ML15323A434

Response (Part 1) to Fourth Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative Source Term

ML15323A434
Person / Time
Site:	Cook
Issue date:	11/16/2015
From:	Gebbie J P American Electric Power, Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2015-110, TAC MF5184, TAC MF5185
Download: ML15323A434 (21)

Text

INDIA NA Indiana Michigan Power MICHIGAN Cook Nuclear Plant POWAER° One Cook Place Bridgman, Ml 49106 A unit of Ameri can Electric Power IndianaMichiganPower.com November 16, 2015 AEP-NRC-201 5-110 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 Response (Part 1) to Fourth 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 (I&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 Nuclear Plant Unit 1 and Unit 2 -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 Alternative Source Term," dated February 12, 2015, ADAMS Accession No. ML15050A247.

3. E-mail capture from A. W. Dietrich, NRC, to H. L. Kish, I&M, "D.C. Cook Nuclear Plant, Units 1 and 2 -ARCB RAI Concerning LAR to Adopt TSTF-490 and Implement Full-Scope AST (TAC NOS. MF5184 and .MF5185)," dated September 15, 2015, ADAMS Accession No.ML1 5259A577.This letter provides Indiana Michigan Power Company's (l&M), licensee for Donald C. Cook Nuclear Plant (CNP) Units 1 and 2, response to the fourth Request for Additional Information (RAI) by the U. S. Nuclear Regulatory Commission (NRC) regarding a license amendment request (LAR) to adopt Technical Specification Task Force (TSTF)-490 and implement Alternative Source Term (AST).By Reference 1, as supplemented by Reference 2, I&M submitted a request to amend the Technical Specifications to CNP Units I and 2 Renewed Facility Operating Licenses DPR-58 and DPR-74.l&M proposes to adopt TSTF-490, Revision 0, and implement full scope AST radiological analysis U. S. Nuclear Regulatory Commission AEP-NRC-2015-1 10 Page 2 methodology.

By Reference 3, the NRC transmitted an RAI from the Radiation Protection and Consequence Branch (ARCB) regarding the LAR submitted by I&M in Reference

1. This RAI contains eight separate items for which additional information is requested.

As specified in Reference 3, l&M agreed to provide responses for items RAI-ARCB-1, -3, -4, -5, -6, and -7 in this submittal and will address items RAI-ARCB-2 and -8 by December 16, 2015.Enclosure 1 to this letter provides an affirmation statement.

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

3. Copies of this letter are being transmitted to the Michigan Public Service Commission and Michigan Department of Environmental Quality, in accordance with the requirements of 10 CER 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 Third Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Full-Scope Alternative Source Term c: A. W. Dietrich, NRC, Washington, D.C.R. J. Ancona, MPSC MDEQ -RMD/RPS NRC Resident Inspector C. D. Pederson, NRC, Region III A. J. Williamson, AEP Ft. Wayne, w/o enclosures Enclosure 1Ito AEP-NRC-2015-110 AFFI RMATI ON 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 U{#!4'DAY OF (UvL~w%.t_4.r, 2015 Notary Public My Commission ExpireslC I- li -G:l.ANNE M. PALMA Notary Public. State of Michigan County of Berrien My commission Expires 10-16-2017 Acting In the county o f ..

Enclosure 2 to AEP-NRC-2015-110 Response (Part 1) to Fourth Request for Additional Information Regarding the License Amendment Request to Adopt TSTF-490 and Implement Alternative 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 (I&M), the licensee for the Donald C. Cook Nuclear Plant (CNP), Units 1 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 alternative source term (AST)radiological analysis methodology.

The U.S. Nuclear Regulatory Commission (NRC) staff in the Radiation Protection and Consequence Branch (ARCB) 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 requests for additional information (RAIs) and I&M's responses are provided below.RAI-ARCB-I Indiana Michigan Power Company (l&M, the licensee) has requested to fully implement an alternate source term (AST) methodology at Donald C. Cook Nuclear Plant (CNP), Units 1 and 2. The license amendment request (LAR) provides the evaluation of the radiological consequences of the design basis loss-of-coolant accident (LOCA) for implementation of a fullscope AST under Title 10 of the Code of Federal Regulations (10 CFR) Section 50.67, using the methodology described in Regulatory Guide (RG) 1.183. Previous LARs submitted by the licensee have sought selective implementation of the AST pursuant to RG 1.183, Staff Position 1.2.2, Selective Implementation, to modify the facility design basis that (1) is based on one or more of the characteristics of the AST or (2) entails re-evaluation of a limited subset of the design basis radiological analyses.

The U.S. Nuclear Regulatory Commission (NRC) safety evaluations (SEs) approving the selective implementations of the AST can be found as follows:*By letter dated October 24, 2001, the NRC issued amendment Nos. 256 for Unit I and 239 for Unit 2 (ADAMS Accession No. ML012690136).

The amendments approved the October 24, 2000, LAR to incorporate a supplemental methodology for the analysis of steam generator overfill following a steam generator tube rupture;*By letter dated October 24, 2001, the NRC issued amendment Nos. 257 for Unit 1 and 240 for Unit 2 (ADAMS Accession No. ML0 12630334).

The amendments approved a portion of the June 12, 2000, LAR related to Generic Letter 99-02, "Laboratory Testing of Nuclear-grade Activated Charcoal;"*By letter dated November 13, 2001, the NRC issued amendment Nos. 258 for Unit I and 241 for Unit 2 (ADAMS Accession No. ML012980378).

