AEP-NRC-2015-75, Response to Second Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term

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Response to Second Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term
ML15238A726
Person / Time
Site: Cook  American Electric Power icon.png
Issue date: 08/24/2015
From: Gebbie J
American Electric Power, Indiana Michigan Power Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2015-75, TAC MF5184, TAC MF5185
Download: ML15238A726 (25)
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Text

INDIA NA Indiana Michigan Power MICHIGAN Cook Nuclear Plant PO WERE One Cook Place A unit of American Electric Power Inridgana MIcia Powerc August 24, 2015 AEP-NRC-201 5-75 10 CFR 50.90.Docket Nos. 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-000 1 Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Response to Second Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term

References:

1. Letter from J. P. Gebbie, Indiana Michigan Power Company (l&M), to U. S. Nuclear Regulatory Commission (NRC), "Donald C. Cook Nuclear Plant Units 1 and 2, License Amendment Request to Adopt TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification" and Implement Full-Scope Alternative Source Term," dated November 14, 2014, Agencywide Documents Access and Management System (ADAMS) Accession No. ML14324A209.
2. Letter from J. P. Gebbie, l&M, to NRC, "Donald C. Cook, Unit 1 and Unit 2 -Supplemental Information for the License Amendment Request to Adopt TSTF-490, Revision 0, "Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification" and Implement Full-Scope Alternative Source Term," dated February 12, 2015, ADAMS Accession No. ML15050A247.
3. E-mail capture from A. W. Dietrich, NRC, to T. L. Curtiss, I&M, "D.C. Cook Units 1 and 2 -SRXB RAI Concerning LAR to Adopt TSTF-490 and Implement Full-Scope AST (TAC NOS. MF5184 AND MF5185)," dated July 14, 2015, ADAMS Accession No. ML15195A698.

This letter provides Indiana Michigan Power Company's (I&M), licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, response to the second Request for Additional Information (RAl) by the U. S. Nuclear Regulatory Commission (NRC) regarding a license amendment request (LAIR) to adopt Technical Specification Task Force (TSTF)-490 and implement Alternative Source Term.By Reference 1, as supplemented by Reference 2, I&M submitted a request to amend the Technical Specifications to CNP Units 1 and 2 Renewed Facility Operating Licenses DPR-58 and DPR-74.I&M proposes to adopt TSTF-490, Revision 0, and implement full scope alternative source term U. S. Nuclear Regulatory Commission AEP-NRC-201 5-75 Page 2 radiological analysis methodology.

By Reference 3, the NRC transmitted an RAI regarding the LAR submitted by l&M in Reference 1.Enclosure 1 to this letter provides an affirmation statement.

Enclosure 2 to this letter provides l&M's response to the NRC's RAl in Reference

3. Copies of this letter and its enclosures are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CFR 50.91.There are no new regulatory commitments made in this letter. Should you have any questions, please contact Mr. Michael K. Scarpello, Regulatory Affairs Manager, at (269) 466-2649.Sincerely, Joel P. Gebbie Site Vice President TLC/ams

Enclosures:

1. Affirmation
2. Response to the Second Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Full-Scope Alternate Source Term c: A. W. Dietrich, NRC, Washington, D.C.J. T. King -MPSC MDEQ -RMD/RPS NRC Resident Inspector C. D. Pederson, NRC, Region Ill A. J. Williamson, AEP Ft. Wayne, w/o enclosures Enclosure I to AEP-NRC-2015-75 AFFIRMATION I, Joel P. Gebbie, being duly sworn, state that I am Site Vice President of Indiana Michigan Power Company (I&M), that I am authorized to sign and file this request with the U. S. Nuclear Regulatory Commission on behalf of I&M, and that the statements made and the matters set forth herein pertaining to I&M are true and correct to the best of my knowledge, information, and belief.Indiana Michigan Power Company Joel P. Gebbie Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THISA&\ DAY OF ,2015 My Commission Expires -DANI4LLE BURGOYNE Notary Public, State of Michigan County of Berrien My Commission Expires 04-04-2018 Acting in the County V\

Enclosure 2 to AEP-NRC-2015-75 Response to the Second Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Full-Scope Alternate Source Term By letter dated November 14, 2014 (Agencywide DocUments Access and Management System (ADAMS) Accession No. ML14324A209), as supplemented by letter dated February 12, 2015 (ADAMS Accession No. ML15050A247), Indiana Michigan Power Company (l&M), the licensee for the Donald C. Cook Nuclear Plant (CNP), Units I and 2, submitted a license amendment request (LAR). The proposed amendment consists of adoption of Technical Specifications Task Force (TSTF)-490, Revision 0, and implementation of a full Scope alternate source term radiological analysis methodology.

The U.S. Nuclear Regulatory Commission (NRC) staff in the Reactor Systems Branch (SRXB)of the Office of Nuclear Reactor Regulation is currently reviewing the submittal, as supplemented, and has determined that additional information is needed in order to complete the review. The text of the request for additional information (RAI) and I&M's response is provided below.RAI -S RXB-1 The license amendment request (LAR) dated November 14, 2014 (Agencywvide Documents Access and Management System (ADAMS) Accession No. ML14324A209) listed in Enclosure 12 the various new alternative source term (AST) input parameter values, including those based on reactor coolant system (RCS) performance, for offsite and control room (CR) habitability doses for each accident analysis.

The supplement dated February 12, 2015 (ADAMS Accession No. ML15050A247) provided information about the thermal-hydraulic (TH) analyses that were applied as the basis for the values.As stated in the supplement, "The current licensing basis (CLB) thermal hydraulic (TH)calculations were used to provide input to most of the new dose analyses, although some of those inputs are different from previous inputs that were~ derived from the same TH calculations." The supplement also stated that the majority of the input parameters originated from calculations performed for previous license amendments for Donald C.Cook Nuclear Plant Units 1 and 2, such as license amendment Nos. 271 and 252 for implementation of AST for CR habitability, and license amendment Nos. 256 and 239 to address steam generator tube rupture (SG TR) overfill.

Other inputs were obtained from projects implemented under Title 10 of the Code of Federal Regulations (10 CFR), Section 50.59, from information obtained from actual plant post-trip data, and from simulator data representing a Unit 1 SG TR transient.

The supplement provided additional descriptions Of the TH analysis for each accident considered in the AST analysis.

