ML18270A237
Text
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 1
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- 01 UFSAR Effective Date: 05/04/2016 Revise Sections 8.1.2 and 8.3.3 to remove excess detail.
This change removes excessive details regarding the voltage and frequency ranges of the Control Room Instrument Distribution (CRID) panels and inverters from Sections 8.1.2 and 8.3.3 of the UFSAR.
Approved:5/4/2016 Affected Unit: Both Units Justification:
Action Request (AR) 2015-14607-10 PMP-2350-SAR-001 - UFSAR Update Process This UFSAR change does not alter the Current Licensing Basis.
UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
This change removes excessive details regarding the voltage and frequency ranges of the CRID panels and inverters from Sections 8.1.2 and 8.3.3 of the UFSAR. The level of detail provided for these Structures, Systems, and Components (SSCs) is considerably beyond what is discussed for every other electrical system in the UFSAR.
This change meets PMP-2350-SAR-001 Attachment 1 Criterion #7, Removal of excessive detail.
Summary of Change:
50.71(e) Basis:
NDM Effective Date: 5/5/2016 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 2
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- 01 UFSAR Effective Date: 05/04/2016 List of Affected Items:
50.59: N/A None.
Comments:
08.01.02 Functional Criteria C
U12 Section 08.03.03 120 Volt AC Vital Instrument Bus System C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 3
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- 02 UFSAR Effective Date: 05/18/2016 Revise Sections 5.2.4 and 5.2.4.2 to correct UFSAR information to make it consistent with current licensing information.
Containment Penetrations for piping are only welded closed on the containment side of the penetration. The other side is allowed to grow thermally through a guided connection.
Welded penetrations are not applicable to all penetrations, just to piping penetrations.
Approved:5/18/2016 Affected Unit: Both Units Justification:
PMP-2350-SAR-001 - UFSAR Update Process SS-SE-2016-0148-00 Engineering Change (EC)-54911 - Remove Bellows on 1-CPN-85 for Piping from 1-WCR-928 This UFSAR change does not alter the Current Licensing Basis.
UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
These UFSAR changes fall under the guidance of PMP-2350-SAR-001, Attachment 1, list Item #8:
Correction of UFSAR information to make it consistent with current licensing basis.
Summary of Change:
50.71(e) Basis:
NDM Effective Date: 5/19/2016 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 4
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- 02 UFSAR Effective Date: 05/18/2016 List of Affected Items:
50.59: N/A Temporary Modification 1-TM-16-19 R0 and Engineering Change-54911 modified containment penetration 1-CPN-85. During this process, it was noted that the UFSAR descriptions of penetrations were inaccurate.
Comments:
05.02.04 Penetrations C
U12 Section 05.02.04.02 Piping Penetrations C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 5
Page UCR-2124 Rev. 0
- 03 UFSAR Effective Date: 06/09/2016 EC-54394 is the INDUS EC corresponding to RTO #1 (of four RTOs) associated with hardcopy EC-54079 (12-TR-9 Transformer 9 Installation - Integration Modification).
EC-54394 disconnects the 34.5kV Train B reserve feed from existing Breaker 12AB (12-52-12AB) and reconnects it to new 345kV-34.5kV Transformer #9 (12-TR-9) via new 345kV Breaker J2 (12-52-J2) and new 34.5kV Breaker BF (12-52-BF). This EC includes installation of breaker control, breaker position indication, and associated annunciator alarms and Plant Process Computer (PPC) inputs in both control rooms.
Precursor Event Required:
After Return To Operations (RTO) of EC-54394 TR-9 Reconnect the 34.5 kilovolt (kV)
Train B Line to the New Switching Structure and Place TR9 and CB-BF in Service (RTO#1)
Precursor Completed On: 6/9/2016 Approved:9/3/2015 Affected Unit: Both Units EC-54394 TR-9 Reconnect the 34.5kv Train B Line to the New Switching Structure and Place TR9 and CB-BF in Service (RTO#1)
Summary of Change:
Justification:
NDM Effective Date: 9/17/2015 List of Affected Items:
50.59: SS-SE-2015-0283-00 Comments:
None.
C U12 Figure 08.01-01A Main Auxiliary One-Line Diagram Bus 'A' & 'B' Engineered Safety System C
U12 Figure 08.01-01B Main Auxiliary One-Line Diagram Bus 'C' & 'D' Engineered Safety System C
U12 Figure 08.01-02A Simplified Offsite Power Sources C
U12 Figure 08.02-01 Switching Arrangements Donald C. Cook Nuclear Plant and Neighboring Stations 08.00 Electrical Systems C
U12 Section 08.01.02 Functional Criteria C
U12 Section 08.02 Network Interconnections C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 6
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- 04 UFSAR Effective Date: 06/09/2016 EC-54395 is the INDUS EC corresponding to RTO #2 (of four RTOs) associated with hardcopy EC-54079 - (12-TR-9 Transformer 9 Installation - Integration Modification).
EC-54395 connects Breaker 12AB to the new switching structure allowing Train B 34.5kV reserve feed to be powered from Transformer 9 or Transformer 5.
Precursor Event Required:
After RTO of EC-54395 12-TR-9 Connect CB-12AB to the New Switching Structure Allowing Train B to be Powered from TR9 or TR5 (RTO#2).
Precursor Completed On: 6/9/2016 Approved:9/3/2015 Affected Unit: Both Units Justification:
EC-54395 12-TR-9 Connect CB-12AB to the New Switching Structure Allowing Train B to be Powered from TR9 or TR5 (RTO#2).
Summary of Change:
NDM Effective Date: 9/17/2015 List of Affected Items:
50.59: SS-SE-2015-0283-00 Comments:
None.
C U12 Figure 08.01-01A Main Auxiliary One-Line Diagram Bus 'A' & 'B' Engineered Safety System C
U12 Figure 08.01-02A Simplified Offsite Power Sources 08.00 Electrical Systems C
U12 Section 08.01.02 Functional Criteria C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 7
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- 05 UFSAR Effective Date: 06/17/2016 Revise Section 9.2.2 - "System Design and Operation - Electrical Heat Tracing" to match the current configuration in the plant for the Boric Acid Heat Trace status.
The Electrical Heat Tracing section in 9.2.2 is being changed from have been abandoned in place or removed to have been functionally abandoned in place or removed.
Additionally, the sentence Power to the heat trace circuits is determinated and abandoned in place, is to be deleted.
Approved:6/17/2016 Affected Unit: Both Units AR 2015-13310-2 12-MOD-35230, Boric Acid Heat Trace Abandonment (BAHT), not being completed justifies the need to correct the UFSAR to match the current configuration of the plant.
PMP-2350-SAR-001 - "UFSAR Update Process" This UFSAR change does not alter the Current License Basis.
This UCR change request has a basis from modification 12-MOD-35230, Rev-0 for "Boric Acid Heat Trace Abandonment" and 50.59 Tracking No. 2003-1498-00.
The Boric Acid concentration was reduced from 12 % to 4 % under modifications 1-DCP-120 and 2-DCP-642. Because the concentration was reduced, the Boric Acid solution was not required to be maintained at an elevated temperature with the BAHT circuits to be abandoned. The abandonment of the BAHT system or circuits is covered under modification 12-MOD-35230 Rev-0.
This UFSAR change meets the requirements under procedure PMP-2350-SAR-001 Rev-12 Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 6/17/2016 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 8
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- 05 UFSAR Effective Date: 06/17/2016 for 50.71e Criteria under Item 1 - "Clarification change made to improve or correct the description of an SSC that does not change the meaning or intent of the SSC". This change is meant to clarify the changes to the Boric Acid Heat System which is no longer required.
List of Affected Items:
50.59: N/A This is correcting a change that was initiated from a previous Modification that never got implemented.
Comments:
09.02.02 System Design and Operation - Electrical Heat Tracing C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 9
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- 06 UFSAR Effective Date: 06/22/2016 EC-54396 is the INDUS EC corresponding to RTO #3 (of four RTOs) associated with hardcopy EC-54079 TR-9 Transformer 9 Installation (Integration Modification).
EC-54396 disconnects the 34.5kV Train A reserve feed from existing Breaker 12CD (12-52-12CD) and reconnects it to 345kV-34.5kV Transformer #9 (12-TR-9) via 345kV Breaker J2 (12-52-J2) and new 34.5kV Breaker BG (12-52-BG). This EC includes installation of breaker control, breaker position indication, and associated annunciator alarms and PPC inputs in both control rooms.
Precursor Event Required:
After RTO of EC-54396 12-TR-9 Reconnect the 34.5kV Train A Line to the New Switching Structure and Change Feed Source to Station Service (RTO#3)
Precursor Completed On: 6/22/2016 Approved:9/3/2015 Affected Unit: Both Units Justification:
EC-54396 12-TR-9 Reconnect the 34.5kV Train A Line to the New Switching Structure and Change Feed Source to Station Service (RTO#3)
Summary of Change:
NDM Effective Date: 9/17/2015 List of Affected Items:
50.59: SS-SE-2015-0283-00 Comments:
None.
C U12 Figure 08.01-01B Main Auxiliary One-Line Diagram Bus 'C' & 'D' Engineered Safety System C
U12 Figure 08.01-02A Simplified Offsite Power Sources 08.00 Electrical Systems C
U12 Section 08.01.02 Functional Criteria C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 10 Page UCR-2127 Rev. 0
- 07 UFSAR Effective Date: 06/23/2016 Asset Suite EC-54079 TR-9 Connect CB-12CD to the New Switching Structure Allowing Train A to be Powered from TR9 or TR4 (RTO#4) is the EC corresponding to RTO#4 (of four RTOs) which completes the installation of Transformer 9 associated with hardcopy EC-54079 12-TR-9 Transformer 9 Installation (Integration Modification).
Asset Suite EC-54079 connects Breaker 12CD to the new switching structure allowing Train A 34.5kV reserve feed to be powered from Transformer 9 or Transformer 4.
Precursor Event Required:
After RTO of INDUS EC-54079 12-TR-9 Connect CB-12CD to the New Switching Structure Allowing Train A to be Powered from TR9 or TR4 (RTO#4)
Precursor Completed On: 6/23/2016 Approved:9/3/2015 Affected Unit: Both Units Justification:
EC-54079 TR-9 Connect CB-12CD to the New Switching Structure Allowing Train A to be Powered from TR9 or TR4 (RTO#4)
Hardcopy EC-54079 TR-9 Transformer 9 Installation (Integration Modification)
Summary of Change:
NDM Effective Date: 9/17/2015 List of Affected Items:
50.59: SS-SE-2015-0283-00 Comments:
None.
C U12 Figure 08.01-01B Main Auxiliary One-Line Diagram Bus 'C' & 'D' Engineered Safety System C
U12 Figure 08.01-02A Simplified Offsite Power Sources 08.00 Electrical Systems C
U12 Section 08.01.02 Functional Criteria C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 11 Page UCR-2147 Rev. 0
- 08 UFSAR Effective Date: 06/29/2016 UFSAR Section 14.3.4 (U2) - Containment Integrity Analysis The Unit 2 LOCA M/E and containment analyses are being redone utilizing the new Westinghouse WCAP-17721 methodology. For UCR-2147, the Unit 2 analysis is complete (AEP-15-61) and being implemented via EC-54698. The current UFSAR analysis is documented in UFSAR Unit 2 Section 14.3.4. Unit 2 Section 14.3.4 was originally created by UCR-2138 which corresponded to the Unit 1 Containment Analysis implemented via EC-54591.
Unit 2 Section 14.3.4 will be updated to be consistent with the new WCAP-17721 analysis.
The text updates for Unit 2 due to the new methodology are extensive and affect a large portion of Section 14.3.4.
Precursor Event Required:
After RTO of EC-54698 "Implementation of Unit 2 Loss of Coolant Accident (LOCA)-
Containment Integrity Analysis using WCOBRA / TRAC Mass and Energy Releases (WCAP-17721-P-A)"
Precursor Completed On: 6/29/2016 Approved:6/3/2016 Affected Unit: Unit Two EC-54698 Summary of Change:
Justification:
NDM Effective Date: 6/8/2016 (Sorted by the "UFSAR Effective Date")
Comments:
None.
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 12 Page UCR-2147 Rev. 0
- 08 UFSAR Effective Date: 06/29/2016 List of Affected Items:
50.59: SS-SE-2016-0139-00 14.03.04 (U2)
Containment Integrity Analysis C
U2 Section 14.03.04.01.01 (U2)
Design Basis C
U2 Section 14.03.04.01.02 (U2)
Design Features C
U2 Section 14.03.04.01.03.01.03 (U2)
Peak Containment Pressure Transient C
U2 Section 14.03.04.01.03.01.05.01 (U2)
Effect of Increased Essential Service Water (ESW) Temperature and Reduced ESW Flow D
U2 Section 14.03.04.01.03.01.06 (U2)
Conclusions C
U2 Section 14.03.04.03 (U2)
Mass and Energy Release Analysis for Postulated Loss-Of-Coolant Accidents C
U2 Section 14.03.04.03.01.02.01 (U2)
Application of Single Failure Analysis C
U2 Section 14.03.04.03.01.02.02 (U2)
Mass and Energy Release Data N
U2 Section 14.03.04.03.01.02.02 (U2)
Blowdown Mass and Energy Release Data D
U2 Section 14.03.04.03.01.02.03 (U2)
Reflood Mass and Energy Release Data D
U2 Section 14.03.04.03.01.02.04 (U2)
Post-Reflood Mass and Energy Release Data D
U2 Section 14.03.04.03.01.02.05 (U2)
Sources of Mass and Energy C
U2 Section 14.03.04.03.01.02.06 (U2)
Significant Modeling Assumptions C
U2 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 13 Page UCR-2133 Rev. 0
- 09 UFSAR Effective Date: 08/18/2016 Summary of Change:
Revise Sections 11.4.1 and 11.4.3 to provide clarity and correctness to existing text to reflect current Radiation Protection (RP) practices.
See 50.71(e) Basis below for a more detailed discussion of the changes.
Approved:8/18/2016 Affected Unit: Both Units AR 2015-7184-1, Request RP to review Chapter 11.0 thru 11.4 of the current UFSAR PMP-2350-SAR-001 - UFSAR Update Process UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
The changes to Chapter 11 of the Donald C. Cook Nuclear Plant (CNP) UFSAR listed below do not alter the Current Licensing Basis.
Section 11.4.1a is being revised to delete the sentence A locker room where personnel change into protective clothing is also available. This is extraneous information not required in the UFSAR and meets the requirements for a 50.71(e) based revision under revision Criteria 7, removal of excessive detail.
Section 11.4.1b is being revised to clarify the disposition process for protective clothing.
Protective clothing may either be sent off-site for laundering or disposal processing. This meets the requirements for a 50.71(e) based revision under revision Criteria 8, Correction of UFSAR information to make it consistent with current licensing basis. The sentence:
Protective clothing meeting the release limits specified is received from the vendor and placed in the change-out facilities for personnel use. is being deleted to remove excessive, unnecessary detail as is allowed per criteria 7.
Section 11.4.1f is being revised to delete the list of calibration sources used by RP. This is extraneous information not required in the UFSAR and meets the Justification:
50.71(e) Basis:
NDM Effective Date: 8/19/2016 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 14 Page UCR-2133 Rev. 0
- 09 UFSAR Effective Date: 08/18/2016 requirements for a 50.71e based revision under revision criteria 7, removal of excessive detail.
Section 11.4.3 is being revised to replace designation: with designation(s): in the section describing radiological CAUTION, DANGER, OR GRAVE DANGER postings.
Additionally, at the list of the designations, the and is to be replaced by and/or.
Multiple of the below designations can be used for each posting. This is an editorial change and requested under 50.71(e) based change Criteria 4, Editorial changes consisting of the following: grammatical changes made to improve readability in accordance with standard rules of grammar.
List of Affected Items:
50.59: N/A Comments:
None.
11.04.01 Facilities C
U12 Section 11.04.03 Access Control C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 15 Page UCR-2167 Rev. 0
- 10 UFSAR Effective Date: 09/07/2016 Summary of Change:
Revise Section 9.3.1 per AR 2015-15099, Evaluate Auxiliary Feedwater (AFW) Mission Time Clarified Section 9.3.1, Design Bases, for the Residual Heat Removal (RHR) system to state that the single train cooldown evaluation is performed to confirm the capability to perform a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown and is not part of the Chapter 14 accident analysis.
Approved:9/7/2016 Affected Unit: Both Units AR 2015-15099, Evaluate AFW Mission Time General Tracking (GT) 2015-11788, Nuclear Regulatory Affairs (NRA) Compliance Tracker for Nuclear Regulatory Commission (NRC) Resident Questions AR 2016-4795, AFW mission time Green Non-cited violation Nuclear Regulatory Commission CNP Confirmatory Action Letter (CAL) closure letter dated February 2, 2000 for the basis of the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> single train RHR cooldown calculation.
This UFSAR change does not alter the Current Licensing Basis.
UFSAR change request (UCR) 2167 makes a modification to UFSAR Section 9.3.1, Design Bases, for the RHR system to clarify that the single train Reactor Coolant System (RCS) cooldown described therein is not part of the UFSAR Chapter 14 accident analyses. This alteration originates from the investigation of AFW mission time performed in response to NRC questions on the subject. The existing wording in the UFSAR also states that the licensing basis cooldown requirements are contained in Technical Specification (TS) 3.0.
Since TS 3.0.3 contains the only required cooldown limits in the TS 3.0 section, the verbiage in the UFSAR is also clarified to specifically refer to TS 3.0.3.
The alterations to the words in UFSAR Section 9.3.1 are clarifications based on already existing information and are not a change to the current license basis. The statement added to Section 9.3.1 says the following:
The single train cooldown evaluation is performed to demonstrate the plant design capability to achieve Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, as may be required to comply with various plant TS Actions. The single train cooldown evaluation is not part of the Justification:
50.71(e) Basis:
NDM Effective Date: 9/8/2016 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 16 Page UCR-2167 Rev. 0
- 10 UFSAR Effective Date: 09/07/2016 Chapter 14 accident analyses.
The descriptions in UFSAR Section 9.3.1 describe the design capability of the RHR system to allow a single train (RHR and Component Cooling Water (CCW)) 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown to 200°F with essential service water temperature equal to 88.9°F. This design capability calculation is based on the CNP CAL item #3 closure resolution contained in Reference 1.
