the KPS will include (1)

ISFSI equipment design is such that there are no ISFSI-related SFP prohibitions on the use of operations that have the potential to cause a rapid drain down event. The pumps that lack adequate DEK procedure governing "procedure use and adherence" requires siphon protection or (2) procedure users to conform to process requirements established to control controls for pump suction specific activities. Moreover, work activities, whether performed under a and discharge points, specific procedure designed to control that activity, or under the general The functionality of anti-process controls of the work control process, are subject to the DEK siphon devices will be integrated risk management procedure, wherein appropriate measures are periodically verified, employed to assess and manage the risk associated with such activities (e.g., address the affects upon SFP cooling availability).

8. An onsite restoration plan DEK practices align with this description. DEK maintains in place will be in place to provide contingency work orders to support the repair and replacement of key SFP repair of the SFP cooling cooling components. Similarly, the necessary work orders are maintained in systems or to provide place to support the alignment of a Residual Heat Removal system heat access for makeup water exchanger to allow for its use to cool the SFP should it be needed.

to the SFP. The plan will Additionally, there are procedures in place that provide for the use of a

Serial No. 13-390A Page 16 of 22 IDC IDC Description KPSIDEK Alignment to IDC Description No.

provide for remote backup means (beyond the NRC Safety Guide 13 capabilities required by alignment of the makeup the KPS design basis for fuel pool makeup) of SFP water makeup which can source to the SFP without be executed without requiring entry to the refuel floor.

requiring entry to the refuel floor.

9. Procedures will be in DEK practices align with this description. The KPS SFP design precludes an place to control SFP operationally induced rapid decrease in SFP inventory. DEK employs a operations that have the formal procedure to allow for pumping down specified volumes within the potential to rapidly SFP. Similarly, the KPS ISFSI equipment design is such that there are no decrease SFP inventory.

ISFSI-related SFP operations that have the potential to cause a rapid drain These administrative down event. The DEK process governing "procedure use and adherence" controls may require requires procedure users to conform to process requirements established to additional operations or control specific activities. Moreover, work activities, whether performed management review, under a specific procedure designed to control that activity, or under the management physical general process controls of the work control process, are subject to the DEK presence for designated integrated risk management procedure wherein appropriate measures are operations or employed to assess and manage the risk associated with such activities. For administrative limitations example, the integrated risk management procedure requires DEK such as restrictions on management to consider requiring direct supervisory oversight or heavy load movements, engineered mitigation methods. Also, the DEK "Decommissioning Safety Assessment Checklist" provides guidance relative to heavy loads at the station. Finally, implementing procedures related to the use of cranes at KPS provide the needed restrictions upon the handling of heavy loads, with specific requirements related to the handling of such loads in the vicinity of the SFP.

10. Routine testing of the DEK practices align with this description. The KPS design basis credits the alternative fuel pool station's seismically designed service water (SW) system for meeting the makeup system NRC Safety Guide 13 requirements for SFP makeup. The SW system has components will be redundant pumping capability and redundant power supplies adequate to performed and support the SFP makeup function. The station SW system is continuously administrative controls for operating, allowing for continuous monitoring for proper operation, and equipment out of service provides pressurized water from Lake Michigan to the inlet of a single will be implemented to manual isolation valve that can be repositioned locally to supply makeup provide added assurance water to the SFP. For defense-in-depth, the station also has available, with that the components supporting procedures for its use, an engine driven emergency makeup would be Available, if pump capable of delivering Lake Michigan water to the SFP. By procedure, needed.

both the design basis and defense-in-depth capabilities are routinely tested to ensure their ongoing availability. The process for testing the design basis makeup water supply isolation valve includes an explicit step requiring that a test failure be addressed via the corrective action program. This complies with the corrective action program's general requirement that plant System, Structure, or Component (SSC) failures be entered into the program.

Serial No. 13-390A Page 17 of 22 TABLE 2 Staff Decommissioning Assumptions (SDAs) Comparison SDA SDA Description KPS/DEK Alignment to SDA Description No.

I.

Licensee's SFP cooling KPS design aligns with the intent of this description. The KPS SFP design will be at least as Cooling System design is based, in part, on NRC Safety Guide 13. Safety capable as that assumed Guide 13 requires a seismic category 1 system for providing makeup in the risk assessment, water to the SFP. This design basis requirement for SFP cooling is including instrumentation.

provided by the SW system, which is a Nuclear Safety Design Class 1*

Licensees will have at system (i.e., it is designed to withstand design basis earthquake least one motor-driven and seismically induced load) protected by a Nuclear Safety Design Class I one diesel-driven fire structure. The SW system has redundant pumping capability and is pump capable of delivering provided with redundant power supplies adequate to provide SFP makeup inventory to the SFP.

at the required capacity.

The SW pumps are normally powered from offsite power, but can be supplied with backup power from the emergency diesel generators (EDGs), which are also Class 1 components housed in a Class 1 structure.

The station design also includes two motor driven fire pumps, each with the ability to be powered from either off site power or from either of two EDGs. The fire pumps have the capability to deliver water to the SFP for makeup.

Finally, the station also maintains available, with supporting procedures for its use, a diesel engine powered emergency makeup pump capable of supplying water from Lake Michigan to the SFP.

2.

Walk-downs of SFP DEK practices align with the intent of this description. Station procedures systems will be performed require a member of the staff to tour the SFP area each shift. Proper at least once per shift by system operation is verified once per shift by verifying and recording the operators.

normal SFP level and temperature. Additional verifications are performed daily by direct observation of the proper operation and status of SFP Procedures will be pumps. The normal and alternate SFP makeup water sources (including developed for and related tank levels) are also verified daily to ensure that they remain employed by the operators available.

to provide guidance on the capability and availability Station procedures require continual verification that the SW system of onsite and offsite (design basis makeup source) is in operation; thereby ensuring that SFP inventory makeup sources emergency makeup is available.

and time available to initiate these sources for Moreover, the KPS guidance in place for "Recovery Plan for Catastrophic various loss of cooling or Event" includes instructions for the following specific actions/events:

inventory events.

SEP makeup Alternate method to ventilate the SFP area SFP leakage control Additional resources Finally, KPS procedure "Validation of Time Sensitive Operator Actions,"

provides the guidance necessary to ensure that operators have sufficient

Serial No. 13-390A Page 18 of 22 SDA SDA Description KPS/DEK Alignment to SDA Description No.

time available to initiate makeup water sources for various loss of cooling or inventory events.

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.

KPS design aligns with the intent of this description. Item 7 in KPS Updated Safety Analysis Report (USAR), Table 9.5-2, "Design Conformance with Safety Guide 13, states that "Level measuring instrumentation and radiation monitoring equipment are provided which alarm both locally and in the Control Room." Additionally, DEK has initiated actions to enhance SFP level monitoring instrumentation. These enhancements are intended to provide an additional level monitoring capability (radar). The planned level instrumentation includes two channels that provide local pool level indication (in the SFP heat exchanger room, from which mitigating action would be taken). Each level channel also provides input to a low level alarm in the Control Room. DEK processes in place to respond to an abnormally low level in the SFP direct the plant staff to take appropriate actions to provide SFP makeup, first through normal means, then by utilizing available onsite resources, including both design basis and defense-in-depth capabilities. Ultimately, if the use of onsite means fails to restore SFP inventory to an acceptable level, processes would direct the plant staff to access offsite resources.

To facilitate accessing offsite resources, DEK maintains an Emergency Telephone Directory (ETD). The ETD provides the information necessary to access necessary offsite resources in a timely manner. Appropriate station personnel are trained to use the ETD to obtain offsite resources when needed to support onsite resources. ETD subsection ETD 02 lists contacts for government agencies, emergency equipment contacts (e.g.,

for fuel, electrical power, makeup water, firefighting equipment). It also identifies private agencies that would be capable of bringing resources when needed such as INPO, NEI, Bartlett Nuclear, and Point Beach Nuclear Plant. ETD subsection ETD 03 has a section specifically for emergency contacts. It lists items such as a portable diesel driven pump, diesel fuel, construction and lifting equipment, firefighting equipment, electrical power equipment, compressed gas and air, and it identifies the Pooled Equipment Inventory Co., to which DEK subscribes.

Regarding the declaration of a general emergency, KPS will be employing Shutdown EALs using an approved NRC EAL Scheme. Consistent with that scheme, there are no conditions that have the capacity to reach any threshold requiring the declaration of a general emergency.

-I.

I 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.

KPS design aligns with this description. Nominal SFP water level is approximately 27 feet above the stored fuel. As stated in KPS USAR, Section 9.3.2.3, "The SFP pump suction lines are located well above the fuel assemblies and a system failure cannot result in loss of pool water.

The return lines enter the pool above the top of the fuel assemblies and the lines contain check valves at the point of entry into the pool shielding concrete. Thus, line failure outside of the SFP cannot cause a loss of pool water due to siphon action." The SSCs relied upon to preserve the SFP inventory are seismically designed SSCs that preclude any significant loss in fuel pool inventory, via the SFP cooling water suction lines, under design basis conditions. Additionally, the SFP cooling water return lines

Serial No. 13-390A Page 19 of 22 SDA SDA Description KPS/DEK Alignment to SDA Description No.

(also seismically qualified) terminate approximately 15 feet above the stored fuel. This limits any reduction in SFP level via the return lines to approximately 12 feet.

5.

Load Drop consequence KPS design aligns with this description. The auxiliary building crane, which analyses will be performed is employed for the lifting and handling of heavy loads in the vicinity of the for facilities with non single spent fuel storage pool, is of a single failure proof design. The NRC failure-proof systems. The Safety Evaluation Report (SER) for KPS License Amendment 200 analyses and any documents the single failure proof design of the auxiliary building crane mitigative actions (SER dated November 20, 2008 (ML082971079)). Additionally, DEK necessary to preclude controls the handling of heavy loads within the protected area by an catastrophic damage to administrative procedure designed to meet the guidance provided in the SFP that would lead to NUREG-0612.

a rapid pool draining would be sufficient to Control of heavy loads is governed by the Technical Requirements Manual demonstrate that there is (TRM), specifically, TRM 8.9.1, "Spent Fuel Pool - Control of Heavy high confidence in the Loads," which is subject to the requirements of 10 CFR 50.59.

facilities ability to withstand a heavy load drop.

6.

Each decommissioning Item 10 of the seismic checklist provides an alternative wherein the plant will successfully licensee delays request for a licensing "waiver" (i.e. License Amendment complete the seismic Request) for Emergency Planning until the plant specific zirconium fire is checklist provided in no longer a credible concern. DEK has performed an analysis (Calculation Appendix 2B to this study.