The amendments approved a portion of the June 12, 2000, LAR to revise the fuel-handling accident with an AST pursuant to 10 CFR 50.67 using the methodology described in RG 1.183 Enclosure 2 to AEP-NRC-2015-1 10 Pg Page 2*By letter dated November 14, 2002, the NRC issued amendment Nos. 271 for Unit 1 and 252 for Units 2 (ADAMS Accession No. ML0229806 19). The amendments approved a portion of the June 12, 2000, LAR that replaced the TID-14844 accident source term used in design-basis radiological analyses for control room (CR) habitability with an AST pursuant to IO.CFR 50.67 using the methodology described in RG 1.183; and*By letters dated December 20, 2002, and May 2, 2003, the NRC issued amendment Nos. 273 for Unit 1 and 259 for Unit 2 (ADAMS Accession Nos. ML023470126 and ML030990094).

The amendments approved the June 28, 2002, and November 15, 2002, LARs that requested an increase of the licensed reactor core power level by 1.66 percent from 3250 megawatts thermal (MWt) to 3304 MWt for each unit.As discussed in amendment Nos. 273 and 259, the NRC staff reviewed the impact of the proposed changes on design basis accident (DBA) radiological analyses as they pertained to an increase of the licensed reactor core power level. The NRC staff concluded that for a power uprate, the radiological consequences of the DBAs would continue to be bound by the doses estimated in amendment Nos. 271 and 252. Thus, the NRC staff reaffirmed the DBA parameters applied in amendment Nos. 271 and 252 as the current licensing basis (CLB).In reviewing and verifying the DBA radiological analyses proposed in the current LAR, the NRC staff referred to amendment Nos. 256/239 and 2 71/252 and the existing CNP Updated Final Safety Analysis Report, Chapter 14, radiological analysis source terms and steam release assumptions.

However, the sources of the proposed parameters and modeling assumptions in the current LAR remain unclear.As stated by the licensee: 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 for implementation of AST for control room (CR) habitability and license amendment Nos. 256 and 239 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, as well as values currently presented in the CNP Units 1 and 2 Updated Final Safety Analysis Report. 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.Furthermore:

Errors were introduced into the CLB dose consequence re-analyses performed in the 2007-20 10 time-frame at CNP, subsequent to the previous control room habitability AST LAR (ADAMS Accession No. ML022980619)

[amendment Nos. 271 and 252]. The errors are being managed via D.C. Cook's corrective actions program and have been appropriately assessed for operability.

Enclosure 2 to AEP-NRC-201 5-110 Pg Page 3 The NRC staff requests the following in formation in a table format: a) List the CR habitability and offsite dose consequences parameters and modeling assumptions for each amendment and each subsequent 10 CFR 50.59 change after the approved amendments discussed above;b) Explain, ]ustify, and provide a reference (ADAMS Accession number if available) for each value in Table 1: Control Room Parameters that was obtained from projects implemented through the use of the 10 CFR 50.59 process that were not already approved in the amendments discussed above;c) Explain how errors introduced into the CLB radiological consequence re-analyses performed in the 2007-2010 time-frame subsequent to amendment Nos. 271 and 252 affect the values in Table 1: Control Room Parameters; and, d) Describe the new CR ventilation modeling in each mode of operation utilizing the CR parameters, since the LAR modeling description differs from that discussed in amendment Nos. 271 and 252.I&M Response to RAI-ARCB-la) and -Ib): Table I outlines the control room parameters utilized in each of the submittals referenced in RAI-ARCB-1 as well as the "CLB [Current Licensing Basisi Value" and "New AST Value" as shown in Enclosure 12 of AEP-NRC-2014-65 (Reference 8). The "Reference

'X' Value" columns correspond to the documents listed in the References section at the end of this enclosure.

The bulleted references identified in the text of RAI-ARCB-1 are summarized below, with reference numbers that also correspond to those listed the References section: Reference 1: NRC letter dated October 24, 2001, issued amendment Nos. 256 for Unit I and 239 for Unit 2;Reference 2: NRC letter dated October 24, 2001, issued amendment Nos. 257 for Unit I and 240 for Unit 2;Reference 3: NRC letter dated November 13, 2001, issued amendment Nos. 258 for Unit 1 and 241 for Unit 2;Reference 4: NRC letter dated November 14, 2002, issued amendment Nos. 271 for Un~it I and 252 for Units 2; and References 5 and 6: NRC letters dated December 20, 2002, and May 2, 2003, issued amendment Nos. 273 for Unit I and 259 for Unit 2 Enclosure 2 to AEP-NRC-201 5-110 Page 4 Table I: Dose Consequence Analysis Control Room Parameters Input/ Reference IReference 2 Reference 3 Reference 4 References CLB Value New AST Vale 1 Vaue VaueVaue516Vaue (efeene )5Value Comments Assumption VleVau VleVle16Vle (Rfrne7s(Reference 8)* .. " ; : :: These values* ..>° ',, '..... .'-. .originate from C o n tr o l o o " R o o m # '. ... .. 5 0 ,6 1 6 f"t .. .... 5 0 ,6 1 6 ft... .. 5 0 , 3 6ft5 0 ,0 00f 3 ft, 3 5 0 ,6 1 6 ft 3 .. ..H V -0 055,0 0-tN50 , 1 6 ft H -0 5 -Volume b] * .. .

9). The* -": :" 'rounded down.Normal Operation Filtered Make-.. : ; 0 cubic feet -Nofledma-u Flee -.',.. .... No filtered aeu Flow Rate normal operation.:

÷...... .... -: Unileed 100cf 00 cm88 fm8OcmSee Section 1.1 Make-up Flow below..: : 10 fm10 fm80cm 8 f Unfiltered 80- cf.. ..m'"40, ,-cfm ...See Section 1.2.n.eaka.e cfm Not- Modeled +below.Emergency Operation Recirculation Mode: ReicltoFiltered Mk 1000 cfm-,:' ..... "..."ii' .... (1 fan) 880 cfm 880 cfm Section 1.1 upFowR t cfm ' : .. .. 'below.Flow Rate 8800.=* cfmn)4520.

cfm .4520...cfm, below.Unfiltered...

