AEP-NRC-2015-19, Supplemental Information for the License Amendment Request to Adopt TSTF-490, Rev 0, Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification and Implement Full-Scope.

Supplemental Information for the License Amendment Request to Adopt TSTF-490, Rev 0, Deletion of E Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification and Implement Full-Scope.

ML15050A247
Person / Time
Site:	Cook
Issue date:	02/12/2015
From:	Gebbie J Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2015-19
Download: ML15050A247 (12)

Text

INDIANA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWERe One Cook Place A unit ofAmerican Electric Power India naMichigan Powercom February 12, 2015 AEP-NRC-2015-19 10 CFR 50.90 Docket Nos. 50-315 50-316 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 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

References:

1. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC) Document Control Desk, "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 Number ML14324A209.

2. Letter from NRC to I&M, Donald C. Cook Nuclear Plant, Units 1 And 2 - Supplemental Information Needed for Acceptance of Requested Licensing Action to Adopt TSTF-490, Rev. 0, "Deletion of E-Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification" and Implement Full-Scope Alternate Source Term (TAC Nos. MF5184 AND MF5185), dated January 16, 2015, ADAMS Accession Number ML14363A491.

By Reference 1, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant Unit 1 and Unit 2, requested a license amendment to adopt TSTF-490 and implement full-scope alternative source term. By Reference 2, the U. S. Nuclear Regulatory Commission (NRC) delineated the supplemental information that is necessary in order to begin a detailed technical review of the license amendment request (LAR). In that letter, the NRC requested that I&M supplement the application to address the requested information by February 13, 2015. to this letter provides an affirmation statement pertaining to the information contained herein. Enclosure 2 contains the supplemental information requested by the NRC. The conclusions reached in the original determination that this LAR contains No Significant Hazards

U. S. Nuclear Regulatory Commission AEP-NRC-2015-19 Page 2 Considerations and the basis for the categorical exclusion from performing an Environmental Impact Statement have not changed as a result of this supplement.

In Reference 2, the NRC also requested an opportunity to conduct a site audit of the analyses to verify the information provided by the licensee. Upon acceptance of this LAR for review by the NRC, I&M will work with the NRC to schedule an on-site audit of the analyses that support this request.

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/amp

Enclosures:

1. Affirmation

2. Supplemental Information for the License Amendment Request to Adopt TSTF-490 and Implement Full-Scope Alternative Source Term c: M. L. Chawla, NRC Washington, D.C.

J. T. King - MPSC MDEQ - RMD/RPS NRC Resident Inspector C. D. Pederson, NRC Region III A. J. Williamson, AEP Ft. Wayne, w/o enclosures

Enclosure 1 to AEP-NRC-2015-19 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 THIS DAY OF , 2015 Notary Public My Commission Expires I I\pYý

Enclosure 2 to AEP-NRC-2015-19 Supplemental Information for the License Amendment Request to Adopt TSTF-490 and Implement Full-Scope Alternative Source Term

1.0 INTRODUCTION

Pursuant to 10 CFR 50.90, by letter dated November 14, 2014, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 1 and Unit 2, submitted a license amendment request (LAR) to adopt TSTF-490, Revision 0, and implement full-scope alternative source term (AST) (Reference 1).

The U. S. Nuclear Regulatory Commission (NRC) staff reviewed the application and concluded that additional information is necessary to enable them to make an independent assessment regarding the acceptability of the proposed amendment in terms of regulatory requirements and the protection of public health and safety and the environment. The NRC requested that I&M supplement the application to address the requested information (Reference 2).

2.0 REQUESTED INFORMATION The following is a restatement of the information requested by the NRC in Reference 2:

In the LAR, the thermal-hydraulic inputs to the accident dose calculations have changed from the values in the current licensing basis (CLB). However, there is no description of the thermal hydraulic (TH) analysis and no references are provided for the CLB. On December 12, 2014, a teleconference was held between the U.S. Nuclear Regulatory Commission (NRC) staff and representatives of the licensee to get further clarification on the CLB and the various values utilized in calculation of the AST. However, during the discussion, the licensee only provided a general methodology of adoption of the various terms from the CLB. The licensee stated that they are still relying on their existing CLB and TH analyses; however they are extracting different values to input into the dose calculations.