2013-11284, "Maximum Cladding and Fuel Temperature Analysis for If the checklist cannot be Uncovered Spent Fuel Pool"), which concludes that, about 17 months after successfully completed, reactor shutdown, decay heat cannot raise the spent fuel cladding the decommissioning plant temperature sufficiently to cause clad failure (565°C) if all water is drained will perform a plant specific from the SFP. Therefore, as of October 2014, when the requested seismic risk assessment of changes will be implemented, the plant specific zirconium fire will no the SFP and demonstrate longer be a credible concern.

that SFP seismically induced structural failure Additionally, KPS is located in a geologically stable region whose seismic and rapid loss of inventory hazard risk is very low as documented in recent seismic hazard estimates is less than the generic (based on U.S. Geological Survey (USGS) of 2008). Geologic bounding estimates investigations throughout the Lake Michigan basin have not found any provided in this study (<1 indication of fault movement in the recent geologic past. As shown in x105 per year including Figure 2 of Generic Issue 199 (GI-1 99, August 2010), the peak horizontal non-seismic events),

acceleration (%g) for 2-percent probability of exceedance in 50 years, for the geographic region where KPS is located, is in the second lowest region of the conterminous United States (between 0.02 and 0.03 g).

Based on existing knowledge of the SFP structural capabilities, the SFP seismically induced structural failure and rapid loss of SFP inventory is very unlikely.

Finally, KPS has procedures in place to ensure successful implementation of mitigation measures to supply alternate cooling water using portable equipment. As a result, no radiological releases with offsite consequence is expected or should occur following a severe earthquake because KPS has been permanently shutdown since May 2013, and mitigation measures for cooling water are in place.

Serial No. 13-390A Page 20 of 22 SDA SDA Description KPS/DEK Alignment to SDA Description No.

7.

Licensees will maintain a DEK procedures align with this description. KPS does not utilize Boraflex program to provide in any of its spent fuel storage racks. Rather, the KPS design employs surveillance and boron carbide (B4C) in the spent fuel storage racks in the north and south monitoring of Boraflex in pools and Boral neutron absorber material in the north canal spent fuel high-density spent fuel storage racks.

racks until such time as spent fuel is no longer stored in these high-density racks.

NRC Question MF2567-RAII-ORLOB-Norris-013 Page 43 of 55 references a site-specific adiabatic heat up to address a partial drain down of the SFP (identified as Reference 12), assuming no air-cooling it states that the time necessary for the hottest fuel assembly to reach the critical temperature of 5650C is six hours after the fuel rods have become uncovered. Some previous exemptions were granted based [on] time to reach the cladding auto-ignition temperature of 9000C.

Based on 17 months of decay time, at what time after the fuel is uncovered, assuming an adiabatic heatup, would the hottest fuel assembly reach 900°C? Please provide a copy of this analysis and the existing analysis for the Partial Loss of Cooling Water Inventory with No Air Cooling.

Response

Calculation 2013-07050, "Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with No Air Cooling," shows that after approximately 17 months of decay time, the hottest fuel assembly in the spent fuel pool would reach 9000C in 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the fuel is uncovered, assuming no air cooling (adiabatic heatup)

(documented in Section 7, "Results" and Figure 7-1, "Heat-Up Time vs. Decay Time").

Seventeen months of decay time will occur on October 21, 2014.

A copy of this site-specific adiabatic heat up analysis to address a partial drain down of the SFP, assuming no air-cooling, is provided in Enclosure 1 to this submittal (Calculation 2013-07050, Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with No Air Cooling). This site-specific quantitative heat up analysis includes both the time to reach 5650C and the time to reach 9000C (the previous analysis for the heat up to 5650C, identified as Reference 12 in the exemption request, was a qualitative analysis whose results have been subsumed into the current analysis).

Serial No. 13-390A Page 21 of 22 NRC Question MF2567-RAII-ORLOB-Norris-014 Page 43 of 55 references spent fuel pool inventory makeup strategies. Please provide additional information related to:

a. What is the availability of trained personnel to perform the required actions?
b. How is the referenced equipment maintained and tested?
c. Are there procedures developed to perform this task and how are they controlled?
d. Will these procedures and equipment be referenced in the emergency plan since the basis for this exemption, in part, is the existence of these mitigative strategies, until such time that the spent fuel has decayed to a point where they are no longer needed or the spent fuel is placed in a dry ISFSI?

Response

a. Availability of trained personnel to perform the required actions The on-shift Plant Operators and Fire Brigade members are appropriately trained on the various actions needed to provide makeup to the spent fuel pool (SFP) based on a systematic approach to training. Because KPS is no longer operating, maintaining SFP cooling and inventory would be the highest priority activity; therefore, the personnel needed to perform these actions are available at all times.
b. Referenced equipment maintained and tested Existing plant systems used for SFP makeup are maintained and tested using plant procedures in accordance with the KPS preventive maintenance program.

This includes testing the capability to align emergency makeup via the service water system. The diesel-driven portable pump is maintained and tested using plant procedures in accordance with the KPS preventative maintenance program.

In addition, flow testing of external makeup capacity is performed periodically to validate that the specified actions can be completed in a timely manner.

c. Procedures developed to perform this task and how they are controlled Operating procedures (NOP-SFP-001, "Spent Fuel Pool Cooling and Cleanup System" and AOP-SFP-001, "Abnormal Spent Fuel Pool Cooling and Cleanup System Operation") provide direction for supplying makeup water to the spent fuel pool using existing plant systems in the event of a loss of level. If these procedurally directed strategies do not result in restoration of level, then response plans (which are in place to address large area fires) would be implemented which direct personnel to provide external makeup water via a portable diesel-driven pump.

Administrative controls (GNP-03.01.01, "Directive, Implementing Document, and Procedure Administrative Controls") are in place to ensure that procedures are

Serial No. 13-390A Page 22 of 22 maintained and implemented, and that any changes to them are appropriately reviewed and approved (including any applicable requirements of 10 CFR 50.59).

d. Referencing these procedures and equipment in the emergency plan These procedures and equipment are not specifically referenced in the existing KPS Emergency Plan and are not included in the planned Permanently Defueled Emergency Plan (to be submitted for NRC approval).

These procedures are required by TS 5.4.1.a, which directs establishing, implementing, and maintaining applicable procedures recommended in RG 1.33, Revision 2, Appendix A.

Therefore, it is not necessary for them to be specifically referenced in the Emergency Plan. Equipment requirements are specified in the pertinent procedures.

References

1. Letter from A. J. Jordan (DEK) to NRC Document Control Desk, "Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E," dated July 31, 2013.
2. Email from Dr. Karl D. Feintuch (NRC) to Margaret Earle, Jack Gadzala, Craig Sly, et al (DEK), "MF2567 Kewaunee Emergency Plan Requests for Exemption MF2567-RAII-ORLT-Norris-001 to -014 8 October 2013," dated October 8, 2013.

Serial No. 13-390A ATTACHMENT 2 SUPPLEMENT I TO DEK REQUEST FOR EXEMPTIONS:

REQUEST FOR EXEMPTIONS FROM PORTIONS OF 10 CFR 50.47(b),

10 CFR 50.47(c)(2), AND 10 CFR 50, APPENDIX E, SECTION IV KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

Serial No. 13-390A Page 1 of 10 Supplement 1:

Request for Exemptions from Portions of 10 CFR 50.47(b),

10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV I.

DESCRIPTION By application dated July 31, 2013 (Reference 1), Dominion Energy Kewaunee, Inc.

(DEK) requested exemptions, pursuant to 10 CFR 50.12 "Specific exemptions," from portions of 10 CFR 50.47(b), 10 CFR 50.47(c)(2), and 10 CFR 50, Appendix E, Section IV, for Kewaunee Power Station (KPS). The requested exemptions would allow DEK to reduce emergency planning requirements and subsequently revise the KPS Emergency Plan consistent with the permanently defueled condition of the station.

In response to the staff's comments, DEK is revising the originally proposed exemption request. Attachment 2 to this letter provides a supplement to the proposed exemption request. The analyses provided in Reference 1 remain applicable and bounding to this revised request. The conclusions of the no significant hazards consideration and the environmental considerations contained in Reference 1 are not affected by, and remain applicable to, this revised request.

A. Revised Exemptions Requested from 10 CFR 50.47 Table 1 (Revised) below lists the pertinent portions of 10 CFR 50.47(b) and 10 CFR 50.47(c)(2) in the left column.

The specific portion of the requirement within the regulation from which exemption is being requested is emphasized (bold/underlined).

The basis for the exemption from the specific portion of each requirement is provided in the corresponding row of the column on the right.

The table below shows only the regulations for which a revision to the originally requested exemption (Reference 1) is being proposed. The rows shown in Table 1 (Revised), below, replace the corresponding rows listed in Table 1 of Reference 1 in their entirety. The requested exemptions from all other regulations shown in Table 1 of Reference 1 remains in effect.

Serial No. 13-390A Page 2 of 10 TABLE 1 (Revised)

Exemptions Requested from 10 CFR 50.47 Regulation Basis for Requested Exemption (portion being exempted shown emphasized) 10 CFR 50.47(b)(1) - Primary responsibilities for Revised radiological analyses have been emergency response by the nuclear facility licensee developed that show that, 90 days after shutdown, and by State and local organizations within the the radiological consequences of design basis Emergency Planning Zones have been assigned, accidents will not exceed the limits of the EPA the emergency responsibilities of the various Protective Action Guides at the EAB. In addition, supporting organizations have been specifically analyses have been developed for beyond design established, and each principal response basis events related to the spent fuel pool which organization has staff to respond and to augment its show that, within 17 months after shutdown, the initial response on a continuous basis.

analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.

Therefore, there will no longer be a need for Emergency Planning Zones.

State and local government agency response will be in accordance with each agency's plans and procedures, and commensurate with the hazard posed by the emergency.

10 CFR 50.47(b)(7)

Information is made Revised radiological analyses have been available to the public on a periodic basis on developed that show that, 90 days after shutdown, how they will be notified and what their initial the radiological consequences of design basis actions should be in an emergency (e.g.,

accidents will not exceed the limits of the EPA listening to a local broadcast station and Protective Action Guides at the EAB. In addition, remaining indoors), the principal points of contact analyses have been developed for beyond design with the news media for dissemination of information basis events related to the spent fuel pool which during an emergency (including the Physical show that, within 17 months after shutdown, the location or locations) are established in advance, analyzed event is either not credible, is capable of and procedures for coordinated dissemination of being mitigated, or the event's radiological information to the public are established, consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. There will be no need for the public to take any protective actions in the event of an emergency at KPS. Therefore, there will no longer be any need for information to be made available to the public about how they will be notified and what their initial protective actions should be.

Serial No. 13-390A Page 3 of 10 Regulation (portion being exempted shown emphasized)

Basis for Requested Exemption 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.