-"No unfiltered make-Make-up Flow ..-, .., ,,÷ 0 cfm 0 cfm ,.- .-.:."0 cfm 0 cfm up in emergency Enclosure 2 to AEP-NRC-2015-1 10 Pg Page 5 Table 1: Dose Consequence Analysis Control Room Parameters Input! Reference 1 Reference 2 Reference 3 Reference 4 References CLB Value NeAS 123 Value Comments Assumption Value1 Value2 Value Value 5/6 Value3 (Reference 7)5 (Reference

8) _________Unflteakaed

- ;... ... 98 cfm 98 cfm ' *80 cfm 40 cfm below.Filter Efficiencies Elmetl " ' 95% (1 fan) 95% (1 fan) , , Elemental..80%

(2 fan) 4 95%______80%

(2 fan) 4 ,,-*.:-., .;,94.05%

94.05%Orai 95% (1 fan) 95% (1 fan) See Section 1.4 Organic (2 fan) 4 ___95% ___80% (2 fan) 4 .,. ..94.05% 94.05% blw Particulate ,: .. 98% 98% (1 fan) 9.....98.01%

__ __ __ _ ....__ __ __, __.. ., __ 98% (2 fan)4 ....__ __ ____...._

._ ___99% 98.01%__ __Occupancy 0-24 hrs 1 1 ,- ,',1 1 The newAST 1-4 ays. :., : " '0.60.6 , ... 0.60.6values are taken 1-:as0. .0.6 0. directly from RG 4-30 days .-. ' 0.4 0.4 ' -' " 0.4 0.4 113(eeec.... , , r- , .. ... _ ' ":10, Section 4.2.6)..... ' '. ,-, ', :,,o :,, ,N ote that the: 'i,,i",-'

4t; ' Reference 3" Breathing

..."':"3.47E-4 3.47E-4 3 3:- values can be found Rate m 3 ,. ....m:.sec3.5E-4 m3/sec 3.5E-4 m3/sec inTbe3o ,, ..... .. .......' ,: -:' ' 'A ttachm ent 6 to ,,, , ,, : " ...

from....__ ______ _____ _ .....______ _____ ..... ... .___ ___ ___.__ ___ ___ __ Reference 11.Notes: 'The Reference 1 submittal contains information regarding Steam Generator Tube Rupture Margin to Overfill, which does not include control room ventilation parameters.

2The Reference 2 submittal requested changes to charcoal testing Technical.

Specification (TS) surveillance requirements (SR). No dose-related parameters were altered as part of the Reference 2 submittal.

Discussions of alternative source term dose implementation was deferred to a later date per Reference 2.3 Per Section 3.5 of the Reference 5 submittal, no new dose analyses were performed as part of the effort of measurement uncertainty recapture.

The NRC staff concluded that for a 1.66% increase to the rated core power level, the radiological consequences of Design Basis Accidents would continue to be bounded by the doses estimated in previously performed and accepted analyses.4 See Section 1.5 below.values were utilized in the series of calculations implemented via 10 CFR 50.59 (Reference 7). The values displayed in this column can be found in Enclosure 12 of AEP-N RC-201 4-65 (Reference 8).

Enclosure 2 to AEP-NRC-201 5-110 Pg Page 6 Information Supporting the Remarks in the Table I "Comments" Column 1.1 Make-up Flow Rate A make-up flow rate of 1,000 cfm was utilized in the calculations performed in preparation of the Reference 3 and Reference 4 submittals.

This value was chosen due to the following historical TS SR: "Verifying that the system maintains the control room envelope/pressure boundary at a positive pressure of greater than or equal to 1/16 inch W.G. relative to the outside atmosphere at a system flow rate of 6000 cfm plus or minus 10% with a makeup air flow rate of < 1000 cfm" (Ref.11).The make-up flow rate of 1,000 cfm does not represent the actual .flow rate of the system. The value provided excessive operational margin as evidenced by calculation MD-12-HV-006-N, Rev. 2 (Reference 21), which concluded that the minimum outside air required to maintain the control room pressure boundary at a positive differential pressure of 0.0625 inches water gauge with respect to the outside environment is 516 cfm when including a 20% safety margin.This surveillance requirement no longer exists in the current revision of the TS. Per Reference 12, the deletion of the surveillance requirement was due to industry unfiltered air leakage measurements which showed that a basic assumption of the SR, an essentially leak-tight control room envelope boundary, was incorrect.

The design make-up flow rate of 800 cfm is provided on plant flow diagrams (References 13 and 14) and calculation MD-I12-HV-017-N, Rev. 2 (Reference

15) states that the use of 880 cfm as the maximum flow rate is consistent with the industry balancing practice of +/-10%.1.2 Unfiltered Inleakage An unfiltered inleakage value of 98 cfm was utilized in the calculations performed in preparation of the Reference 3 and Reference 4 submittals.

This value was determined through the use of tracer gas testing (Reference 4).Subsequent tracer gas testing, after completion of system improvements, shows a limiting CNP Unit 2 unfiltered inleakage of 9 cfm from Reference

16. Calculation MD-12-HV-052-N, Rev. I (Reference

17) provides a conservative unfiltered inleakage value of 40 cfm which would account for additional ingress or egress through the control room doors. Note that the "CLB Value" doubled this value to 80 cfm for additional conservatism.