The source documents for most of the accident analyses, such as the volume data provided by Westinghouse Electric Company during a steam generator replacement project, are not available for the NRC staff to verify the proper incorporation of the values as input parameters to the AST analysis.

Thus, the NRC staff cannot verify the authenticity of the RCS input parameter values to their Enclosure 2 to AEP-NRC-2015-75 Pg Page 2 source documentation, along with whether the sources, other than the previous license amendments, were reviewed and approved by the NRC.a. Provide information for all of the in put parameter values provided in Enclosure 12 of the LAR connecting each value to its respective source documentation.

Additionally, provide the source documentation that produced the input parameter values that were applied in each accident analysis.

The source documentation can be submitted on the docket or made available for staff inspection via an audit. If placed on the docket and considered a proprietary document, submit a redacted version along with the proprietary version in accordance with 10 CFR 2.390.I&M's Response to RAI-SRXB-1:

The following information is being provided in response to RAI-SRXB-1 regarding the CNP LAR to adopt TSTF-490, Revision 0, and implement full-scope alternative source term (AST)(Reference 1). Information presented in the following tables corresponds to the information provided in Enclosure 12 of Reference I for the revised dose consequence analyses input parameters

("AST Value").The source document for each new AST input value is also identified in the tables below. A list of all the source documents is provided after the tables in the "References" section. Any of the input parameter source documents that are not currently available on the NRC docket for CNP will be made available during the audit scheduled for September 2015. The tables below contain input values and source document information for the following accident scenarios:

Table 1 -Control Room Parameters Table 2 -Loss of Coolant Accident (LOCA)Table 3 -Fuel Handling Accident (FHA)Table 4 -Main Steam Line Break (MSLB)Table 5 -Steam Generator Tube Rupture (SGTR)Table 6 -Locked Rotor Accident (LRA)Table 7 -Control Rod Ejection (CRE)Table 8 -Waste Gas Decay Tank (WGDT) Rupture Table 9 -Volume Control Tank (VCT) Rupture Enclosure 2 to AEP-NRC-2015-75 Pg Page 3 Table 1: Control Room Parameters Input/Assumption New AST Value Reference and/or Additional Comments Reference 30 Contol oom(CR Volme 0,66 f3 Contol oom CR)Volme 5,61 ft This value is consistent with the CR habitability AST LAR___________________________________approved via Reference 50 in 2002.Normal Operation Filtered Make-up Flow Rate 0 cubic feet per N itaindrn omloeain_______________

minute (cfm) N itaindrn omloeain Filtered Recirculation Flow Rate 0 cfm No filtration during normal operation.

References 14, 15, and 31 Unfitere Mae-upFlowRat 880cfm The design makeup flow rate is given as 800 cfm in References 14 and 15. Reference 31 provides an expected operating range of between 720 cfm and 880 cfm. Use of a maximum make-up flow rate is conservative.

Reference 32 Unfiltered Inleakage 40 cfm The value provided in Reference 32 of 40 cfm was confirmed to be conservative via gas tracer testing outlined in Reference 59.Emergency Operation Recirculation Mode: References 14, 15, and 31 Rate 80 cfm The design makeup flow rate is given as 800 cfm in References FiltredMakeup low14 and 15. Reference 31 provides an expected operating range of between 720 cfm and 880 cfm. Use of a maximum make-up flow rate is conservative.

Reference 6, Section 5.5.9 FiltredRecrcuatin Flw Rte 520cfm This value represents the total value of 5400 cfm from Reference 6 minus the 880 cfm listed above. The minimum recirculation flow rate reduces the radionuclide removal by the control room filters in the dose consequence analyses.Unfiltered Make-up Flow Rate 0 cfm Filtration activated during emergency operation.

Reference 32 Unfitere Ineakae 40cfm The value provided in Reference 32 of 40 cfm was confirmed to___________________________________be conservative via gas tracer testing outlined in Reference 59.Filter Efficiencies Elemental 94.05% Reference 1 (proposed change to Technical Specifications Organic 94.05% (TS)) and References 3 and 6 The maximum allowable methyl iodide penetration is 2.5% from Reference

1. This corresponds to a filter efficiency of 95% for elemental iodine and organic iodide per Reference
3. The Particulate 98.01% HEPA filter efficiency is 99%Y for radioactive particulates per Reference
6. The filter efficiencies are adjusted to account for the HEPA removal efficiency requirement from Reference 6, Section 5.5.9 (1% adjustment for bypass).Occupancy 0-24 hrs 1Rfrne2 eto ..1-4 days0.4-30 days0.Breathing Rate 3.5 x 10-4 m3/sec Reference 2, Section 4.2.6 Enclosure 2 to AEP-NRC-2015-75 Pg Page 4 Table 2: LOCA Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments Containment Purge Iodine Chemical Form 95% aerosol, 4.85% elemental, Reference 2 0.15% organic Reference 28 This value represents the sum of the upper containment (sprayed/unsprayed), lower Containment Volume 1,066,352 ft 3 containment (sprayed/unsprayed), lower containment fan room (sprayed), ice condenser (unsprayed), and lower containment dead-ended (unsprayed) volumes.References 16, 17, and 19 Containment Purge 36,300 cfm Increased by 10% for conservatism from the Flow Rate source document value.Reference 8 (U 1), Table 14.3.1-8 References 16, 17, and 41 Reference 6, Table 3.3.6-1, Item #4 Representative safety injection signal of Containment Purge 15 seconds approximately 5 seconds from Reference 8 Isolation Time increased to 10 seconds for conservatism.

The containment purge isolation valves are fully closed within 5 seconds of receipt of a safety injection signal for a total isolation time of 15 seconds.References 16, 17, and 18 Containment Purge 0% The referenced documents show that the Filtration purge exhaust fans discharge to the plant vent without filtration.

RmvlbWalNone Not credited.Deposition Removal by Sprays None Not credited.Containment Leakage Iodine Chemical Form 95% aerosol, 4.85% elemental, Reference 2 0.15% organic Reference 27 Continmet Sup pH>7.0An analysis has been performed (Reference

27) to include strong acid generation due to radiation effects.Compartment Volumes (max) Reference 28, Tables 5.2.5.1 & 5.4.4.1 Upper Containment62,8ft (Spryed)The values represent a 2% increase for Lower Compartment 103,770 ft 3 conservatism.(Sprayed)Fan Rooms (Sprayed) 48,913 ft 3 Upper Containment12,0ft (Unsprayed)12,0f3 Ice Condenser10,7ft (Unsprayed)10,7f3 Lower Containment 6,8 t (Unsprayed) 6,8 t Dead-End (Unsprayed) 18,663 ft 3 Enclosure 2 to AEP-NRC-2015-75 Pg Page 5 Table 2: LOCA Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments Reference 43 This value represents a conservative input assumption which will support future plant Containment activities.