Revised radiological analyses have been developed that show that, 90 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.

Therefore, assessing and monitoring of offsite consequences of radiological emergency conditions will no longer be required.

Since a need for monitoring and assessing will no longer exist, DEK no longer intends to maintain the capability to deploy field teams for assessing and monitoring offsite radiological conditions.

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 from 10 CFR 50.47(b)(14) is being proposed.

B. Exemptions Requested from 10 CFR 50, Appendix E Table 2 (Revised) below lists the pertinent portions of 10 CFR 50, Appendix E, Section IV, in the left column. The specific portion of the requirement within the regulation from which exemption is being requested is emphasized (bold/underlined). The basis for the exemption from the specific portion of each requirement is provided in the corresponding row of the column on the right.

The table below shows only the regulations for which a revision to the originally requested exemption (Reference 1) is being proposed. The rows shown in Table 2 (Revised), below, replace the corresponding rows listed in Table 2 of Reference 1 in their entirety. The requested exemptions from all other regulations shown in Table 2 of Reference 1 remains in effect.

Serial No. 13-390A Page 4 of 10 TABLE 2 (Revised)

Exemptions Requested from 10 CFR 50, Appendix E Regulation (10 CFR 50, Appendix E)

(portion being exempted shown emphasized)

Basis for Requested Exemption i

§ IV.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 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.

Revised radiological analyses have been developed that show that, 90 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. Therefore, offsite emergency response plans for local government authorities will no longer be necessary.

Since offsite emergency plans will no longer be necessary, and based on the significantly reduced scope of EALs for the permanently defueled facility, the scope of the annual review of EALs with State and local governmental authorities is expected to be commensurately reduced (e.g.,

informational mailings, etc.).

Justification from the requirements in Appendix E related to a "hostile action" is provided in the Basis for the requested exemption from § IV. 1 above.

§ IV.F.1 - In addition, a radiological orientation training program shall be made available to local services personnel; e.g.,

local emergency serviceslCivil Defense, local law enforcement personnel, local news media persons.

Revised radiological analyses have been developed that show that, 90 days. after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. Therefore, offsite emergency response plans will no longer be necessary.

Local news media persons no longer need radiological orientation training since they will not be called upon to respond to a radiological event.

The term "Civil Defense" is no longer commonly used; therefore, reference to this term in the examples provided in the regulation is not needed.

Serial No. 13-390A Page 5 of 10 Regulation (10 CFR 50, Appendix E)

(portion being exempted shown emphasized)

Basis for Requested Exemption

§ IV.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.

Revised radiological analyses have been developed that show that, 90 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.

There will be no need for the public to take any protective actions in the event of an emergency at KPS. Therefore, participation by offsite entities will no longer be necessary, public alert and notification system will no longer be required.

Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility.

Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.

Serial No. 13-390A Page 6 of 10 Regulation (10 CFR 50, Appendix E)

Basis for Requested Exemption (portion being exempted shown emphasized)

§ IV.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.

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.

Revised radiological analyses have been developed that show that, 90 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB. There will be no need for the public to take any protective actions in the event of an emergency at KPS.

Therefore, participation by offsite entities will no longer be necessary and associated exercises will no longer need to be conducted.

Although NUREG-0654 states that "an exercise shall include mobilization of State and local personnel and resources adequate to verify the capability to respond to an accident scenario requiring response," such an exercise scenario scope is not necessary for a permanently defueled facility.

Performance of reduced scope exercises is sufficient to maintain and assess the capability of the emergency response organization to properly perform activities.

Offsite emergency response plans will no longer be necessary and there will be no required response by offsite agencies to the EOF. An EOF will no longer be maintained.

An onsite facility (whether the control room or a facility similar to the technical support center) would continue to be maintained, from which effective control can be exercised during an emergency.

Serial No. 13-390A Page 7 of 10 Regulation (10 CFR 50, Appendix E)

Basis for Requested Exemption (portion being exempted shown emphasized)

§ IV.F.2.e - Licensees shall enable any State or Revised radiological analyses have been local government located within the plume developed that show that, 90 days after shutdown, exposure Pathway EPZ to participate in the the radiological consequences of design basis licensee's drills when requested by such State or accidents will not exceed the limits of the EPA local government.

Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.

Therefore, the plume exposure pathway emergency planning zone, and offsite plans and drills will no longer be necessary. In the context of this paragraph of the regulation, "any State" means Wisconsin and "local government" means the organizations that provide emergency support services (i.e. ambulance, fire, police) to KPS upon request.

§ IV.F.2.f - Remedial exercises will be required if Revised radiological analyses have been the emergency plan is not satisfactorily tested developed that show that, 90 days after shutdown, during the biennial exercise, such that NRC, in the radiological consequences of design basis consultation with FEMA.

cannot (1) find accidents will not exceed the limits of the EPA reasonable assurance that adequate protective Protective Action Guides at the EAB. In addition, measures can and will be taken in the event of a analyses have been developed for beyond design radiological emergency or (2) determine that the basis events related to the spent fuel pool which Emergency Response Organization (ERO) has show that, within 17 months after shutdown, the maintained key skills specific to emergency analyzed event is either not credible, is capable of response. The extent of State and local being mitigated, or the event's radiological participation in remedial exercises must be consequences will not exceed the limits of the EPA sufficient to show that appropriate corrective Protective Action Guides at the EAB. Therefore, measures have been taken regqarding the offsite emergency response plans will no longer be elements of the plan not Properly tested in the necessary and the scope of exercises can be previous exercises.

commensurately reduced.

Serial No. 13-390A Page 8 of 10 Regulation (10 CFR 50, Appendix E)

Basis for Requested Exemption (portion being exempted shown emphasized)

§ IV.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.

Revised radiological analyses have been developed that show that, 90 days after shutdown, the radiological consequences of design basis accidents will not exceed the limits of the EPA Protective Action Guides at the EAB. In addition, analyses have been developed for beyond design basis events related to the spent fuel pool which show that, within 17 months after shutdown, the analyzed event is either not credible, is capable of being mitigated, or the event's radiological consequences will not exceed the limits of the EPA Protective Action Guides at the EAB.

Requirements for offsite planning will no longer be necessary. Therefore, the scope of exercises can be commensurately reduced.

Following docketing of its "Certification of Permanent Removal of Fuel from the Reactor Vessel,"

dated May 14,

2013, KPS is a

permanently shutdown facility with spent fuel stored in the spent fuel pool and ISFSI. In the EP Final Rule (76 FR 72596, Nov. 23, 2011), the Commission defined "hostile action" as, in part, an act directed toward a nuclear power plant or its personnel. The NRC excluded non-power reactors (NPR) 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.

Likewise, spent fuel pools and ISFSIs are not a nuclear power plant.

The following similarities between the KPS facility and NPRs show that the KPS facility should be treated similarly to NPRs. Similar to NPRs, KPS poses lower radiological risks to the public from accidents than do power reactors because: (1)

KPS is a permanently shutdown facility (with fuel stored in the spent fuel pool and ISFSI) and no longer generates fission products; 2) Fuel stored in the KPS SFP has lower decay heat, resulting in lower risk of fission product release in the event of a non-credible boil off or draindown event; and 3) no credible accident at KPS will result in radiological releases requiring offsite protective actions. NPRs have lower decay heat associated with a lower risk of core melt and fission product release in a loss-of-coolant accident.

Likewise, KPS has a low likelihood of a credible accident resulting in radiological releases requiring offsite protective actions.

Serial No. 13-390A Page 9 of 10 The portions of 10 CFR 50.47 and 10 CFR 50, Appendix E that are not identified in Tables 1 and 2 of Reference 1, as modified in the two tables above (i.e., those portions for which exemption is not being requested), will remain applicable to KPS.

II.

BACKGROUND The background information contained in Reference 1 remains applicable to this supplement.

As discussed in Reference 1, an analysis of the potential radiological impact of a design basis accident at KPS in a permanently defueled condition indicates that any potential radiological releases beyond the site boundary would be below the EPA PAG exposure levels, as detailed in the EPA's "Protective Action Guide and Planning Guidance for Radiological Incidents," Draft for Interim Use and Public Comment dated March 2013 (PAG Manual).

As stated in Reference 1, the KPS USAR contains the following description regarding spent fuel pool indication available to operators for responding to a postulated loss of heat removal capability for the spent fuel pool.

Both temperature and level indicators in the pool would alert operators to a loss of cooling. Local and remote alarms are provided. This allows the operator to take corrective measures in a timely manner to restore cooling capability to the spent fuel pool cooling loop.

A recent review of this USAR description revealed that the sentence regarding local and remote alarms is ambiguous and could imply that local and remote alarms are provided both for pool temperature and for pool level. Although installation of a new spent fuel pool level indication system will include both a local and remote alarm, only a remote alarm is provided for spent fuel pool temperature. These indications and alarms are considered adequate to allow timely operator response to an abnormal spent fuel pool condition.

II1. JUSTIFICATION FOR EXEMPTIONS AND SPECIAL CIRCUMSTANCES 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 Part 50 which are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the defense and security.

10 CFR 50.12 also states that the Commission will not consider granting an exemption unless special circumstances are present.

The justification for exemptions and special circumstances contained in Reference 1 are not affected by, and remain applicable to, this supplement. Therefore, this exemption request satisfies the provisions of Section 50.12.

Serial No. 13-390A Page 10 of 10 IV. ENVIRONMENTAL CONSIDERATION The conclusions of the environmental considerations contained in Reference 1 are not affected by, and remain applicable to, this supplement.

V.

CONCLUSION The conclusions contained in Reference 1 are not affected by, and remain applicable to, this supplement. Therefore, the requested exemptions, as supplemented herein, 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, and special circumstances are present as set forth in 10 CFR 50.12(a)(2).

REFERENCES

1. Letter from A. J. Jordan (DEK) to NRC Document Control Desk, "Request for Exemptions from Portions of 10 CFR 50.47 and 10 CFR 50, Appendix E," dated July 31, 2013
2. Email from Dr. Karl D. Feintuch (NRC) to Margaret Earle, Jack Gadzala, Craig Sly, et al (DEK), "MF2567 Kewaunee Emergency Plan Requests for Exemption MF2567-RAII-ORLT-Norris-001 to -014 8 October 2013," dated October 8, 2013.

Serial No. 13-390A ENCLOSURE 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION:

REQUEST FOR EXEMPTIONS FROM PORTIONS OF 10 CFR 50.47(b),

10 CFR 50.47(c)(2), AND 10 CFR 50, APPENDIX E, SECTION IV SUPPORTING CALCULATION

1. CALCULATION 2013-07050, MAXIMUM CLADDING TEMPERATURE ANALYSIS FOR AN UNCOVERED SPENT FUEL POOL WITH NO AIR COOLING KEWAUNEE POWER STATION DOMINION ENERGY KEWAUNEE, INC.