Reference 1 states the following:
The statement does not imply [Referring to a statement in the CNP 1973 licensing SER.]
that one train of the CCW system alone is always sufficient to satisfy the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cool down time limit of TS 3.0.3. In the event TS 3.0.3 requires a cool down of the unit to 200°F and only one CCW train is operable, operators may use other non-safety systems to reach MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Therefore, the NRC staff finds that there is no licensing or design basis requirement to be able to cool down a unit at CNP within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> as specified in TS 3.0.3 utilizing only one train of CCW.
Reference 1 goes on to state:
However; the capability to meet all TS action requirement time limits, including those contained in TS 3.0.3, continues to be a requirement of the facility operating license.
CNP performs a calculation to demonstrate the design capability to perform a single train 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> cooldown assuming elevated Essential Service Water (ESW) temperature. As stated previously, this cooldown is not part of the CNP design and license basis and is therefore not part of the UFSAR Chapter 14 accident analyses. The proposed alteration to UFSAR Section 9.3.1 makes that clear.
Per Attachment 1 of PMP-2350-SAR-001, UFSAR Update Process, these two changes to UFSAR Section 9.3.1 are Clarification changes made to improve or correct the description of an SSC that do not change the meaning or intent of the SSC.
Reference:
- 1. Nuclear Regulatory Commission CNP CAL closure Letter, Dated February 2, 2000 for the basis of the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> single train RHR cooldown calculation.
(Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 17 Page UCR-2167 Rev. 0
- 10 UFSAR Effective Date: 09/07/2016 List of Affected Items:
50.59: N/A Comments:
None.
09.03.01 Design Bases C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 18 Page UCR-2169 Rev. 0
- 11 UFSAR Effective Date: 09/26/2016 Revise Section 3.2.2.2 (U2) regarding material make up of Unit 2 Reactor Vessel baffle-former bolts.
This UFSAR change is being made to correct some information pertaining to the material make-up of Unit 2 Reactor Vessel baffle-former bolts.
Add the following two sentences to the third paragraph of Unit 2 Section 3.2.2.2 (U2),
Descriptions and Drawings as new third and fourth sentences:
Originally installed baffle-former bolts are Type 347 Stainless Steel. Replacement baffle-former bolts are Type 316 Stainless Steel.
Approved:9/26/2016 Affected Unit: Unit Two AR 2016-9067 Potential minor change to information in UFSAR AS 2016-9067-1 Determine needed changes to UFSAR PMP-2350-SAR-001 - UFSAR Update Process This UFSAR change does not alter the Current Licensing Basis.
AR 2016-9067 was generated to revise some information in the Unit 2 UFSAR pertaining to the material of the reactor vessel baffle-former bolts.
The text in the UFSAR Section 3.2.2.2 is correct as written, however, some additional detail is being added to improve and enhance the descriptions of the materials found in the reactor vessel internals. As described in EC-50972 and EC-52196, the originally installed Baffle-Former Bolts are made from Type 347 annealed stainless steel, and the replacement bolts (installed during the U2C19 refueling outage) are type 316. Approximately 52 bolts out of 832 bolts were replaced during the U2C19 refueling outage.
This UFSAR change meets Criteria 1 and 2 of Attachment 1, 50.71(e) Criteria of PMP-2350-SAR-001, UFSAR Update Process, because it consists of clarification changes made to Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 10/4/2016 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 19 Page UCR-2169 Rev. 0
- 11 UFSAR Effective Date: 09/26/2016 improve or correct the description of the baffle-former bolts in the Unit 2 Reactor Vessel. The meaning or intent of the SSC is not impacted. This change also meets Item 2 of Attachment 1, because the replacement baffle bolts (Type 316) were installed via EC-50972, which contained a 50.59 review (2010-0297-02).
List of Affected Items:
50.59: N/A Comments:
None.
03.02.02.02 (U2)
Description and Drawings C
U2 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 20 Page UCR-2164 Rev. 0
- 12 UFSAR Effective Date: 10/07/2016 Add the following text to the last paragraph of Section 6.2.6 Programmatic Controls:
"and in accordance with "Guidelines for Effective Prevention and Management of System Gas Accumulation" NEI 09-10 [Rev 1a-A] (reference 6.2.7.4) and (reference 6.2.7.5)"
Add the following References to Section 6.2.7:
- 4. NEI 09-10 [Rev. 1a-A], "Guidelines for Effective Prevention and Management of System Gas Accumulation," dated April, 2013
- 5. AEP-15-46, "American Electric Power Donald C. Cook Units l and 2 Emergency Core Cooling System, Residual Heat Removal System and Containment Spray System Gas Accumulation Evaluation for D. C. Cook Units 1 and 2 Approved:10/7/2016 Affected Unit: Both Units EHI-5202 Rev.9 "Gas Accumulation Condition Monitoring Program" Summary of Change:
Justification:
NDM Effective Date: 10/19/2016 List of Affected Items:
50.59: SS-SE-2016-0224-00 Procedure EHI-5202,"Gas Accumulation Condition Monitoring Program identifies the administrative requirements for the Gas Accumulation Condition Monitoring Program (GACM) in Emergency Core Cooling System (ECCS), Decay Heat Removal, and Containment Spray System (CTS) for CNP. The purpose of this program is to ensure that these fluid systems, which could be susceptible to gas accumulation, are maintained in a state within their design basis and ready to perform their intended function. Revion 9 incorporates guielines of NEI 09-10 as implemented and is documented for CNP in AEP-15-46 Evaluation.
Comments:
06.02.06 Programmatic Controls C
U12 Section 06.02.07 References for Section 6.2 C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 21 Page UCR-2162 Rev. 0
- 13 UFSAR Effective Date: 10/17/2016 The NRC stated in a Non-cited Violation ML15042A380, dated 2/10/2015) that the removal of the Concrete Shadow Shield for Equipment Access Hatch (CSSEAH) in front of the Unit 1 and 2 equipment hatches in 2010 by EC-49191 resulted in a violation of 10 CFR 50 Appendix B, Criterion III, Design Control. The NRC position is that the CSSEAH removal on the 650' Elevation level of the auxiliary building will allow a significant increase in dose rates in the area in the event of a large break loss of coolant accident (LBLOCA); therefore, personnel in the area at the time of the accident would receive an occupational dose not in compliance with certain requirements of 10 CFR 20.
An engineering evaluation (documented in AR assignment 2014-13016-5) reviewed 10 CFR 20, 10 CFR 100, 10 CFR 50.34, NUREG/CR-6204, Reg Guides 8.8, 8.34, 1.187, 8.19, Standard Review Plan 12.2, and CNP licensing basis for plant radiation shielding contained in the original FSAR, original NRC Safety Evaluation Report (dated 9/10/1973) and the current UFSAR.
The engineering evaluation concluded:
"Use of an assumed TID-I4844 source term [or AST Source Term] for design basis accidents is specifically required by GDC-I9, NUREG-0737 Item II.B.2, and 10 CFR 100 to demonstrate that 1) vital plant equipment is capable of performing its design function(s) post-accident, 2) plant personnel inside and outside of the control room, who are required to mitigate the consequences of the accident, do not receive an accident dose greater than 5 Rem Total Effective Dose Equivalent (TEDE), and 3) members of the public do not receive accident doses in excess of the limits of 10 CFR 100. These regulatory requirements and the plant's shielding design that supports the acceptance criteria contained in them ensure that the acceptable design basis accident mitigation is possible and the public will be protected from unacceptable radiation exposure.
Since the auxiliary missile blocks were designed as Accident Shields to limit offsite dose following a design basis accident to within the limits of 10 CFR 100, a TID-14844 source term would be assumed to demonstrate continued compliance with GDC-I9, NUREG-0737 Item II.B.2 commitments, and 10 CFR 100."
The above summary of the licensing basis is also consistently reflected in UFSAR Chapter 11.2.1:
"Radiation shielding is designed for operation at maximum calculated thermal power and Approved:10/17/2016 Affected Unit: Both Units Summary of Change:
NDM Effective Date: 10/18/2016 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 22 Page UCR-2162 Rev. 0
- 13 UFSAR Effective Date: 10/17/2016 limits the normal operation radiation levels at the site boundary to below those levels allowed for continuous unrestricted exposure as set forth in 10 CFR 20. The plant is capable of continued safe operation with 1% fuel element defects.
In addition, the shielding provided ensures that in the unlikely event of a maximum design accident, the contained activity does not result in off-site radiation doses in excess of those given in 10 CFR 100."
The above excerpt makes the clear distinction between radiation shielding requirements for normal operation (assuming 1%) failed fuel compliant to 10 CFR 20 limits, and postulated maximum design accident scenario compliant to 10 CFR 100 limits for off-site radiation dose. However, the UFSAR subsection 11.2.1.1.4 Accident Shield, only states the following:
"The main purpose of the accident shield is to ensure safe radiation levels outside the containment building following a maximum design accident."
Taken out of context this subsection is vague and additional clarification needs to be added to improve the description of the accident shield and make it consistent with D. C. Cook licensing basis.
The first change will be to revise UFSAR Section 11.2.1 to clarify that all radiation shielding, including but not limited to the accident shields, are designed to ensure compliance with 10 CFR 20.
New Section 11.2.1 wording below:
"All Radiation shielding is designed for operation at maximum calculated thermal power and limits the normal operation radiation levels at the site boundary to below those levels allowed for continuous unrestricted exposure as set forth in 10 CFR 20. The plant is capable of continued safe operation with 1% fuel element defects."
The second change will be to Subsection 11.2.1.1.4 which will be revised as follows to provide clarification:
"The main purpose of the accident shield is to achieve the following after a maximum design accident: (a) prevent off-site radiation exposures in excess of 10 CFR 100; (b) to limit exposure to control room operators."
This change in wording is acceptable because it does not change the purpose of the accident (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 of Summary 106 23 Page UCR-2162 Rev. 0
- 13 UFSAR Effective Date: 10/17/2016 shield but instead provides a more specific description of its purpose. The more specific description is being pulled from CNP current licensing basis, Chapter 11.2.1 (which 11.2.1.1.4 is a subsection of) and regulatory documents, GDC-19, NUREG-0737 Item II.B.2, and 10 CFR 100.
Finally, there has been some confusion regarding the wording in Section 11.2.1.2.3 pertaining to the difference between the two personnel airlock shields. The CSSEAH that was removed was located on the 650' elevation, but there is a second personnel airlock on the 609' elevation that still has 3ft of concrete shielding. As part of the original modification that removed the CSSEAH, UCR-1936 was performed which made the following UFSAR change to Section 11.2.1.2.3:
"... Supplemental shielding has been provided for containment penetrations. This includes a 3 ft. - 6 in. concrete shadow shield for the equipment access hatch and a 3ft. concrete shield for the personnel lock."
Became:
"... Supplemental shielding has been provided for containment penetrations, which includes a 3ft. concrete shield for the personnel lock."
In reference to the CSSEAH being removed, this UCR made sense; however, as a standalone description it can be ambiguous because both the CSSEAH and the 609' elevation concrete shields provided shielding for a personnel airlock.
In order to clarify, the following change will be made to 11.2.1.2.3:
"Supplemental shielding has been provided for containment penetrations, which includes a 3 ft. concrete shield for the personnel lock."
Will become:
"Supplemental shielding has been provided for containment penetrations, which includes a 3 ft. concrete shield for the 609' elevation personnel lock."
Again, this is an administrative change. The proposed edit does not change which concrete shield is being described, but instead clarifies that the concrete shield being described is in fact the 609' elevation concrete shield and not the 650' elevation concrete shield.
(Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 24 of Summary 106 UCR-2162 Rev. 0
- 13 UFSAR Effective Date: 10/17/2016 This is an administrative change. The proposed edit does not change which concrete shield is being described, but instead clarifies that the concrete shield being described is in fact the 609' elevation concrete shield and not the 650' elevation concrete shield.
Per PMP-2350-SAR-001, Attachment 1, Criterion l, this change is a clarification change made to improve or correct the description of an SSC that do not change the meaning or intent of the SSC.
The UFSAR changes do not alter the current licensing basis.
List of Affected Items:
50.59: N/A Comments:
Final UFSAR search confirmed on 10/18/2016.
11.02.01 Design Basis C
U12 Section 11.02.01.01.04 Accident Shielding C
U12 Section 11.02.01.02.03 Accident Shield C
U12 Section (Sorted by the "UFSAR Effective Date")
50.71(e) Basis:
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 25 of Summary 106 UCR-2076 Rev. 0
- 14 UFSAR Effective Date: 11/05/2016 In Figure 6.2-1 place a cloud around the valve that represents SI-127 and add Note #7 explaining that EC-53594 removed 2-SI-127 and that 1-SI-127 remains in place. EC-53594 is going to remove 2-SI-127 permanently as it is no longer needed. The valve is leaking-by and represents unnecessary maintenance and housekeeping.
Precursor Event Required:
After Maintenance Release Only (MRO) of EC-53594 "Permanently Remove 2-SI-127 from Service and Replace with Pipe Cap" Precursor Completed On: 11/5/2016 Approved:6/9/2014 Affected Unit: Unit Two Justification:
AR#20008547 and EC-53594 "Permanently Remove the Boron Injection Tank 2-TK-11 Bypass Shutoff Valve 2-SI-127."
Summary of Change:
NDM Effective Date: 6/11/2014 List of Affected Items:
50.59: SS-SE-2014-0170-00 Comments:
C U12 Figure 06.02-01 Flow Diagram Emergency Core Cooling (SIS) Unit No. 1 or 2 (Sorted by the "UFSAR Effective Date")
None.
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 26 of Summary 106 UCR-2173 Rev. 0
- 15 UFSAR Effective Date: 11/11/2016 Update Unit 2 Section 14.3.1.6.1 sixth bullet for Steam Generator Tube Plugging (SGTP) level from 1.0% to 1.5%.
This change is made as a result of the Westinghouse evaluation performed in AEP-16-33, AEP CNP Unit 2, CNP Unit 2 Large-Break LOCA Steam Generator Tube Plugging Level Evaluation.
Approved:11/11/2016 Affected Unit: Unit Two Justification:
AEP-16-33, AEP CNP Unit 2, CNP Unit 2 Large-Break LOCA Steam Generator Tube Plugging Level Evaluation Summary of Change:
NDM Effective Date: 11/16/2016 List of Affected Items:
50.59: SS-SE-2012-0078-02 Comments:
None.
14.03.01.06.01 (U2)
Thermal Conductivity Degradation Error Resolution C
U2 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 27 of Summary 106 UCR-2175 Rev. 0
- 16 UFSAR Effective Date: 11/11/2016 Summary of Change:
Revise Section 14.3.4.1.3.1.3 (U1), Item #2, to insert text to complete a sentence.
UCR-2138, Rev. 0 "UFSAR Section 14.3.4 (Both Units) - Containment Integrity Analysis" approved a number of changes, including a revision of Section 14.3.4.1.3.1.3 (U1), Item #2 that required a deletion of some text and an insertion of some text to replace the deleted text.
The text to be deleted was deleted; however, the text to be inserted was not, leaving an incomplete sentence. This UCR inserts the missing text that was supposed to be inserted by UCR-2138.
Approved:11/11/2016 Affected Unit: Unit One UCR-2138, Rev. 0, "UFSAR Section 14.3.4 (Both Units) - Containment Integrity Analysis" EC-54591 - Implementation of Unit 1 LOCA-Containment Integrity Analysis using WCOBRA/TRAC Mass and Energy Releases (WCAP-17721-P-A)
Justification:
NDM Effective Date: 11/15/2016 List of Affected Items:
50.59: SS-SE-2015-0322-00 Comments:
None.
14.03.04.01.03.01.03 (U1)
Peak Containment Pressure Transient C
U1 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 28 of Summary 106 UCR-2058 Rev. 0
- 17 UFSAR Effective Date: 12/15/2016 Update Table 6.3-2 per EC-0000051369 HE-18E, Unit 2 East Containment Spray Heat Exchanger Replacement EC-51369, Rev. 0, replaced the Unit 2 East Containment Sprat Heat Exchanger (CTS) with a new heat exchanger. The new heat exchanger provides more cooling capacity than the existing heat exchanger. The new heat exchanger also contains additional manways for inspection, testing, and preventative maintenance. Revise Table 6.3-2 to reflect the installation of the new heat exchanger in Unit 2.
Precursor Event Required:
After RTO of EC-51369 HE-18E, Unit 2 East Containment Spray Heat Exchanger Replacement Precursor Completed On: 12/15/2016 Approved:10/24/2016 Affected Unit: Unit Two Justification:
EC-51369 HE-18E, Unit 2 East Containment Spray Heat Exchanger Replacement Summary of Change:
NDM Effective Date: 10/25/2016 List of Affected Items:
50.59: SS-SE-2013-0468-00 Comments:
None.
C U12 Table 06.03-02 Containment Spray Heat Exchanger Design Parameters (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 29 of Summary 106 UCR-2059 Rev. 0
- 18 UFSAR Effective Date: 12/15/2016 Update Table 6.3-2 per EC-51370 HE-18W, Unit 2 West Containment Spray Heat Exchanger Replacement EC-51370, Rev. 0, replaced the Unit 2 West CTS Heat Exchanger with a new heat exchanger. The new heat exchanger provides more cooling capacity than the existing heat exchanger. The new heat exchanger also contains additional manways for inspection, testing, and preventative maintenance. Revise Table 6.3-2 to reflect the installation of the new heat exchanger in Unit 2.
Precursor Event Required:
After RTO of EC-51370 HE-18W, Unit 2 West Containment Spray Heat Exchanger Replacement Precursor Completed On: 12/15/2016 Approved:10/24/2016 Affected Unit: Unit Two Justification:
EC-51370 - "2-HE-18W - U2 West CTS Hx Replacement" Summary of Change:
NDM Effective Date: 10/25/2016 List of Affected Items:
50.59: SS-SE-2013-0468-00 Comments:
None.
C U12 Table 06.03-02 Containment Spray Heat Exchanger Design Parameters (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 30 of Summary 106 UCR-2174 Rev. 0
- 19 UFSAR Effective Date: 12/29/2016 This UCR is a 50.71(e) evaluation type UCR to make editorial and other acceptable corrections to the UFSAR found during the revision process for UFSAR Revision 28.0.
- 1. UCR-2138, Rev. 0 was found to be missing three words on page 164 of 616 (for Section 14.3.4.3.1.2.2 (U1)) in numbered line 2. Line 2 should read "Maximized the Refueling Water Storage Tank (RWST) temperature (105°F, technical specification maximum)" as it is correctly worded in UFSAR Revision 27.0. The UCR is missing the words "technical specification maximum." It was determined that the error occurred during one of the last sessions of converting all the pages into one pdf document by reviewing the documents used in preparing the UCR. However, the three words are actually on the document, but are invisible - this was proved by copying the subject section in the pdf and then pasting it into a Word document. (The mechanism of this error is not understood.) The purpose of this discussion is to document that UFSAR Revision 27.0, Section 14.3.4.3.1.2.2 (U1) was correctly revised by UCR-2138, Rev. 0.