Delaying the ventilation start time Vetlto tr ie300 seconds is conservative for the purposes of dose consequence analyses.

This value is consistent with the input utilized in the CNP normal operating pressure / normal operating temperature (NOP/NOT) submittal_________________(see Enclosure 8 of Reference 57).Containment Ventilation Flow Rate Fan Rooms to Lower Containment (Unsprayed)

Fan Rooms to Lower Containment (Sprayed)Lower Containment (Unsprayed) to Dead -End Dead-End to Fan Rooms Lower Containment (Unsprayed ) to Fan Rooms Lower Containment (Unsprayed) to Ice Condenser Lower Containment (Sprayed) to Ice Condenser ice Condenser to Upper Containment (Sprayed)Ice Condenser to Upper Containment (Unsprayed)

Upper Containment (Sprayed) to Fan Rooms Upper Containment (Unsprayed) to Fan Rooms Lower Containment

-Sprayed to/from Unsprayed Upper Containment

-Sprayed to/from Unsprayed 14,580.5 cfm 22,859.5 cfm 90 cfm 90 cfm 1,350 cfm 13,140.5 cfm 22,859.5 cfm 30,072.3 cfm 5,927.7 cfm 30,072.3 cfm 5,927.7 cfm 2206.3 cfm (spray induced circulation) 4086.7 cfm (spray induced circulation)

References 16, 17, and 20 Values taken from the Reference 16 and 17 drawings were conservatively decreased by 10% and then modified in Reference 20 to ratio flows to the appropriate compartments within the model.Sprayed/Unspraed Miin 2 Turnovers of Unsprayed Volume Indued MiigCompartment/hour Reference 2, Appendix A, Section 3.3 Enclosure 2 to AEP-NRC-2015-75 Pg Page 6 Table 2: LOCA Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments Reference 43 This value represents a conservative input assumption which will support future plant Containment Spray 300 seconds activities.

Delaying the spray start time is Start Time conservative for the purposes of dose consequence analyses.

This value is consistent with the input utilized in the CNP NOP/NOT submittal (see Enclosure 8 of Reference 57).References 10, 11, and 20 Continmnt pra 0.39-0426hous an afer 4 hurs The spray interruption corresponds to pump Stopntainmen Spa .1-.2 or n fe 4hus suction realignment.

The values utilized in StopTimethe analyses bound the source documentation value of a 5 minute interruption.

Containment Spray References 29 and 42 Flow Rate Minimum spray flow rates conservatively Upper Containment 1466 gallons per minute (gpm) minimize the removal coefficients.

Lower Containment 660 gpm Fan Rooms 201 gpm Containment Spray-Drop Fall Height Reference 28 The spray drop fall height is used in the calculation of the aerosol iodine removal Upper Containment 58.6 feet (ft) coefficient.

Lower Containment 28.5 ft Fan Rooms 20.1 ft Containment Spray Mean Drop Diameter Reference 29 Upper Containment 609 microns This parameter is used in the calculation of the elemental iodine spray removal Lower Containment 671 microns coefficient.

Fan Rooms 671 microns Elemental Iodine Spray 20 hr-I, with a total decontamination Rfrne4 eto l..~Removal Coefficient factor of 200 Rfrne4 eto 1..~-Time that Total Rfrne2 Elemental Referncer2 decontamination factor2hor reaches 200 Output of RADTRAD computer run.Aerosol Spray Removal CoeffcientReference 20 Upper Containment 5.06 hour-i The coefficients are calculated using the Lower Containment 6.65 hour-I volumes provided above.Fan Rooms 3.03 hour-I TimethatTota AersolReference 20 TimethatTota Aersol2.32 hour3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />s___F ____reaches_______________50_____Output of RADTRAD computer run.

Enclosure 2 to AEP-NRC-2015-75 Pg Page 7 Table 2: LOCA Inputs and Assumptions InputlAssumption New AST Value Reference and/or Additional Comments Reference 4 Organic Iodine Spray None Reference 4 states that it is conservative to Removalassume that organic iodides are not removed by either spray or wall deposition.

Elemental, Organic Iodine -None Not credited.Natural Deposition Aerosols -0.1 hr-i in unsprayed regions Reference 5 only Containment Leakage Rate Reference 1 (proposed change to TS)o to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.18 %/day 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 30 days 0.09 %/day____________________

Containment Leakage 0% Not credited.Filtration___________________

_________________

____Engineered Safety Feature (ESF) Leakage to the Auxiliary Building Iodine Chemical Form 0% aerosol, 97% elemental, 3% organic Reference 2, Appendix A, Section 5.6 Reference 29 This value represents the minimum sump Containment Sump 5095f3 volume at the time of switchover to Volume 5095~recirculation, which conservatively maximizes the radionuclide concentration for the ESF leakage outside of containment.

Emergency Core Reference 47 Cooling System 18. eod Recirculation Start 1384scnsA minimum switchover time increases the Time amount of ESF leakage outside containment.

Reference 12 ESF Leakage Flow 0.2 gpm (two times the allowable value) Per Reference 2, Appendix A, Section 5.2, Rate this value represents double the allowable value of 0.1 gpm from the CNP Leakage_________________Monitoring Program.ESF Leakage Flashing 10% Reference 2 Fraction Auxiliary Building 0 o rdtd Ventilation Filtration 0 o rdtd ESF Leakage to the Refueling Water Storage Tank (Not Explicitly Modeled in Current Licensing Basis)

Enclosure 2 to AEP-NRC-2015-75 Pg Page 8 Table 3: FHA Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments Iodine Cheical Form0% aerosol, 99.85% elemental, Rfrne2 pedxB eto Iodine Chemical Form 0.15% organic Rfrne2 pedxB eto Number of Fuel Assemblies1Rernc2 Damaged Percentage of Fuel Rods Failed 100% Reference 8, Section 14.2.1.4 No. of rods exceeding 6.3 Reference 19 kilowatts per foot (kwlft) above 150 Assumed value to provide margin for future 54 GW/MTUcore designs (conservative assumption).

