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{{Adams
#REDIRECT [[AEP-NRC-2009-23, License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology]]
| number = ML090930453
| issue date = 03/19/2009
| title = Donald C. Cook, Unit 2 - License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology
| author name = Weber L J
| author affiliation = Indiana Michigan Power Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000316
| license number = DPR-074
| contact person =
| case reference number = AEP-NRC-2009-23, TAC MB9483, TAC MD7556
| document type = Letter, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specifications
| page count = 40
| project = TAC:MB9483, TAC:MD7556
| stage = Other
}}
 
=Text=
{{#Wiki_filter:INDIANA MICHIGANM POWER Indiana Michigan Power Company Nuclear Generation Group One Cook Place Bridgman, Ml 49106 aep.com AEP-NRC-2009-23 10 CFR 50.90 March 19, 2009 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
 
==SUBJECT:==
 
==References:==
 
Donald C. Cook Nuclear Plant Unit 2 Docket No. 50-316 License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology
: 1) Letter from H. N. Berkow, U. S. Nuclear Regulatory Commission (NRC) , to J. A. Gresham, Westinghouse Electric Company, "Final Safety Evaluation for WCAP-16009-P, Revision 0, 'Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)' (TAC No. MB9483)," dated November 5, 2004 (ADAMS Accession Number ML043100073).
: 2) Letter from J. N. Jensen, Indiana Michigan Power Company (I&M), to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Docket No.50-315, License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology," dated December 27, 2007 (ADAMS Accession Number ML080090268).
: 3) Letter from T. A. Beltz, NRC, to M. W. Rencheck, I&M, "Donald C. Cook Nuclear Plant, Unit 1 -Issuance of Amendment to Renewed Facility Operating License Regarding Use of The Westinghouse ASTRUM Large Break Loss-of-Coolant Accident Analysis Methodology (TAC NO. MD7556)," dated October 17, 2008 (ADAMS Accession Number ML082670351).
 
==Dear Sir or Madam:==
Pursuant to 10 CFR 50.90, Indiana Michigan Power Company (I&M), the licensee for Donald C. Cook Nuclear Plant (CNP) Unit 2, proposes to amend Appendix A, Technical Specification (TS), to Facility Operating License DPR-74. I&M proposes to modify TS 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," TS 3.5.2, "ECCS-Operating," and TS 5.6.5, "Core Operating Limits Report (COLR)." ,4-oo1 U. S. Nuclear Regulatory Commission AEP-NRC-2009-23 Page 2 I&M is also requesting Nuclear Regulatory Commission (NRC) approval to adopt a new analysis of a large break loss-of-coolant accident (LBLOCA) for CNP Unit 2. The new analysis uses a plant-specific adaptation of the methodology documented in Westinghouse Topical Report WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)." NRC-approval of the ASTRUM methodology is documented in Reference
: 1. The plant-specific adaptation of the ASTRUM methodology employs an increased number of circumferential noding stacks in the downcomer.
Validation of the plant-specific adaptation for CNP Unit 1 and Unit 2 was documented in Enclosure 3 to Reference 2.Use of a plant-specific adaption of the ASTRUM methodology for CNP Unit 1 was approved by the NRC via Reference 3.Enclosure 1 to this letter provides an affirmation statement.
Enclosure 2 provides I&M's evaluation of the proposed change. The attachment to this letter provides the CNP Unit 2 TS pages marked to show changes. Clean copies of the affected TS pages with the proposed changes incorporated will be provided to the NRC Licensing Project Manager upon request.I&M requests approval of the proposed amendment in accordance with the normal NRC review schedule.
The proposed changes to the CNP Unit 2 TS will be implemented within 90 days of approval.Copies of this letter and its attachment 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. This letter contains no new or modified NRC commitments.
Should you have any questions, please contact Mr. John A. Zwolinski, Regulatory Affairs Manager, at (269) 466-2478.Sincerely, Lawrence J. Weber Site Vice President JRW/rdw
 
==Enclosures:==
: 1. Affirmation
: 2. Indiana Michigan Power Company's Evaluation
 