1.3 Recirculation

Flow Rate The recirculation flow rate represents the system flow rate provided in TS 5.5.9 (5,400 cfm) less the filtered make-up flow rate. The values utilized in the referenced submittals are consistent with the "New AST" value when accounting for the difference in filtered make-up flow. rate.

Enclosure 2 to AEP-NRC-2015-1 10 Pg Page 7 1.4 Filter Efficiencies The filter efficiencies are derived from Section 5.5.9 of the CNP TS (Reference 18). Note that the "New AST" values differ slightly from the earlier submittals due to the inclusion of filter bypass (1%). The filter bypass was applied using the following equation: Design Filter Efficiency x 0.99 =Filter Efficiency Including Bypass 1.5 Multiple Fan Operation Two values are provided in the Reference 2 and 4 submittals for single and two fan operation for various parameters.

The operation of two fans can potentially decrease the effectiveness of the charcoal adsorbers.

Per procedure 1-OHP-4023-E-0 (Reference 19), following system actuation, one of the two fans is secured. This is done to maintain lower air velocities through the charcoal adsorbers to ensure a minimum iodine residence time. Therefore, modeling single fan operation accurately represents the system alignment during accident conditions.

I&M Response to RAI-ARCB-l c): The only error introduced in the 2007-2010 time-frame subsequent to amendment Nos. 271 and 252 affecting the values in Table I of Enclosure 12 (Reference

8) was the decrease in control room volume. The value used in the calculations that supported the Reference 3 and Reference 4 submittals (50,61.6 ft 3) was utilized in the current LAR to correct this error. The other differences outlined in Table 1 above are unrelated to the errors introduced in the 2007-2010 time-frame (see the response to Items RAI-ARCB-la) and -lb) for further information).

I&M Response to RAI-ARCB-ld):

The primary difference, as outlined in the response to Items RAI-ARCB-la) and -Ib) above, is the representation of single fan operation only. Per Reference 19, following system actuation, one of the two fans is secured. This is done to maintain lower air velocities through the charcoal adsorbers to ensure a minimum iodine residence time. Therefore, modeling single fan operation accurately represents the system alignment during accident conditions.

The remaining differences are explained in the response to Items RAI-ARCB-la) and -lb)above, with the majority of the input values originating directly from the CNP Units 1 and 2 TS.RAI-ARCB-3 The following statement is made in Section 2. 4 of the LAR: In addition to the dose from contamination of the control room atmosphere by intake or infiltration, the total control room dose also requires consideration of direct shine dose contributions from control room filters, from the external radiation plume, and from radioactive material in the containment building.

The filter shine dose is calculated by Enclosure 2 to AEP-NRC-2015-110 ag Page 8 first determining the maximum activity loading on the control room ventilation system filters during the LOCA event. This is done by considering the control room ventilation maximum fan capacity flow rate along with filter efficiencies of 100%. The activities from the recirculation filter edit of the PAD TRAD output files are then input into a Micro.Shield

[version]

8.03 model that reflects the geometry of the control room filter housing and the recirculation air handler unit position with respect to the control room.Credit is taken for shielding by structural materials and attenuation in air. An integrated 30-day dose is calculated for control room personnel.

In evaluating the LAR, the NRC staff could not thoroughly review and perform confirmatory calculations with the assumptions and methodologies described in the licensee's computation of the direct shine dose to CR personnel.

Therefore, provide the following additional information:

a) The source terms (nuclide activity on the CR ventilation system filter at the same timesteps considered in the direct shine dose analysis), and, b) A thorough description of the CR geometry used in the computation of the direct shine dose to CR personnel.

l&M Response to RAI-ARCB-3:

a) The nuclide activities on the control room ventilation system filter are presented in Table 2 below.Table 2: Control Room Filter Nuclide Activities (Cu ries)Nuclide 2 Hours 8 Hours 24 Hours 96 Hours 720 Hours Co-58 5.31 8E-05 7.850E-05 7.920E-05 7.691E-05 5.963E-05 Co-60 4.071E-05 6.024E-05 6.116E-05 6.I11E-05 6.053E-05 Kr-85 0.000E+00 0.000E+00 0.000E+OO 0.000E+00 0.OOOE+OO Kr-85m 0.O00E+00 0.000E+00 O.000E+OO O.000E+00 O.O00E+OO Kr-87 0.000E+00 0.000E+00 O.O00E+00 O.O00E+OO O.O00E+0O Kr-88 0.000E+00 0.000E+00 O.000E+00 O.OOOE+OO 0.O00E+00 Rb-86 2.372E-03 3.027E-03 2.985E-03 2.671E-03 1.017E-03 Sr-89 4.158E-02 6.130E-02 6.168E-02 5.920E-02 4.143E-02 Sr-90 4.802E-03 7.106E-03 7.216E-03 7.215E-03 7.203E-03 Sr-91 4.556E-02 4.352E-02 1.375E-02 7.193E-05 1.213E-24 Sr-92 3.402E-02 1.085E-02 1.840E-04 1.849E-12 8.961E-82 Y-90* 9.727E-05 5.742E-04 1.638E-03 4.659E-03 7.203E-03 Y-91 6.827E-04 1.102E-03 1.325E-03 1.374E-03 1.010E-03 Y-92 6.703E-03 1.811E-02 1.770E-03 1.783E-09 1.547E-62 Y-93 5.683E-04 5.568E-04 1.884E-04 1.346 E-06 3.395E-25 Zr-95 9.244E-04 1.281E-03 1.288E-03 1.247E-03 9.412E-04 Zr-97 7.022E-04 8.116E-04 4.276E-04 2.227E-05 1.709E-16 Enclosure 2 to AEP-NRC-2015-1 10 Pg Page 9 Table 2: Control Room Filter Nuclide Activities (Curies)Nuclide 2 Hours 8 Hours 24 Hours 96 Hours 720 Hours Nb-95 1.008E-03 1.371E-03 i.387E-03 1.380E-03 1.258E-03 Mo-99 2.491E-02 2.799E-02 2.385E-02 1.120E-02 1.596E-05 Tc-99m 2.330E-02 2.648E-02 2.293E-02 1.080E-02 1.540E-05 Ru-103 1.121E-02 1.645 E-02 i.650E-02 1.565E-02 9.895E-03 Ru-105 6.537E-03 3.791E-03 3.167E-04 4.160E-09 2.052E-5i Ru-106 5.920E-03 8.640E-03 8.760 E-03 8.711E-03 8.295E-03 Rh-105 7.839E-03 1.100E-02 8.523E-03 2.089E-03 i.018E-08 Sb-127 1.259 E-02 1.781iE-02 i.605 E-02 9.350E-03 8.668 E-05 Sb-129 2.795 E-02 1.579E-02 1.231 E-03 i.183E-08 3.899E-52 Te-127 1.442E-02 2.028E-02 1.939E-02 i.350 E-02 4.311E-03 Te-127m 3.992E-03 5.047E-03 5.089E-03 5.034E-03 4.3i7E-03 Te-129 3.304E-02 2.506E-02 9.679E-03 7.868E-03 4.602E-03 Te-129m 9.481E-03 i.279E-02 1.279E-02 1.203E-02 7.034E-03 Te-131m 2.464E-02 3.158E-02 2.215E-02 4.198E-03 2.299E-09 Te-132 i.686E-01 2.334E-01 2.056E-01 1.086E-01 4.304E-04 1-131 1.076E+00 1.974E+00 2.391E+00 3.i33E+00 1.264E+00 1-132 1.177E+00 5.989E-01 2.307E-01 i.i78E-01 4.613E-04 1-133 2.001E+00 3.078E+O0 2.3i5E+00 3.562E-01 1.258E-09 1-134 7.631E-01 1.4i5E-02 6.006E-08 i.945E-32 0.000E+00 1-135 i.631E+00 1.632E+00 3.908E-01 3.485E-04 5.049E-32 Xe-i133 0.000E+00 0.O00E+00 0.000E+00 0.000E+0O 0.000E+00 Xe-i135 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.O00E+00 Cs-i134 4.345E-01 5.1 05E-0i 5.1 39E-01 5.125E-01 5.004E-01 Cs-i136 6.455E-02 8.262E-02 8.066E-02 6.882E-02 1 .739E-02 Cs-i37 2.358E-01 2.758E-01 2.777E-01 2.777E-01 2.772E-O1 Ba-139 2.968E-02 2.149E-03 6.990E-07 1.317E-22 0.000E+00 Ba-140 7.821 E-02 1 .142E-O1 1.11 8E-01 9.497E-02 2.308E-02.