To begin a review of this amendment, the NRC staff requires a description of the TH analysis. The licensee will need to describe how they used their CLB analysis to generate new inputs to the accident dose calculations.

The information contained in this enclosure provides the requested supplemental information.

3.0 OVERVIEW 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 following sections provide a discussion of thermal hydraulic and associated input parameters utilized in the AST dose to AEP-NRC-2015-19 Page 2 consequence analyses outlined in AEP-NRC-2014-65 (Reference 1). Where appropriate, explanations are provided for differences between CLB dose consequence analysis values and the values utilized in the Reference 1 analyses. A comparison of input values was provided in 2 of Reference 1.

The majority of the input parameters originate from calculations performed for previous submittals such as CNP Units 1 and 2 license amendment numbers (Nos.) 271 and 252 (Reference 3) for implementation of AST for control room (CR) habitability and license amendment Nos. 256 and 239 (Reference 4) to address steam generator tube rupture (SGTR) overfill. Other inputs are obtained from projects implemented under 10 CFR 50.59, such as the Unit 1 Replacement Steam Generator (SG) modification as documented in the applicable annual report (Reference 5), as well as values currently presented in the CNP Units 1 and 2 Updated Final Safety Analysis Report (Reference 7). The CR habitability and offsite dose consequence analyses were revised in 2011 and implemented under 10 CFR 50.59 as documented in the applicable annual report (Reference 6).

The remaining inputs originate from information obtained from actual plant post-trip data recorded by the CNP Units 1 and 2 plant process computer (PPC) and simulator data representing a Unit 1 SGTR transient including operator actions. The simulator event data are documented in a CNP engineering document. CNP's simulator is certified to American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.5 (Reference 8) and appropriate conservatisms were applied to the PPC and simulator data to bound both units when applicable as discussed in the following sections.

4.0 INPUT PARAMETERS COMMON TO MULTIPLE ANALYSES 4.1 Minimum Reactor Coolant System (RCS) Mass A minimum RCS mass of 466,141.5 pounds-mass (Ibm) is utilized in the revised dose consequence analyses outlined in Reference 1 for the Main Steam Line Break (MSLB), SGTR, Locked Rotor Accident (LRA), and Control Rod Ejection (CRE) postulated event scenarios. The minimum value was derived from volume data provided by Westinghouse Electric Company (WEC) during replacement of the Unit 1 SGs, which was implemented by 10 CFR 50.59 evaluation. The modification document for SG replacement is identified in Reference 5. The minimum RCS mass value was calculated by subtracting the larger pressurizer volume (Unit 1) of 1834.4 cubic feet (ft3) from the lowest RCS volume (Unit 2) of 12,144.3 ft3 and converting to mass via a density of 45.213 Ibm/ft 3 (Unit 2 operating conditions of 574 degrees Fahrenheit ('F) and 2250 pounds per square inch absolute (psia)).

Likewise, the CLB value also was derived from the information provided by WEC. The CLB value of 499,325 Ibm (converted from 2.2649E+08 grams (g) for comparison purposes) was calculated using a Unit 2 RCS volume of 12,470 ft 3 (without tube plugging), subtracting the Unit 2 total pressurizer volume of 1800 ft3 , and adding a pressurizer liquid volume of 454 ft 3 for a total RCS volume of 11,124 ft 3. This volume was converted to grams using a density of 2.05114E+04 g/ft3 for the RCS and 1.68133E+04 g/ft3 for the pressurizer. As the Unit 2 volume to AEP-NRC-2015-19 Page 3 of 11,124 ft3 was less than the value of 11,487 ft3 calculated for Unit 1 for the CLB, the lesser value was used to calculate the RCS radionuclide concentration and iodine appearance rates.

The difference noted in Enclosure 12 of Reference 1 for the new minimum RCS mass is due to the selection of parameters such that the most conservative values were used in calculations.