Reference 56 The high burnup adjustment involves High burnup multiplier applied to 2 doubling the gap inventory of all of the rods in gap fractions the affected assembly.

Since all of the rods in the dropped assembly are assumed to fail, the entire source term for this event is_____________________

_____________________increased by a factor of two.Water Level Above Damaged 23 feet Reference 6, Sections 3.7.14 and 3.9.6 Fuel Pool Decontamination Factors Elmna 25References 2 and 58 Organic -1.0 Reference 7 Delay Before Fuel Movement 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> TRO 8.9.2 states that the reactor shall be subcritical for at least 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> prior to the movement of irradiated fuel assemblies.

Containment Release Filtration 0% Not credited.References 3 and 6 The maximum allowable methyl iodide penetration is 5% from Reference

6. This Aerosol -98.01% corresponds to a filter efficiency of 90% for Fuel Handling Area Exhaust Elemental

-89.1% elemental iodine and organic iodide per Ventlatin Fitraton 8.1%Reference

3. The HEPA filter efficiency is VetltinFltainOrganic

-8.%99% for radioactive particulates per Reference

6. The filter efficiencies are adjusted to account for the HEPA removal efficiency requirement from Section 5.5.9 (1%adjustment for bypass).Table 4: MSLB Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments Maximum Pre-Accident Iodine Spike Concentration 60 pCi/gm Dose Equivalent 1-131 Reference 6, Section 3.4.16 Concurrent Iodine Spike 500x Equilibrium Reference 2 Appearance Rate Iniia StamGenraor 0.1 pCi/gm Dose Equivalent 1-131 Reference 6, Section 3.7.17 Iodine Source Term Iodine Chemical Form 0% aerosol, 97% elemental, Rfrne2 pedxE eto 3% organic Rfrne2 pedxE eto References 38 and 39 Pecnagled oFulRd0%Sufficient DNB margin exists to prevent fuel failures for this event.

Enclosure 2 to AEP-NRC-2015-75 Pg Page 9 Table 4: MSLB Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments References 23 and 44 The Reference 44 information was provided by Westinghouse Electric Company (WEC)Reactor Coolant System 461.5pud-as(b) during replacement of the CNP Unit 1 steam (RCS) Mass 46115pud-as(b) generators (SG(s)). The modification document for the SO replacement is identified in Reference

51. Section 4.1 of Reference 52 provides further information on calculation of this value.Reference 34 97,515.7 Ibm/SG The SG secondary liquid mass values are obtained from the steam release calculation SO Secondary Liquid Mass documented in Reference
34. This calculation was performed for CNP as part of 161,000 Ibm/SO the CR habitability AST LAR approved via Reference 50 in 2002. Section 4.3 of Reference 52 provides further information on these values.0 -2 hours: 456,000 Ibm Reference 34 Intact SG Steam Release 2 -8 hours: 1,186,000 ibm The intact SO steam release values are obtained from the steam release calculation documented in Reference
34. This 8 -24 hours: 1,347,000 Ibm calculation was performed for CNP as part of the CR habitability AST LAR approved via Reference 50 in 2002.Primary-Secondary Leak 0.25 gpm to each SG Reference 1 (proposed change to TS)Rate References 2, 13, and 19 RG 1.183, Appendix E, Section 5.2 requires that the density used in converting volumetric leak rates to mass leak rates be consistent Densty Ued fr Lekagewith the surveillance tests and facility Douensty-MUse for.Leakaget3 instrumentation used to show co'mpliance Volume-t-asio 6.Ib/twith leak rate technical specifications.

Per Reference 13 the required fluid conditions for reported RCS leakage by the reactor coolant leak rate monitoring program is 70°F. At atmospheric pressure, the corresponding fluid density is 62.30 ibm/ft3 References 19 and 42 This value was derived from simulator data representing a Unit 1 post-SGTR transient Duration of intact SO Tube with applicable operator actions. CNP's Unco ery Aftr R acto 40 mintessimulator is certified to Reference

53. From Trip the simulator data it was shown that the intact S~s return to their initial levels at approximately 20 minutes following a reactor trip. A tube recovery time of 40 minutes is utilized to bound both units.

Enclosure 2 to AEP-NRC-2015-75Pae1 Page 10 Table 4: MSLB Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments 0-60secnds:16%References 19 and 42 0-60secnds:16%Reference 24 Tube Leakage Flashing 60- 300 seconds: 6%FrctonDuin Ucoer'300-1200 seconds: 5% This information is derived from plant simulator data and actual plant post-trip data.A discussion of how these values were 1200 seconds-40 mai: 4% derived is provided in Section 4.6 of Reference 52.Reference 46 This is an assumed value with the Reference 46 calculation providing confidence in the assumption.

The Reference 46 calculation performed to Timeto oolRCSto 22F 4 hurssupport CNP's ultimate heat sink (UHS)program, which utilizes conservative assumptions including UHS temperature of 90.1 °F, shows that a temperature of 212°F is reached at approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.Intact SG Iodine Partition Unflashed Leakage -100 Coefficient Reference 2, Appendix E, Section 5.5.4 Flashed Leakage -'0 Reference 49, Section 8.2.1, Unit 1 Reference 45, Unit 2 Intact SG Moisture 0.2% (Particulate Partition Carryover Fraction Coefficient

= 500) SG moisture carryover fractions for Unit 1 and Unit 2 conservatively increased to 0.2%for use in the analysis to bound both units.Table 5: SGTR Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments Maximum Pre-Accident Iodine Spike 60 pCi/gm Dose Equivalent 1-131 Reference 6, Section 3.4.16 Concentration Concurrent Iodine Spike 35 qiiru eeec Appearance Rate 35 qiiru eeec Iniia S Idie ouce 0.1 IpCi/gm Dose Equivalent 1-131 Reference 6, Section 3.7.17 Term Iodine Chemical Form 0% aerosol, 97% elemental, Reference 2 3% organic References 38 and 39 Pecnagled ofFe os0% Sufficient DNB margin exists to prevent fuel failures for this event.References 24 and 44 The Reference 44 information was provided by WEC during replacement of the CNP Unit 1 RCS Mass 466,141.5 Ibm SGs. The modification document for the SG replacement is identified in Reference 51.Section 4.1 of Reference 52 provides further information on calculation of this value.