La-140* 2.032E-03 i.380E-02 3.803E-02 8.311E-02 2.659E-02 La-141 1.051 E-03 5.293E-04 3.191 E-05 9.749E-i11 1 .556E-58 La-142 2.876E-04 2.867E-05 2.187E-08 1.910E-22 0.000E+00 Ce-141 1.959E-03 2.838E-03 2.842E-03 2.666E-03 1.531iE-03 Ce-143 1.658E-03 2.161E-03 1.568E-03 3.457E-04 7.022E-10 Ce-144 i.859E-03 2.602E-03 2.633E-03 2.615E-03 2.454E-03 Pr-i43 7.204E-04 1.058E-03 1.101E-03 i.057E-03 2.902E-04 Nd-147 2.945E-04 4.290E-04 4.177E-04 3.457E-04 6.696E-05 N p-239 3.050E-02 4.i93E-02 3.500E-02 i.447E-02 6.873 E-06 Enclosure 2 to AEP-NRC-201 5-110Pae1 Page 10 Table 2: Control Room Filter Nuclide Activities (Curies)Nuclide 2 Hours 8 Hours 24 Hours 96 Hours 720 Hours Pu-238 4.792E-06 7.072 E-06 7.1 93 E-06 7.223E-06 7.262E-06 Pu-239 0.000E+O0 6.388E-07 6.506E-07 6.560E-07 6.597E-07 Pu-240 7.698E-07 1.136E-06 1.153E-06 1.153E-06 1.153E-06 Pu-241 1.921 E-04 2.842E-04 2.886E-04 2.885 E-04 2.875E-04 Am-241 0.000E+00 0.000E+0O 0.000E+00 0.000E+00 0.000E+00 Cm-242 3.555E-05 5.260E-05 5.337E-05 5.279E-05 4.726E-05 Cm-244 8.527E-06 1.258E-05 1.277E-05 1.277 E-05 1.273E-05 Kr-83m 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 Br-82 4.217E-03 7.017E-03 6.568E-03 2.706E-03 4.892E-08 Br-83 6.796E-02 2.239E-02 2.772E-04 4.01 3E-13 3.865E-91 Br-84 1.610E-02 1.183E-05 1.239E-14 2.679E-55 0.000E+00 Rb-89 2.675E-03 2.765E-10 2.725E-29 0.OO0E+00 0.000E+00 Y-91 m 1.403E-02 2.759E-02 8.769E-03 4.586 E-05 7.737E-25 Y-95 0.000E+00 0.000E+00 0.000E+00 0.O00E+00 0.000E+00 Nb-95m 2.920E-07 3.207E-17 3.195E-44 0.000E+00 0.000E+00 Nb-97 9.289E-06 1.299E-05 1.261 E-05 1.072 E-05 6.550E-06 Rh-i103m 2.676E-04 5.702E-05 2.447E-05 1 .277E-06 9.806E-1 8 Pd-109 1.121E-02 1.652E-02 1.657E-02 1.572E-02 9.936E-03 Sb-124 3.106E-03 3.372 E-03 1.499E-03 3.645E-05 3.730E-19 Sb-125 1.71 7E-04 2.533E-04 2.552 E-04 2.466E-04 1.828E-04 Sb-i126 1 .475E-03 2.1 82E-03 2.21 5E-03 2.211 E-03 2.171 E-03 Te-125m 7.004E-05 1.022 E-04 9.999E-05 8.456E-05 1.977E-05 Te-131 5.039E-04 6.600E-04 6.662E-04 6.606E-04 6.175E-04 Te-133 .8.885E-03 7.211E-03 5.057E-03 9.582E-04 5.248E-10 Te-133m 4.192E-03 6.783E-05 4.179E-10 1.404E-33 0.000E+00 Te-134 2.401E-02 3.930E-04 2.425E-09 8.145E-33 0.000E+0O I-130 2.786 E-02 1 .053E-04 I1.305E-1 1 1.010OE-42 0.000 E+00 Xe-131m 3.864E-02 5.186E-02 2.712E-02 8.IIIE-04 1.952E-18 Xe-I133m 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0O Xe-i135m 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+0O Xe-i138 0.000E+00 0.000E+00 0.000E+00 0.000E+0O 0.000E+00 Cs-134m 2.797E-02 9.279E-03 2.054E-04 6.899E-12 1.163E-76 Cs-138 2.837E-01 2.421E-04 2.596E-13 1.066E-53 0.000E+00 Ba-141 7.681E-04 1.327E-09 2.037E-25 1.308E-96 0.000E+00 La-143 1.971E-06 7.094E-14 3.580E-34 0.000E+00 0.000E+00 Pm-147 9.318E-05 1.379 E-04 1.402 E-04 1.407E-04 1.413E-04 Enclosure 2 to AEP-NRC-201 5-1 10Paei Page 11 Table 2: Control Room Filter Nuclide Activities (Curies)Nuclide 2 Hours 8 Hours 24 Hours 96 Hours 720 Hours Pm-148 8.730E-05 1.251 E-04 1.i67E-04 7.973E-05 3.806E-06 Pm-i48m 2.255 E-05 3.322E-05 3.336E-05 3.173E-05 2.051E-05 Pm-149 2.916E-04 3.989E-04 3.287E-04 1.284E-04 3.713E-08 Pm-151 9.937E-05 1.270E-04 8.728E-05 1.506E-05 3.661E-12 Sm-153 3.162E-04 4.280E-04 3.428E-04 1.177E-04 l.119E-08 Eu-154 4.434E-06 6.540E-06 6.640E-06 6.636E-06 6.599E-06 Eu-i155 2.054E-06 3.029E-06 3.075E-06 3.072E-06 3.042E-06 Eu-156 2.291E-04 3.351E-04 3.301E-04 2.879 E-04 8.790E-05 N p-238 6.022E-04 8.210E-04 6.703E-04 2.510E-04 5.042E-08 Pu 243 1.044E-03 6.677E-04 7.235E-05 3.064E-09 3.846E-47 Am-242 5.046E-05 5.759E-05 2.927E-05 1.299E-06 2.443E-18* Filter activities for La-140 and Y-90 are modified and new activities for Ba-i137m and Ru-i106 are created within Microshield to satisfy equilibrium daughter product activity standards.

These values represent data taken directly from the computer code RADTRAD prior to MicroShield's internal manipulation of the data.b) The computer program MicroShield is used to replicate the geometry of the control room ventilation filters with respect to a dose location in the control room. The activity in the filters is modeled as a rectangular volume source. The air handler unit and building structural materials are modeled as radiation shields.The origin of the rectangular volume source in the MicroShield model is at the top, southeast corner of the filter unit. The source 'length' is in the x-direction and must be oriented towards the dose point. Since the dose point represents the operator in the control room below the filter unit, the x-direction is downward.

The y-direction is west from the origin and represents the source 'height' (actual filter width). The source 'width' runs north along the centerline of the filter unit in the z-direction (actual filter depth). Figure 1 shows the source dimensions used in the model. The filter source material is assumed to be air, which minimizes self-shiel ding within the filter unit. An air density of 0.00122 g/cm 3 is specified for the source material Enclosure 2 to AEP-NRC-2015-1 10Pae1 Page 12 I Height = 54" (W)Y Width = 12" (N)z Length =78" (Down)Figure 1: MicroShield Rectangular Volume Source Orientation A dose point is specified below the center of the source and 6 feet above the 633' elevation control room floor. Plant references show that the bottom of the filter is located 19" above the 650' elevation floor. Therefore, based upon a filter 'length' of 78", the x-distance from the top of the filter to the dose point is: in-Dose Point x Distance = 78 in+1.9 in + (650 ft -(633 +6) ft) x = 229 in The dose point is positioned below the center of the source by providing offsets in the 'y'and 'z' directions equal to one-half the source height (27 in) and width (6 in).A series of shields are placed between the source and the dose point as illustrated in Figure 2. The shield material, density, and thickness are summarized in Table 3. Since the x-distance from the source origin to the dose point is greater than the sum of the shield thicknesses, MicroShield automatically creates an 'air gap' between the final shield and the dose point. A density of 0.00122 g/cm 3 is specified for this gap.