The CLB value was derived using nominal parameters. The Unit 2 value was chosen as the minimum value to bound both units. The newly performed analyses presented in Reference 1 utilize values that are conservatively biased low by using bounding parameters for RCS volume and pressurizer volume. A minimum mass is used to maximize radionuclide concentrations.

4.2 Maximum RCS Mass A maximum RCS mass of 275,460,950 g is utilized in the dose consequence analyses outlined in Reference 1 for the postulated Waste Gas Decay Tank and Volume Control Tank rupture scenarios. The maximum value was derived from volume data provided for CNP by WEC during the SG replacement project, the modification package for which is identified in Reference 5. The value was calculated by applying a three percent (%) multiplier (accounting for volumetric expansion under hot conditions) to the highest provided RCS volume of 12,535.4 ft 3 (Unit 1, no tube plugging, full pressurizer volume), converting to mass via a density of 47.035 Ibm/ft 3 (no load temperature of 547°F and pressure of 2250 psia), and utilizing a conversion factor of 453.59 g/lbm. A maximum RCS mass maximizes the quantity of nuclides for a given RCS specific activity for these postulated ruptures.

4.3 SG Secondary Liquid Mass The SG secondary liquid mass values (minimum and maximum) for both the new analyses and the CLB were obtained from TH and steam release calculations performed for CNP as part of the CR habitability AST LAR approved by Reference 3 in 2002. The maximum SG secondary liquid mass of 161,000 Ibm/SG originates from a steam release calculation for dose consequence analysis and was utilized consistently between the CLB and newly performed analyses for the MSLB accident scenario. This value was then used in the newly performed analyses for the SGTR, LRA, and CRE event scenarios. The maximum SG inventory is conservatively applied in an evaluation of the dose contribution from the release of iodine initially present in the SG secondary at the beginning of each postulated event.

The minimum value of 91,000 Ibm/SG for the CLB was taken from the SGTR TH calculation and then applied to all accident scenarios. In contrast, the minimum value of 97,515.7 Ibm/SG utilized in the newly performed analyses was taken from the steam release calculation, discussed in the paragraph above, for consistency between minimum and maximum input documentation. Beyond providing consistency between inputs, the steam release calculation provided secondary liquid mass for both hot full power and hot zero power conditions, and is therefore appropriate for use for minimum and maximum values.

to AEP-NRC-2015-19 Page 4 4.4 Primary to Secondary Leak Rate As described in Enclosure 12 of Reference 1, the CLB analyses incorrectly utilize the operational primary to secondary leakage from Technical Specification (TS) 3.4.13 of 150 gallons per day/SG. The newly performed analyses use the accident-induced value of 1 gallon per minute (gpm) for all SGs from TS 5.5.7.b.2. CNP has proposed a revision to TS 5.5.7.b.2 to clarify that the accident induced primary to secondary leakage rate is evenly divided between the four SGs (i.e., 0.25 gpm/SG). The non-conservatism identified in the CLB analyses is being addressed in CNP's corrective action program.

4.5 Duration of Intact SG Tube Uncovery after Reactor Trip The SG tube recovery time of 40 minutes utilized in the dose consequence analyses described in Reference 1 was derived from the Unit 1 simulator data, which are documented in a CNP engineering document. The simulator showed that levels in the intact SGs return to their initial values approximately 20 minutes following the reactor trip. A tube recovery time of 40 minutes is utilized to bound both units. As discussed in section 3.0, the CNP simulator is certified to ANSI/ANS 3.5.

4.6 Tube Leakage Flashing Fraction during Uncovery The time dependent flashing fractions, outlined in Reference 1 for the postulated MSLB, SGTR, LRA, and CRE event scenarios, are derived from the plant simulator data and actual plant post-trip data. The flashing fraction is calculated using the difference of the RCS hot leg enthalpy and saturated liquid enthalpy of the SG divided by the heat of vaporization at the SG conditions. This equation is derived from a heat balance for the adiabatic process. System conditions (i.e., RCS hot leg temperature, RCS pressure, and steam pressure) from the simulator data, which represents Unit 1, were utilized to determine the aforementioned enthalpies and calculate Unit 1 flashing fractions. Plant post-trip data was then utilized to calculate Unit 1 flashing fractions for comparison with the simulator-derived values. The calculated values from the simulator and post-trip data were found to be comparable.