Enclosure 2 to AEP-NRC-2015-75Pae1 Page 11 Table 5: SGTR Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments SG Secondary Liquid 97,515.7 Ibm/SG Reference 34 Mass The SG secondary liquid mass values are obtained from the steam release calculation documented in Reference

34. This calculation 161,000 Ibm/SG was performed for CNP as part of the CR habitability AST LAR approved via Reference 50 in 2002. Section 4.3 of Reference 52 provides further information on these values.Intact SG Steam Release 0 -30 mmn: 198,515 Ibm Reference 35 30 m -2 our: 31,432Ibm The intact SG steam release values are 2 -8 hours: 1,367,475 Ibm obtained from the SGTR thrust (TH)calculation documented in Reference
35. This calculation was performed for CNP as part of 8 -24 hours: 1,347,000 Ibm the CR habitability AST LAR approved via Reference 50 in 2002. Section 5.2.1 of Reference 52 provides further information on these values.Reference 35 The ruptured SG steam release values are RuptredSG Seamobtained from the SGTR TH calculation RupuedeasGe ta 0-30 min -66,l711bm documented in Reference
35. This calculation was performed for CNP as part of the CR habitability AST LAR approved via Reference 50 in 2002. Section 5.2.2 of Reference 52 provides further information on these values.Pre-Trip Total Steam Flow 17,153,800 Ibm/hr References 8 and 40 Reference 33 This value was taken from a supporting Time of Reactor Trip 101 seconds calculation, Reference 33, to address SGTR overfill (Reference 55). Section 5.2.4 of Reference 52 provides further information about this value.PriarySecnday Lak 0.25 gpm to each steam generator Reference 1 (proposed change to TS)Rate Reference 2 Reference 13 RG 1.183, Appendix E, Section 5.2 requires that the density used in converting volumetric Densty Ued fr Lekageleak rates to mass leak rates be consistent Douensty-MUse for.Leakaget3
  • with the surveillance tests and facility Volume-tonMs 6.Ib/tinstrumentation used to show compliance with leak rate technical specifications.

Per Reference 13 the required fluid conditions for reported RCS leakage by the reactor coolant leak rate monitoring program is 70°F. At atmospheric pressure, the corresponding fluid density is 62.30 Ibm/ft3 Enclosure 2 to AEP-NRC-2015-75Pae1 Page 12 Table 5: SGTR Inputs and Assumptions InputlAssumption New AST Value Reference andlor Additional Comments Reference 35 The ruptured, tube break flow value was obtained from the SGTR TH calculation Ruptured Tube Break 146,704 Ibm documented in Reference

35. This calculation Flow was performed for CNP as part of the CR habitability AST LAR approved via Reference 50 in 2002. Section 5.2.5 of Reference 52 provides further information on this value.Reference 35 This value is taken from the SGTR TH Duration of Ruptured Tube 30 minutes calculation documented in Reference
35. This BreakFlowcalculation was performed for CNP as part of the CR habitability AST LAR approved via Reference 50 in 2002.Break Flow Flashing Pre-Trip:

16% References 19 and 42 Fraction Post rip:Reference 24 0-60 seconds: 16% This information is derived from plant simulator 60-300 seconds: 6% data and actual plant post-trip data. A 300-1200 seconds: 5% discussion of how these values were derived is 1200 seconds-30 mai: 4% " provided in Section 4.6 of Reference 52.References 19 and 42 This value was derived from simulator data representing a Unit 1 post-SGTR transient with Duration of Intact SG applicable operator actions. CNP's simulator Tube Uncovery After 40 minutes is certified to Reference

53. From the Reactor Trip simulator data it was shown that the intact SGs return to. their initial levels at approximately 20 minutes following a reactor trip. A tube recovery time of 40 minutes is utilized to bound both units.Tube Leakage Flashing 06seod:6%References 1 9and 42 Fraction During Uncovery 06seod:1%Reference 24 60- 300 seconds: 6% This information is derived, from plant simulator 300-1200 seconds: 5% data and actual plant post-trip data. A discussion of how these values were derived is 1200 seconds-40 min: 4% provided in Section 4.6 of Reference 52.Reference 46 This is an assumed value with the Reference 46 calculation providing confidence Time to Cool RCS to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the assumption.

The Reference 46 212°F calculation performed to support CNP's UHS program, which utilizes conservative assumptions including an UHS temperature of 90.1 0 F, shows that a temperature of 21 2°F is reached at approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.CGodiinenPatito Unflashed Leakage -100 Reference 2, Appendix E, Section 5.5.4 Flashed Leakage -0 CnesrPriin100 Reference 35, Section 6.3.4 Coefficient Enclosure 2 to AEP-NRC-2015-75Pae1 Page 13 Table 5: SGTR Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments Reference 49, Section 8.2.1, Unit 1 Reference 45, Unit 2 Intact SG Moisture 0.2% (Particulate Partition Coefficient Carryover Fraction = 500) SG moisture carryover fractions for Unit 1 and Unit 2 conservatively increased to 0.2% for___________________

__________________________use in the analysis to bound both units.Table 6: LRA Inputs and Assumptions Input/Assumption New AsT Value Reference and/or Additional Comments Fuel Rod Gap Fractions 1-131 -0.08 Kr-85 -0.10 Other Noble Gases -0.05Rernc2 Other Halogens -0.05 Alkali Metals -0.12 PretgofFeRos11%

Reference 8 (U2), Section 14.1.6.2.2.8 Failed References 9 and 54 FuelRod eakig Fator .65This value was conservatively chosen to bound both units.Reference 19 No. of rods exceeding 6.3 150 rods in two assemblies Asmdvlet rvd agnfrftr kw/ft above 54 GWD/MTU Asmdvlet rvd agnfrftr core designs (conservative assumption).

Reference 22 High burnup multiplier 1.0104 This value is calculated using the high burnup applied to gap fractions core fraction (2 assemblies/i193 assemblies

=0.0104) and doubling the gap fractions for this portion of the core.Iniia S Idie ouce 0.1 IJCi/gm Dose Equivalent 1-131 Reference 6, Section 3.7.17 Term Iodine Chemical Form 0% aerosol, 97% elemental, Reference 2 3% organic References 22 and 44 The Reference 44 information was provided by WEC during replacement of CNP Unit 1 SGs.RCS Mass 466,141.5 Ibm The modification document for the SG replacement is identified in Reference 51.Section 4.1 of Reference 52 provides further____________________information on calculation of this value.