SFDominion Calculation Cover Sheet I

CMAACC-0 ATAHET26ae1o Note: This form is only applicable to Revision 6 of this procedure.

Complete the fields with text or an X as required.

Calculation Number:

2013-07050 Vendor (If not Dominion):

Sargent & Lundy Calculation Preparation Risk: OLow E]Medium

-]High HU-RISK-20130082 Vendor Proprietary: E]Yes SNo Pre-Job Brief Completed: ElYes EINo ENA Calculation Quality Class:

E-Safety Related ONSQ E]Non-Safety Related Subject (Calculation Title):

Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with no Air Cooling Addendum

Title:

N/A Station(s) and Unit(s): NA Eli E12 1773 Affected System(s), Structure(s), or Component(s):

SU [-1l E-2 KW 0 MP Ell 1[2 Ei3 21-SFP-Spent Fuel Cooling EICO (Note: If both SU and NA then only check CO)

Purpose (Executive Summary):

This vendor calculation was performed by Sargent & Lundy to their Quality Program. It evaluates the length of time it takes for uncovered spent fuel assemblies to reach the temperature where the zircaloy might ignite for the unlikely postulated event of a loss of spent fuel pool inventory. The analysis assumes no air cooling of the fuel assemblies.

Originator (Qual. Required): Printed Name (1) (3)

Signature: (1) (3)

Date: (1) (3)

N/A N/A N/A Reviewer (Qual. Required): Printed Name (1)

Type of Review: (2)

Signature:

Date:

M. S. Lico OwnerNm Sgare g /u Approve r: Printed Namew.J ai Sign. -*/b*at ure-/

Date: *:*t**//_*

(i Note: Physical or electronic signatures are acceptable.

Note: At the discretion of the originator, a facsimile of this cover sheet that does not contain the "CM-AA-CLC-301"or "Attachment 1" headers may be used. Facsimiles must contain all of the elements of the cover sheet in the current revision of CM-AA-CLC-301. (1) Add lines for additional originators or reviewers as necessary. (2) Note if reviews are "Independent," "Peer", "Subject Matter Expert", "Supervisor", or "Owner's". (3) Enter N/A for Owner's Review of Vendor Calculation.

731189 (Mar 2012)

JFDominion Calculation Review Checklist I

CMAACC-0 ATAHET4 ae1o Calculation # 2013-07050 Rev. 2 Add. N/A NOTE: If "Yes" is not answered, an explanation may be provided below. Reference may be made to explanations contained in the calculation or addendum.

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

I 1 ]

2. Are the sources of design inputs up-to-date and retrievable/attached to the calculation?

[X]

[ ]

3. Where appropriate, have the other disciplines reviewed or provided the design inputs for which they are responsible?

[X 1 [

4. Have design inputs been confirmed by analysis, test, measurement, field walkdown, or other pertinent means as appropriate for the configuration analyzed?

I I V ]

5. Have the bases for assumptions been adequately and clearly presented and are they bounded by the Station Design Basis?

[X I I

6. Were appropriate calculation/analytic methods used and are outputs reasonable when compared to inputs?

[X 1

[

7.

Are computations technically accurate?

[X I

[

8. Has the calculation made appropriate allowances for instrument errors and calibration equipment errors?

[X]

[

9.

Have those computer codes used in the analysis been referenced in the calculation?

[X ]

[ I

10. Have all exceptions to station design basis criteria and regulatory requirements been identified and justified in accordance with NQA-1-1994?

I I

[X]

11. Has the design authority/original preparer for this calculation been informed of its revision or addendum, if required?

I I

[X]

Comments provided to S & L resolved in the final draft of the calculation to Dominion's satisfaction.

Item # 4 - Input data is from the literature or provided in ETE-NAF-2013-0077, Rev. 0 (documented in S & L calc).

  1. 10 - Calculation provides information and is not used for design basis criteria.
  1. 11 - The original preparer of Rev. 2 is the same as Rev. 0 and 1.

Signature:

N/A Date:

N/A (Preparer)

Signature:

M. S. Lico Date:

9"16 F/f3 (Reviewer)

Note: Physical or electronic signatures are acceptable.

731190 (Mar 2012)

DESIGN CONTROL

SUMMARY

CLIENT:

Dominion UNIT:

1 PAGE NO.:

1 PROJECT NAME:

Kewaunee Power Station S&L NUCLEAR OA PROGRAM PROJECT NO.:

11862-198 APPLICABLE El YES [l NO CALC. NO.:

2013-07050 SAFETY RELATED E: YES Z NO TITLE:

Maximum Cladding Temperature Analysis for an Uncovered Spent Fuel Pool with no Air Cooling EQUIPMENT NO.:

IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDNOIDED & REVIEW METHOD Initial Issue. The main body is pages 1 through 10. The final page is INPUTS/ASSUMPTIONS Attachment C, page C2.

M VERIFIED El UNVERIFIED REVIEW METHOD:

Detailed Review REV.:

0 STATUS:

Z APPROVED El SUPERSEDED BY CALCULATION NO.

Ql VOID DATE FOR REV.:

7-22-13 PREPARER:

Matthew M. Ross Signature on file DATE:

7-22-13 REVIEWER:

Joseph J. Pawasarat Signature on file DATE:

7-22-13 APPROVER:

Robert J. Peterson Signature on file DATE:

7-22-13 IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDED/VOIDED & REVIEW METHOD Revision 1 corrects a typo in Figure 7-1 and incorporates minor editorial changes. Pages 1 through 10.are revised and the changes are tracked with INPUTS/ASSUMPTIONS revision bars. None of the attachments are revised. The main body is pages 1 2 VERIFIED through 10. The final page is Attachment C, page C2.

El UNVERIFIED REVIEW METHOD:

Detailed Review REV.:

1 STATUS:

E APPROVED

[E SUPERSEDED BY CALCULATION NO.

El VOID DATE FOR REV.:

7-23-13 PREPARER:

Matthew M. Ross Signature on file DATE:

7-23-13 REVIEWER:

Joseph J. Pawasarat Signature on file DATE:

7-23-13 APPROVER:

Robert J. Peterson Signature on file DATE:

7-23-13 IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDIVOIDED & REVIEW METHOD Revision 2 calculates the date when it will take 2, 4, 6, or 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> for the spent fuel to heat up from 32 0C to 900 0. Changes to the main body are tracked with revision bars. Changes to Attachment B are tracked with revision bars.

INPUTS/ASSUMPTIONS There are no changes to Attachments A and C. The main body is pages 1 Z VERIFIED through 10. The final page of the calculation is Attachment C, page C2.

El UNVERIFIED REVIEW METHOD:

Detailed Review REV.:

2 STATUS:

[0 APPROVED El SUPERSEDED BY ALCULATION NO.

El VOID DATE FOR REV.:

7-,

ý-

PREPARER:

Matthew M. Ross DATE:

Z,9/tg/-

0 I REVIEWER:

Joseph J. Pawasarat DATE:,-Z-

_j APPROVER:

Robert J. Peterson DATE:

Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 2 of 10i S. Purpose and Scope.............................................................................................................................

3

2.

References...........................................................................................................................................

4

3.

Definitions...........................................................................................................................................

4

4.

Input Data...........................................................................................................................................

5

5.

Assum ptions........................................................................................................................................

6

6.

M ethodology.......................................................................................................................................

7 7.

R e su lts.................................................................................................................................................

9

8.

Conclusions and R ecom m endations...........................................................................................

10 Attachments:

No. of Pages:

Attachm ent A : Generation Rate vs. Decay Tim e (Reference 2.5)............................................................

6 Attachm ent B : Analysis................................................................................................................................

2 Attachm ent C: Current SFP Tem perature.................................................................................................

2 I--jfr~

&~vi LS d

Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 3 of 10

1. Purpose and Scope 1.1.

Purpose The purpose of this calculation is to conservatively evaluate the length of time (number of hours) it takes for uncovered spent fuel assemblies to reach the temperature where the zirconium cladding would fail. This analysis conservatively assumes that there is no air cooling of the assemblies: the flow paths that would provide natural circulation cooling are assumed to be blocked.

1.2. Scope The length of time for the fuel to heat up (the heat-up time) is determined as a function of the day that the analysis is performed (the decay time). The heat load from Westinghouse 422V+ fuel is used in this analysis (Reference 2.5 and Assumption 5.1).

The zirconium cladding must remain below the temperature where it will fail. Per NUREG/CR-6451 (Ref. 2.1, see Design Input 4.1), 565 'C (1049 'F) is the lowest temperature where incipient cladding failure might occur. NUREG-1738 (Ref. 2.7, pg.

3-7) states that runaway oxidation of zirconium occurs at 900 'C. For this analysis, the NUREG/CR-6451 temperature (565 'C, 1049 'F) and the NUREG-1738 temperature (900 'C, 1652 'F) are the temperatures of interest for the zirconium cladding.

There are no specific acceptance criteria for this analysis, however, SECY-99-168 (Ref.

2.4) suggests that "10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> (is) sufficient time to take mitigative action" and that for PWRs, 2.5 years is expected to be the decay time needed to reach a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat-up time from 30 'C to 900 'C. NUREG-1738 shows that a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat up time to 900 'C for a PWR would occur at less than 2 years (Ref. 2.7, Fig. 2-2).

-,mrar tnt C. L.In l'y

Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 4 of 10

2. References 2.1.

NUREG/CR-6451, "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants," August 1997.

2.2. Incropera, Frank P., and David P. DeWitt, Introduction to Heat Transfer, Fourth Edition, John Wiley & Sons.

2.3. Kewaunee USAR, Chapter 3: Reactor, Revision 24.02 - Updated Online 04/15/13.

2.4. SECY-99-168, "Improving Decommissioning Regulations for Nuclear Power Plants,"

June 30, 1999.

2.5. Document No. ETE-NAF-2013-0077, "Information for Kewaunee Spent Fuel Pool Postulated Loss of Inventory Calculation," Rev. 0, July 10, 2013.

2.6. Email from Michael Lico (Dominion) to Matthew Ross (S&L), "KPS sfp temp today,"

July 2 2nd, 2013. Included as Attachment C.