- 2. New Table 14.3.4-9 (U1) contained two typographical errors in the last two rows in the "Break Size" column. The 1.4ft2 values are replaced with the correct values of 1.0ft2 as directed by UCR-2054, Rev. 0.
- 3. New Table 14.3.4-10 (U1) contained four typographical errors - in four places "MSN" is replaced by the correct acronym "MSIV" as directed by UCR-2054.
- 4. Table 14.1-1 (U1) contained one value that did not get updated. The "Vessel /
Core Inlet" parameter for the "MUR Power Uprate - Case 2" value is changed to "541.7" as directed by UCR-2054, Rev. 0.
- 5. Section 14.3.4.1.3.2.3 (U2) delete the second of a double "the" in the fifth paragraph, fourth sentence
- 6. Section 14.1.0.3 (U1) - add the missing word "RCS" in the second sentence of the last paragraph as directed by UCR-2054, Rev. 0.
Approved:12/29/2016 Affected Unit: Both Units
- 1. UCR-2138, Rev. 0 - approved on November 12, 2015 Summary of Change:
Justification:
NDM Effective Date: 1/4/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 31 of Summary 106 UCR-2174 Rev. 0
- 19 UFSAR Effective Date: 12/29/2016
- 2. UCR-2054, Rev. 0 - approved on April 07, 2016
- 3. UCR-2054, Rev. 0 - approved on April 07, 2016
- 4. UCR-2054, Rev. 0 - approved on April 07, 2016 and e-mail " Kingseed to Kettle, 11-29-16 regarding Table 14.1-1 (U1)
- 5. UCR-2054, Rev. 0 - approved on April 07, 2016
- 6. UCR-2054, Rev. 0 - approved on April 07, 2016 This UFSAR change does not alter the Current License Basis.
Change #1 is only a discussion to clarify a possibly confusing situation.
For changes #2 though #6:
PMP-2350-SAR-001 "UFSAR Update Process", Attachment 1, 50.71(e) Criteria:
Item #4. Editorial Changes consisting of the following:
Typographical Changes that are corrections of errors introduced during the production (typing/printing/magnetic storage) of the document or changes resulting from repagination.
50.71(e) Basis:
List of Affected Items:
50.59: N/A None.
Comments:
C U1 Table 14.01-01 (U1)
Unit 1 Design Power Capability Parameters Used in Non-LOCA Safety Analyses C
U1 Table 14.03.04-09 (U1)
Double-Ended Rupture Steamline Breaks C
U1 Table 14.03.04-10 (U1)
Steamline Ruptures 14.01 (U2)
Core and Coolant Boundary Protection Analysis C
U1 Section 14.03.04.01.03.02.03 (U2)
Sensitivity of the Results C
U1 Section 14.03.04.03.01.02.02 (U1)
Mass and Energy Release Data R
U1 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 32 of Summary 106 UCR-2176 Rev. 0
- 20 UFSAR Effective Date: 01/09/2017 A total of twelve (12) stainless steel rods have been placed into fuel assemblies XX16 and ZZ51 to mitigate the risk of both baffle jetting and debris defects during Unit 2 Cycle 23 per EC-54548. This type of reconstitution activity is currently discussed in Unit 2 Section 3.2.1.2.2 Fuel Assembly Structure subsection Top Nozzle; however, Unit 2 Chapter 3 is being enhanced to include additional information regarding stainless steel filler rods, similar to wording that is currently included in the Unit 1 UFSAR Section 3.5.1.3. In addition, the fuel assembly reconstitution evaluation methodology (WCAP-13060-P-A) is being added to the References section for Unit 2 Section 3.2.
Approved:1/9/2017 Affected Unit: Unit Two EC-54548 Revision 0 Unit 2 Cycle 23 Core Reload Summary of Change:
Justification:
NDM Effective Date: 1/10/2017 List of Affected Items:
50.59: SS-SE-2016-0371-00 None.
Comments:
03.02.01.02.01 (U2)
Fuel Rods C
U2 Section 03.02.04 (U2)
References for Section 3.2 C
U2 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 33 of Summary 106 UCR-2178 Rev. 0
- 21 UFSAR Effective Date: 01/13/2017 Summary of Change:
Reformat the document properties for the List of Affected Figures below to meet NRC minimum standards for electronic UFSAR submission.
This is an administrative UCR, a 50.71(e) based change, which only reformats the document properties for the figures, no technical changes were made.
Approved:1/13/2017 Affected Unit: Unit One Justification:
PMP-2350-SAR-001 "UFSAR Update Process" - Attachment 1, Acceptable Changes to the UFSAR Controlled via 50.71(e) Evaluation UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
The changes made by this UCR meet the requirements of PMP-2350-SAR-001,, Item 4, Bullet 1:
- 4. Editorial Changes consisting of the following:
Format Changes made to improve layout or appearance.
This UFSAR change does not alter the Current License Basis.
50.71(e) Basis:
NDM Effective Date: 1/16/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 34 of Summary 106 UCR-2178 Rev. 0
- 21 UFSAR Effective Date: 01/13/2017 List of Affected Items:
50.59: N/A Comments:
None.
F U1 Figure 14.03.01-01D (U1)
D. C. Cook Unit 1 Limiting Case Broken and Intact Loop Pump Void Fraction F
U1 Figure 14.03.01-01DA (U1)
D. C. Cook Unit 1 Limiting Case Broken and Intact Loop Pump Void Fraction for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-01E (U1)
D. C. Cook Unit 1 Limiting Case Hot Assembly Top of Core Vapor Flow F
U1 Figure 14.03.01-01EA (U1)
D. C. Cook Unit 1 Limiting Case Hot Assembly Top of Core Vapor Flow for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-01F (U1)
D. C. Cook Unit 1 Limiting Case Pressurizer Pressure F
U1 Figure 14.03.01-01FA (U1)
D. C. Cook Unit 1 Limiting Case Pressurizer Pressure for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-01G (U1)
D. C. Cook Unit 1 Limiting PCT Lower Plenum Collapsed Liquid Level*
F U1 Figure 14.03.01-01GA (U1)
D. C. Cook Unit 1 Limiting PCT Lower Plenum Collapsed Liquid Level for the Return to RCS NOP/NOT Evaluation*
F U1 Figure 14.03.01-01H (U1)
D. C. Cook Unit 1 Limiting Case Vessel Fluid Mass F
U1 Figure 14.03.01-01HA (U1)
D. C. Cook Unit 1 Limiting Case Vessel Fluid Mass for the Return to RCS NOP/NOT Evaluation (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 35 of Summary 106 UCR-2182 Rev. 0
- 22 UFSAR Effective Date: 01/19/2017 Approved Alternative Source Term (AST) dose consequence analysis has reduced the maximum leakage allowable from containment from 0.25 wt % / day to 0.18 wt % /
day. The NRC Safety Evaluation for Amendment No. 332 to license DPR-58 and Amendment No. 314 to license DPR-74 dated October 20, 2016.
Approved:1/19/2017 Affected Unit: Both Units Justification:
SER dated October 20th, 2016.
UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
The changes made by this UCR meet the requirements of PMP-2350-SAR-001, Attachment 1, Item 2:
- 2. Changes made to the UFSAR that incorporate information contained in pre-approved NRC documents or already covered by an existing 50.59 Review.
This UFSAR change does not alter the Current License Basis.
Summary of Change:
50.71(e) Basis:
NDM Effective Date: 1/20/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 36 of Summary 106 UCR-2182 Rev. 0
- 22 UFSAR Effective Date: 01/19/2017 List of Affected Items:
50.59: N/A Tables 14.3.5-1 and 14.3.5-2 are assigned is to be revised under Configuration Document Impact (CDI) 2012-14068-14 to Nuclear Safety Analysis (NSA) under 2014-12706-12.
Table 14.2.6-2 of the UFSAR is to be revised by a UCR generated by NSA.
Comments:
01.02.05 Ice Condenser Containment Structure C
U12 Section 01.04.07 Engineered Safety Features - Criterion 52 - Containment Heat Removal Systems C
U12 Section 01.04.11 References for Section 1.4 C
U12 Section 05.02 Application of Design Criteria to the Containment Structure C
U12 Section 05.07.02 Initial Containment (Pre-Operational) Leakage Rate Tests - Integrated Leakage Rate Tests C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 37 of Summary 106 UCR-2183 Rev. 0
- 23 UFSAR Effective Date: 01/30/2017 Revise the first sentence in the last paragraph of 6.2.6 "Programmatic Controls" and add two new references to Section 6.2.7 "References for Section 6.2."
The first sentence of the last paragraph of Section 6.2.6 should be revised to read as follows:
Plant programs, process, procedures, and Technical Specifications also exist to ensure ECCS and CTS are maintained sufficiently filled with water to enable all safety related functions to be accomplished in both the injection and recirculation modes of operation in accordance with Cook Plant commitments to Generic Letter 2008-01 (reference 6.2.7.3, 6.2.7.6, and 6.2.7.7).
Add two new references to Section 6.2.7; these will be numbered as references #6 and
- 7; (UCR-2164 added new References #4 and #5).
Precursor Event Required:
The implementation of Amendment No. 331 to DPR-58 (Unit 1) and the implementation of Amendment No. 312 to DRP-74 (Unit 2).
Precursor Completed On: 1/30/2017 Approved:1/27/2017 Affected Unit: Both Units AEP-NRC-2016-07, CNP Unit 1 and Unit 2, License Amendment Request to Revise Technical Specifications to Adopt Technical Specifications Task Force-523, "Generic Letter 2008-01, Managing Gas Accumulation," Using the Consolidated Line Item Improvement Process.
A0816 CNP, Units 1 and 2, Issuance of Amendments to Revise TS to Adopt TS Task Force - 523, Generic Letter 2008-01, Managing Gas Accumulation.
The above documents resulted in the issuance of Amendment No. 331 to DPR-58 (Unit 1) and Amendment No. 312 to DRP-74 (Unit 2).
Summary of Change:
Justification:
NDM Effective Date: 1/27/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 38 of Summary 106 UCR-2183 Rev. 0
- 23 UFSAR Effective Date: 01/30/2017 UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
These changes are consistent with PMP-2350-SAR-001 "UFSAR Update Process",, Acceptable Changes to the UFSAR Controlled via 50.71(e) Evaluation" -
Item 2: "Changes made to the UFSAR that incorporate information contained in pre-approved NRC documents or already covered by an existing 50.59 Review."
This UFSAR change does not alter the Current Licensing Basis.
50.71(e) Basis:
List of Affected Items:
50.59: N/A The TS and Bases are being revised to include Surveillance Requirements with wording similar to verify Residual Heat Removal (RHR), Emergency Core Cooling System (ECCS), and Containment Spray (CTS) system locations susceptible to gas accumulation are sufficiently filled with water.
Comments:
06.02.06 Programmatic Controls C
U12 Section 06.02.07 References for Section 6.2 C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 39 of Summary 106 UCR-2180 Rev. 0
- 24 UFSAR Effective Date: 02/07/2017 Reformat the document properties for the List of Affected Figures below to meet NRC minimum standards for electronic UFSAR submission. This is an administrative UCR, a 50.71(e) based change, which only reformats the document properties for the figures, no technical changes were made.
Approved:2/7/2017 Affected Unit: Unit One Justification:
PMP-2350-SAR-001 "UFSAR Update Process" - Attachment 1, Acceptable Changes to the Updated Final Safety Analysis Report (UFSAR) Controlled via 50.71(e) Evaluation UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
The changes made by this UCR meet the requirements of PMP-2350-SAR-001,, Item 4, Bullet l:
- 4. Editorial Changes consisting of the following:
Format Changes made to improve layout or appearance.
This UFSAR change does not alter the Current License Basis.
Summary of Change:
50.71(e) Basis:
NDM Effective Date: 2/15/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 40 of Summary 106 UCR-2180 Rev. 0
- 24 UFSAR Effective Date: 02/07/2017 List of Affected Items:
50.59: N/A Comments:
None.
F U1 Figure 14.03.01-01N (U1)
D. C. Cook Unit 1 Limiting Case PCT Location*
F U1 Figure 14.03.01-01NA (U1)
D. C. Cook Unit 1 Limiting Case PCT Location for the Return to RCS NOP/NOT Evaluation*
F U1 Figure 14.03.01-02 (U1)
D. C. Cook Unit 1 BELOCA Analysis Axial Power Shape Operating Space Envelope F
U1 Figure 14.03.01-03 (U1)
D.C. Cook Unit 1 Lower Bound Containment Pressure F
U1 Figure 14.03.01-03A (U1)
D.C. Cook Unit 1 Lower Bound Containment Pressure for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-04 (U1)
D.C. Cook Unit 1 Containment Pressure*
F U1 Figure 14.03.01-04A (U1)
D.C. Cook Unit 1 Containment Pressure for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-05 (U1)
D.C. Cook Unit 1 Structural Heat Removal Rate F
U1 Figure 14.03.01-05A (U1)
D.C. Cook Unit 1 Structural Heat Removal Rate for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-06 (U1)
D.C. Cook Unit 1 Lower Compartment Structural Heat Removal Rate (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 41 of Summary 106 UCR-2181 Rev. 0
- 25 UFSAR Effective Date: 02/07/2017 Reformat the document properties for the List of Affected Figures below to meet NRC minimum standards for electronic UFSAR submission. This is an administrative UCR, a 50.71(e) based change, which only reformats the document properties for the figures, no technical changes were made.
Approved:2/7/2017 Affected Unit: Unit One Justification:
PMP-2350-SAR-001 "UFSAR Update Process" - Attachment 1, Acceptable Changes to the UFSAR Controlled via 50.71(e) Evaluation UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
The changes made by this UCR meet the requirements of PMP-2350-SAR-001,, Item 4, Bullet l:
- 4. Editorial Changes consisting of the following:
Format Changes made to improve layout or appearance.
This UFSAR change does not alter the Current License Basis.
Summary of Change:
50.71(e) Basis:
NDM Effective Date: 2/15/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 42 of Summary 106 UCR-2181 Rev. 0
- 25 UFSAR Effective Date: 02/07/2017 List of Affected Items:
50.59: N/A Comments:
None.
F U1 Figure 14.03.01-06A (U1)
D.C. Cook Unit 1 Lower Compartment Structure Heat Removal for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-07 (U1)
D.C. Cook Unit 1 Heat Removal by Sump F
U1 Figure 14.03.01-07A (U1)
D.C. Cook Unit 1 Heat Removal by Symp for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-08 (U1)
D.C. Cook Unit 1 Heat Removal by Lower Compartment Spray F
U1 Figure 14.03.01-08A (U1)
D.C. Cook Unit 1 Heat Removal by Lower Compartment Spray for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-09 (U1)
D.C. Cook Unit 1 Containment Temperatures*
F U1 Figure 14.03.01-09A (U1)
D.C. Cook Unit 1 Containment Temperatures for the Return to RCS NOP/NOT Evaluation (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 43 of Summary 106 UCR-2200 Rev. 0
- 26 UFSAR Effective Date: 02/07/2017 Revise Section 9.1 to replace a reference to Chapter 14 in the first sentence to correctly reference Section 1.4.
UCR-1024, Rev. 0 revised Section 9.1 extensively when relocating all the Plant Specific Design Criteria (formally the General Design Criteria) from individual chapters and subsections to a consolidated listing in Section 1.4 of Chapter 1. The markup clearly shows the section to be "1.4" but the resultant revised UFSAR (Revision 17.0) contained "Chapter 14" instead. This UCR corrects the reference back to Section 1.4.
The other reference to Chapter 14 in Section 9.1 is correct per the mark up of Section 9.1 in UCR-1024, Rev. 0.
Approved:2/7/2017 Affected Unit: Both Units Justification:
UCR-1024, Rev. 0 (page 115 of 136)
PMP-2350-SAR-001 "UFSAR Update Process" Attachment 1, Acceptable Changes to the UFSAR Controlled via 50.71(e) Evaluation" UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
This UFSAR change does not alter the Current License Basis.
This change falls under the guidance of PMP-2350-SAR-001, Attachment 1, Item 4, Bullet 4:
- 4. Editorial Changes consisting of the following:
Typographical Changes that are corrections of errors introduced during the production (typing/printing/magnetic storage) of the document or changes resulting from repagination.
Summary of Change:
50.71(e) Basis:
NDM Effective Date: 2/9/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 44 of Summary 106 UCR-2200 Rev. 0
- 26 UFSAR Effective Date: 02/07/2017 List of Affected Items:
50.59: N/A This typographical error occurred when revising UFSAR Revision 16.6 to UFSAR Revision 17.0.
Comments:
09.01 Application of Plant Design Criteria C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 45 of Summary 106 UCR-2179 Rev. 0
- 27 UFSAR Effective Date: 02/08/2017 Reformat the document properties for the List of Affected Figures below to meet NRC minimum standards for electronic UFSAR submission. This is an administrative UCR, a 50.71(e) based change, which only reformats the document properties for the figures, no technical changes were made.
Approved:2/8/2017 Affected Unit: Unit One Justification:
PMP-2350-SAR-001 "UFSAR Update Process" - Attachment 1, Acceptable Changes to the UFSAR Controlled via 50.71(e) Evaluation.
UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
The changes made by this UCR meet the requirements of PMP-2350-SAR-001,, Item 4, Bullet l:
- 4. Editorial Changes consisting of the following:
Format Changes made to improve layout or appearance.
This UFSAR change does not alter the Current License Basis.
Summary of Change:
50.71(e) Basis:
NDM Effective Date: 2/13/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 46 of Summary 106 UCR-2179 Rev. 0
- 27 UFSAR Effective Date: 02/08/2017 List of Affected Items:
50.59: N/A Comments:
None.