As the core power level and RCS average temperatures are higher for Unit 2, and noting that these parameters could contribute to higher flashing fractions early in the post-trip portion of the postulated event before the RCS reaches no-load conditions, actual plant trip data for Unit 2 was utilized to calculate flashing fractions for comparison with the Unit 1 fractions. As expected, the calculated flashing fractions for Unit 2 were consistently higher post-trip. The difference between the Unit 2 and Unit 1 flashing fractions derived from actual plant trip data were then added to the Unit 1 flashing fractions derived from the simulator data to arrive at bounding flashing fraction values applicable to both units. The simulator data are documented in a CNP engineering document.

4.7 Time to Cool RCS to 212°F As documented in Enclosure 12 of Reference 1, the CLB dose consequence analyses utilized a duration of eight hours (hr) to cool the RCS to 212'F. As there was not adequate justification to AEP-NRC-2015-19 Page 5 provided for limiting the cool down duration to eight hours, the issue was entered into the CNP corrective action program. The newly performed analyses assume a time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to cool the RCS to 212°F for steam release termination. The assumption is supported by a cool down calculation prepared for CNP by WEC. The calculation models a single train residual heat removal cool down with power operated relief valves open down to 250 0 F. Conservative assumptions, including an ultimate heat sink temperature of 90.1 0 F, are included in the calculation, which provides justification for the assumed cool down time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5.0 EVENT-SPECIFIC INPUT PARAMETERS 5.1 MSLB - Intact SG Steam Release The intact SG steam release input values for both the CLB and newly performed dose analyses originate from a background calculation prepared for the CR habitability AST approved in 2002 (Reference 3). The steam releases are consistent with the values listed in Reference 3 for the eight-hour time period. See Section 4.7 for a discussion of the duration of the steam release.

5.2 SGTR - Miscellaneous Inputs 5.2.1 Intact SG Steam Release The intact SG steam release input values for both the CLB and newly performed dose analyses originate from the SGTR TH calculation prepared for the CR habitability AST approved by Reference 3 in 2002. Guidance provided in the TH calculation prepared for CNP by WEC recommends that the calculated mass releases be increased by 10% to minimize the need to perform future radiological calculations. The application of the 10%

increase in steam release is a vendor recommendation to provide future analytical margin.

Therefore, the limiting calculated values were utilized directly in the SGTR dose consequence analyses described in Reference 1. The additional input conservatism was not desired for the new analyses.

The intact SG steam release values from Reference 3 include the 10% conservatism and additional rounding, but are otherwise consistent with the values utilized in the SGTR dose consequence analysis described in Reference 1. Additionally, the CLB values listed in Enclosure 12 of Reference 1 are consistent with the values listed in Reference 3.

5.2.2 Ruptured SG Steam Release The ruptured SG steam release input values for both the CLB and newly performed dose analyses originate from the same SGTR TH calculation as described in section 5.2.1. This calculation was prepared for the CR habitability AST that was approved by Reference 3.

The calculation evaluated a spectrum of cases (34 total) at various TH conditions to determine a bounding ruptured SG steam release. The limiting case produced a steam release of 66,171 Ibm. The value of 66,171 Ibm was utilized directly in the SGTR dose consequence analysis outlined in Reference 1, while the CLB analysis used a value of 73,000 Ibm.

to AEP-NRC-2015-19 Page 6 As described previously, WEC recommends that the calculated mass releases be increased by 10% to minimize the need to perform future radiological calculations. Applying a 10%

increase to the limiting case results in a ruptured SG steam release of 73,000 Ibm, which was then used in the CLB SGTR dose consequence analyses.

As the application of the 10% increase in steam release is a vendor recommendation, the limiting calculated value of 66,171 Ibm was utilized in the SGTR dose consequence analyses described in Reference 1 without alteration. The additional input conservatism was not desired for the new analyses.