Enclosure 2 to AEP-NRC-2015-75Pae1 Page 14 Table 6: LRA Inputs and Assumptions lnput/Assumption New AST Value Reference and/or Additional Comments SG Secondary Liquid Reference 34 Mass 97,515.7 lbmISG The SG secondary liquid mass values are obtained from the steam release calculation documented in Reference

34. This calculation was performed for CNP as part of the CR habitability AST LAR approved via Reference 161,000 lbm/SG 50 in 2002. Section 4.3 of Reference 52 provides further information on these values.Primary-Secondary Leak Rate0.25gpm o eah SGReference 1 (proposed change to TS)References 2, 13, and 19 RG 1.183, Appendix E, Section 5.2 requires that the density used in converting volumetric leak rates to mass leak rates be consistent Density Used for Leakage with the surveillance tests and facility Volume-to-Mass 623Imf 3 instrumentation used to show compliance with Conversion 623Imf3leak rate TS. Per Reference 13 the required fluid conditions for reported RCS leakage by the reactor coolant leak rate monitoring program is 70°F. At atmospheric pressure, the corresponding fluid density is 62.30 Ibm/ft.Secondary Steam Reference 34 Release 0 -2 hours: 460,000 Ibm The secondary steam release values are 2 -8 hours: 1,256,000 Ibm obtained from the steam release calculation documented in Reference
34. This calculation was performed for CNP as part of the CR 8 -24 hours: 1,347,000 Ibm habitability AST LAR approved via Reference 50 in 2002.Reference 46 This is an assumed value with the Reference 46 calculation providing confidence Time to Cool RCS to in the assumption.

The Reference 46 212 0 24 ourscalculation performed to support CNP's UHS program, which utilizes conservative assumptions including a UHS temperature of 90.1 0 F, shows that a temperature of 212°F is reached at approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Enclosure 2 to AEP-NRC-2015-75Pge1 Page 15 Table 6: LRA Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments References 19 and 42 This value was derived from simulator data representing a Unit 1 post-SGTR transient with Duration of SG Tube applicable operator actions. CNP's simulator Uncovery Following 40 minutes is certified to Reference

53. From the Reactor Trip simulator data it was shown that the intact SGs return to their initial levels at approximately 20 minutes following a reactor trip. A tube recovery time of 40 minutes is utilized to bound both units.Intact Tube Leakage Flashing Fraction During 0-60 seconds: 16% References 19 and 42 Uncovery Reference 24 60- 300 seconds: 6%This information is derived from plant simulator 300-1200 seconds: 5% data and actual plant post-trip data. A discussion of how these values were derived is 1200 seconds-40 mai: 4% provided in Section 4.6 of Reference 52.SG Iodine Partition Coefficient Unflashed Leakage -100Rernc2 Flashed Leakage -0 Reference 49, Section 8.2.1, Unit 1 Reference 45, Unit 2 SG Moisture Carryover 0.2% (Particulate Partition S osuecryvrfatosfrUi n FractionCofiin=50)SmosuecryvrfatosfrUi1ad Coeficiet

= 00)Unit 2 conservatively increased to 0.2% for____________________

__________________________use in the analysis to bound both units.Table 7: CRE Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments Fuel Rod Gap Fractions Noble Gases -0.10 Other Halogens -0.10 Reference 2 Alkali Metals -0.12 PretgofFeRos10%

References 36 and 37 Failed Percentage of Fuel That 0.25% References 36 and 37 Experiences Melting Reference 2 Reference 19 No. of rods exceeding 6.3 150 rods in two assemblies Asmdvlet rvd agnfrftr kw/ft above 54 GWD/MTU Asmdvlet rvd agnfrftr core designs (conservative assumption).

Reference 25 This value is calculated using the high burnup High burnup multiplier 1.0104 core fraction (2 assemblies/i193 assemblies

=appled t gapfracions0.0104) and doubling the gap fractions for this portion of the core.References 9 and 54 FuelRod eakig Fator .65This value was conservatively chosen to bound____________________

_________________________

both units.Inital G Idin Sorce 0.1 pCi/gm Dose Equivalent 1-131 Reference 6, Section 3.7.17 Term _______________

____________________

Enclosure 2 to AEP-NRC-2015-75Pae1 Page 16 Table 7: CRE Inputs and Assumptions Input/Assumption New AST Value Reference and/or Additional Comments Iodine Chemical Form -0% aerosol, 97% elemental, Reference 2 Secondary Release 3% organic Iodine Chemical Form -95% aerosol, 4.85% elemental, Reference 2 Containment Release 0.15% organic Reference 28 This value represents the sum of the upper containment (sprayed/unsprayed), lower Containment Volume 1,066,352 ft 3 containment (sprayed/unsprayed), lower containment fan room (sprayed), ice condenser (unsprayed), and lower containment dead-ended (unsprayed) volumes.Reference I (proposed change to TS)Containment Leakage Rate.0 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> .0.18 %/day 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 30 days 0.09 %/day Containment Leakage Not credited.Filtration 0%Natural Deposition in Containment Element Iodine -NoneNocrdt.

Aerosols -0.,1 hr-i after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Reference 5 Iodine/Particulate RemovalNocrdt.

by Containment Sprays NoneNocrdt.

References 25 and 44 The Reference 44 information was provided by WEC during replacement of the CNP Unit I RCS MassSGs. The modification document for the SG RC as466,141.5 Ibm replacement is identified in Reference 51.Section 4.1 of Reference 52 provides further information on calculation of this value.Reference 34 SG Secondary Liquid Mass 97,515.7 Ibm/SG The SG liquid mass values are obtained from the steam release calculation documented in Reference

34. This calculation was performed for CNP as part of the CR 161,000 Ibm/SG habitability AST LAR approved via Reference 50 in 2002. Section 4.3 of Reference 52 provides further information on these values.Primary-Secondary Leak Rate0.25gpm o eah SGReference 1 (proposed change to TS)References 2, 13, and 19 Density Used for Leakage RG 1.183, Appendix E, Section 5.2 requires Volume-to-Mass Conversion 62.3 Ibm/ft 3 that the density used in converting volumetric leak rates to mass leak rates be consistent with the surveillance tests and facility_____________________instrumentation used to show compliance with Enclosure 2 to AEP-NRC-201 5-75Pae1 Page 17 Table 7: CRE Inputs and Assumptions Input/Assumption New AST Value Reference andlor Additional Comments leak rate TS. Per Reference 13 the required fluid conditions for reported RCS leakage by the reactor coolant leak rate monitoring program is 70°F. At atmospheric pressure, the corresponding fluid density is 62.30 Ibm/ft 3.Reference 34 Secondary Steam Release 0 -2 hours: 460,000 Ibm The secondary steam release values are 2 -8 hours: 1,256,000 Ibm obtained from the steam release calculation documented in Reference
34. This calculation 8 -24 hours: 1,347,000 Ibm was performed for CNP as part of the CR habitability AST LAR approved via Reference 50 in 2002. Section 5.4 of Reference 52 provides further information on these values.Reference 46 This is an assumed value with the Reference 46 calculation providing confidence in the assumption.