F U1 Figure 14.03.01-01I (U1)
D. C. Cook Unit 1 Limiting Case Loop 2 Accumulator Flow F
U1 Figure 14.03.01-01IA (U1)
D. C. Cook Unit 1 Limiting Case Loop 2 Accumulator Flow for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-01J (U1)
D. C. Cook Unit 1 Limiting Case Loop 2 Charging Safety Injection Flow F
U1 Figure 14.03.01-01JA (U1)
D. C. Cook Unit 1 Limiting Case Loop 2 Charging Safety Injection Flow for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-01K (U1)
D. C. Cook Unit 1 Limiting Case Loop 2 RHR and HHSI Flow F
U1 Figure 14.03.01-01KA (U1)
D. C. Cook Unit 1 Limiting Case Loop 2 RHR and HHSI Flow for the Return to RCS NOP/NOT Evaluation F
U1 Figure 14.03.01-01L (U1)
D. C. Cook Unit 1 Limiting Case Core Average Channel Collapsed Liquid Level*
F U1 Figure 14.03.01-01LA (U1)
D. C. Cook Unit 1 Limiting Case Core Average Channel Collapsed Liquid Level for the Return to RCS NOP/NOT Evaluation*
F U1 Figure 14.03.01-01M (U1)
D. C. Cook Unit 1 Limiting Case Loop 2 Downcomer Collapsed Liquid Level*
F U1 Figure 14.03.01-01MA (U1)
D. C. Cook Unit 1 Limiting Case Loop 2 Downcomer Collapsed Liquid Level for the Return to RCS NOP/NOT Evaluation*
(Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 47 of Summary 106 UCR-2184 Rev. 0
- 28 UFSAR Effective Date: 02/15/2017 Summary of Change:
Revise Table 9.8-5 per Condition Evaluation (CE) 2016-9695-2.
During the Auxiliary Feedwater (AFW) Vertical Slice Review a discrepancy between the required AFW flowrate when pulling from the Essential Service Water (ESW) supply was noted. Based on CE 2016-9695-2, the second sentence of UFSAR Table 9.8.5, Note 5, which states:
The required [AFW] flow is reduced to 250 gpm corresponding to heat sink requirement following depletion of CST inventory.
This sentence does not accurately account for the potential realignment of AFW to ESW after 33 minutes and should be removed. Calculations show that an ESW flowrate in excess of 450 gpm can be obtained to meet the required AFW flowrate of 450 gpm.
Therefore, the second sentence of Table 9.8.5, Note 5, does not apply.
Approved:2/15/2017 Affected Unit: Both Units Condition Evaluation AR 2016-9695-2 PMP-2350-SAR-001 "UFSAR Update Process", Attachment 1, Acceptable Changes to the UFSAR Controlled via 50.71(e) Evaluation UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
This change meets the criteria in PMP-2350-SAR-001 "UFSAR Update Process", "Acceptable Changes to the Updated Final Safety Analysis Report (UFSAR) Controlled via 50.71(e) Evaluation" Item 8 "Correction of UFSAR information to make it consistent with current licensing basis."
The deletion of reduced ESW required flow rate in Note 5 of UFSAR Table 9.8.5 is a Justification:
50.71(e) Basis:
NDM Effective Date: 2/17/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 48 of Summary 106 UCR-2184 Rev. 0
- 28 UFSAR Effective Date: 02/15/2017 removal of inaccurate information. Removing this information does not change any requirements of ESW supply to AFW; instead, it removes incorrect information for the required ESW flowrate following CST depletion.
Removal of this information does not alter the CNP license basis of the AFW system.
This UFSAR change does not alter the Current License Basis.
List of Affected Items:
50.59: N/A Comments:
None.
C U12 Table 09.08-05 Essential Service Water System Flow Requirements per Train (GPM)
(Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 49 of Summary 106 UCR-2129 Rev. 0
- 29 UFSAR Effective Date: 03/02/2017 Westinghouse issued a Nuclear Safety Analysis Letter (NSAL 14-5) which explained that there were issues with Westinghouses critical heat flux (CHF) correlations (also referred to as departure from nucleate boiling or DNB). NSAL 14-5 identified that DNB correlations used in the Unit 1 (Westinghouse Rod Bundle (WRB)-1) and Unit 2 (WRB-2) DNB analysis are potentially impacted at conditions of high quality. Westinghouse has performed the DNB safety analysis for both Unit 1 and Unit 2 and confirms DNB statepoints as part of core reload on a cycle by cycle basis. The NSAL states the following regarding long term corrective actions (LTCAs) for plants where Westinghouse performs the DNB safety analyses:
"For those plants where Westinghouse performs the DNB safety analyses on behalf of the customer, the CHF non-conservatism would not create a significant safety hazard (SSH) for one or both of the following reasons:
- 1. For the analysis of design basis accident (DBA) transients which are based on the DNB design criterion, the plant does not reach those conditions at which the affected correlation was found to be non-conservative. Therefore, the safety analysis results would be unaffected. This applies to all plants except those with 14 foot 17x17 Robust Fuel Assembly (RFA) fuel without IFM grids. (LTCA-1)
- 2. The available DNB margin in the analysis of the DBA transients is sufficient to more than compensate for the CHF correlation non-conservatism. The use of some DNB margin was only necessary for plants with 14 foot 17x17 RFA fuel without Intermediate Flow Mixnig (IFM) grids. (LTCA-2)
CNP does not use 14 foot 17x17 RFA fuel. Therefore, the existing CNP safety analyses results are unaffected per LTCA-1.
Long Term Actions:
As discussed, the CNP DNB correlations WRB-1 and WRB-2 are impacted by the issue identified in NSAL 14-5; however, the actual conditions where the correlation is non conservative is not applicable for either Unit 1 or Unit 2 design basis DNB event analysis results. It should also be noted that the WRB-3 DNB correlation is also used in both Unit 1 and Unit 2 (e.g., for steam line break statepoint evaluation) but is unaffected by the issue identified in NSAL 14-5.
AEP-15-2, Table 4, indicates that the impacted quality regime is above 25% quality for Unit 1 15x15 fuel and above 26% quality for Unit 2 17x17 fuel for the applicable correlations.
Approved:3/2/2017 Affected Unit: Unit One Summary of Change:
NDM Effective Date: 3/9/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 50 of Summary 106 UCR-2129 Rev. 0
- 29 UFSAR Effective Date: 03/02/2017 CNP UFSAR Sections 3.5.3.3.1 (U1) and 3.4.2.3.1 (U2) both list the applicable range of variables for the WRB-1 and WRB-2 correlations. Included in the list of variables is local quality, which is limited to less than 30% for both WRB-1 and WRB-2. Resolution of NSAL 14-5 involves applying a penalty if local quality is 0.25 for Unit 1 and 0.26 for Unit 2. UFSAR Section 3 needs to be changed to reflect the revised quality regimes.
Since the WRB-1 and WRB-2 correlations were licensed for use by the NRC, the UFSAR should be modified to note that NSAL 14-5 places more restrictive limits on quality of 25% (U1) and 26% (U2). The existing value of 30% should be retained in the UFSAR as the licensed limit applicable to both WRB-1 and WRB-2. Additionally, AEP-15-59 will be added to Chapter 3 of the UFSAR as a reference.
Justification:
Westinghouse NSAL 14-5 Lower Than Expected Critical Heat Flux Results Obtained During Departure from Nucleate Boiling Testing AEP-15-59, American Electric Power, CNP Units 1 & 2, Westinghouse Resolution Plan and Technical Basis for NSAL-14-5, "Lower Than Expected Critical Heat Flux Results Obtained During DNB Testing," December 1, 2015.
List of Affected Items:
50.59: SS-SE-2017-0012-00 Comments:
None.
03.05.03.03.01 (U1)
Critical Heat Flux Ratio or Departure from Nucleate Boiling Ration and Mixing Technology (Upgrade)
C U1 Section 03.05.03.06 (U1)
References for Section 3.5.3 (Upgrade)
C U1 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 51 of Summary 106 UCR-2150 Rev. 0
- 30 UFSAR Effective Date: 03/02/2017 Westinghouse issued a Nuclear Safety Analysis Letter (NSAL 14-5) which explained that there were issues with Westinghouses CHF correlations (also referred to as DNB). NSAL 14-5 identified that DNB correlations used in the Unit 1 (WRB-1) and Unit 2 (WRB-2)
DNB analysis are potentially impacted at conditions of high quality. Westinghouse has performed the DNB safety analysis for both Unit 1 and Unit 2 and confirms DNB statepoints as part of core reload on a cycle by cycle basis. The NSAL states the following regarding LTCAs for plants where Westinghouse performs the DNB safety analyses:
For those plants where Westinghouse performs the DNB safety analyses on behalf of the customer, the CHF non-conservatism would not create a significant safety hazard (SSH) for one or both of the following reasons:
- 1. For the analysis of design basis accident (DBA) transients which are based on the DNB design criterion, the plant does not reach those conditions at which the affected correlation was found to be non-conservative. Therefore, the safety analysis results would be unaffected. This applies to all plants except those with 14 foot 17x17 RFA fuel without IFM grids. (LTCA-1)
- 2. The available DNB margin in the analysis of the DBA transients is sufficient to more than compensate for the CHF correlation non-conservatism. The use of some DNB margin was only necessary for plants with 14 foot 17x17 RFA fuel without IFM grids. (LTCA-2)
CNP does not use 14 foot 17x17 RFA fuel. Therefore, the existing CNP safety analyses results are unaffected per LCTA-1.
Long Term Actions:
As discussed, the CNP DNB correlations WRB-1 and WRB-2 are impacted by the issue identified in NSAL 14-5, however the actual conditions where the correlation is non conservative is not applicable for either Unit 1 or Unit 2 design basis DNB event analysis results. It should also be noted that the WRB-3 DNB correlation is also used in both Unit 1 and Unit 2 (e.g., for steam line break statepoint evaluation) but is unaffected by the issue identified in NSAL 14-5.
AEP-15-2, Table 4, indicates that the impacted quality regime is above 25% quality for Unit 1 Approved:3/2/2017 Affected Unit: Unit Two Summary of Change:
NDM Effective Date: 3/8/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 52 of Summary 106 UCR-2150 Rev. 0
- 30 UFSAR Effective Date: 03/02/2017 15x15 fuel and above 26% quality for Unit 2 17x17 fuel for the applicable correlations.
CNP UFSAR Sections 3.5.3.3.1 (U1) and 3.4.2.3.1 (U2) both list the applicable range of variables for the WRB-1 and WRB-2 correlations. Included in the list of variables is local quality, which is limited to less than 30% for both WRB-1 and WRB-2. Resolution of NSAL 14-5 involves applying a penalty if local quality is 0.25 for Unit 1 and 0.26 for Unit 2. UFSAR Section 3 needs to be changed to reflect the revised quality regimes.
Since the WRB-1 and WRB-2 correlations were licensed for use by the NRC the UFSAR should be modified to note that NSAL 14-5 places more restrictive limits on quality of 25% (U1) and 26% (U2). The existing value of 30% should be retained in the UFSAR as the licensed limit applicable to both WRB-1 and WRB-2. Additionally, AEP-15-59 will be added to Chapter 3 of the UFSAR as a reference.
Westinghouse NSAL 14-5 Lower Than Expected Critical Heat Flux Results Obtained During Departure from Nucleate Boiling Testing AEP-15-59, American Electric Power, CNP Units 1 & 2, Westinghouse Resolution Plan and Technical Basis for NSAL-14-5, "Lower Than Expected Critical Heat Flux Results Obtained During DNB Testing," December 1, 2015.
Justification:
List of Affected Items:
50.59: SS-SE-2017-0012-00 Comments:
None.
03.04.02.03.01 (U2)
Departure from Nucleate Boiling Technology C
U2 Section 03.04.06 (U2)
References for Section 3.4 C
U2 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 53 of Summary 106 UCR-2168 Rev. 0
- 31 UFSAR Effective Date: 03/03/2017 Summary of Change:
Table 9.8.3 - revise Non-essential Service Water (NESW) pump Casing material from: "cast iron" to be "cast iron or cast steel" Table 9.8.3 - change NESW Strainers from: Duplex-automatic backwashing to Automatic - Self Cleaning Figures 9.8-4 and 9.8-5, change NESW Unit 1 and 2, to depict single stage (vs duplex)
- strainer, and change label from Duplex Strainer to Precursor Event Required:
After the RTO of the first of any of the following four INDUS ECs for the "NESW Pumps and Discharge Strainers" modifications:
54253 (Unit 1 North) 54254 (Unit 1 South) 54255 (Unit 2 North) 54256 (Unit 2 South)
Precursor Completed On: 3/3/2017 Approved:2/10/2017 Affected Unit: Both Units NDM Effective Date: 2/14/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 54 of Summary 106 UCR-2168 Rev. 0
- 31 UFSAR Effective Date: 03/03/2017 Duplex Strainer or Automatic Strainer.
Change Section 9.8.3.2 "Description" - "Non-Essential Service Water System" second paragraph, third sentence, from:
All motor-operated valves on the main non-essential service water system are operated from the station battery system with the exception of the inlet and outlet valve motor operators for each of the main NESW duplex strainers, which are powered from the 600 VAC auxiliary bus.
Revise to state the following:
All motor-operated valves on the main non-essential water system are operated from the station battery system with the exception of the inlet and outlet valve motor operators for each of the main NESW duplex strainers, or the motor operated strainer discharge motor operators for each of the NESW automatic strainers, which are powered from the 600 VAC auxiliary bus.
Change Section 9.8.3.2 "Description" - "Non-Essential Service Water System" third paragraph, from:
The discharge strainers of the pumps are of duplex construction, with automatic backwashing. Each strainer is effectively two strainers in one casing with flow directed through one half, while slide gates block off the other half. When the strainer is in service and if it becomes dirty or clogged, a high differential pressure signal initiates a shift of the slide gates blocking the flow to the dirty basket and directing it through the clean basket.
The dirty basket is then backwashed and is ready for re-use.
to read as follows (revise sentence 1, add sentences 5, 6, & 7):
The discharge strainers of the pumps are of duplex construction, with automatic backwashing, or a single basket automatic strainer. Each duplex strainer is effectively two strainers in one casing with flow directed through one half, while slide gates block off the other half. When the strainer is in service and if it becomes dirty or clogged, a high differential pressure signal initiates a shift of the slide gates blocking the flow to the dirty basket and directing it through the clean basket. The dirty basket is then backwashed and is ready for re-use. The automatic discharge strainers of the pumps are provided with automatic backwashing. Each strainer contains an internal arm that cleans a segment (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 55 of Summary 106 UCR-2168 Rev. 0
- 31 UFSAR Effective Date: 03/03/2017 of the strainer during backwash cycles. When the strainer is in service and if it becomes dirty or clogged, a high differential pressure signal initiates rotation of the internal arm that directs a small portion of the NESW flow into reverse flow across a basket segment, and directs the reverse flow and debris to a drain, maintaining the strainer in service during the backwash cycle.
EC-54253 "NESW Pump and Strainer Replacement" Justification:
List of Affected Items:
50.59: SS-SE-2016-0190-01 This is the first of two sequential UCRs associated with the EC-54253 modification of NESW pumps and discharge strainers. This first UCR modifies the UFSAR to reflect that strainers are either duplex or single basket automatic style (following the first train RTO), and the second UCR following the forth train RTO, eliminates reference to duplex strainers.
Comments:
C U1 Figure 09.08-04 Non-Essential Service Water Unit No. 1 C
U2 Figure 09.08-05 Non Essential Service Water Unit No. 2 C
U12 Table 09.08-03 Service Water Systems Components Design Data 09.08.03.02 Description - Non-Essential Service Water System C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 56 of Summary 106 UCR-2201 Rev. 0
- 32 UFSAR Effective Date: 03/06/2017 UCR-2201, Rev. 0 is an Administrative UCR to fix or enhance items found while revising the UFSAR to Revision 28.0; this one focuses on Chapter 9, Sections 9.0, 9.1, &
9.2.
Sections 9.0, 9.1, & 9.2 Number all unnumbered, but titled subsections.
Section 9.2.2.5 - "Makeup" Convert the four numbered items in the second paragraph into a numbered list Section 9.2.2.7 - "Boration" Convert the three numbered items in the first paragraph into a numbered list Section 9.2.2.9 - "Charging Pump Control" Convert the
- footnote into a numbered footnote 1 Section 9.2.2.19 - "Charging Pumps" Convert the two
- footnotes into a numbered footnote 2 Section 9.2.3.5 - "Malfunction Analysis" Convert the two lettered items in the sixth paragraph into a lettered list Section 9.2.3.6 - "Galvanic Corrosion" Convert the superscripted (1) found in the first paragraph, second sentence next Approved:3/6/2017 Affected Unit: Both Units Summary of Change:
NDM Effective Date: 3/10/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 57 of Summary 106 UCR-2201 Rev. 0
- 32 UFSAR Effective Date: 03/06/2017 from:
"have been shown(1)"
to "have been shown (Ref. 1)"
- the superscripted (1) could be interpreted as a footnote marker and there is not a footnote associated with this Section; this was meant to refer the reader to Reference 1 in Section 9.2.3.1 "References for Section 9.2."
PMP-2350-SAR-001 "UFSAR Update Process" PMP-2350-SAR-001, Attachment 1, Acceptable Changes to the Updated Final Safety Analysis Report (UFSAR) Controlled via 50.71(e) Evaluation UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
This UFSAR change does not alter the Current License Basis.
Adding numbers to titled, but unnumbered sections:
- 4. Editorial Changes consisting of the following:
Standardization Changes made to establish a consistent means of presenting information.
Changing
- Footnotes into numbered Footnotes:
- 4. Editorial Changes consisting of the following:
Standardization Changes made to establish a consistent means of presenting information.
Changing numbered or lettered lists in sentences to show as numbered or lettered lists:
- 4. Editorial Changes consisting of the following:
Justification:
50.71(e) Basis:
(Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 58 of Summary 106 UCR-2201 Rev. 0
- 32 UFSAR Effective Date: 03/06/2017 Standardization Changes made to establish a consistent means of presenting information.
Changing the Reference Number from a Footnote appearance into a clear reference to a Reference Number:
- 4. Editorial Changes consisting of the following:
Standardization Changes made to establish a consistent means of presenting information.
(Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 59 of Summary 106 UCR-2201 Rev. 0
- 32 UFSAR Effective Date: 03/06/2017 List of Affected Items:
50.59: N/A Comments:
None.