Enclosure 2 to AEP-NRC-201 5-110Pae1 Page 13 Shield #2: Galv. Steel Housing Shield #4: Concrete Floor -k Source-Shield #1:Air Shield #3: Skid (Air)Figure 2: MicroShield Shield Configuration Table 3: Shield Inputs Shield Materil Density Thickness No. era (g/cm 3) (in)1 Air 0.00122 13 2 Steel 7.86 0.1382 3 Air 0.00122 6 4 Concrete 2.35 18 The resulting MicroShield source, shield, and dose point configuration is shown in Figure 3 (rotated 90o from the orientation displayed in Figure 1).

page 14 Enclosure 2 to AEP-NRC-2 0 1 5-110 Figure 3: MicroShield Control Room Filter Model RAI-A RAt -A At IfC- tedphowaeabvthdmaed fuel is 23 feet or greater, the decontamination factors [DESth for wther elemental and ogrganic species are 500 and "1, resPetitel iving an verll ffetiv dcontamination factor of 200 (i e., 99.5 percent of h oalidn relacrse /DliSfferomthedmgede rod isrtie te water). This difference in decontaomintion factr fo elementa (99.85 perent) and orgni ioin (0.1 p rcen t),"" spece resus in th iodn e a bto ve t wa. 9ter being com posed of 57, percnelm ta and 43 percent organic species. lf the depth of water is not 23 feet, the decontamination factor will have to be determined on a case-by-case methOd.Regulatory summar'y 2006-04, ,,Experience with implementation of Alternative Source Terms, c arie tht o nrnrides assumlptionls for evaluating the radiological

..... ÷, hnveI the damaged fuelis I Appendix B_ P,,e 1.handling accident.

If_ .÷t"the wa er ontmlnau " -..vr 23fetorse u Reg.ulatoIy Position 2 state ta "the resecntaivey iiga elemental and organic [iodine) species are 500 and 1,repcilY gvngaoeal effective decontamination factor of 200." 50 However, an overall DF of 200 is achieved when the DE for elemental iodine is 285, not 50 The licensee's analysis for the fuel handling accident credits a DF of 285 for elemental iodine and 1.0 for organic iodine with 23 feet of water level above damaged fuel. Confirm: a) The DEs applied in the analysis; and, b.Th.dpt.o.wte auuv .da a e u lfr oh a do n h o tim n t building and in the Auxiliary 5uilding.

Enclosure 2 to AEP-NRC-2015-1 10Pae1 Page 15 I&M Response to RAI-ARCB-4:

a) The decontamination factors have been confirmed to be 285 for elemental iodine and 1.0 for organic iodine in the vendor-prepared calculation.

b) During movement of irradiated fuel assemblies, the water level is maintained to -> 23 ft per the CNP Units 1 and 2 TS (Reference 18). The water level is maintained in the auxiliary building and containment by TS 3.7.14 and 3.9.6, respectively.

RAI-ARCB-5 a) Provide the RADTRAD input files, in electronic format, for each of the AST DBAs described in the LAR.I&M Response to RAI-ARCB-5:

The RADTRAD input files will be affected by an error that was recently found in the meteorological input data. Therefore, current RADTRAD input files are not accurate and will not be provided at this time. After the meteorological input data has been corrected and the related calculations and the RADTRAD input files have been updated, the RADTRAD files will be provided electronically via a CD-ROM attachment.

RAI-ARCB-6 Licensees who have requested approval to fully implement an AST using the methodology described in Regulatory Guide 1.183 have also proposed modifications to the technical specifications (TSs) definition of dose equivalent I-13 1. Some have modified the definition to base it upon the thyroid dose conversion factors of International Commission on Radiation Protection (ICRP) Publication 2, "Report of Committee II on Permissible Dose for Internal Radiation" or ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers." Others have proposed a definition which is a combination of different iodine dose conversion factors, (e.g., RG 1.109, Revision 1, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," ICRP Publication 2, and Federal Guidance Report 11, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion." Although different references are available for dose conversion factors, the TS definition should be based on the same dose conversion factors that are used in the determination of the reactor coolant dose equivalent iodine curie content for the main steam line break and steam generator tube rupture accident analyses.The proposed definition of dose equivalent 1-131 in the LAR specifies the use of thyroid dose conversion factors from Federal Guidance Report 11. These same dose conversion factors are used in the analysis to establish the equilibrium iodine activities in the reactor coolant system source term and to determine the iodine appearance rates in the main steam line break and steam generator tube rupture events.a) List each of the dose conversion factors being used to define the dose equivalent 1-13 1.

Enclosure 2 to AEP-NRC-2015-1 10Pae1 Page 16 I&M Response to RAI-ARCB-6:

Table 12 presents the dose conversion factors utilized to define the Dose Equivalent 1-131. The dose conversion factors (inhalation) are taken from the Thyroid column of Table 2.1 of Federal Guidance Report 11 (Reference 20).ITable 12: Iodine Dose Conversion Factors DCF Isotope (Sv/Bq)I-131 2.92 E-07 1-132 1.74E-09 1-133 4.861E-08 I-134 2.88E-10 I-135 8.46E-09 RAI-ARCB-7 The licensee proposed changes to TS Limiting Condition for Operation 3.4.16 that are based on the generic changes including those in Task Specifications Task Force (TSTF), Subject TSTF-490, Revision 0, "Deletion of E bar Definition and Revision of~ RCS Specific Activity Tech Spec." The licensee reviewed a number of previously approved TSTF-490 requests, corresponding NRC staff requests for additional information and subsequent letters for issuance of the amendments.

The licensee states that, "as a result of those reviews, a deviation was taken from TSTF-490 [as it] relates to mode applicability of surveillance requirements." The licensee elaborates that, "For Surveillance Requirement (SR) 3.4.16.1, the TSTF-490 proposed note "Only required to be performed -in Mode 1" will not be added." a) Explain why the TSTF-490 proposed note for SR 3.4.16.1, "Only required to be performed in Mode 1," will not be added.I&M Response to RAI-ARCB-7:

As stated in Reference 8, several previously approved TSTF-490 LARs from other stations were reviewed, along with the associated RAIs, prior to submitting the l&M request to adopt TSTF-490.