==Attachment:==
 
Donald C. Cook Nuclear Plant Unit 2 Technical Specification Pages Marked To Show Changes U. S. Nuclear Regulatory Commission AEP-NRC-2009-23 Page 3 c: T. A. Beltz -NRC Washington DC, K. D. Curry, Ft. Wayne AEP, w/o enclosures/attachment J. T. King, MPSC MDEQ -WHMD/RPS NRC Resident Inspector M. A. Satorius, NRC Region III Enclosure 1 to AEP-NRC-2009-23 AFFIRMATION I, Lawrence J. Weber, 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 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 Lawrence J. Weber Site Vice President SWORN TO AND SUBSCRIBED BEFORE ME THIS ___ DAY OF [VLUJ ,2009 My Commission Expires G 1o 7o/-Vr ..\,
Enclosure 2 to AEP-NRC-2009-23 INDIANA MICHIGAN POWER COMPANY'S EVALUATION
 
==Subject:==
License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology
 
==1.0 DESCRIPTION==
 
==2.0 PROPOSED CHANGE==
 
==3.0 BACKGROUND==
 
===3.1 Change===
to TS 3.4.1 3.2 Change to TS 3.5.2 3.3 Change to TS 5.6.5.b.4 3.4 Change to CNP Unit 2 LBLOCA Analysis Methodology
 
===4.0 TECHNICAL===
 
ANALYSIS 4.1 Change to TS 3.4.1 4.2 Change to TS 3.5.2 4.3 Change to TS 5.6.5.b.4 4.4 Change to CNP Unit 2 LBLOCA Analysis Methodology
 
===5.0 REGULATORY===
 
SAFETY ANALYSIS 5.1 No Significant Hazards Consideration
 
===5.2 Applicable===
 
Regulatory Requirements
/ Criteria 6.0 ENVIRONMENTAL CONSIDERATIONS
 
==7.0 REFERENCES==
 
===8.0 PRECEDENTS===
 
Enclosure 2 to AEP-NRC-2009-23 Page 2
 
==1.0 DESCRIPTION==
 
Indiana Michigan Power Company (I&M) proposes to amend Appendix A, Technical Specification (TS), to Donald C. Cook Nuclear Plant (CNP) Unit 2 Facility Operating License, DPR-74. I&M proposes to modify TS 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," TS 3.5.2, "ECCS -Operating," and TS 5.6.5, "Core Operating Limits Report (COLR)." -I&M also requests Nuclear Regulatory Commission (NRC)approval to adopt a new analysis of a large break loss-of-coolant accident (LBLOCA) for CNP Unit 2.
 
==2.0 PROPOSED CHANGE==
As shown in the attachment to this letter, I&M proposes to: Change the minimum Reactor Coolant System (RCS) total flow rate specified in TS Limiting Condition for Operation (LCO) 3.4.1.c and TS Surveillance Requirements 3.4.1.3 and 3.4.1.4 from 366,400 gallons per minute (gpm), to 354,000 gpm.Delete Condition D "One or more Safety Injection (SI) System cross tie valves closed," and delete reference to Condition D in Conditions A and C from the TS 3.5.2 Actions.Replace the reference to the existing LBLOCA analysis methodology identified in TS 5.6.5.b.4 with a reference to a plant-specific adaptation of the methodology documented in Westinghouse Topical Report WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRUM)." I&M also proposes to adopt a new CNP Unit 2 LBLOCA analysis which uses a plant-specific adaptation of the ASTRUM methodology documented in WCAP-16009-P-A.
The plant-specific adaptation uses twelve circumferential noding stacks in the downcomer as compared to the four stacks used for four-loop plants in the WCAP-1 6009-P-A methodology.
Changes to TS Bases will be needed to reflect adoption of the new LBLOCA analysis.
The Bases changes will be made in accordance with the CNP Technical Specification Bases Control Program.
 