09.02.02 System Design and Operation - Condensate Filter D
U12 Section 09.02.02 System Design and Operation - Batching Tank D
U12 Section 09.02.02 System Design and Operation - Boric Acid Tank Heaters D
U12 Section 09.02.02 System Design and Operation - Boric Acid Transfer Pumps D
U12 Section 09.02.02 System Design and Operation - Boric Acid Blender D
U12 Section 09.02.02 System Design and Operation - Holdup Tanks D
U12 Section 09.02.02 System Design and Operation - Boric Acid Reserve Tank D
U12 Section 09.02.02 System Design and Operation - Holdup Tank Recirculation Pump D
U12 Section 09.02.02 System Design and Operation - Boric Acid Evaporator Feed Pumps D
U12 Section 09.02.02 System Design and Operation - Evaporator Feed Ion Exchangers D
U12 Section 09.02.02 System Design and Operation - Ion Exchanger Filters D
U12 Section 09.02.02 System Design and Operation - Boric Acid Filter D
U12 Section 09.02.02 System Design and Operation - Evaporator Condensate Demineralizers D
U12 Section 09.02.02 System Design and Operation - Seal Water Injection Filters D
U12 Section 09.02.02 System Design and Operation - Monitor Tanks D
U12 Section 09.02.02 System Design and Operation - Monitor Tank Pumps D
U12 Section 09.02.02 System Design and Operation - Deborating Demineralizers D
U12 Section 09.02.02 System Design and Operation - Concentrates Filter D
U12 Section 09.02.02 System Design and Operation - Concentrates Holding Tank D
U12 Section 09.02.02 System Design and Operation - Concentrates Holding Tank Transfer Pumps D
U12 Section 09.02.02 System Design and Operation - Electrical Heat Tracing D
U12 Section 09.02.02 System Design and Operation - Valves D
U12 Section 09.02.02 System Design and Operation - Piping D
U12 Section 09.02.02 System Design and Operation C
U12 Section 09.02.02 System Design and Operation - Boric Acid Evaporators D
U12 Section 09.02.02 System Design and Operation - Letdown Orfices D
U12 Section 09.02.02 System Design and Operation - System Description D
U12 Section 09.02.02 System Design and Operation - Expected Operating Conditions D
U12 Section 09.02.02 System Design and Operation - Reactor Coolant Activity Concentration D
U12 Section 09.02.02 System Design and Operation - Reactor Makeup Control Modes D
U12 Section 09.02.02 System Design and Operation - Makeup D
U12 Section 09.02.02 System Design and Operation - Dilution D
U12 Section 09.02.02 System Design and Operation - Boration D
U12 Section 09.02.02 System Design and Operation - Alarm Functions D
U12 Section 09.02.02 System Design and Operation - Charging Pump Control D
U12 Section 09.02.02 System Design and Operation - Centrifugal Charging Pumps D
U12 Section 09.02.02 System Design and Operation - Boric Acid Tanks D
U12 Section 09.02.02 System Design and Operation - Regenerative Heat Exchanger D
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 60 of Summary 106 UCR-2201 Rev. 0
- 32 UFSAR Effective Date: 03/06/2017 09.02.02 System Design and Operation - Letdown Heat Exchanger D
U12 Section 09.02.02 System Design and Operation - Mixed Bed Demineralizers D
U12 Section 09.02.02 System Design and Operation - Cation Bed Demineralizer D
U12 Section 09.02.02 System Design and Operation - Reactor Coolant Filter D
U12 Section 09.02.02 System Design and Operation - Volume Control Tank D
U12 Section 09.02.02 System Design and Operation - Charging Pumps D
U12 Section 09.02.02 System Design and Operation - Chemical Mixing Tank D
U12 Section 09.02.02 System Design and Operation - Excess Letdown Heat Exchanger D
U12 Section 09.02.02 System Design and Operation - Seal Water Heat Exchanger D
U12 Section 09.02.02 System Design and Operation - Seal Water Filter D
U12 Section 09.02.02 System Design and Operation - Components D
U12 Section 09.02.02.01
System Description
N U12 Section 09.02.02.02 Expected Operating Conditions N
U12 Section 09.02.02.03 Reactor Coolant Activity Concentration N
U12 Section 09.02.02.04 Reactor Makeup Control Modes N
U12 Section 09.02.02.05 Makeup N
U12 Section 09.02.02.05 Makeup F
U12 Section 09.02.02.06 Dilution N
U12 Section 09.02.02.07 Boration N
U12 Section 09.02.02.07 Boration F
U12 Section 09.02.02.08 Alarm Functions N
U12 Section 09.02.02.09 Charging Pump Control F
U12 Section 09.02.02.09 Charging Pump Control N
U12 Section 09.02.02.10 Centrifugal Charging Pumps N
U12 Section 09.02.02.11 Components N
U12 Section 09.02.02.12 Regenerative Heat Exchanger N
U12 Section 09.02.02.13 Letdown Orifices N
U12 Section 09.02.02.14 Letdown Heat Exchanger N
U12 Section 09.02.02.15 Mixed Bed Demineralizers N
U12 Section 09.02.02.16 Cation Bed Demineralizer N
U12 Section 09.02.02.17 Reactor Coolant Filter N
U12 Section 09.02.02.18 Volume Control Tank N
U12 Section 09.02.02.19 Charging Pumps N
U12 Section 09.02.02.19 Charging Pumps F
U12 Section 09.02.02.20 Chemical Mixing Tank N
U12 Section 09.02.02.21 Excess Letdown Heat Exchanger N
U12 Section 09.02.02.22 Seal Water Heat Exchanger N
U12 Section 09.02.02.23 Seal Water Filter N
U12 Section 09.02.02.24 Seal Water Injection Filters N
U12 Section 09.02.02.25 Boric Acid Filter N
U12 Section 09.02.02.26 Boric Acid Tanks N
U12 Section 09.02.02.27 Batching Tank N
U12 Section 09.02.02.28 Boric Acid Tank Heaters N
U12 Section 09.02.02.29 Boric Acid Transfer Pumps N
U12 Section 09.02.02.30 Boric Acid Blender N
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 61 of Summary 106 UCR-2201 Rev. 0
- 32 UFSAR Effective Date: 03/06/2017 09.02.02.31 Holdup Tanks N
U12 Section 09.02.02.32 Boric Acid Reserve Tank N
U12 Section 09.02.02.33 Holdup Tank Recirculation Pump N
U12 Section 09.02.02.34 Boric Acid Evaporator Feed Pumps N
U12 Section 09.02.02.35 Evaporator Feed Ion Exchangers N
U12 Section 09.02.02.36 Ion Exchanger Filters N
U12 Section 09.02.02.37 Boric Acid Evaporators N
U12 Section 09.02.02.38 Evaporator Condensate Demineralizers N
U12 Section 09.02.02.39 Condensate Filter N
U12 Section 09.02.02.40 Monitor Tanks N
U12 Section 09.02.02.41 Monitor Tank Pumps N
U12 Section 09.02.02.42 Deborating Demineralizers N
U12 Section 09.02.02.43 Concentrates Filter N
U12 Section 09.02.02.44 Concentrates Holding Tank N
U12 Section 09.02.02.45 Concentrates Holding Tank Transfer Pumps N
U12 Section 09.02.02.46 Electrical Heat Tracing N
U12 Section 09.02.02.47 Valves N
U12 Section 09.02.02.48 Piping N
U12 Section 09.02.03 System Design Evaluation - Malfunction Analysis D
U12 Section 09.02.03 System Design Evaluation - Incident Control D
U12 Section 09.02.03 System Design Evaluation - Availability and Reliability D
U12 Section 09.02.03 System Design Evaluation - Leakage Provisions D
U12 Section 09.02.03 System Design Evaluation - Control of Tritium D
U12 Section 09.02.03 System Design Evaluation - Galvanic Corrosion D
U12 Section 09.02.03 System Design Evaluation - Tests and Inspections D
U12 Section 09.02.03.01 Availability and Reliability N
U12 Section 09.02.03.01 References for Section 9.2 D
U12 Section 09.02.03.02 Control of Tritium N
U12 Section 09.02.03.03 Leakage Provisions N
U12 Section 09.02.03.04 Incident Control N
U12 Section 09.02.03.05 Malfunction Analysis N
U12 Section 09.02.03.05 Malfunction Analysis F
U12 Section 09.02.03.06 Galvanic Corrosion N
U12 Section 09.02.03.06 Galvanic Corrosion F
U12 Section 09.02.03.07 Tests and Inspections N
U12 Section 09.02.04 References for Section 9.2 N
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 62 of Summary 106 UCR-2122 Rev. 0
- 33 UFSAR Effective Date: 04/03/2017 Revise selective sections of Chapter 1, Chapter 10 and Section 14.1 (U1) as required by EC-52519 - Unit 2 High Pressure Turbine and Low Pressure Turbines Retrofit with Alstom Turbines.
Incorporate EC-52519 Unit 2 High Pressure Turbine and Low Pressure Turbines Retrofit with Alstom Turbines for the Unit 2 main turbine retrofit into the UFSAR.
Precursor Event Required:
After RTO of EC-52519 Unit 2 High Pressure Turbine and Low Pressure Turbines Retrofit with Alstom Turbines Precursor Completed On: 4/3/2017 Approved:10/3/2016 Affected Unit: Unit Two EC-52519 Unit 2 High Pressure Turbine and Low Pressure Turbines Retrofit with Alstom Turbines Summary of Change:
Justification:
NDM Effective Date: 10/12/2016 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 63 of Summary 106 UCR-2122 Rev. 0
- 33 UFSAR Effective Date: 04/03/2017 List of Affected Items:
50.59: SS-SE-2014-0043-00 EC-52519 retrofits the Unit 2 high and low pressure turbines with Alstom turbines. The proposed UCR revises the UFSAR description for the Unit 2 high and low pressure turbines.
The changes are needed to reflect the retrofit design completed by EC-52519.
The change also incorporates NRC approved missile probability analysis instead of mechanistic analysis that was used for the Brown Boveri Corporation (BBC) turbines.
Comments:
C U2 Figure 14.01.13-01 (U2)?
Cross Section - L.P. Turbine Rotor (Unit 2)
C U2 Figure 14.01.13-02 (U2)?
Stress in L-P Turbine Rotor Discs at 176% Speed (Unit 2)
C U2 Figure 14.01.13-03 (U2)?
Calculated Failure Speeds of L-P Turbine Rotor Discs (Unit 2)
C U2 Figure 14.01.13-05 (U2)?
Sketch of Low Pressure Turbine Rotor and Casing Unit No. 2 C
U2 Table 05.01-01 Potential Missiles Considered in Class I (Seismic) Structure Design C
U1 Table 14.01.13-01 (U1)
Potential Turbine-Generator Missiles 01.00 Introduction and Summary C
U12 Section 01.04.07 Engineered Safety Features - Criterion 40 - Missile Protection C
U12 Section 01.04.07 Engineered Safety Features - Criterion 40 - Missile Protection - Potential Missiles C
U12 Section 01.04.07 Engineered Safety Features - Criterion 40 - Missile Protection -
Probability Analysis of Missile Generation Occurrence C
U12 Section 01.04.11 References for Section 1.4 C
U12 Section 10.01 General Descriptions C
U12 Section 10.03.02 Equipment Description C
U12 Section 14.01.13.01 (U1)
Introduction C
U1 Section 14.01.13.01.02 (U1)
Unit No. 2 - Turbine - Units 1 and 2 N
U1 Section 14.01.13.01.02 (U1)
Unit No. 2 - Brown-Boveri D
U1 Section 14.01.13.01.02 (U1)
Unit No. 2 - Brown-Boveri - Units 1 and 2 D
U1 Section 14.01.13.01.02 (U1)
Unit No. 2 - Turbine N
U1 Section 14.01.13.02 (U1)
Failure of Turbine-Generator Rotating Elements C
U1 Section 14.01.13.02.01 (U1)
Unit 1 (General Electric ^4)
C U1 Section 14.01.13.02.02 (U1)
Unit 1 (Brown-Boveri)
D U1 Section 14.01.13.02.02 (U1)
Unit 1 (Brown-Boveri**)
N U1 Section 14.01.13.02.02.01 (U1)
Vane of Last Stage Blade - Unit No. 2 (See Figure 14.1.13-5)
D U1 Section 14.01.13.02.02.01 (U1)
Vane of Last Stage Blade - Unit No. 2 (See Figure 14.1.13-5)**
N U1 Section 14.01.13.02.02.02 (U1)
Fragment of Next to Last Stage Disc - Unit 2 D
U1 Section 14.01.13.02.02.02 (U1)
Fragment of Next to Last Stage Disc - Unit 2**
N U1 Section 14.01.13.03 (U1)
References for Section 14.1.13 C
U1 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 64 of Summary 106 UCR-2202 Rev. 0
- 34 UFSAR Effective Date: 04/10/2017 The primary change driver for UCR-2202, Rev. 0 is the control room habitability and offsite radiological dose consequence analysis project using the AST methodology outlined in Regulatory Guide (RG) 1.183.
This project was initiated to resolve numerous technical issues identified in AR #s 2011-2817, 2012-12451, 2012-14068, and 2012-14106, which are present in the current dose consequence analyses of record (AOR) described in the UFSAR. The new dose analyses were performed for the following accident scenarios:
Loss of Coolant Accident (LOCA)
Fuel Handling Accident (FHA)
Main Steam Line Break (MSLB)
Steam Generator Tube Rupture (SGTR)
Locked Rotor Accident (LRA)
Control Rod Ejection (CRE)
Waste Gas Decay Tank (WGDT) Rupture Volume Control Tank (VCT) Rupture Additionally, the atmospheric dispersion factors and the core RCS source terms utilized as inputs in the above analyses have been revised. The revised AST dose analyses were submitted via a License Amendment Request (LAR) to the NRC and were approved. Note that, along with the dose consequence analyses, TSTF-490 will be implemented as part of the LAR.
Precursor Event Required:
After RTO of EC-53882 "Implementation of Offsite & Control Room Alternative Source Term (AST) Analyses" Precursor Completed On: 4/10/2017 Approved:3/30/2017 Affected Unit: Both Units EC-53882 "Implementation of Offsite & Control Room Alternative Source Term (AST) Analyses" Summary of Change:
Justification:
NDM Effective Date: 4/3/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 65 of Summary 106 UCR-2202 Rev. 0
- 34 UFSAR Effective Date: 04/10/2017 TSTF-490 (Deletion of E Bar Definition and Revision to RCS Specific Activity TS)
(Sorted by the "UFSAR Effective Date")
Comments:
None.
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 66 of Summary 106 UCR-2202 Rev. 0
- 34 UFSAR Effective Date: 04/10/2017 List of Affected Items:
50.59: SS-SE-2017-0071-00 01.00 Introduction and Summary C
U12 Section 01.03.08 Safety Features C
U12 Section 01.04.03 Nuclear and Radiation Controls - Critierion 11 - Control Room C
U12 Section 01.04.07 Engineered Safety Features - Criterion 40 - Missile Protection - Missile Protection Methods C
U12 Section 01.04.07 Engineered Safety Features (PSDC 37 - PSDC 65)
C U12 Section 01.04.07 Engineered Safety Features - Criterion 40 - Missile Protection -
Probabilistic Methodology for Determining Risk from Tornado Generated Missiles C
U12 Section 01.04.07 Engineered Safety Features - Criterion 49 - Reactor Containment Design Basis C
U12 Section 01.04.07 Engineered Safety Features - Criterion 54 - Initial Leak Rate Testing for Containment C
U12 Section 01.04.07 Engineered Safety Features - Criterion 40 - Missile Protection C
U12 Section 01.04.08 Fuel and Waste Storage Systems - Criterion 69 - Protection Against Radioactivity Release from Spent Fuel and Waste Storage C
U12 Section 01.04.08 Fuel and Waste Storage Systems (PSDC 66 - PSDC 69)
C U12 Section 01.04.09 Effluents - Criterion 70 - Control of Releases of Radioactivity to the Environment C
U12 Section 06.01 Application of ESF Design Criteria C
U12 Section 06.03.01 Application of Design Criteria C
U12 Section 06.03.02 System Design - System Description C
U12 Section 07.02.01.01.11 Protective Actions C
U12 Section 11.02.01 Design Basis C
U12 Section 11.03.01 Application of Design Criteria C
U12 Section 14.00 (U1)
Safety Analysis [Unit 1]
C U1 Section 14.00 (U2)
Safety Analysis [Unit 2]
C U2 Section 14.01.06.04.08 (U1)
Locked Rotor Radiological Consequence Analysis C
U1 Section 14.01.06.04.09 (U1)
Conclusions C
U1 Section 14.02.01.06 (U1)
Radiological Consequence Analysis C
U1 Section 14.02.03.01 (U1)
Volume Control Tank C
U1 Section 14.02.03.02 (U1)
Gas Decay Tanks C
U1 Section 14.02.04.05 (U1)
Radiological Consequence Analysis C
U1 Section 14.02.04.06 (U1)
Conclusion C
U1 Section 14.02.05.03 (U1)
Radiological Consequence Analysis C
U1 Section 14.02.05.05 (U1)
Conclusions C
U1 Section 14.02.06.21 (U1)
Radiological Consequence Analysis C
U1 Section 14.02.06.22 (U1)
Conclusions C
U1 Section 14.03.05 (U1)
Environmental Consequences of a Loss-Of-Coolant-Accident C
U1 Section 14.03.05.02 (U1)
Basic Events and Release Fractions C
U1 Section 14.03.05.03 (U1)
Containment Iodine Removal Capabilities C
U1 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 67 of Summary 106 UCR-2202 Rev. 0
- 34 UFSAR Effective Date: 04/10/2017 C
U12 Table 02.02-11 Offsite Atmospheric Dispersion Factors C
U12 Table 02.02-12 Control Room Atmospheric Dispersion Factors N
U1 Table 14.01.06-02 (U1)
Parameters Used for the Radiological Consequence Analysis of A Locked Rotor Event D
U1 Table 14.01.06-02 (U1)
Parameters Used for the Control Room Radiological Consequence Analysis of A Locked Rotor Event D
U1 Table 14.02.01-01 (U1)
Parameters Used for the Offsite Radiological Consequence Analysis of a Fuel Handling Accident D
U1 Table 14.02.01-02 (U1)
Parameters Used for the Control Room Radiological Consequence Analysis of a Fuel Handling Accident N
U1 Table 14.02.01-02 (U1)
Parameters Used for the Radiological Consequence Analysis of a Fuel Handling Accident N
U1 Table 14.02.04-01 (U1)
Parameters Used for the Radiological Consequence Analysis of a Steam Generator Tube Rupture D
U1 Table 14.02.04-01 (U1)
Parameters Used for the Offsite Radiological Consequence Analysis of a Steam Generator Tube Rupture D
U1 Table 14.02.04-02 (U1)
Parameters Used for the Control Room Radiological Consequence Analysis of a Steam Generator Tube Rupture D
U1 Table 14.02.05-03 (U1)
Parameters Used for the Offsite Radiological Consequence Analysis of a Main Steam Line Break N
U1 Table 14.02.05-03 (U1)
Parameters Used for the Radiological Consequence Analysis of a Main Steam Line Break D
U1 Table 14.02.05-04 (U1)
Parameters Used for the Control Room Radiological Consequence Analysis of a Main Steam Line Break D
U1 Table 14.02.06-02 (U1)
Parameters Used for the Control Room Radiological Consequence Analysis of a Rod Ejection Accident N
U1 Table 14.02.06-02 (U1)
Parameters Used for the Radiological Consequence Analysis of a Rod Ejection Accident N
U1 Table 14.03.05-01 (U1)
Parameters Used for the Radiological Consequence Analysis of a Loss of Coolant Accident D
U1 Table 14.03.05-01 (U1)
Parameters Used for the Offsite Radiological Consequence Analysis of A Loss Of Coolant Accident D
U1 Table 14.03.05-02 (U1)
Parameters Used for the Control Room Radiological Consequence Analysis of A Loss Of Coolant Accident 14.03.05.03.01 (U1)
Elemental Iodine Spray Removal Coefficients N
U1 Section 14.03.05.03.01 (U1)
Elemental Iodine Spray Removal Coefficients During Injection Phase D
U1 Section 14.03.05.03.02 (U1)
Elemental Iodine Spray Removal Coefficients During Recirculation Phase D
U1 Section 14.03.05.03.03 (U1)
Maximum Elemental Iodine Decontamination Factor C
U1 Section 14.03.05.03.04 (U1)
Particulate Iodine Spray Removal Coefficients C
U1 Section 14.03.05.04 (U1)
Containment Model C
U1 Section 14.03.05.05 (U1)
Engineered Safety Feature Leakage Outside Containment Model C
U1 Section 14.03.05.07 (U1)
Control Room Model C
U1 Section 14.03.05.08 (U1)
Atmospheric Dispersion Factors C
U1 Section 14.03.05.09 (U1)
Radiological Consequence Analysis Results C
U1 Section 14.03.05.10 (U1)
Population Center Considerations C
U1 Section 14.A.01 (U1)
Core Activities C
U1 Section 14.A.02 (U1)
Fuel Handling Activities C
U1 Section 14.A.03 (U1)
Reactor Coolant Activities C
U1 Section 14.A.05 (U1)
Volume Control Tank Activities C
U1 Section 14.A.06 (U1)
Gas Decay Tank Activities C
U1 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 68 of Summary 106 UCR-2202 Rev. 0
- 34 UFSAR Effective Date: 04/10/2017 D
U1 Table 14.A.01-01 (U1)
Core Activities for Offsite Radiological Consequence Analysis N
U1 Table 14.A.01-01 (U1)
Core Activities for Radiological Consequence Analysis D
U1 Table 14.A.01-02 (U1)
Core Activities for Control Room Radiological Consequence Analysis D
U1 Table 14.A.02-01 (U1)
Activities in the Highest Rated Discharged Assembly for Offsite Radiological Consequence Analyses N
U1 Table 14.A.02-01 (U1)
Core Activities in the Highest Discharged Assembly for Radiological Consequence Analysis D
U1 Table 14.A.02-02 (U1)
Activities in the Highest Rated Discharged Assembly for Control Room Radiological Consequence Analyses C
U1 Table 14.A.03-01 (U1)
Parameters Used in the Calculation of Reactor Coolant Fission Product Concentrations D
U1 Table 14.A.03-02 (U1)
Parameters Used in the Calculation of Reactor Coolant Corrosion Product Activities C
U1 Table 14.A.03-03 (U1)
Reactor Coolant Equilibrium Fission Product Concentrations D
U1 Table 14.A.03-04 (U1)
Reactor Coolant Equilibrium Corrosion Product Activities C
U1 Table 14.A.03-05 (U1)
Reactor Coolant Iodine Appearance Rates (Ci/sec)
D U1 Table 14.A.05-01 (U1)
Volume Control Tank Activities N
U1 Table 14.A.05-01A (U1)
Volume Control Tank Activities for Offsite Dose Consequence Analysis N
U1 Table 14.A.05-01B (U1)
Volume Control Tank Activities for Control Room Dose Consequence Analysis N
U1 Table 14.A.05-02 (U1)
Waste Gas Decay Tank Activities (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 69 of Summary 106 UCR-2159 Rev. 0
- 35 UFSAR Effective Date: 04/12/2017 Revise Figure 8.1-2B per Indus EC-55009 Re-Power Visitor Center Transformers Removed from 12-TR12EP-2 now Powered from New Source 12-TR-7 EC-55009 is the INDUS EC associated with one of the RTOs in hardcopy EC-54867 (Re-Powering of Loads from 12-TR12EP-2 to 12-TR-7). EC-55009 will re-power the two Visitors Center 750 Kilovolt Ampere (kVA) transformers (B346-156 and B346-157) currently fed from Transformer 12-TR12EP-2 to Transformer #7 (12-TR-7).