2.7. NUREG-1738, "Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power Plants," February 2001.

3. Definitions 3.1. Decay Time The decay time is the time since the reactor was shut down (May 7 th, 2013).

3.2. Heat-up Time The heat-up time is the amount of time between when the fuel becomes uncovered and when the zirconium cladding reaches the failure temperatures of interest, 565 'C (1049

'F) and 900 'C (1652 'F).

sAr,0orn rStucl

Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 5 of 10

4. Input Data 4.1.

Maximum Zirconium Temperature Several studies are presented in NUREG/CR-6451 (Ref. 2.1) discussing the maximum allowable temperature of zirconium cladding that will ensure that failure of the zirconium cladding will not occur. Per NUREG/CR-6451 (Ref. 2. 1, see Design Input 4.1), 565 -C (1049 OF) is the lowest temperature where incipient cladding failure might occur. NUREG-1738 uses 900 'C (1652 °F) as the temperature where "runaway oxidation" is expected to occur (Ref. 2.7, pg. 3-7). These two temperatures are the failure temperatures of interest for this calculation 4.2. Zirconium Properties The specific heat of zirconium at 600 K (620 OF) is 322 J/kg-K and the density of zirconium is 6570 kg/m3 (Ref. 2.2, pg. 822). A temperature of 620 °F is in the temperature range (less than the midpoint for both ranges) of this analysis. From Reference 2.2, the specific heat slightly increases with an increase in temperature. At higher temperatures, the zirconium would heat up more slowly. This temperature is representative of the full temperature range for this analysis.

4.3. Uranium Properties The specific heat of uranium at 600 K (620 °F) is 146 J/kg-K and the density of uranium is 19070 kg/m 3 (Ref. 2.2, pg. 822). A temperature of 620 OF is in the temperature range (less than the midpoint for both ranges) of this analysis. From Reference 2.2, the specific heat slightly increases with an increase in temperature. At higher temperatures, the uranium would heat up more slowly. This temperature is representative of the full temperature range for this analysis.

4.4. Geometry for Westinghouse 422V+ Assemblies The table below shows the geometry inputs for the fuel assemblies used in this analysis.

Table 4-1: Fuel Assembly Inputs (from USAR Table 3.2-8, Ref. 2.3)

Uranium Pellet Diameter 0.3659 inches Inner Diameter of Cladding 0.3734 inches Outer Diameter of Cladding 0.422 inches Rod Configuration and Total Rods 14 x 14, 196 total spaces Number of Guide Tubes, Instrument Tubes 16 guide, I instrument Total Number of Heated Rods 179 rods Inner Diameter of Guide Tubes (Above Dashpot) 0.492 inches Outer Diameter of Guide Tubes (Above Dashpot) 0.526 inches snrvg~nlt Cý L..ndjv,

Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 6 of 10 Table 4-1 Continued Heated Height of Rods 143.25 inches Cladding and Guide Tube Material ZIRLO Zirconium Theoretical Uranium Density Percentage 96.56%

4.5. Heat Load Reference 2.5 determines the maximum heat load from a single assembly. The assembly with the highest heat load will have the shortest heat-up time. The table showing the maximum fuel assembly heat generation rate for several years is located in Attachment A. The heat generation rates were calculated using the computer program HEATUP. Per Reference 2.5, the results in HEATUP are conservative compared to ORIGEN models.

5. Assumptions 5.1.

All of the fuel assemblies are assumed to be Westinghouse 422V+ fuel. This is appropriate because the most recent design consisted of a full core of 422V+ assemblies (Ref. 2.3, pg. 3.2-22). The most recently offloaded assemblies are limiting in terns of heat generation.

5.2. The properties of pure zirconium are used for the specific heat and deilsity of the zirconium alloy cladding. Based on an examination of alloys of some metals (e.g.

aluminum, nickel, or steel) in Table A. 1 of Reference 2.2, the density and specific heat are not significantly impacted by alloying.

5.3. Details of the thermal mass of the instrument tube are unavailable. For simplicity, the instrument tube is assumed to be identical to the guide tubes. This is appropriate because there are 16 guide tubes and one instrument tube, and the guide tubes are hollow while the instrument tube may have other thermal mass of the instruments.

5.4. The starting temperature for the heat-up analysis is assumed to be uniform and 90 'F (32 'C). A temperature of 90 'F is selected as representative of the current pool conditions (see Attachment C). The water temperature in the pool will continue to decrease over time due to a reduction in the heat load. It is appropriate to use a realistic value for the initial temperature due to the inherently conservative methodology (i.e. no heat transfer to the environment). In addition, this temperature is consistent with the sample analysis performed in SECY-99-168, where the starting temperature was 30 'C (86 'F).

5.5. The heat-up time is assumed to start when the spent fuel pool has been completely drained. This is conservative. It is likely that site personnel will start to respond to an incident when draindown starts.

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Document No. 2013-07050 Revision 2 Kewaunee Power Station Paqe7 ofl1 I

6. Methodology This analysis determines the heat-up time of the fuel assembly using the thermal capacity of materials (Based on Section 2.3 of.Ref. 2.2).

4=P x V x Cp x AT Equation 6-1 Where:

4 is the heat generation rate in BTU/hr p is the density of the material in lb/ft3 Vis the volume of the material in ft3 cp is the specific heat in BTU/lb-°F AT is the temperature increase in 'F t is the heat-up time in hr For this analysis, there are two materials being heated: the uranium fuel pellets and the ZIRLO zirconium alloy cladding. The zirconium is in the cladding and the instrument tubes, which are also being heated. The zirconium and the uranium are modeled as heating up at the same rate, so the AT/t will be the same for both materials.

q=AT x ýpx ýx cP'+

P:x Vx Cp)

I Where:

X,, signifies the property is for uranium XK signifies the property is for zirconium Equation 6-2 This calculation seeks the heat-up time, so Equation 6-2 is solved for t.

AT( x 1

, x

+ p xV Equation 6-3 The volume of uranium is given below.

r=

1x DD2jNhr, xL Where:

D, is the diameter of the uranium pellet Ni,. is the number of heated rods L is the heated length of the rods Equation 6-4 S*rrj" C6 L, Indy z

Document No. 2013-07050 Kewaunee Power Station Revision 2 Page 8 of 10 The volumes of zirconium in the heated rods and in the guide tubes are given below. The length of the cladding and guide tubes that are heated is conservatively modeled as being the same as the heated length of uranium. The guide tubes and cladding are longer than the length of the uranium pellets.

D.,°2 _D j "i V=

=7 x c

Nh,. x L Equation 6-5 D* o" - Dgj '"

V

[g

= K x Ng, x L Equation 6-6 4

Vý = Vg + V_."

Equation 6-7 Where:

J'%, is the volume of zirconium in the cladding of heated tubes VP.g is the volume of zirconium in the guide tubes D0,o is the outer diameter of the cladding Dcj is the inner diameter of the cladding Dgo is the outer diameter of the guide tubes Dgji is the inner diameter of the guide tubes Ng, is the number of guide tubes The temperature increase (AT) for this analysis is taken to be from the initial temperature of the pool, 90 'F (Assumption 5.4), to the zirconium cladding failure temperatures of interest, 1049 'F and 1652 'F (Input 4.1).

The heat-up time is calculated as a function of the decay time.

To avoid rounding, the Hottest Assembly column is recalculated in Attachment A based on the equations presented in Reference 2.5. Per Reference 2.5, the hottest assembly is calculated as:

Hottest Assembly (Heat Load from Cycle 32 Discharge Assemblies x 1.449 HtsA b 121 f_--g~n L&c L..ndy

Document No. 2013-07050 Revision 2 Kewaunee Power Station Paqe9of 10

7. Results The results are shown in Table 7-1 below (from Attachment B).

Table 7-1: Results Date End Temperature Decay Time Heat-Up Time (0C, 'F)

(months)

(hours)

October 4th, 2013 565, 1049

-5 2.0 April 8th, 2014 565, 1049

- 11 4.0 July 7tt, 2014 565, 1049 14 4.9 October 7t, 2014 565, 1049 17 6.0 August 21s t, 2015 565, 1049

-28 10.0 July 18th, 2013 900, 1652

-2 2.0 November 16t', 2013 900, 1.652

-6 4.0 March 11"h, 2014 900, 1652

-10 6.0 October 21 st, 2014 900, 1652

-17 10.0 The 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat-up time to a temperature of 565 'C (1049 'F) occurs at a decay time of under 2.5 years, which is the expected decay time to a temperature of 900 'C (1652 'F) stated in SECY-99-168 (Ref. 2.4). The 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> heat-up time to a temperature of 900 'C (I 652 'F) occurs at a decay time of roughly 1.5 years, which is less than the expected decay time calculated in NUREG-1738 (Ref. 2.7, pg. 2-3).

A plot showing the heat-up time to the temperatures of interest as a function of decay time is Figure 7-1.

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Document No. 2013-07050 Revision 2 Kewaunee Power Station Paae 10 of 10 Figure 7-1: Heat-Up Time vs. Decay Time 12 0) 10 16 8

E 0

D 0 U

C4, E

4 2

U.

0E CL R0 0'

1/1/2013 7/2/2013 1/1/2014 7/2/2014 1/1/2015 7/3/2015 1/1/2016 7/2/2016 Day (Shutdown was May 7, 2013)

8. Conclusions and Recommendations The Kewaunee results are more favorable than the analyses performed for SECY-99-168 (Ref 2.4) and NUREG-1738 (Ref 2.7). There are no acceptance criteria for this analysis.

There are no specific recommendations for this analysis.

The primary input to this analysis is the heat generation rate, which is conservative. The heat generation rates were calculated using the computer program HEATUP. Per Reference 2.5, the results in HEATUP are conservative compared to ORIGEN models.