The Reference 46 calculation Time to Cool RCS to 212°F 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> performed to support CNP's UHS program, which utilizes conservative assumptions including a UHS temperature of 90.1 0 F, shows that a temperature of 212°F is reached at approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.References 19 and 42 This value was derived from simulator data representing a Unit 1 post-SGTR transient with Duration of SG Tube applicable operator actions. CNP's simulator is Uncovery Following Reactor 40mntscertified to Reference

53. From the simulator Trip 40mntsdata it was shown that the intact SGs return to their initial levels at approximately 20 minutes following a reactor trip. A tube recovery time of 40 minutes is utilized to bound both units.Tube Leakage FlashingRerncs1ad4 Fraction During Uncovery 0-60 seconds: 16% Reference 24 This information is derived from plant simulator 60- 300 seconds: 6% data and actual plant post-trip data. A discussion of how these values were derived is 300-1200 seconds: 5% provided in Section 4.6 of Reference 52.1200 seconds-40 min: 4%SGIoin PrttinUnflashed Leakage -100 Reference 2, Appendix E, Section 5.5.4 Coefficient

____________________Flashed Leakage -0 Reference 49, Section 8.2.1, Unit 1 Reference 45, Unit 2 SG Moisture Carryover 0.2% (Particulate Partition S osuecryvrfatosfrUi n Fracion oeficiet = 00)Unit 2 conservatively increased to 0.2% for use in the analysis to bound both units.

Enclosure 2 to AEP-NRC-2015-75Pae1 Page 18 Table 8: WGDT Inputs and Assumptions Input/Assumption New Analytical Value Reference and/or Additional Comments References 21 and 44 The Reference 44 information was provided by RCS Mass 275,460,950 gm WEC during replacement of the CNP Unit 1 SGs. The modification document for the SG replacement is identified in Reference 51.Section 4.2 of Reference 52 provides further information on calculation of this value.Reference 21 Tank Volume 500 ft 3 Arbitrarily chosen volume modeled such that 100% of the tank volume is released instantaneously.

Reference 21 Tank Release Rate 1,000,000 cfm Conservatively high flow rate is used to model the complete and instantaneous release of all of____________________________________________the activity in the WGDT.Table 9: VCT Inputs and Assumptions Input/Assumption New Analytical Value Reference and/or Additional Comments References 26 and 44 The Reference 44 information was provided by RCS Mass 275,460,950 gm WEC during replacement of the CNP Unit 1 SG.The modification document for the SG replacement is identified in Reference 51.Section 4.2 of Reference 52 provides further information on calculation of this value.Tank Volume Liquid Volume 267 ft 3 Reference 48 Reference 26 Tank Volume Vapor Volume 500 ft 3 Arbitrarily chosen volume modeled such that 100% of the tank volume is released instantaneously.

Reference 26 VCT Release Rate 1,000,000 cfm Conservatively high flow rate is used to model the complete and instantaneous release of all of the activity in the VCT.References 8 and 19 Letdown Flow Rate 132 gpm This value is increased by 10% from the source documentation for conservatism.

Letdown Isolation Time 15 minutes Reference 8, Section 14.2.3.1 Enclosure 2 to AEP-NRC-2015-75Pae1 Page 19 References

1. AEP-NRC-2014-65, "License Amendment Request To Adopt TSTF-490, Revision 0,"Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification and Implement Full-Scope Alternative Source Term," November 14, 2014.2. U. S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.3. NRC RG 1.52, Rev. 2, "Design, Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants," March 1978.4. NRC Standard Review Plan NUREG-0800, Rev. 4, Section 6.5.2, "Containment Spray as a Fission Product Cleanup System." 5. NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments," July 1996.6. Donald C. Cook Nuclear Plant (CNP) Units 1 and 2 Technical Specifications, Revs. 43 (U1) and 41 (U2)7. CNP Units 1 and 2 Technical Requirements Manual (TRM), Revs. 37 (U1) and 36 (U2)S. CNP Updated Final Safety Analysis Report, Revision 25 9. ITSA-U2 COLR-CYCLE 19 10. Procedure 12-EHP-4075-TCA-001, Rev. 4, "Operator Time Critical Actions," August 2013.11. Procedure 1 2-OHP-4023-E-1, Rev. 11, "Loss of Reactor or Secondary Coolant," March 2013.12. Procedure OHI-4032, Rev. 14, "Leakage Monitoring Program" 13. Procedure PMP-5076-ULR-001, Rev. 4, "Reactor Coolant System Leakage Monitoring Program" 14. Drawing OP-1-5149-46, "Flow Diagram, Control Room Ventilation, Unit No. 1" 15. Drawing OP-2-5149-54, "Flow Diagram, Control Room Ventilation, Unit No. 2" 16. Drawing OP-1-5147A-39, "Flow Diagram, Containment Ventilation, Unit No. 1" 17. Drawing OP-2-5147A-46, "Flow Diagram, Containment Ventilation, Unit No. 2" 18. Drawing OP-12-5148-63, "Flow Diagram, Auxiliary Building Ventilation Units #1 and #2" 19. Calculation RWA-1313-001, Rev. 0, "Cook Nuclear Plant AST Radiological Analysis Input Parameter Development," July 2014.