Precursor Event Required:
After RTO of EC-55009 - Re-Power Visitor Center Transformers Removed from 12-TR12EP-2 now Powered from New Source 12-TR-7.
Precursor Completed On: 4/12/2017 Approved:11/18/2016 Affected Unit: Both Units Justification:
EC-55009 - Re-Power Visitor Center Transformers Removed from 12-TR12EP-2 now Powered from New Source 12-TR-7.
(hardcopy) EC-54867 - Re-Powering of Loads from 12-TR12EP-2 to 12-TR-7.
Summary of Change:
NDM Effective Date: 11/23/2016 List of Affected Items:
50.59: SS-SE-2016-0222-00 Comments:
None.
C U12 Figure 08.01-02B Simplified Offsite Power Sources (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 70 of Summary 106 UCR-2205 Rev. 0
- 36 UFSAR Effective Date: 04/25/2017 Replace the last two paragraphs of Section 2.9.3. The new paragraphs will read as follows:
When Seismic Class II or III components (source for seismic interaction) are located within the interaction zone (defined zone where failed components are expected to fall) of a Seismic Class I component (target), the support interface for those Seismic Class II or III components is designed and installed in such a way that there is no potential for the subject Seismic Class II or III components by adversely interacting with them during a design basis seismic event. This has been accomplished by one of the following:
- Physical separation between the source and the target (source is remotely located with regard to the target)
- Barrier between the source and the target
- Evaluated of the effects of the potential impact on the target
- The support interface of the Seismic Class II or III components with Seismic Class I structures shall bedesigned and installed to Seismic Class I criteria, which Approved:4/25/2017 Affected Unit: Both Units Summary of Change:
NDM Effective Date: 4/26/2017 (Sorted by the "UFSAR Effective Date")
Justification:
To provide clarificaion to AR 2016-12455.
requires qualifying the support system for Seismic Class I loads, qualifying the support anchors in accordance with a QCN specification. The other materials required for Seismic II and III supports can be procured to standard grade and installed either to nuclear or standard grade depending upon the quality level assigned to the system or equipment.
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Page 71 of Revision: 28.0 Summary 106 50.71(e) Basis:
The change made in Item #1 provides clarification by rephrasing the current information.
With respect to anchor bolts, the requirement is to use safety related anchor bolts for anchoring the supports to preclude a seismic interaction concern when seismic class II and III components interface with seismic class I structures. the change made in Item #2 clarifies this requirement. Since the safety classification of the anchor bolts required for safety interface applications is clearly defned, "procuring nuclear grade anchors exclusively" is considered as excessive information that can be removed.
Therefor, it is concluded that the above mentioned changes to the UFSAR do not alter the Current Licensing Basis and the subject changes can be performed under 50.71(e) as delineated in Attachement 1 of procedure PMP-2350-SAR-001, Rev. 13, "UFSAR Update Process."
- 36 UCR-2205 Rev. 0 UFSAR Effective Date: 04/25/2017 List of Affected Items:
50.59: N/A Comments:
None.
C U12 Section 02.09.03 Seismic Design Criteria for Seismic Class I and II Piping (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 72 of Summary 106 UCR-2207 Rev. 0
- 37 UFSAR Effective Date: 04/27/2017 Revise existing description in Section 9.8.3.2 of ESW supply to the Emergency Diesel Generators (EDGs). Existing UFSAR Section 9.8.3.2 states that ESW is normally supplied to CTS Heat Exchangers and EDGs only when these systems are in service; however, a subsequent sentence states that ESW can be manually aligned to the EDG to help establish minimum flow requirements. The change moves information for ESW flow to the EDGs to a stand-alone sentence. In addition, a new description is added for aligning ESW flow to EDGs to provide flow for prevention of microbiological growth in ESW supply piping.
Approved:4/27/2017 Affected Unit: Both Units AR 2016-13049 Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 4/28/2017 List of Affected Items:
50.59: SS-SE-2017-0125-00 Comments:
The change clarifies existing information in Section 9.8.3.2 and adds an additional reason for aligning ESW flow to the EDGs.
09.08.03.02 Description - Essential Service Water System C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 73 of Summary 106 UCR-2161 Rev. 0
- 38 UFSAR Effective Date: 05/10/2017 Summary of Change:
Revise Figure 8.1-2B per INDUS EC-54867 Re-Power Guardhouse Transformer removed from 12-TR12EP-2 now Powered from New Source 12-TR-7 EC-54867 is the INDUS EC associated with one of the RTOs in hardcopy EC-54867 -"Re-Powering of Loads from 12-TR12EP-2 to 12-TR-7." EC-54867 will re-power the existing 750 kVA, 4kV-600V indoor Guard House transformer 12-TR-6EPS (B346-365) currently fed from Transformer 12-TR12EP-2 to Transformer #7 (12-TR-7). This EC includes installation of three new 7.2 kV to 2.4kV step-down transformers and a new 3-phase line.
Precursor Event Required:
After RTO of EC-54867 -"Re-Power Guardhouse Transformer removed from 12-TR12EP-2 now Powered from New Source 12-TR-7" (RTO-3)
Precursor Completed On: 5/10/2017 Approved:11/18/2016 Affected Unit: Both Units EC-54867 -"Re-Power Guardhouse Transformer removed from 12-TR12EP-2 now Powered from New Source 12-TR-7" (hardcopy) EC-54867 - Re-Powering of Loads from 12-TR12EP-2 to 12-TR-7 Justification:
NDM Effective Date: 11/23/2016 List of Affected Items:
50.59: SS-SE-2016-0222-00 (None)
Comments:
C U12 Figure 08.01-02B Simplified Offsite Power Sources (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 74 of Summary 106 UCR-2160 Rev. 0
- 39 UFSAR Effective Date: 05/12/2017 Summary of Change:
Revise Figure 8.1-2B per INDUS EC-55010 - Re-Power Single Phase Loads Removed from 12-TR12EP-2 now Powered from New Source 12-TR-7 EC-55010 is the INDUS EC associated with one of the RTOs in hardcopy EC-54867 (Re-Powering of Loads from 12-TR12EP-2 to 12-TR-7).
EC-55010 will re-power the single phase loads for roadway lighting, Post 1, and Sewage Treatment Building (25kVA transformers B346-138 and B346-139, 75kVA transformer B346-137 and 50kVA transformer B346-154) currently fed from Transformer 12-TR12EP-2 to Transformer #7 (12-TR-7).
Precursor Event Required:
After RTO of EC-55010 - Re-Power Single Phase Loads Removed from 12-TR12EP-2 now Powered from New Source 12-TR-7 Precursor Completed On: 5/12/2017 Approved:11/18/2016 Affected Unit: Both Units EC-55010 - Re-Power Single Phase Loads Removed from 12-TR12EP-2 now Powered from New Source 12-TR-7 (hardcopy) EC-54867 - Re-Powering of Loads from 12-TR12EP-2 to 12-TR-7 Justification:
NDM Effective Date: 11/23/2016 List of Affected Items:
50.59: SS-SE-2016-0222-00 (None)
Comments:
C U12 Figure 08.01-02B Simplified Offsite Power Sources (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 75 of Summary 106 UCR-2170 Rev. 0
EC-55173 is the INDUS EC associated with one of the RTOs in hardcopy EC-54649 (Replacement of Emergency Power (EP) Transformers 12-TR12EP-1 and 12-TR12EP-2).
EC-55173 will remove the existing 69kV switchyard Emergency Power Bus #1 and Bus #2 crosstie disconnect switch 12-152-TR12EP as well as removing the transformer 12-TR12EP-2 connection to the 69kV bus (including line side disconnect switch 12-152-X2).
Precursor Event Required:
After RTO of EC-55173 - "Replacement of EP Transformers 12-TR12EP-1 &
12-TR12EP-2; RTO #1: Remove Crosstie Disconnect Switch 12-152-TR12EP" Hard Copy Mod: EC-54649, Rev. 0 Precursor Completed On: 5/25/2017 Approved:3/2/2017 Affected Unit: Both Units EC-55173 - Replacement of EP Transformers 12-TR12EP-1 & 12-TR12EP-2; RTO #1:
Remove Crosstie Disc Switch 12-152-TR12EP Hard Copy Mod: EC-54649, Rev. 0 - Replacement of EP Transformers 12-TR12EP-1 &
12-TR12EP-2; RTO #3: Remove Old EP-1 and Install New Spare EP-2 in its Place Summary of Change:
Justification:
NDM Effective Date: 3/24/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 76 of Summary 106 UCR-2170 Rev. 0
- 40 UFSAR Effective Date: 05/25/2017 List of Affected Items:
50.59: SS-SE-2016-0372-00 Comments:
None.
C U12 Figure 08.01-01A Main Auxiliary One-Line Diagram Bus 'A' & 'B' Engineered Safety System C
U12 Figure 08.01-02B Simplified Offsite Power Sources (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 77 of Summary 106 UCR-2208 Rev. 0
- 41 UFSAR Effective Date: 05/30/2017 Revise Table 1.0-1 to include the full revision date for Revision 26.0 and add a row to include the full revision date for Revision 27.0.
Document that the correct entry for Table 6.3.2 revised by UCR-2059 should be:
"2 (1-HE-18E / W) / 2 (2-HE-18 E / W)"
The last "/ W" was dropped during the UCR package preparation. The original markup provided by the Responsible Engineer and the final package markup are attached for review.
Approved:5/30/2017 Affected Unit: Both Units PMP-2350-SAR-001 "UFSAR Update Process" UCR-2059 - Update Table 6.3-2 per EC-51370 - "2-HE-18W, Unit 2 West Containment Spray Heat Exchanger Replacement" UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
This UFSAR change does not alter the Current License Basis.
The Table 1.0-1 change meets the 50.71(e) evaluation criteria presented in PMP-2350-SAR-001, Attachment 1, Item 4, Bullet 5:
- 4. Editorial Changes consisting of the following:
Addition of general references.
The Table 6.3-2 change meets the 50.71(e) evaluation criteria presented in PMP-2350-Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 6/1/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 78 of Summary 106 UCR-2208 Rev. 0
- 41 UFSAR Effective Date: 05/30/2017 SAR-001, Attachment 1, Item 4, bullet 4:
- 4. Editorial Changes consisting of the following:
Typographical Changes that are corrections of errors introduced during the production (typing/printing/magnetic storage) of the document List of Affected Items:
50.59: N/A C
U12 Table 01.00-01 UFSAR Revision History C
U12 Table 06.03-02 Containment Spray Heat Exchanger Design Parameters (Sorted by the "UFSAR Effective Date")
Comments:
None.
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 79 of Summary 106 UCR-2210 Rev. 0
- 42 UFSAR Effective Date: 06/05/2017 Summary of Change:
Replace Unit 2 generator Megawatt electric (MWe) gross and MWe net predicated output values with values based on actual test results.
Revise Unit 2 generator MWe gross and MWe net after testing of EC-52519, Unit 2 High Pressure Turbine and Low Pressure Turbines Retrofit with Alstom Turbines. UCR-2122, Rev. 0 made changes to the UFSAR based on predicted Unit 2 main generator MWe output.
This UCR updates those predicted values based on actual test results.
Approved:6/5/2017 Affected Unit: Unit Two Justification:
EC-52519 special test procedure EC-52519-TP-001 (generation of 1255 MWe gross minus 35 MWe station load equals 1220 Mwe net).
NDM Effective Date: 6/6/2017 List of Affected Items:
50.59: SS-SE-2014-0043-00 Comments:
None.
01.00 Introduction and Summary C
U12 Section 10.01 General Descriptions C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 80 of Summary 106 UCR-2168 Rev. 1
- 43 UFSAR Effective Date: 06/14/2017 Summary of Change:
Table 9.8.3 - revise NESW pump Casing material from:
"cast iron or cast steel" [Done per UCR-2168, Rev. 0]
to be "Cast Steel" Table 9.8.3 - change Non-Essential Service Water Strainers from:
Duplex-automatic backwashing to Automatic - Self Cleaning [Done per UCR-2168, Rev. 0]
Figures 9.8-4 and 9.8-5, change NESW Unit 1 and 2, to depict single stage (vs duplex) strainer, and change label from Duplex Strainer or Automatic Strainer" [Done per UCR-2168, Rev. 0]
to Precursor Event Required:
After the RTO of the last of the INDUS ECs for the "NESW Pumps and Discharge Strainers" modifications which is: EC-54254 (Unit 1 South)
Precursor Completed On: 6/14/2017 Approved:5/31/2017 Affected Unit: Both Units NDM Effective Date: 6/2/2017 (Sorted by the "UFSAR Effective Date")
Automatic Strainer
[Done per UCR-2168, Rev. 0]
[Done per UCR-2168, Rev. 0]
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 81 of Summary 106 UCR-2168 Rev. 1
- 43 UFSAR Effective Date: 06/14/2017 Table 9.8-4, add footnote number 5 next to Minimum Design Flow Total of 9,564.
"5 Actual operational data with cold NESW temperatures indicated nominal minimal flow is approximately 8000 gpm."
Change Section 9.8.3.2 "Description" - "Non-Essential Service Water System" second paragraph, third sentence, from:
All motor-operated valves on the main non-essential water system are operated from the station battery system with the exception of the inlet and outlet valve motor operators for each of the main NESW duplex strainers, or the motor operated strainer discharge motor operators for each of the NESW automatic strainers, which are powered from the 600 VAC auxiliary bus.
[Done per UCR-2168, Rev. 0]
Revise to state the following:
All motor-operated valves on the main non-essential water system are operated from the station battery system with the exception of the motor operated strainer discharge motor operators for each of the NESW automatic strainers, which are powered from the 600 VAC auxiliary bus.
Change Section 9.8.3.2 "Description" - "Non-Essential Service Water System" third paragraph, from:
The discharge strainers of the pumps are of duplex construction, with automatic backwashing, or a single basket automatic strainer. Each duplex strainer is effectively two strainers in one casing with flow directed through one half, while slide gates block off the other half. When the strainer is in service and if it becomes dirty or clogged, a high differential pressure signal initiates a shift of the slide gates blocking the flow to the dirty basket and directing it through the clean basket. The dirty basket is then backwashed and is ready for re-use. The automatic discharge strainers of the pumps are provided with automatic backwashing. Each strainer contains an internal arm that cleans a segment of the strainer during backwash cycles.
(Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 82 of Summary 106 UCR-2168 Rev. 1
- 43 UFSAR Effective Date: 06/14/2017 When the strainer is in service and if it becomes dirty or clogged, a high differential pressure signal initiates rotation of the internal arm that directs a small portion of the NESW flow into reverse flow across a basket segment, and directs the reverse flow and debris to a drain, maintaining the strainer in service during the backwash cycle.
[Done per UCR-2168, Rev. 0]
to read as follows (delete sentences 1,2,3 & 4):
The automatic discharge strainers of the pumps are provided with automatic backwashing.
Each strainer contains an internal arm that cleans a segment of the strainer during backwash cycles. When the strainer is in service and if it becomes dirty or clogged, a high differential pressure signal initiates rotation of the internal arm that directs a small portion of the NESW flow into reverse flow across a basket segment, and directs the reverse flow and debris to a drain, maintaining the strainer in service during the backwash cycle.
EC-0000054253 "NESW Pump and Strainer Replacement" Justification:
List of Affected Items:
50.59: SS-SE-2016-0190-01 This is the second of two sequential UCRs associated with the EC-54253 modification of the NESW pumps and discharge strainers. This second UCR modifies the UFSAR to reflect that all strainers are now single basket automatic style (following the last train RTO EC-54254), eliminating all reference to the duplex strainers.
Comments:
C U1 Figure 09.08-04 Non-Essential Service Water Unit No. 1 C
U2 Figure 09.08-05 Non Essential Service Water Unit No. 2 C
U12 Table 09.08-03 Service Water Systems Components Design Data C
U12 Table 09.08-04 Non-Essential Service Water Requirements 09.08.03.02 Description - Non-Essential Service Water System C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 83 of Summary 106 UCR-2201 Rev. 1
- 44 UFSAR Effective Date: 06/22/2017 UCR-2201, Rev. 1 is an Administrative UCR to fix or enhance items found while revising the UFSAR to Revision 28.0; this one focuses on Chapter 9, Sections 9.0, 9.1, &
9.2. UCR-2201, Rev. 1 supercedes UCR-2201, Rev. 0 in its entirety. (UCR-2201, Rev. 0 didn't list the impacts in Section 9.0 and Section 9.2.1).
Sections 9.0, 9.1, & 9.2 Added bullet points to Auxiliary and Emergency Systems Summary section (9.0.1)
Section 9.2.2.5 - "Makeup" Convert the four numbered items in the second paragraph into a numbered list Section 9.2.2.7 - "Boration" Convert the three numbered items in the first paragraph into a numbered list Section 9.2.2.9 - "Charging Pump Control" Convert the
- footnote into a numbered footnote 1 Section 9.2.2.19 - "Charging Pumps" Convert the two
- footnotes into a numbered footnote 2 Section 9.2.3.5 - "Malfunction Analysis" Convert the two lettered items in the sixth paragraph into a lettered list Approved:6/22/2017 Affected Unit: Both Units Summary of Change:
NDM Effective Date: 6/23/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 84 of Summary 106 UCR-2201 Rev. 1
- 44 UFSAR Effective Date: 06/22/2017 Section 9.2.3.6 - "Galvanic Corrosion" Convert the superscripted (1) found in the first paragraph, second sentence next from:
"have been shown(1)"
to "have been shown (Ref. 1)"
- the superscripted (1) could be interpreted as a footnote marker and there is not a footnote associated with this Section; this was meant to refer the reader to Reference 1 in Section 9.2.3.1 "References for Section 9.2."
PMP-2350-SAR-001 "UFSAR Update Process" PMP-2350-SAR-001, Attachment 1 "Acceptable Changes to the Updated Final Safety Analysis Report (UFSAR) Controlled via 50.71(e) Evaluation" UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
This UFSAR change does not alter the Current License Basis.
Adding numbers to titled, but unnumbered sections, changing
- footnotes into numbered footnotes, changing numbered or lettered lists in sentances into numbered or lettered lists, and changing reference number from a footnote appearance into a clear reference number sections:
- 4. Editorial Changes consisting of the following:
Standardization Changes made to establish a consistent means of presenting Justification:
50.71(e) Basis:
(Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 85 of Summary 106 UCR-2201 Rev. 1
- 44 UFSAR Effective Date: 06/22/2017 List of Affected Items:
50.59: N/A Comments:
None.
09.00.01 Summary - Chemical and Volume Control D
U12 Section 09.00.01 Summary - Residual Heat Removal System D
U12 Section 09.00.01 Summary - Spent Fuel Pool Cooling System D
U12 Section 09.00.01 Summary - Component Cooling System D
U12 Section 09.00.01 Summary - Sampling Systems D
U12 Section 09.00.01 Summary - Fuel Handling System D
U12 Section 09.00.01 Summary - Facility Service Systems D
U12 Section 09.02.01 Design Bases - Codes and Classifications D
U12 Section 09.02.01.01 Codes and Classifications N
U12 Section 09.02.02 System Design and Operation - Evaporator Feed Ion Exchangers D
U12 Section 09.02.02 System Design and Operation - Boric Acid Evaporator Feed Pumps D
U12 Section 09.02.02 System Design and Operation - Holdup Tank Recirculation Pump D
U12 Section 09.02.02 System Design and Operation - Boric Acid Reserve Tank D
U12 Section 09.02.02 System Design and Operation - Holdup Tanks D
U12 Section 09.02.02 System Design and Operation - Boric Acid Blender D
U12 Section 09.02.02 System Design and Operation - Boric Acid Transfer Pumps D
U12 Section 09.02.02 System Design and Operation - Boric Acid Tank Heaters D
U12 Section 09.02.02 System Design and Operation - Ion Exchanger Filters D
U12 Section 09.02.02 System Design and Operation - Boric Acid Tanks D
U12 Section 09.02.02 System Design and Operation - Deborating Demineralizers D
U12 Section 09.02.02 System Design and Operation - Batching Tank D
U12 Section 09.02.02 System Design and Operation - Boric Acid Evaporators D
U12 Section 09.02.02 System Design and Operation - Evaporator Condensate Demineralizers D
U12 Section 09.02.02 System Design and Operation - Condensate Filter D
U12 Section 09.02.02 System Design and Operation - Boric Acid Filter D
U12 Section 09.02.02 System Design and Operation - Monitor Tank Pumps D
U12 Section 09.02.02 System Design and Operation - Seal Water Heat Exchanger D
U12 Section 09.02.02 System Design and Operation - Concentrates Filter D
U12 Section 09.02.02 System Design and Operation - Concentrates Holding Tank D
U12 Section 09.02.02 System Design and Operation - Concentrates Holding Tank Transfer Pumps D
U12 Section 09.02.02 System Design and Operation - Electrical Heat Tracing D
U12 Section 09.02.02 System Design and Operation - Valves D
U12 Section 09.02.02 System Design and Operation - Piping D
U12 Section 09.02.02 System Design and Operation - Monitor Tanks D
U12 Section 09.02.02 System Design and Operation - Centrifugal Charging Pumps D
U12 Section 09.02.02 System Design and Operation C
U12 Section 09.02.02 System Design and Operation - Boration D
U12 Section 09.02.02 System Design and Operation - System Description D
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 86 of Summary 106 UCR-2201 Rev. 1
- 44 UFSAR Effective Date: 06/22/2017 09.02.02 System Design and Operation - Expected Operating Conditions D
U12 Section 09.02.02 System Design and Operation - Reactor Coolant Activity Concentration D
U12 Section 09.02.02 System Design and Operation - Reactor Makeup Control Modes D
U12 Section 09.02.02 System Design and Operation - Makeup D
U12 Section 09.02.02 System Design and Operation - Dilution D
U12 Section 09.02.02 System Design and Operation - Seal Water Injection Filters D
U12 Section 09.02.02 System Design and Operation - Charging Pump Control D
U12 Section 09.02.02 System Design and Operation - Components D
U12 Section 09.02.02 System Design and Operation - Charging Pumps D
U12 Section 09.02.02 System Design and Operation - Seal Water Filter D
U12 Section 09.02.02 System Design and Operation - Alarm Functions D
U12 Section 09.02.02 System Design and Operation - Chemical Mixing Tank D
U12 Section 09.02.02 System Design and Operation - Regenerative Heat Exchanger D
U12 Section 09.02.02 System Design and Operation - Volume Control Tank D
U12 Section 09.02.02 System Design and Operation - Reactor Coolant Filter D
U12 Section 09.02.02 System Design and Operation - Cation Bed Demineralizer D
U12 Section 09.02.02 System Design and Operation - Mixed Bed Demineralizers D
U12 Section 09.02.02 System Design and Operation - Letdown Heat Exchanger D
U12 Section 09.02.02 System Design and Operation - Letdown Orfices D
U12 Section 09.02.02 System Design and Operation - Excess Letdown Heat Exchanger D
U12 Section 09.02.02.01
System Description
N U12 Section 09.02.02.02 Expected Operating Conditions N
U12 Section 09.02.02.03 Reactor Coolant Activity Concentration N
U12 Section 09.02.02.04 Reactor Makeup Control Modes N
U12 Section 09.02.02.05 Makeup F
U12 Section 09.02.02.05 Makeup N
U12 Section 09.02.02.06 Dilution N
U12 Section 09.02.02.07 Boration F
U12 Section 09.02.02.07 Boration N
U12 Section 09.02.02.08 Alarm Functions N
U12 Section 09.02.02.09 Charging Pump Control F
U12 Section 09.02.02.09 Charging Pump Control N
U12 Section 09.02.02.10 Centrifugal Charging Pumps N
U12 Section 09.02.02.11 Components N
U12 Section 09.02.02.12 Regenerative Heat Exchanger N
U12 Section 09.02.02.13 Letdown Orifices N
U12 Section 09.02.02.14 Letdown Heat Exchanger N
U12 Section 09.02.02.15 Mixed Bed Demineralizers N
U12 Section 09.02.02.16 Cation Bed Demineralizer N
U12 Section 09.02.02.17 Reactor Coolant Filter N
U12 Section 09.02.02.18 Volume Control Tank N
U12 Section 09.02.02.19 Charging Pumps F
U12 Section 09.02.02.19 Charging Pumps N
U12 Section 09.02.02.20 Chemical Mixing Tank N
U12 Section 09.02.02.21 Excess Letdown Heat Exchanger N
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 87 of Summary 106 UCR-2201 Rev. 1
- 44 UFSAR Effective Date: 06/22/2017 09.02.02.22 Seal Water Heat Exchanger N
U12 Section 09.02.02.23 Seal Water Filter N
U12 Section 09.02.02.24 Seal Water Injection Filters N
U12 Section 09.02.02.25 Boric Acid Filter N
U12 Section 09.02.02.26 Boric Acid Tanks N
U12 Section 09.02.02.27 Batching Tank N
U12 Section 09.02.02.28 Boric Acid Tank Heaters N
U12 Section 09.02.02.29 Boric Acid Transfer Pumps N
U12 Section 09.02.02.30 Boric Acid Blender N
U12 Section 09.02.02.31 Holdup Tanks N
U12 Section 09.02.02.32 Boric Acid Reserve Tank N
U12 Section 09.02.02.33 Holdup Tank Recirculation Pump N
U12 Section 09.02.02.34 Boric Acid Evaporator Feed Pumps N
U12 Section 09.02.02.35 Evaporator Feed Ion Exchangers N
U12 Section 09.02.02.36 Ion Exchanger Filters N
U12 Section 09.02.02.37 Boric Acid Evaporators N
U12 Section 09.02.02.38 Evaporator Condensate Demineralizers N
U12 Section 09.02.02.39 Condensate Filter N
U12 Section 09.02.02.40 Monitor Tanks N
U12 Section 09.02.02.41 Monitor Tank Pumps N
U12 Section 09.02.02.42 Deborating Demineralizers N
U12 Section 09.02.02.43 Concentrates Filter N
U12 Section 09.02.02.44 Concentrates Holding Tank N
U12 Section 09.02.02.45 Concentrates Holding Tank Transfer Pumps N
U12 Section 09.02.02.46 Electrical Heat Tracing N
U12 Section 09.02.02.47 Valves N
U12 Section 09.02.02.48 Piping N
U12 Section 09.02.03 System Design Evaluation - Availability and Reliability D
U12 Section 09.02.03 System Design Evaluation - Tests and Inspections D
U12 Section 09.02.03 System Design Evaluation - Galvanic Corrosion D
U12 Section 09.02.03 System Design Evaluation - Malfunction Analysis D
U12 Section 09.02.03 System Design Evaluation - Incident Control D
U12 Section 09.02.03 System Design Evaluation - Leakage Provisions D
U12 Section 09.02.03 System Design Evaluation - Control of Tritium D
U12 Section 09.02.03.01 Availability and Reliability N
U12 Section 09.02.03.01 References for Section 9.2 D
U12 Section 09.02.03.02 Control of Tritium N
U12 Section 09.02.03.03 Leakage Provisions N
U12 Section 09.02.03.04 Incident Control N
U12 Section 09.02.03.05 Malfunction Analysis N
U12 Section 09.02.03.05 Malfunction Analysis F
U12 Section 09.02.03.06 Galvanic Corrosion N
U12 Section 09.02.03.06 Galvanic Corrosion F
U12 Section 09.02.03.07 Tests and Inspections N
U12 Section 09.02.04 References for Section 9.2 N
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 88 of Summary 106 UCR-2185 Rev. 0
- 45 UFSAR Effective Date: 07/25/2017 Summary of Change:
Reformat the document properties for the List of Affected Figures below to meet NRC minimum standards for electronic UFSAR submission.
This is an administrative UCR, a 50.71(e) based change, which only reformats the document properties for the figures listed below, no technical changes were made.
Approved:7/25/2017 Affected Unit: Unit Two PMP-2350-SAR-001 "UFSAR Update Process" - Attachment 1 " Acceptable Changes to the Updated Final Safety Analysis Report (UFSAR) Controlled via 50.71(e) Evaluation.
UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
The changes made by this UCR meet the requirements of PMP-2350-SAR-001, Attachment 1, Item 4, bullet 1:
- 4. Editorial Changes consisting of the following:
Format Changes made to improve layout or appearance.
This UFSAR change does not alter the Current License Basis.
Justification:
50.71(e) Basis:
NDM Effective Date: 7/26/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 89 of Summary 106 UCR-2185 Rev. 0
- 45 UFSAR Effective Date: 07/25/2017 List of Affected Items:
50.59: N/A Comments:
None.
F U2 Figure 14.03.04-013 (U2)
Worst Break Lower Compartment Temperature Comparison F
U2 Figure 14.03.04-014 (U2)
Upper Compartment Temperature (30% Power Level)
F U2 Figure 14.03.04-015 (U2)
Lower Compartment Pressure (30% Power Level)
F U2 Figure 14.03.04-016 (U2)
Lower Compartment Temperature (30% Power Level)
F U2 Figure 14.03.04-017 (U2)
Worst Break - Lower Compartment Temperature Comparison Generic Analysis F
U2 Figure 14.03.04-018 (U2)
Cook Unit 1 & 2 - TMD Model Plan at Equipment Rooms Elevation F
U2 Figure 14.03.04-019 (U2)
Cook Unit 1 & 2 - TMD Model Containment Section View F
U2 Figure 14.03.04-020 (U2)
Cook Unit 1 & 2 - Plan View at Ice Condenser Elevation Ice Condenser Compartments F
U2 Figure 14.03.04-021 (U2)
Cook Unit 1 & 2 - TMD Model Layout of Containment Shell F
U2 Figure 14.03.04-022 (U2)
Steam Generator Enclosure Above Elevation 665 ft.
(Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 90 of Summary 106 UCR-2186 Rev. 0
- 46 UFSAR Effective Date: 07/25/2017 Summary of Change:
Reformat the document properties for the List of Affected Figures below to meet NRC minimum standards for electronic UFSAR submission.
This is an administrative UCR, a 50.71(e) based change, which only reformats the document properties for the figures listed below, no technical changes were made.
Figure 14.3.4-24 (Sheet 1 of 2) is titled "Steam Generator Enclosure Cut-Open View of the Steam Generator Enclosure" and Figure 14.3.4-24 (Sheet 2 of 2) is titled "Schematic of the Cook Unit 1 & 2 9 Node TMD Steam Generator Enclosure Model" are being seperated.
Sheet 2 of 2 for Figure 14.3.4-24 is now Sheet 1 of 2 in Figure 14.3.4-25 also titled "Schematic of the Cook Unit 1 & 2 9 Node TMD Steam Generator Enclosure Model".
Therefore, Figure 24 is now 1 page and Figure 25 has 2 pages.
Approved:7/25/2017 Affected Unit: Unit Two PMP-2350-SAR-001 "UFSAR Update Process" - Attachment 1 " Acceptable Changes to the Updated Final Safety Analysis Report (UFSAR) Controlled via 50.71(e) Evaluation.
UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
The changes made by this UCR meet the requirements of PMP-2350-SAR-001, Attachment 1, Item 4, bullet 1:
- 4. Editorial Changes consisting of the following:
Format Changes made to improve layout or appearance.
This UFSAR change does not alter the Current License Basis.
Justification:
50.71(e) Basis:
NDM Effective Date: 7/26/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 91 of Summary 106 UCR-2186 Rev. 0
- 46 UFSAR Effective Date: 07/25/2017 List of Affected Items:
50.59: N/A Comments:
None.
F U2 Figure 14.03.04-023 (U2)
Steam Generator Enclosure Below Elevation 665 ft.
F U2 Figure 14.03.04-024 (U2)
Steam Generator Enclosure - Cut-Open View of the Steam Generator Enclosure F
U2 Figure 14.03.04-025 (U2)
Schematic of the Cook Unit 1 & 2 9 Node TMD Steam Generator Enclosure Model F
U2 Figure 14.03.04-026 (U2)
Cook Unit 1 & 2 - Pressurizer Enclosure TMD Model Noding Network F
U2 Figure 14.03.04-027 (U2)
Cook Unit 1 & 2 - TMD Model with Pressurizer Enclosure Nodalization Network F
U2 Figure 14.03.04-028 (U2)
Cook Unit 1 & 2 - Fan / Accumulator Room TMD Model Plan at Equipment Rooms Elevation F
U2 Figure 14.03.04-029 (U2)
Cook Unit 1 & 2 - Fan / Accumulator Room TMD Model Layout of Containment Shell F
U2 Figure 14.03.04-030 (U2)
Cook Unit 1 & 2 - TMD Model with Fan / Accumulator Room Nodalization Network F
U2 Figure 14.03.04-031 (U2)
Cook Unit 1 & 2 - Loop Subcompartment TMD Model Nodalization Network (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 92 of Summary 106 UCR-2206 Rev. 0
- 47 UFSAR Effective Date: 08/17/2017 Added section numbers to two titled, but previously unnumbered sections which fall under changes permitted via a 50.71(e) evaluation, resulted in two "new" numbered sections.