Since 2010, there have been no TSTF-490 LARs approved by the NRC that contained the SR note "Only required to be performed in Mode 1." During that time, the following plants have had TSTF-490 LARs approved with the note "Only required to be performed in Mode 1" either deleted from the SR or modified to include multiple modes:* Arkansas Nuclear One -Approved March 18, 2010 (ML1 00610687)* Braidwood

-Approved March 23, 2010 (ML1 00690386)* Turkey Point -Approved June 23, 2011 (ML1.10800666)

• Catawba, McGuire, Oconee (Duke) -Approved June 25, 2012 (ML1 20760079)* Prairie Island -Approved January 22, 2013 (ML112521289)

• Palo Verde -November 25, 2013 (ML13294A576)

Enclosure 2 to AEP-NRC-2015-1 10Pae1 Page 17 The common RAI received by each licensee who proposed incorporating the MODE I applicablity note into their TSTF-490 LAR was a request from the NRC to provide an explanation of the apparent disparity between the limiting condition for operation (LCO) modes of applicability and the limited mode (MODE 1) under which the surveillance is required.

I&M reviewed those RAIs, TSTF-490, and our current TS, and determined that the Mode 1 applicability note also would create a disparity between the LCO and SR mode applicability for CNP. In addition, exclusion of the note is conservative because it applies to more operational modes than the proposed TSTF-490 SR. The wording and justification for exclusion of the note from the CNP LAR is the same as that provided in the LAR that was submitted for Palo Verde Nuclear Generating Station on December 12, 2012, and was approved by the NRC on January 22, 2013.Based on the observation that previous NRC reviewers have questioned the inclusion of this note to the extent that it was subsequently removed from each approved TSTF-490 LAR in the last six years, and reviews of those LARs confirmed that a similar disparity in mode applicability would affect the CNP TS, I&M determined that the SR note for Mode 1 applicability should be omitted in the LAR for CNP.REFERENCES

1. Letter from U. S. Nuclear Regulatory Commission (NRC) to Indiana Michigan Power Company (I&M), "Donald C. Cook Nuclear Plant,- Units 1 and 2 -Issuance of Amendments (TAC Nos. MB0739 and MB0740)," Dated October 24, 2001, Agencywide Documents Access and Management System (ADAMS) Accession No. ML012690136.

2. Letter from NRC to I&M, "Donald C. Cook Nuclear Plant, Units 1 and 2 -Issuance of Amendments (TAC NOS. MA9394 and MA9395)," dated October 24, 2001, ADAMS Accession No. ML012630334.

3. Letter from NRC to I&M, "Donald C. Cook Nuclear Plant, Units I and 2 -Issuance of Amendments (TAC NOS. MA9394 and MA9395)," dated November 13, 2001, ADAMS Accession No. ML012980378.

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

5. Letter from NRC to l&M, "Donald C. Cook Nuclear Plant, Unit I -Issuance of Amendment 273 Regarding Measurement Uncertainty Recapture-Power Uprate (TAC NO. MB5498)," dated December 20, 2002, ADAMS Accession No. ML023470126.

6. Letter from NRC to l&M, "Donald C. Cook Nuclear Plant, Unit 2 -Issuance of Amendment 273 Regarding Measurement Uncertainty Recapture Power Uprate (TAC NO. MB6751)," dated May 2, 2003, ADAMS Accession No. ML030990094.

7. SS-SE-2010-0016-00,'"UFSAR Change to Incorporate Updated Radiological Dose Accident Analyses," February 2010.

Enclosure 2 to AEP-NRC-201 5-110Pae1 Page 18 8. AEP-NRC-2014-65, "License Ammendment 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 2014, ADAMS Accession No. ML14324A209.

9. Calculation MD-12-HV-005-N, Rev. 0, "Control Room Pressure Boundary Volume," September 1999.10. USNRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000.11. Letter from l&M to NRC, 'Donald C. Cook Nuclear Plant Units I and 2 License Amendment Request for Control Room Habitability and Generic Letter 99-02 Requirements," dated June 12; 2000, ADAMS Accession No. ML003724470.

12. Letter from NRC to I&M, "Donald C. Cook Nuclear Plant, Units 1 And 2 -Issuance of Amendments Regarding Incorpation of TSTF-448, Revision 3, "Control Room Habitability" (TAC NOS. MD7554 and MD7555)," dated December 30, 2008, ADAMS Accession No. ML083430469.

13. Drawing OP-1-5149-46, "Flow Diagram, Control Room Ventilation, Unit No. 1," September 2005.14. Drawing OP-2-5149-54, "Flow Diagram, Control Room Ventilation, Unit No. 2," September 2005.15. 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," June 2011.16. NCS Corporation Data Report, "Unit 1 and Unit 2 Control Room Tracer Gas Testing Results," December 2010.17. Calculation MD-12-HV-052-N, Rev. 1, "Control Room Ventilation Flow Rates and Charcoal Filter Efficiencies for Radiological Consequence Accident Analyses," October 2009.18. D.C. Cook Units I and 2 Technical Specifications, Revs. 51 (U1) and 48 (U2).19. Procedure 1-OHP-4023-E-0, Rev. 38, "Reactor Trip or Safety Injection," September 2015.20. EPA-520/1-88-020, "Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," Federal Guidance Report No. 11, September 1988.21. Calculation MD-12-HV-006-N, Rev. 2, "Control Room Pressure Boundary Minimum Outside Air Requirement," May 2007.