5.2.3 Pre-Trip Total Steam Flow Rate through the Condenser The pre-trip total steam flow rate through the condenser of 17,153,800 Ibm/hr utilized in the SGTR dose consequence analysis presented in Reference 1 represents the combined relieving capacity of all Main Steam Safety Valves (MSSV) and is conservatively high with respect to full-power steam flow as described below. This value is taken directly from Section 10.2.2 of Reference 7. Per Reference 7, this capacity is sufficient for the steam generation rate at maximum calculated conditions. Additionally, this value is at least 105%

of the maximum secondary steam flow rate at 100% rated thermal power.

As documented in the associated CNP dose calculation, the CLB rounded up the relieving capacity to 17,200,000 Ibm/hr. This additional conservatism was not applied in the newly performed analyses.

5.2.4 Time of Reactor Trip The CLB for the SGTR dose consequence analysis assumed a reactor trip time of 120 seconds due to low pressurizer pressure, which was conservative when compared to the reactor trip time provided in Table 4.2-1 of WCAP-1 0698-P-A (Reference 9).

In contrast, the reactor trip time utilized in the newly performed analysis of 101 seconds was taken from a supporting calculation performed to address SGTR overfill (Reference 4). This calculation also demonstrates that the 30 minute "hand calculation" analytical approach used in the SGTR TH calculation, which was prepared for the CR habitability AST approved by Reference 3, is more limiting from a dose standpoint than a more realistic analysis modeling operator actions. The time of reactor trip is biased low to conservatively increase the amount of radionuclide activity released to the environment (i.e., minimizing iodine partitioning by the condenser prior to the trip).

5.2.5 Ruptured Tube Break Flow The ruptured tube break flow input values for both the CLB and newly performed dose analyses originate from the SGTR TH calculation that was prepared for the CR habitability AST (Reference 3). This calculation evaluated a spectrum of cases (34 total) at various TH conditions to determine a bounding ruptured tube break flow. The limiting case produced a total break flow of 146,704 Ibm. The value of 146,704 Ibm was utilized directly in the SGTR to AEP-NRC-2015-19 Page 7 dose consequence analysis outlined in Reference 1, while the CLB analysis used a value of 162,000 Ibm.

As described previously, WEC recommends that the calculated mass releases be increased by 10% to minimize the need to perform future radiological calculations. Applying a 10%

increase to the limiting case results in a ruptured SG steam release of 162,000 Ibm, which was then used in the CLB SGTR dose consequence analyses.

As the application of the 10% increase in steam release is a vendor recommendation to provide analytical margin, the limiting calculated value of 146,704 Ibm was utilized in the SGTR dose consequence analyses described in Reference 1 without alteration. The additional input conservatism was not desired for the new analyses.

5.2.6 Duration of Ruptured Tube Break Flow The duration of ruptured tube break flow of 30 minutes utilized in both the newly performed analyses and the CLB is consistent with the SGTR TH analysis performed for the CR habitability AST that was approved by Reference 3 in 2002.

5.2.7 Break Flow FlashingFraction See Section 4.6 for a discussion of how the break flow flashing fractions were developed for the newly performed analyses outlined in Reference 1. The newly calculated flashing fractions do not reflect the plant cooldown which occurs after approximately 1200 seconds.

Maintaining the pre-cooldown flashing fraction value until the break flow is isolated removes any timing requirement for the cooldown from the analysis, and results in conservatively higher steam releases. Such a conservative approach allows for the use of a single set of flashing fractions for both faulted and intact SGs.

The CLB break flow flashing fractions were derived using the same relationship as described in Section 4.6 (i.e., flashing fractions derived using enthalpies of the RCS and SG). Conservative assumptions and inputs representing RCS and SG conditions were utilized to arrive at bounding flashing fractions. Actual plant data was used to derive the conservatively bounding flashing fraction values utilized in the new analyses, while more conservative inputs were used in the derivation of the CLB values. Therefore, the CLB flashing fractions are expectedly higher than those for the new analyses for the majority of the transient, as shown in Enclosure 12 of Reference 1.