Added new Section 14.1.10.2.2.3 (U1) "Evaluation to Support 240% Nominal Main Feedwater Flow."
Added new Section 14.1.10.2.2.2.1 (U2) "Evaluation to Support 240% Nominal Main Feedwater Flow."
Approved:8/17/2017 Affected Unit: Both Units Westinghouse Letter AEP-17-5, AMERICAN ELECTRIC POWER DONALD C. COOK UNITS 1 AND 2 Feedwater Malfunction Evaluation to Support an Increased Feedwater Flow Rate (240% of Nominal)
UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
This UFSAR change does not alter the Current License Basis.
Adding section numbers to titled, but previously unnumbered sections falls under changes permitted via a 50.71(e) evaluation as discussed in PMP-2350-SAR-001, Attachment 1, Item 4, bullet 1:
- 4. Editorial Changes consisting of the following:
Format Changes made to improve layout or appearance Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 8/21/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 93 of Summary 106 UCR-2206 Rev. 0
- 47 UFSAR Effective Date: 08/17/2017 List of Affected Items:
50.59: SS-SE-2017-0120-00 Comments:
No changes to the UFSAR are required except those outlined.
14.01.10.02.02 (U1)
Results - Effect of the Return to RCS NOP/NOT Program on Unit 1 D
U1 Section 14.01.10.02.02 (U1)
Results - Effect of the MUR Program on Unit 1 D
U1 Section 14.01.10.02.02.01 (U1)
Effect of the MUR Program on Unit 1 N
U1 Section 14.01.10.02.02.02 (U1)
Effect of the Return to RCS NOP/NOT Program on Unit 1 N
U1 Section 14.01.10.02.02.02.01 (U2)
Evaluation to Support 240% Nominal Main Feedwater Flow N
U2 Section 14.01.10.02.02.03 (U1)
Evaluation to Support 240% Nominal Main Feedwater Flow N
U1 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 94 of Summary 106 UCR-2213 Rev. 0
- 48 UFSAR Effective Date: 08/29/2017 Replace Unit 2 Table 14.3.1-6, Peak Cladding Temperature (PCT) Including All Penalties and Benefits, Best-Estimate Large Break LOCA (BE LOCA) for CNP Unit 2, in its entirety with new Table 14.3.1-6, Peak Cladding Temperature, including all Penalties and Benefits, Best-Estimate Large Break LOCA (BE LOCA) for D. C. Cook Unit 2.
This change updates the UFSAR for identified PCT penalties and benefits made to the large break LOCA analysis, and incorporates the penalty from stainless steel filler rods used during Unit 2 Cycle 23.
Note that the source document also transmitted a Unit 1 PCT rackup sheet but no changes were necessary to the UFSAR table for Unit 1.
Approved:8/29/2017 Affected Unit: Unit Two Westinghouse Letter LTR-LIS-17-31, D. C. Cook Units 1 and 2 10 CFR 50.46 Annual Notification and Reporting for 2016, dated February 8 2017.
EC-4548, Rev. 0, Unit 2 Cycle 23 Core Reload, RTO was on 1/4/2017.
N/A Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 8/29/2017 List of Affected Items:
50.59: SS-SE-2017-0236-00 Routine update to the Unit 2 LBLOCA PCT rack-up table in the UFSAR. This UCR includes impacts from Stainless Steel filler rods installed in U2C23 in the Unit 2 rack-ups in UFSAR Table 14.3.1-6.
Comments:
C U2 Table 14.03.01-06 (U2)
Peak Cladding Temperature Including All Penalties and Benefits, Best-Estimate Large Break LOCA (BE LOCA) for D. C. Cook Unit 2 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 95 of Summary 106 UCR-2166 Rev. 0
- 49 UFSAR Effective Date: 09/01/2017 Revise Section 8.2 and Figures 8.1-1A and -2A per EC-54575 - Installation of 345KV Breaker 12-52-J1.
EC-54575 will replace the existing motorized disconnect J1S1 with a double ended disconnect and install new Breaker J1 on the primary side of Transformer #5 in between the new double ended disconnect J1S1 and TR5.
Precursor Event Required:
After RTO of EC-54575 - Installation of 345KV Breaker 12-52-J1 Precursor Completed On: 9/1/2017 Approved:8/29/2017 Affected Unit: Both Units EC-54575 - Installation of 345KV Breaker 12-52-J1 Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 8/30/2017 List of Affected Items:
50.59: SS-SE-2016-0318-00 Comments:
None.
C U12 Figure 08.01-01A Main Auxiliary One-Line Diagram Bus 'A' & 'B' Engineered Safety System C
U12 Figure 08.01-02A Simplified Offsite Power Sources 08.02 Network Interconnections C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Page 96 of Summary 106 UCR-2202 Rev. 1
- 50 UFSAR Effective Date: 10/12/2017 UCR-2202, Rev. 1 is an Administrative UCR to fix or enhance items found while revising the UFSAR to Revision 28.0; this one focuses on Chapter 11, Section 11.2.1.1.4. UCR-2202, Rev. 1 supersedes the change found in UCR-2162, Rev. 0 and updates with information from UCR-2202, Rev. 0.
Section 11.2.1.1.4 - Accident Shield in UCR-2162, Rev. 0 added this change:
The main purpose of the accident shield is to achieve the following after a maximum design accident:
- a. prevent off-site radiation exposure in excess of 10 CFR 100;
- b. to limit exposure to control room operators.
UCR-2202, Rev. 0 was the control room habitability and offsite radiological dose consequence analysis project using the AST methodology outlined in RG 1.183. The changes in UCR-2202, Rev. 0 updated and replaced references to 10 CFR 100 with RG 1.183 and 10 CFR 50.67. While revising the UFSAR to Revision 28.0, the changes made by UCR-2162 were noticed and not updated.
UCR-2202, Rev. 1 will update Section 11.2.1.1.4 to update 10 CFR 100 with Regulatory Guide 1.183 and 10 CFR 50.67. All other changes from UCR-2202, Rev. 0 will remain the same.
Approved:10/12/2017 Affected Unit: Both Units PMP-2350-SAR-001 "UFSAR Update Process" PMP-2350-SAR-001, Attachment 1 "Acceptable Changes to the Updated Final Safety Analysis Report (UFSAR) Controlled via 50.71(e) Evaluation" UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 10/15/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Summary Page 97 of 106 UCR-2202 Rev. 1
- 50 UFSAR Effective Date: 10/12/2017 This UCR change is permitted through SAR-001, Attachment 1, Step 8 - Correction of UFSAR information to make it consistent with current licensing basis.
This UFSAR change does not alter the Current License Basis.
List of Affected Items:
50.59: N/A Comments:
None.
11.02.01.01.04 Accident Shielding C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Summary Page 98 of 106 UCR-2214 Rev. 0
- 51 UFSAR Effective Date: 10/20/2017 Revise selective sections of Chapter 6, Chapter 9, and Chapter 11 as required by EC-55737 Change ultimate heat sink design temperature from 88.8°F to 86.9°F.
Precursor Event Required:
After RTO of EC-55737 Change ultimate heat sink design temperature from 88.8°F to 86.9°F Precursor Completed On: 10/20/2017 Approved:10/2/2017 Affected Unit: Both Units EC-55737 - Change ultimate heat sink design temperature from 88.8°F to 86.9°F N/A Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 10/4/2017 List of Affected Items:
50.59: SS-SE-2017-0240-00 Although the design bases value was changed to 86.9°F, some analysis is still based on the bounding input value of 88.8°F. UFSAR sections that describe the CNP design bases are revised. UFSAR sections that describe input values to analysis do not require revision.
Comments:
C U12 Table 06.03-02 Containment Spray Heat Exchanger Design Parameters C
U12 Table 09.05-03 Component Cooling System Component Design Data C
U12 Table 11.05-02 Blowdown Treatment System Components 06.02.02 System Design and Operation - Components - Heat Exchangers C
U12 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Summary Page 99 of 106 UCR-2220 Rev. 0
- 52 UFSAR Effective Date: 10/30/2017 This is an Administrative change to Figures 01.03-01A and 02.01-04B to remove the TSOC building and add the NEST and FLEX buildings.
These changes were missed in UCR-2130 (EC-54252 - Electrical and Fire Water Supply for the Nuclear Engineering Services and Technology (NEST) Building.") and UCR-2081 (EC-53178 - Unit 1 & 2 Design & Construct Flex Equipment Storage Facility"). The updates are as follows:
- 1. Figure 1.3-1A: Removal of Technical Support Office Complex (TSOC) Building
- 2. Figure 2.1-4B: Add the FLEX building and the Nuclear Engineering Services &
Technology (NEST) Building Approved:10/30/2017 Affected Unit: Both Units
- 1. AR 2017-10381
- 2. EC-54252 - Electrical and Fire Water Supply for the Nuclear Engineering Services and Technology (NEST) Building."
- 3. EC-53178 - Unit 1 & 2 Design & Construct Flex Equipment Storage Facility" UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review. The changes are classified as clarification changes and as removal of obsolete information.
These proposed changes improve the information provided in the associated site layout maps. The meaning or intent of the SSCs is not affected by the subject changes. The changes to Figures 1.3-1A and 2.1-4B also meet the 50.71(e) evaluation criteria presented in PMP-2350-SAR-001, Attachment 1. This UFSAR change does not alter or have any impact on the Current Licensing Basis.
Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 11/2/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT UCR-2220 Rev. 0
- 52 UFSAR Effective Date: 10/30/2017 List of Affected Items:
50.59: N/A This is an administrative update to the FSAR only. EC-53178 and EC-54252 already provided updates to the UFSAR as a result of the modifications (per UCR-2081 and UCR-2130). However, the removal of the TSOC Building on Figure 1.3-1A and addition of the FLEX Building and NEST Building on Figure 2.1-4B were missed. This UCR incorporates those missing drawing updates.
Comments:
C U12 Figure 01.03-01A Key Plan C
U12 Figure 02.01-04B Donald C. Cook Nuclear Plant Topographic Map (Sorted by the "UFSAR Effective Date")
Revision: 28.0 Summary Page 100 of 106
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Summary Page 101 of 106 UCR-2222 Rev. 0
- 53 UFSAR Effective Date: 11/28/2017 This is an Administrative change to Section 14.3.1.8 - References for Section 14.3.1.
This UCR is to include the missing information in References 6 through 12.
These references are missing or incomplete from the current UFSAR. These references were inadvertently missed in the UFSAR Revision 23.0 update (and subsequent revisions thereafter), which was updated by UCR-1974 (EC-50387 - Implementation of Unit 2 Best-Estimate Large Break LOCA Analysis). The references to be fixed and added back in are as follows:
- 6. WCAP-16009-P-A, "Realistic Large Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)," (Westinghouse Proprietary), in conjunction with Licensing Amendment Request "WCOBRA/TRAC Validation with revised Downcomer Noding for D. C. Cook Unit 1 and 2," November 2007.
- 7. WCAP-8355, Supplement 1, May 1975, WCAP-8354 (Proprietary), "Long-Term Ice Condenser Containment LOTIC Code Supplement 1," July 1974.
- 8. WCAP-13451, "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting", October 1992.
- 9. Donald C. Cook, Unit 1 - License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology," ADAMS Accession Number ML080090268, December 27, 2007.
- 10. "Donald C. Cook Nuclear Plant, Unit 1 - Issuance of Amendment to Renewed Facility Operating License Regarding use of the Westinghouse ASTRUM Large Break Loss-of-Coolant Accident Analysis Methodology," TAC MD7556, ADAMS Accession Number ML082670351, October 17, 2008.
- 11. "Donald C. Cook Nuclear Plant, Unit 2 (CNP-2) - Issuance of Amendment to Adopt A New Large-Break Loss-of-Coolant Accident Analysis (TAC No. ME1017),"
ADAMS Accession Number ML110730783, March 31, 2011.
- 12. Westinghouse Letter NF-AE-11-53, "Evaluation Options for the LOTIC2 Safety Injection Spill Mass and Energy Error (IR # 10-218-M021)," Revision 0, May 20, 2011.
Approved:11/28/2017 Affected Unit: Unit Two Summary of Change:
NDM Effective Date: 11/30/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Summary Page 102 of 106 UCR-2222 Rev. 0
- 53 UFSAR Effective Date: 11/28/2017 This UCR is to correct typographical errors. These missing or incomplete references were discovered by the 50.71(e) reviewer and it was determined that they are still required and correct.
AR 2017-11094 UCR-1974 - EC-50387 - Implementation of Unit 2 Best-Estimate Large Break LOCA Analysis UFSAR changes that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review.
This UFSAR change does not alter the Current License Basis.
The Section 14.03.01.08 change meets the 50.71(e) evaluation criteria presented in PMP-2350-SAR-001, Attachment 1, Item 4, bullet 4:
- 4. Editorial Changes consisting of the following:
Typographical Changes that are corrections of errors introduced during the production (typing/printing/magnetic storage) of the document Justification:
50.71(e) Basis:
List of Affected Items:
50.59: N/A Reviewed potential impacts of other pending UCRs to be sure there are no further issues.
Comments:
14.03.01.08 (U2)
References for Section 14.3.1 C
U2 Section (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Summary Page 103 of 106 UCR-2100 Rev. 0
- 54 UFSAR Effective Date: 12/04/2017 This UCR is a result of the changes described in EC-53120, Rev. 1, Technical Support Center/North Access Building (TSC/NAB) Addition Modification.
This UCR revises Figures 1.3-1, 1.3-1A, 2.1-3, and 2.1-4B for the removal of the Outage Management/Storeroom and fire protection valve house buildings and for the addition of the TSC/NAB building in order to reflect the updated building configuration in this area. Also, Figure 9.8-1 Fire Protection Water Units 1 & 2 is updated to reflect the changes to the fire protection water system caused by the addition of the TSC/NAB building and the removal of the fire protection valve house.
Precursor Event Required:
After RTO of EC-53120 "Technical Support Center / North Access Building Addition Modification" Precursor Completed On: 12/4/2017 Approved:9/7/2017 Affected Unit: Both Units EC-53120, Rev. 1 - "Technical Support Center / North Access Building Addition Modification" Summary of Change:
Justification:
NDM Effective Date: 9/11/2017 List of Affected Items:
50.59: SS-SE-2014-0501-06 N/A Comments:
C U12 Figure 01.03-01 Plot Plan C
U12 Figure 01.03-01A Key Plan C
U12 Figure 02.01-03 Topographic Map of Site C
U12 Figure 02.01-04B Donald C. Cook Nuclear Plant Topographic Map C
U12 Figure 09.08-01 Fire Protection Water 1 & 2 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Summary Page 104 of 106 UCR-2219 Rev. 0
- 55 UFSAR Effective Date: 12/04/2017 This UCR is a result of the changes described in EC-53120, Rev 1, Technical Support Center/North Access Building Addition Modification and in FCR-53120-127, Technical Support Center / North Access Building Modification TSC/NAB (RTO-1)
Hard Copy.
This UCR revises Figures 1.3-1, 1.3-1A, 2.1-3, and 2.1-4B for the removal of approximately 550 linear foot section of the railroad tracks from the plant road starting at the covered stairway between the Michigan Lot and Managers Lot and stopping at the current Protected Area (PA) fence in order to reflect the updated configuration in this area due to changes by the TSC/NAB Project.
Precursor Event Required:
After RTO of EC-53120, Rev 1, Technical Support Center/North Access Building Addition Modification Precursor Completed On: 12/4/2017 Approved:10/11/2017 Affected Unit: Both Units EC-53120, Rev 1, Technical Support Center/North Access Building Addition Modification FCR-53120-127, Technical Support Center / North Access Building Modification TSC/NAB (RTO-1) Hard Copy GT-2017-9930 Summary of Change:
Justification:
NDM Effective Date: 10/15/2017 (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Summary Page 105 of 106 UCR-2219 Rev. 0
- 55 UFSAR Effective Date: 12/04/2017 List of Affected Items:
50.59: SS-SE-2014-0501-06 The search criteria resulted in some hits. The verbiage in the UFSAR Sections is below the level of detail shown in the markups of the Figures in this UCR. Therefore, only the Figures are updated by this UCR.
Comments:
C U12 Figure 01.03-01 Plot Plan C
U12 Figure 01.03-01A Key Plan C
U12 Figure 02.01-03 Topographic Map of Site C
U12 Figure 02.01-04B Donald C. Cook Nuclear Plant Topographic Map (Sorted by the "UFSAR Effective Date")
INDIANA AND MICHIGAN POWER D. C. COOK NUCLEAR PLANT UPDATED FINAL SAFETY ANALYSIS REPORT Revision: 28.0 Summary Page 106 of 106 UCR-2224 Rev. 0
- 56 UFSAR Effective Date: 01/11/2018 Update Section 15.1.35 and 15.1.41 to update the Aging Management Programs and Activities.
Updating Sections 15.1.35 - Structures Monitoring - Structures Monitoring Program and 15.1.41 - System Walkdown Program to meet the 10 CFR 54.37(b) requirements.
Approved:1/11/2018 Affected Unit: Both Units DIT-S-06332-00 UFSAR changes to Chapter 15 that do not change the meaning or intent of the current licensing basis, and are not associated with a change, test, or experiment, are considered part of the UFSAR update process, in accordance with 10 CFR 50.71(e), FSAR Update, and do not require a 50.59 Review per PMP-2350-SAR-001. UCR-2224 adds aging management activities for flood protection SSCs in the Turbine Building and Screenhouse that were inadvertently not included in the original License Renewal Application. This update is not associated with a change, test, or experiment. It is a correction of UFSAR information in accordance with 10 CFR 54.37(b) to make it consistent with the current licensing basis since these SSCs require aging management by the renewed license.
Summary of Change:
Justification:
50.71(e) Basis:
NDM Effective Date: 1/18/2018 List of Affected Items:
50.59: N/A Only Sections 15.1.35 and 15.1.41 are impacted by this UCR activity.
Comments:
15.01.35 Structures Monitoring - Structures Monitoring Program C
U12 Section 15.01.41 System Walkdown Program C
U12 Section (Sorted by the "UFSAR Effective Date")