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&. LArndy',

Calculation 2013-07050 Rev. 0 Pg lo Kewaunee Power Station Pg lo Attachment A: Heat Generation Rate vs. DcyTime (from Ref. 2.5 Heat Load from Hottest Fuel Recalculated Cycle 32 Discharge Assembly Hottest Assemblies Only Estimate Date Days since Assembly Date Time (MBTU/hr)

(MBTU/hr)

__(Reprinted)

May 8, 2013 (MBTU/hr) 5/8/2013 0:00 32.25 0.386 5/8/2013 0-0.3862 5/8/2013 18:00 28.23 0.338 1_ 5/8/2013 0.33 0.3381 5/8/2013 16:00 26.62 0.319 5/8/2013 0.67 0.3188 5/9/2013 0:00 25.39 0.304 5/9/2013 1

0.3041 5/9/2013 8:00 24.36 0.292 5/9/2013 1.33 0.2917 5/9/2013 16:00 23.45 0.281 5/9/2013 1.67 0.2808 5/10/2013 0:00 22.62 0.271 5/10/2013 2

0.2709 5/10/2013 8:00 21.86 0.262 1_ 5/10/2013 2.33 0.2618 5/10/2013 16:00 21.15 0.253 5/10/2013 2.67 0.2533 5/11/2013 0.00 20.5 0.246 5/11/2013 3

0.2455 5/11/2013 8:00 19.9 0.238 5/11/2013 3.33 0.2383 5/11/2013 16:00 19.33 0.232 5/11/2013 3.67 0.2315 5/12/2013 0:00 18.81 0.225

__5/12/2013 4

0.2253 5/13/2013 0:00 17.44 0.209 15/13/2013 5

0.2088 5/14/2013 0:00 16.3 0.195 5/14/2013 6

0.1952 5/15/2013 0:00 15.34 0.184 5/15/2013 7

0.1837 5/16/2013 0:00 14.52 0.174 5/16/2013 8

0.1739 5/17/2013 0:00 13.81 0.165 5/17/2013

___9 0.1654 5/18/2013 0:00 13.19 0.158 5/18/2013 10 0.1580 5/19/2013 10:00 12.65 0.151 15/19/2013 11 0.1515 5/20/2013 0:00 12.16 0.146 5/20/2013 12 0.1456 5/21/2013 0:00 11.73 0.140 5/21/2013 13 0.1405 5/22/2013 0:00 11.34 0.136 5/22/2013 14 0.1358 5/23/2013 0:00 10.99 0.132

-5/23/2013 15 0.1316 5/24/2013 0:00 10.67 0.128 5/24/2013 16 0.1278 5/25/2013 10:00 10.38 0.124 1 5/25/2013 17 0.1243 5/26/2013 0:00 10.11 0.121 5/26/2013 18 0.1211 5/27/2013 0:00 9.87 0.118 5/27/2013 19 0.1182 5/28/2013 0:00 9.64 0.115 5/28/2013 20 0.1154 5/29/2013 0:00 9.43 0.113 5/29/2013 21 0.1129 5/30/2013 0:00 9.24 0.111 5/30/2013 22 0.1107 5/31/2013 10:00 9.06 0.108 15/31/2013 23 0.1085 6/1/2013 0:00 8.88 0.106 6/1/2013 24 0.1063 6/3/2013 0:00 8.57 0.103 6/3/2013 26 0.1026 6/5/2013 0:00 8.29 0.099 6/5/2013 28 0.0993 6/7/2013 0:00 8.04 0.096 6/7/2013 30 0.0963 6/9/2013 0:00 7.8 0.093 6/9/2013 32 0.0934 6/11/2013 0:00 7.59 0.091 6/11/2013 34 0.0909 6/13/2013 10:00 7.38 0.088 16/13/2013 36 0.0884 6/15/2013 0:00 7.19 0.086 6/15/2013 38 0.0861 6/17/2013 0:00 7.01 0.084 6/17/2013 40 0.0839 6/19/2013 0:00 6.84 0.082 6/19/2013 42 0.0819 6/21/2013 0:00 6.68 0.080 6/21/2013 44 0.0800 6

./25/2013 0:00 6.38 0.076 6/25/2013 48 0.0764 6/29/2013 0:00 6.11 0.073 6/29/2013 52 0.0732 7/3/2013 0:00 5.86 0.070 7/3/2013 56 0.0702

Calculation 2013-07050 Rev. 0 Kewaunee Power Station Page A2 of 6 Attachment A: Heat Generation Rate vs. DcyTime (from Ref.IL5 Heat Load from Hottest Fuel Recalculated Cycle 32 Discharge Assembly Hottest Assemblies Only Estimate Date Days since Assembly Date Time (MBTU/hr)

(MBTU/hr)

__(Reprinted)

May 8, 2013 (MBTU/hr) 7/7/2013 0:00 5.64 0.068 17/7/2013 60 0.0675 7/11/2013 10:00 5.43 0.065 7/11/2013 64 0.0650 7/15/2013 0:00 5.24 0.063 7/15/2013 68 0.0628 7/19/2013 0:00 5.07 0.061 7/19/2013 72 0.0607 7/23/2013 0:00 4.91 0.059 7/23/2013 76 0.0588 7/27/2013 0:00 4.76 0.057 7/27/2013 80 0.0570 8/6/2013 0:00 4.42 0.053 18/6/2013 90 0.0529 8/16/2013 10:00 4.13 0.049 8/16/2013 100 0.0495 8/26/2013 0:00 3.88 0.046 8/26/2013 110 0.0465 9/5/2013 0:00 3.66 0.044

.9/5/2013 120 0.0438 9/15/2013 0:00 3.46 0.041 9/15/2013 130 0.0414 9/25/2013 0:00 3.27 0.039 9/25/2013 140 0.0392 10/5/2013 0:00 3.11 0.037 10/5/2013 150 0.0372 10/15/20131 0:00 2.96 0.035 110/15/2013 160 0.0354 10/25/2013 0:00 2.82 0.034 10/25/2013 170 0.0338 11/4/2013 0:00 2.69 0.032 11/4/2013 180 0.0322 11/24/2013 0:00 2.46 0.030 11/24/2013 200 0.0295 12/14/2013 0:00 2.27 0.027 12/14/2013 220 0.0272 1/3/2014 0:00 2.1 0.025 1/3/2014 240 0.0251 1/23/2014 10:00

____1.96 0.023 11/23/2014 260 0.0235 2/12/2014 0:00

____1.84 0.022 2/12/2014 280 0.0220 3/4/2014 0:00

____1.73 0.021 3/4/2014 300 0.0207 3/24/2014 0:00

____1.63 0.020 3/24/2014 320 0.0195 4/13/2014 0:00

____1.54 0.018 4/13/2014 340 0.0184 5/3/2014 0:00

____1.47 0.018 5/3/2014 360 0.0176 5/23/2014 0:00 1.4 0.017 15/23/2014 380 0.0168 6/12/2014 0:00 1.33 0.016 6/12/2014 400 0.0159 7/2/2014 0:00 1.28 0.015 7/2/2014 420 0.0153 7/22/2014 0:00

____1.22 0.015 7/22/2014 440 0.0146 8/11/2014 0:00

____1.17 0.014 8/11/2014 460 0.0140 8/31/2014 10:00 1.13 0.013 8/31/2014 480 0.0135 9/20/2014 0:00 1.08 0.013 19/20/2014 500 0.0129 10/10/2014 0:00 1.04 0.012 110/10/2014 520 0.0125 10/30/2014 0:00 1

0.012 10/30/2014 540 0.0120 11/19/2014 0:00 0.97 0.012 11/19/2014 560 0.0116 12/9/2014 0:00 0.93 0.011 12/9/2014 580 0.0111 12/29/2014 10:00 0.9 0.011 12/29/2014 600 0.0108 1/18/2015 0:00 0.87 0.010 1/18/2015 620 0.0104 2/7/2015 0:00 0.84 0.010 12/7/2015 640 0.0101 2/27/2015 0:00 0.81 0.010 2/27/2015 660 0.0097 3/19/2015 0:00 0.79 0.009 3/19/2015 680 0.0095 4/8/2015 0:00 0.76 0.009 4/8/2015 700 0.0091 4/28/2015 10:00 0.74 0.009 4/28/2015 720 0.0089 5/18/2015 0:00 0.72 0.009 5/18/2015 740 0.0086 6/7/2015 0:00 0.7 0.008 116/7/2015 760 0.0084 6/27/2015 0:00 0.68 0.008 116/27/2015 780 0.0081

Calculation 2013-07050 Rev. 0 Kewaunee Power Station Page A3 of 6 Attachment A: Heat Generation Rate vs. Decay Time (from Ref. 2.5)

Heat Load from Hottest Fuel Recalculated Cycle 32 Discharge Assembly Hottest Assemblies Only Estimate Date Days since Assembly Date Time (MBTU/hr)

(MBTU/hr)

(Reprinted) May 8, 2013 (MBTU/hr) 7/17/2015 0:00 0.66 0.008 1 7/17/2015 800 0.0079 8/6/2015 0:00 0.64 0.008 8/6/2015 820 0.0077 8/26/2015 0:00 0.62 0.007 8/26/2015 840 0.0074 9/15/2015 0:00 0.6 0.007 9/15/2015 860 0.0072 10/5/2015 0:00 0.59 0.007 10/5/2015 880 0.0071 10/25/2015 0:00 0.57 0.007 10/25/2015 900 0.0068 11/14/2015 0:00 0.56 0.007 111/14/2015 920 0.0067 12/4/2015 0:00 0.54 0.007 12/4/2015 940 0.0065 12/24/2015 0:00 0.53 0.006 12/24/2015 960 0.0063 1/13/2016 0:00 0.52 0.006 1/13/2016 980 0.0062 2/2/2016 0:00 0.5 0.006 2/2/2016 1000 0.0060 2/22/2016 0:00 0.49 0.006 2/22/2016 1020 0.0059 3/13/2016 0:00 0.48 0.006 3/13/2016 1040 0.0057 4/2/2016 0:00 0.47 0.006 1 4/2/2016 1060 0.0056 4/22/2016 0:00 0.46 0.005 4/22/2016 1080 0.0055

Calculation 2013-07050 Rev. 0 Kewaunee Power Station Page A4 of 6 Attachment A: Heat Generation Rate vs. Deca' Time (from Ref. 2.5 Heat Load from Cycle Hottest Fuel 32 Discharge Assembly Assemblies Only Estimate Date Days since Recalculated Hottest Date Time (MBTU/hr)

(MBTU/hr)

(Reprinted)