5.3 LRA - Secondary Steam Release The secondary steam release input values for both the CLB and newly performed dose analyses originate from a supporting CNP steam release calculation for dose consequence analyses prepared for the CR habitability AST that was approved by Reference 3 in 2002. The steam release duration was increased beyond eight hours to reflect a more accurate cool down duration. The assumption of a 24-hour cool down is supported by a calculation prepared for CNP by WEC, which was discussed in Section 4.7.

to AEP-NRC-2015-19 Page 8 5.4 CRE - Secondary Steam Release The secondary steam release input values for the newly performed dose analyses originate from a supporting CNP calculation prepared for the CR habitability AST approved by Reference 3 in 2002. See Section 4.7 for a discussion of the duration of steam release.

The postulated CRE is not dominated by a forced heat up or cooldown of the nuclear steam supply system, so the integrated steam releases developed for the LRA are assumed to be applicable for the CRE scenario. The CLB used fractional values of the MSSV relieving capacity of 17,200,000 Ibm/hr for the CRE analysis. The release fractions were developed from information provided by WEC and are conservative relative to information provided in ANSI/ANS-5.1 (Reference 10).

5.5 Volume Control Tank - Letdown Flow Rate The letdown flow rate of 132 gpm utilized in the newly performed volume control tank dose consequence analysis was derived from plant system information and industry guidance. A maximum letdown flow rate of 120 gpm is provided in Section 9.2.2 of Reference 7. From recommendations provided in Nuclear Safety Advisory Letter-00-004 (Reference 11), a 10%

uncertainty was applied to this value to arrive at the analytical value of 132 gpm. Reference 11 is a vendor communication of non-conservative analytical assumptions with vendor-provided recommendations for resolution.

The CLB offsite and CR habitability analyses utilize a letdown flow rate of 75 gpm (nominal) and 120 gpm (maximum), respectively. These values are supported by Section 9.2.2 of Reference 7.

6.0 REFERENCES

1. Letter from J. P. Gebbie, Indiana Michigan Power Company (I&M), to U. S. Nuclear Regulatory Commission (NRC) Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Docket Nos. 50-315 and 50-316, 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, AEP-NRC-2014-65, dated November 14, 2014, Agencywide Documents Access and Management System (ADAMS) Accession No. ML14324A209.

2. Letter from NRC to I&M, Donald C. Cook Nuclear Plant, Units 1 And 2 - Supplemental Information Needed for Acceptance of Requested Licensing Action to Adopt TSTF-490, Rev. 0, "Deletion of E-Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Specification" and Implement Full-Scope Alternate Source Term (TAC Nos. MF5184 AND MF5185), dated January 16, 2015, ADAMS Accession Number ML14363A491.

to AEP-NRC-2015-19 Page 9

3. Letter from NRC to I&M, "Donald C. Cook Nuclear Plant, Units 1 AND 2 - Issuance of Amendments (TAC NOS. MB5318 AND MB5319)," License Amendment Nos. 271 and 252, dated November 14, 2002, ADAMS Accession No. ML022980619.

4. 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.

5. Letter from M. W. Rencheck, I&M, to NRC Document Control Desk, "Updated Final Safety Analysis Report Update in Accordance 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.

6. Letter from J. P. Gebbie, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, 10 CFR 50.71(e) Update and Related Site Change Reports,"

AEP-NRC-2012-3, dated April 25, 2012, ADAMS Accession No. ML12142A362.

7. D.C. Cook Units 1 and 2 Updated Final Safety Analysis Report, Revision 25, September 9, 2013.

8. American National Standards Institute/American Nuclear Society (ANSI/ANS)-3.5, "Nuclear Power Plant Simulators for Use in Operator Training and Examination," September 2009.

9. WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill," August 1987 (WCAP-1 0750-A, Non-Proprietary).

10. ANSI/ANS-5. 1, "Decay Heat Power in Light Water Reactors," April 1, 2005.

11. Nuclear Safety Advisory Letter-00-004, "Nonconservatisms in Iodine Spiking Calculations,"

March 7, 2000.