Enclosure 2 to AEP-NRC-2015-75 Page 20 20. Calculation RWA-1 313-006, Rev. 1, "Cook Nuclear Plant LOCA AST Radiological Analysis," September 2014.21. Calculation RWA-1313-008, Rev. 0, "Cook Nuclear Plant WGDT Radiological Analysis as Part of AST Methodology Implementation," July 2014.22. Calculation RWA-1 313-009, Rev. 0, "Cook Nuclear Plant Locked Rotor AST Radiological Analysis," July 2014.23. Calculation RWA-1313-010, Rev. 0, "Cook Nuclear Plant Main Steam Line Break AST Radiological Analysis," July 2014.24. Calculation RWA-1 313-011, Rev. 0, "Cook Nuclear Plant Steam Generator Tube Rupture AST Radiological Analysis," July 2014.25. Calculation RWA-1313-012, Rev. 0, "Cook Nuclear Plant Control Rod Ejection AST Radiological Analysis," July 2014.26. Calculation RWA-1313-013, Rev. 0, "Cook Nuclear Plant VCT Radiological Analysis as Part of AST Methodology Implementation," July 2014.27. Calculation RWA-1 515-001, Rev. 0, "DC Cook Sump pH Analysis Including Radiation Effects Outlined in NUREG/CR-5950," June 2015.28. Calculation PRA-DOSE-Ol11, "containment Sprayed Volumes, Unsprayed Volumes, and Average Spray Fall Heights" 29. Calculation PRA-DOSE-012, Rev. 0, "Iodine Removal by Containment Spray and Residual Heat Removal Spray," April 2008.30. Calculation MD-I12-HV-005-N, Rev. 0, "Control Room Pressure Boundary Volume" 31. Calculation MD-12-HV-017-N, Rev. 2, "Establish outside airflow rates for normal air conditioning system and the pressurization system for the control room"~32. Calculation MD-12-HV-052-N, Rev. 1, "Control Room Ventilation Flow Rates and Charcoal Filter Efficiencies for Radiological Consequence Accident Analyses," October 2009.33. Calculation TH-00-03, Rev. 0, "D.C. Cook Unit 2 Steam Generator Tube Rupture with Operator Actions" 34. Calculation CN-CRA-99-047, Rev. 0, "D.C. Cook Units 1 & 2 Steam Releases for Radiological Dose Calculation" 35. Calculation CN-CRA-99-55, Rev. 1, "Donald C. Cook Steam Generator Tube Rupture T&H Analysis for' NUREG-1 465 Dose Project -Revised" 36. WCAP-9500-A, Volume 4, "Reference Core Report 17x17 Optimized Fuel Assembly," May 1982.37. WCAP-7588, Rev. 1-A, "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods," January 1975.

Enclosure 2 to AEP-NRC-2015-75Pae2 Page 21 38. Engineering Change EC-0000051727, "Unit 1 Cycle 25 Core Reload" 39. Engineering Change EC-0000052225, "Unit 2 Cycle 21 Core Reload" 40. Engineering Evaluation EE-2005-0139, "Steam Generator Safety Valves" 41. Specification ES-CIV-0306-QCN, Rev. 0 (Plus Change Sheets), "Containment Isolation System Licensing/Design Bases Requirements" 42. Design Information Transmittal DIT-B-03594-00, "Miscellaneous Input for Dose Reanalysis Effort (Contract

  1. 01559762)," May 2014.43. Design Information Transmittal DIT-B-03526-02, "Normal Operating Pressure (NOP) /Normal Operating Temperature (NOT) Steam Line Break Mass & Energy Release Analysis," August 2013.44. Design Information Transmittal DIT SGRP 99035-00, Rev. 0, "Reactor Coolant System Volumes," November 1999.45. Design Information Transmittal DIT-SGRP-00064-00, "Unit 2 Steam Generator Design Moisture Carryover," June 2000.46. AEP-13-63, "American Electric Power Donald C. Cook Units 1 and 2 Ultimate Heat Sink Program," August 2013.47. Letter Report AEP-99-277, "Safety Evaluation SECL 99-076 Revision 2 -Containment Modification Evaluation," August 1999.48. Letter Report AEP-88-331, "Radiation Analysis Manual, D.C. Cook Units 1 and 2," July 1988.49. Vendor Technical Document VTD-BAWI-0015, Section 8.2.1, "Babcock and Wilcox Canada, Operating and Maintenance Manual for Unit 1 Replacement Steam Generators, PUB. #222-7803-O&M-1" 50. Letter from NRC to Indiana Michigan Power Company (I&M), "Donald C. Cook Nuclear Plant, Units I AND 2 -Issuance of Amendments (TAC NOS. MB5318 MB5319)," License Amendment Nos. 271 and 252, dated November 14, 2002, ADAMS Accession No. ML022980619.
51. Letter from M.W. Rencheck, l&M, to NRC Document Control Desk, "Updated Final Safety Analysis Report Update in with 10 CFR 50.71(e), Report of Changes, Tests, and Experiments in Accordance with 10 CFR 50.59(d)(2) and Annual Commitment Change Summary Report in Accordance with NEI 99-04," Letter C0601-11, dated June 21, 2001, ADAMS Accession No. ML011760348.
52. AEP-NRC-2015-19, "Donald C. Cook Nuclear Plant Unit 1 and Unit 2 Supplemental Information for the License Amendment Request to Adopt TSTF-490, Revision 0,"Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification" and Implement Full-Scope Alternative Source Term," February 2015.

Enclosure 2 to AEP-NRC-2015-75Pae2 Page 22 53. American National Standards Institute/American Nuclear Society (ANSI/ANS)-3.5,"Nuclear Power Plant Simulators for Use in Operator Training and Examination," September 2009.54. DIT-B-03557-00, 'Core Source Term Input for Dose Reanalysis Effort (Contract

  1. 01559762)," October 2013.55. Letter from NRC to I&M, "Donald C. Cook Nuclear Plant, Units 1 AND 2 -Issuance of Amendments (TAC NOS. MB0739 AND MB0740)," License Amendment Nos. 256 and 239, October 24, 2001, ADAMS Accession No. ML012690136.56. RWA-1313-007, "Cook Nuclear Plant Fuel Handling Accident AST Radiological Analysis," July 2014.57. AEP-NRC-2013-79,- "Donald C. Cook Nuclear Plant Unit 1 Docket No. 50-315 License Amendment Request Regarding Restoration of Normal Reactor Coolant System Operating Pressure and Temperature Consistent With Previously Licensed Conditions," October 2013.58. NRC Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternative Source Terms," March 2006.59. NCS Corporation Data Report, "Unit 1 and Unit 2 Control Room Tracer Gas Testing*Results," December 2010.
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