May 8, 2013 Assembly (MBTU/hr) 41402 0

32.25 0.386

=A3+B3

=0

=(C3/121)*1.449 41402 0.3333333, 28.23 0.338

=A4+64

=F4-F$3

=(C4/121)*1.449 41402 0.6666666E 26.62 0.319

=A5+B5

=F5-F$3

=(C5/121)*1.449 41403 0

25.39 0.304

=A6+B6

=F6-F$3

=(C6/121)*1.449 41403 0.3333333" 24.36 0.292

=A7+B7

=F7-F$3

=(C7/121)'1.449 41403 0.6666666E 23.45 0.281

=A8+B8

=F8-F$3

=(C8/121)*1.449 41404 0

22.62 0.271

=A9+B9

=F9-F$3

=(C9/121)*1.449 41404 0.33333333 21.86 0.262

=A10+B10

=F10-F$3

=(C1O/121)*1.449 41404 0.6666666E 21.15 0.253

=A11+B11

=F11-F$3

=(C11/121)*1.449 41405 0

20.5 0.246

=A12+B12

=F12-F$3

=(C12/121)*1.449 41405 0.33333332 19.9 0.238

=A13+B13

=F13-F$3

=(C13/121)*1.449 41405 0.6666666E 19.33 0.232

=A14+B14

=F14-F$3

=(C14/121)*1.449 41406 0

18.81 0.225

=A15+B15

=F15-F$3

=(C15/121)*1.449 41407 0

17.44 0.209

=A16+B16

=F16-F$3

=(C16/121)*1.449 41408 0

16.3 0.195

=A17+B17

=F17-F$3

=(C17/121)*1.449 41409 0

15.34 0.184

=A18+B18

=F18-F$3

=(C18/121)*1.449 41410 0

14.52 0.174

=A19+B19

=F19-F$3

=(C19/121)*1.449 41411 0

13.81 0.165

=A20+B20

=F20-F$3

=(C20/121)'1.449 41412 0

13.19 0.158 1 =A21+B21

=F21-F$3

=(C21/121)*1.449 41413 0

12.65 0.151

=A22+B22

=F22-F$3

=(C22/121)*1.449 41414 0

12.16 0.146

=A23+B23

=F23-F$3

=(C23/121)*1.449 41415 0

11.73 0.14

=A24+B24

=F24-F$3

=(C24/121)*1.449 41416 0

11.34 0.136

=A25+B25

=F25-F$3

=(C25/121)*1.449 41417 0

10.99 0.132

=A26+B26

=F26-F$3

=(C26/121)*1.449 41418 0

10.67 0.128

=A27+B27

=F27-F$3

=(C27/121)*1.449 41419 0

10.38 0.124

=A28+B28

=F28-F$3

=(C28/121)*1.449 41420 0

10.11 0.121

=A29+B29

=F29-F$3

=(C29/121)'1.449 41421 0

9.87 0.118

=A30+B30

=F30-F$3

=(C30/121)*1.449 41422 0

9.64 0.115

=A31+B31

=F31-F$3

=(C31/121)*1.449 41423 0

9.43 0.113

=A32+B32

=F32-F$3

=(C32/121)*1.449 41424 0

9.24 0.111

=A33+B33

=F33-F$3

=(C33/121)*1.449 41425 0

9.06 0.108

=A34+B34

=F34-F$3

=(C34/121)*1.449 41426 0

8.88 0.106

=A35+B35

=F35-F$3

=(C35/121)*1.449 41428 0

8.57 0.103

=A36+B36

=F36-F$3

=(C36/121)*1.449 41430 0

8.29 0.099

=A37+B37

=F37-F$3

=(C37/121)*1.449 41432 0

8.04 0.096

=A38+B38

=F38-F$3

=(C38/121)*1.449 41434 0

7.8 0.093

=A39+B39

=F39-F$3

=(C39/121)*1.449 41436 0

7.59 0.091

=A40+1B40

=F40-F$3

=(C40/121)*1.449 41438 0

7.38 0.088

=A41+B41

=F41-F$3

=(C41/121)*1.449 41440 0

7.19 0.086

=A42+B42

=F42-F$3

=(C42/121)*1.449 41442 0

7.01 0.084

=A43+B43

=F43-F$3

=(C43/121)*1.449 41444 0

6.84 0.082

=A44+B44

=F44-F$3

=(C44/121)*1.449 41446 0

6.68 0.08

=A45+B45

=F45-F$3

=(C45/121)*1.449 41450 0

6.38 0.076

=A46+B46

=F46-F$3

=(C46/121)*1.449 41454 0

6.11 0.073

=A47+B47

=F47-F$3

=(C47/121)*1.449 41458 0

5.86 0.07

=A48+B48

=F48-F$3

=(C48/121)*1.449 41462 0

5.64 0.068

=A49+B49

=F49-F$3

=(C49/121)*1.449 41466 0

5.43 0.065

=A50+B50

=F50-F$3

=(C50/121)*1.449

Calculation 2013-07050 Rev. 0 Kewaunee Power Station Page A5 of 6 Attachment A: Heat Generation Rate vs. Deca' Time (from Ref. 2.5)

Heat Load from Cycle Hottest Fuel 32 Discharge Assembly Assemblies Only Estimate Date Days since Recalculated Hottest Date Time (MBTU/hr)

(MBTU/hr)

(Reprinted)

May 8, 2013 Assembly (MBTU/hr) 41470 0

5.24 0.063

=A51+B51

=F51-F$3

=(C51/121)*1.449 41474 0

5.07 0.061

=A52+B52

=F52-F$3

=(C52/121)*1.449 41478 0

4.91 0.059

_=A53+B53

=F53-F$3

=(C53/121)'1.449 41482 0

4.76 0.057

=A54+B54

=F54-F$3

=(C54/121)*1.449 41492 0

4.42 0.053

=A55+B55

=F55-F$3

=(C55/121)*1.449 41502 0

4.13 0.049

=A56+B56

=F56-F$3

=(C56/121)*1.449 41512 0

3.88 0.046

=A57+B57

=F57-F$3

=(C57/121)*1.449 41522 0

3.66 0.044

=A58+B58

=F58-F$3

=(C58/121)*1.449 41532 0

3.46 0.041

=A59+B59

=F59-F$3

=(C59/121)*1.449 41542 0

3.27 0.039

=A60+B60

=F60-F$3

=(C60/121)*1.449 41552 0

3.11 0.037

=A61+B61

=F61-F$3

=(C61/121)*1.449 41562 0

2.96 0.035

=A62+B62

=F62-F$3

=(C62/121)*1.449 41572 0

2.82 0.034

=A63+B63

=F63-F$3

=(C63/121)*1.449 41582 0

2.69 0.032

=A64+B64

=F64-F$3

=(C64/121)*1.449 41602 0

2.46 0.03

=A65+B65

=F65-F$3

=(C65/121)*1.449 41622 0

2.27 0.027

=A66+B66

=F66-F$3

=(C66/121)'1.449 41642 0

2.1 0.025

=A67+B67

=F67-F$3

=(C67/121)*1.449 41662 0

1.96 0.023

=A68+B68

=F68-F$3

=(C68/121)*1.449 41682 0

1.84 0.022

=A69+B69

=F69-F$3

=(C69/121)*1.449 41702 0

1.73 0.021

=A70+B70

=F70-F$3

=(C70/121)*1.449 41722 0

1.63 0.02

=A71+B71

=F71-F$3

=(C71/121)*1.449 41742 0

1.54 0.018

=A72+B72

=F72-F$3

=(C72/121)*1.449 41762

.0 1.47 0.018

=A73+B73

=F73-F$3

=(C73/121)*1.449 41782 0

1.4 0.017

=A74+B74

=F74-F$3

=(C74/121)*1.449 41802 0

1.33 0.016

=A75+B75

=F75-F$3

=(C75/121)*1.449 41822 0

1.28 0.015

=A76+B76

=F76-F$3

=(C76/121)*1.449 41842 0

1.22 0.015

=A77+B77

=F77-F$3

=(C77/121)'1.449 41862 0

1.17 0.014

=A78+B78

=F78-F$3

=(C78/121)*1.449 41882 0

1.13 0.013

=A79+B79

=F79-F$3

=(C79/121)*1.449 41902 0

1.08 0.013

=A80+B80

=F80-F$3

=(C80/121)*1.449 41922 0

1.04 0.012

=A81+B81

=F81-F$3

=(C81/121)*1.449 41942 0

1 0.012

=A82+B82

=F82-F$3

=(C82/121)*1.449 41962 0

0.97 0.012

=A83+B83

=F83-F$3

=(C83/121)'1.449 41982 0

0.93 0.011

=A84+B84

=F84-F$3

=(C84/121)*1.449 42002 0

0.9 0.011

=A85+B85

=F85-F$3

=(C85/121)*1.449 42022 0

0.87 0.01

=A86+886

=F86-F$3

=(C86/121)*1.449 42042 0

0.84 0.01

=A87+B87

=F87-F$3

=(C87/121)*1.449 42062 0

0.81 0.01

=A88+B88

=F88-F$3

=(C88/121)*1.449 42082 0

0.79 0.009

=A89+B89

=F89-F$3

=(C89/121)*1.449 42102 0

0.76 0.009

=A90+B90

=F90-F$3

=(C90/121)*1.449 42122 0

0.74 0.009

=A91+B91

=F91-F$3

=(C91/121)*1.449 42142 0

0.72 0.009

=A92+B92

=F92-F$3

=(C92/121)*1.449 42162 0

0.7 0.008

=A93+B93

=F93-F$3

=(C93/121)*1.449 42182 0

0.68 0.008

=A94+B94

=F94-F$3

=(C94/121)*1.449 42202 0

0.66 0.008

=A95+B95

=F95-F$3

=(C95/121)*1.449 42222 0

0.64 0.008

=A96+B96

=F96-F$3

=(C96/121)'1.449 42242 0

0.62 0.007

=A97+B97

=F97-F$3

=(C97/121)*1.449 42262 0

0.6 0.007

=A98+B98

=F98-F$3

=(C98/1211)*1.449

Calculation 2013-07050 Rev. 0 Kewaunee Power Station Page A6 of 6 Attachment A: Heat Generation Rate vs. Decay Time (from Ref. 2.5 Heat Load from Cycle Hottest Fuel 32 Discharge Assembly Assemblies Only Estimate Date Days since Recalculated Hottest Date Time (MBTU/hr)

(MBTU/hr)

(Reprinted)

May 8, 2013 Assembly (MBTU/hr) 42282 0

0.59 0.007

=A99+B99

=F99-F$3

=(C99/121)*1.449 42302 0

0.57 0.007

=A100+B100

=F100-F$3

=(C100/121)*1.449 42322 0

0.56 0.007

=A101+B101

=F101-F$3

=(C101/121)*1.449 42342 0

0.54 0.007

=A102+B102

=F102-F$3

=(C102/121)*1.449 42362 0

0.53 0.006

=A103+B103

=F103-F$3

=(C103/121)'1.449 42382 0

0.52 0.006

=A104+B104

=F104-F$3

=(C104/121)'1.449 42402 0

0.5 0.006

=A105+B105

=F105-F$3

=(C105/121)*1.449 42422 0

0.49 0.006

=A106+B106

=F106-F$3

=(C106/121)*1.449 42442 0

0.48 0.006

=A107+B107

=F107-F$3

=(C107/121)*1.449 42462 0

0.47 0.006

=A108+B108

=F108-F$3

=(C108/121)'1.449 42482 0

0.46 0.005

=A109+B109

=F109-F$3

=(C109/121)*1.449

Calculation 2013-07050 Rev. 2 Kewaunee Power Station Page B1 of 2 Attachment B: Analysis Specific Heat of Uranium Specific Heat of Uranium Specific Heat of Zirconium Specific Heat of Zirconium Diameter of Fuel Uranium Inner Diameter of Zirconium Outer Diameter of Zirconium Heated Rods per Assem Unheated Rods (Guide or Instrument Tubes)

ID of Guide Tubes OD of Guide Tubes Density of Uranium Theoretical Density Density of Uranium Density of Zirconium Density of Zirconium Heated Length of Uranium Initial Temperature Final Temperature Total temperature Increase Volume of Uranium Volume of Zirconium in a Heated Rod Volume of Zirconium in a Guide Tube Total Volume of Zirconium 146 0.035 322 0.077 0.3659 0.3734 0.422 179 17 0.492 0.526 19,070 96.56%

1149.5 6570 410.2 11.9375 90 1049 959 1.560 0.451 0.038 0.489 J/kg-K BTU/Ib-F J/kg-K BTU/Ib-F inches inches inches Rods Tubes inches inches kg/m3 lb/ft3 kg/M 3

lb/ft3 feet F

F F

ft3 ft3 ift3 ft.3 Input 4.3 Conversion Input 4.2 Conversion Input 4.4 Input 4.4 Input 4.4 Input 4.4 Input 4.4 Input 4.4 Input 4.4 Input 4.3 Input 4.4 Conversion Input 4.2 Conversion Input 4.4 Assumption 5.4 Input 4.1 Initial Minus Final Equation 6-4 Equation 6-5 Equation 6-6 Equation 6-7 Assem Heat Generation at 14 Months 0.01515 MBTU/hr Interpolated from Att. A Time to Failure 4.94 hrs Equation 6-3 Assem Heat Generation at 17 Months 0.01253 MBTU/hr Interpolated from Att. A Time to Failure 5.97 hrs Equation 6-3 Heat Generation that Gives 2 Hour Heat-Up 0.03739 MBTU/hr Iterated Time to Failure 2.00 hrs Equation 6-3 Date of Associated Heat Generationr 10/4/2013-1 Interpolated from Att. A Heat Generation that Gives 4 Hour Heat-Up 0.01869 MBTU/hr Iterated Time to Failure 4.00 hrs Equation 6-3 Date of Associated Heat Generationr 4/8/2014 -

Interpolated from Att. A Heat Generation that Gives 10 Hour Heat-Up 0.00748 MBTU/hr Iterated Time to Failure 10.00 hrs Equation 6-3 Date of Associated Heat Generationr 8/21/2015-1 Interpolated from Att. A NUREG-1783 Maximum Temperature (900 C) 1652 F

Input 4.1 Temperature Increase 1562 F

Initial Minus Final Heat Generation that Gives 10 Hour Heat-Up 0.01218 MBTU/hr Iterated Time to Failure 10.00 hrs Equation 6-3 Date of Associated Heat Generation 10/21/2014 Interpolated from Att. A Heat Generation that Gives 6 Hour Heat-Up 0.02030 Time to Failure 6.00 h

Date of Associated Heat Generationr 3/11/2014-1 Heat Generation that Gives 4 Hour Heat-Up 0.03045 Time to Failure 4.00 h"

Date of Associated Heat Generationl 11/16/2013 Heat Generation that Gives 2 Hour Heat-Up 0.06089 Time to Failure 2.00 h

Date of Associated Heat Generation 7/18/2013 ABTU/hr Iterated irs Equation 6-3 Interpolated from Att. A MBTU/hr Iterated irs Equation 6-3 Interpolated from Att. A ABTU/hr Iterated irs Equation 6-3 Interpolated from Att. A

Calculation 2013-07050 Rev. 2 Kewaunee Power Station Page B2 of 2 A

BC J

D E

F 1 Attachment B: Analysis 2

3 Specific Heat of Uranium 146 J/kg-K input4.3 4

Specific Heat of Uranium =B3"0.0009478/2.20462/(9/5)

BTU/lb-F Conversion 5

Specific Heat of Zirconium 322 J/kg-K _jInput 4.2 6

Specific Heat of Zirconium =B5*0.000947812.20462/(915)

BTU/lb-F 'Conversion i

7 Diameter of Fuel Uranium 0.3659 inches I Input 4.4 1

8 Inner Diameter ot Zirconium 0.3734 inches lnnput 4.4 9

Outer Diameter of Zirconium 0.422 inches t Input 4.4 10 Heated Rods erAssem 179 Rods tnaput 4.4 11 Unheated Rods (Guide or Instrument Tubes) 17 Tubes Input 4.4 12 ID of Guide Tubes 0.492 inches Input 4.4 13 OD of Guide Tubes 0.526 inches Input 4.4 14 Density of Uranium 19070 kg/im tInput 4.3 15 Theoretical Density 0.9656 Input 4.4 16 Density of Uranium =814*2.20462/3.280841^3*B15 lb/ft'

!Conversion 17 Density of Zirconium 6570 kg/mf CoInput 4.2 18 Density of Zirconium =B17"2.20462/3.28084^3 lb/hf 3

!Conversion 19 Heated Length of Uranium =143.25/12 feet i tnput 4.4 20 Initial Temperature 90 F

]Assumption 5.4 21 Final Temperature 1049 F

!Input 4.1 22 Total temperature Increase =B21-B20 F

itnitial Minus Final 23 1

24 Volume of Uranium =PIt*B7^214*B19/144*B10 ft 3

Equation 6-4 25 Volume of Zirconium in a Heated Rod =Pl()*(B9^2-B8^2)/4*B19/144*BlO ft3 Equation 6-5 5 26 Volume of Zirconium in a Guide Tube =PlO*(B13^2-B12^2)/4"B19/144*B11 ft3

'Equation 6-6 27 Total Volume of Zirconium =B25+126 ft E

-7io 28 29 Assent Heat Generation at 14 Months ='Attachment A'!H76-(5/20)*('Attachment A'!H76-'Attachment A1H77)

MBTU/hr Ilnterpolated from Al. A 1 30 Time to Failure =$BS22/(B29°10 *6)*(SBS16"$B$24"$B$4+$B$18*SB$27*SBS6) hrs

'Equation 6-3 31 32 Assent Heat Generation at 17 Months ='Attachment A'!H80-(17/20)°(Attachment A'!H80-'Attachment A'1H81)

MBTU/hr IInterpolated from Att. A 33 Time to Failure =$B$22/(B32*10^6)*($BS16*$BS24"$B$4+$B$18"SB$27'$BS6) hrs iEquation 6-3 35 Heat Generation that Gives 2 Hour Heat-Up 0.0373850764604915 MBTU/hr Iterated 36 Time to Failure =SBS22/(B35*10^6)*($B$16*$BS24°$B$4+$B$18*SB$27"$BS6) hrs

'Equation 6-3 37 Date of Associated Heat Generation ='Attachment A'F60+('Attachment A'!H60-B35)/('Attachment A'!H60-'Attachment A'!H61)*10

'Interpolated from Att. A 38 1

39 Heat Generation that Gives 4 Hour Heat-Up 0.0186925354210505 MBTU/hr Iterated 40 Time to Failure =SBS22/(B39°10^6)*($BS16*$B$24*$B$4+$B$18*SBS27*$BS6) hrs Equation 6-3 41 Date of Associated Heat Generation ='Attachment A'IF71+i'Attachment A'tH71-B39)/('Attachment A!H7I-'Attachment A'!H72)*20 Interpolated from Att. A 1 42 _

43 Heat Generation that Gives 10 Hour Heat-Up 0.00747701366999636 MBTU/hr Iterated 44 Time to Failure =SB$221(B43"10^6)*($BS16"$B$24*$B$4+$B$18*$B$27°SBS6) hrs

'Equation 6-3 45 Date of Associated Heat Generation ='Attachment A'!F96+('Attachment Al1H96-B43)/('Attachment Al1H96-'Attachment A:IH97)*20 I Interpolated from Att. A 461i 47 NUREG-1783 Maximum Temperature (900 C) 11652 F

Input 4.1 48 7-B20 F

Ilnitial Minus Final 49 Heat Generatic MBTU/hr Iterated Date of Associated Heat C hirs Equation 6-3

.Interpolated from Att. A 53 Heat Generation that Gives 6 Hour Heat-Up 10.0202973514135572 MBTU/hr Iterated 54 Time to Failure =$6S8/(B5310v6)*(B$16$B$24S$B

$18SB$27S$B$6 551 Date of Associated Heat Generation ='Attachment A'!F70v('Attachment A'!H70-853/('Attachment A'l 561 hrs neuateron 6-3 iterpolated trom Alt. A 57 Heat Generation that Gives 4 Hour Heat-Up 0.0304460281462153 MBTU/hr Iterated 58 Time to Failure =$BS48/[B57*1 4+$B$18"SB$27*$BS6) hrs Equation 6-3 59 Date of Associated Heat Generation 60 61 Heat Generation that Gives 2 Hour Heat-Up 62 Time to Failure 63 Date of Associated Heat Generation rent All ttachment A'-H65)'20

! Interpolated from ACt. A

'U/hr Iterated IEquation 6-3 Ilnteroolated from Att. A

=$BS48/(B61"10 6)(*$BS16"$B$24*$B$4 $B$18*SB$27*$B$6)

='Attachment A'1F51 +('Attachment A!H51-B61 )/('Attachment A'!H51-'Attachment A'lH52)*4

4 L

Calculation 2013-07050 Kewaunee Power Station Page C1 of 2 FW: sfp temp today Michael S Lico (Generation - 6) 07/22/2013 11:28 AM To:

'MATTHEW.M.ROSS@Sargentlundy.com' Show Details

Max, The attached printout from the KPS Plant Parameters List shows the present SFP temperature to be 800 F.

Clearly, the temperature will only continue to drop as the SFP heat load decreases. Thanks.

Mike Lico From: John F Helfenberger (Generation - 4)

Sent: Monday, July 22, 2013 9:18 AM To: Michael S Lico (Generation - 6)

Subject:

sfp temp today FYI.

John F. Helfenberger, Lead Reactor Engineer, Kewaunee Power Station bus. 920.388.8294 Dominion Tie-line - 8.691.8294 pag. 920.704.4471 CONFIDENTIALITY NOTICE: This electronic message contains information which may be legally confidential and/or privileged and does not in any case represent a firm ENERGY COMMODITY bid or offer relating thereto which binds the sender without an additional express written confirmation to that effect. The information is intended solely for the individual or entity named above and access by anyone else is unauthorized. If you are not the intended recipient, any disclosure, copying, distribution, or use of the contents of this information is prohibited and may be unlawful. If you have received this electronic transmission in error, please reply immediately to the sender that you have received the message in error, and delete it. Thank you.

file://C:\\Documents and Settings\\0n7447\\Local Settings\\Temp\\notesFFF692\\-web5264.htm 7/22/2013

-=1~i~I I Eile Edit Trn-oer Reports 2etnp Ntridcm Help V

Control Room Log Description Reading Requir:ed In MKin 2 Day "Shit (06:00- 18:00)

.?; Nght shift (18:00 - 06:00) 1 'ant Mode or Conditorn Defeled Yes

_ ' Tech Spec Tracking 2 ýFP ttboIl (RD 11.2. i S) 53 his Yes Day Shift (06:00 - 18:00) 3 ýw (sFP Heat 5ink) Temp 58 F Yes i...,)'*

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Engineering Log S*'Enginering (06:00- 06:00)

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