==3.0 BACKGROUND==
 
This section provides the background and reason for each of the proposed changes.3.1 Change to TS 3.4.1 The TS 3.4.1 LCO and Surveillance Requirements establish limits on RCS pressure, temperature, and total flow rate to ensure that the minimum departure from nucleate boiling (DNB) ratio will be met for each of the transients analyzed in the safety analysis.
The current value for minimum RCS total flow specified in the TS 3.4.1 LCO and Surveillance Requirements is 366,400 gpm. The 366,400 gpm value is a minimum measured flow value, which includes allowances for flow measurement uncertainty.
As recognized by the NRC in Reference 1, the Enclosure 2 to AEP-NRC-2009-23 Page 3 current practice is that the thermal design flow value, which does not. include allowances for flow measurement uncertainty, be specified in the TS. The minimum measured flow is specified in the COLR. I&M is therefore proposing that the value for minimum RCS total flow specified in the TS 3.4.1 LCO and Surveillance Requirements be changed to the value for thermal design flow, 354,000 gpm. The proposed change will not affect the 354,000 gpm value used in the current and the new LBLOCA analyses.3.2 Change to TS 3.5.2 The TS 3.5.2 LCO requires that two independent and redundant Emergency Core Cooling System (ECCS) trains be operable to ensure that sufficient flow is available, assuming a single failure affecting either ECCS train, to provide core cooling and negative reactivity to ensure that the reactor core is protected after certain accidents, including an LBLOCA and a small break loss of coolant accident (SBLOCA).
The ECCS consists of three separate subsystems:
Centrifugal Charging (high head), Safety Injection (SI) (intermediate head), and Residual Heat Removal (RHR) (low head). Each subsystem consists of two redundant, 100% capacity trains, each with its own pump.One train of the Sl subsystem discharges to RCS loops 1 and 4. The other train of the Sl subsystem discharges into RCS loops 2 and 3. The Sl subsystem trains are normally cross-tied so that if one pump were to fail, SI system flow would still be delivered to all four discharge lines. If a cross-tie valve is closed, SI system flow would be limited to two lines, resulting in lower flow than if the cross-tie valve were open. The current TS 3.5.2 Actions include a Condition D that allows the unit to be in Mode 1, 2, or 3 for an unlimited amount of time if an SI system cross-tie valve is closed, provided that thermal power is reduced to less than or equal to a specified value. This allowance is justified by the current LBLOCA and SBLOCA analyses.However, the proposed new LBLOCA analysis does not include a condition in which an Sl subsystem cross-tie valve is closed. I&M is therefore proposing that Condition D be deleted from the TS 3.5.2 Actions, and reference to Condition D be deleted from Condition A and Condition C.3.3 Change to TS 5.6.5.b.4 TS 5.6.5.b states that the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, and lists analytical methods used. The methodology currently listed in TS 5.6.5.b.4 for the LBLOCA analysis is the BASH evaluation methodology documented in WCAP-10266-P-A, "The 1981 Version of Westinghouse ECCS Evaluation Model Using the BASH Code." I&M is proposing to adopt a new LBLOCA analysis methodology as described below. I&M is therefore proposing that the LBLOCA methodology listed in TS 5.6.5.b.4 be replaced with the new LBLOCA methodology.
 
===3.4 Change===
to CNP Unit 2 LBLOCA Analysis Methodoloqy The regulations specified in 10 CFR 50.46(a)(1) identify calculation methodology requirements for nuclear power plant loss-of-coolant accident (LOCA) analyses.
Sections 10 CFR 50.46(a)(3)(i) and (ii) specify criteria to be applied and actions to be taken when significant changes or errors in parts of the plant-specific LOCA methodology are found to have accumulated.
Enclosure 2 to AEP-NRC-2009-23 Page 4 The current CNP Unit 2 LBLOCA analysis-of-record was performed in 1995 using the BASH evaluation model methodology documented in WCAP-10266-P-A, "The 1981 Version of Westinghouse ECCS Evaluation Model Using the BASH Code." By Reference 2, I&M committed to provide a new LBLOCA analysis for CNP Unit 2 due to a cumulation of changes and errors requiring a reanalysis in accordance with 10 CFR 50.46(a)(3)(ii).
The new analysis has been performed using a plant-specific adaptation of the ASTRUM methodology.
The ASTRUM methodology is documented in Westinghouse Topical Report WCAP-16009-P-A.
As documented in a Safety Evaluation (SE) (Reference 3), the NRC has approved the ASTRUM methodology for meeting the regulatory requirements of 10 CFR 50.46. As also documented in the SE, license amendment requests that deviate from WCAP-16009-P-A are subject to a plant-specific review. I&M is therefore requesting NRC approval to adopt the CNP Unit 2 plant-specific adaptation of the ASTRUM methodology.
 
===4.0 TECHNICAL===
 
ANALYSIS This section provides a technical analysis and justification for each of the proposed changes.4.1 Change to TS 3.4.1 I&M is proposing that the value for minimum RCS total flow specified in the TS 3.4.1 LCO and Surveillance Requirements, 366,400 gpm (minimum measured flow), be changed to 354,000 gpm (thermal design flow). Although the numerical value specified in the TS 3.4.1 LCO and Surveillance Requirements will be changed, the LCO and Surveillance Requirement will continue to require that the limit for minimum measured flow will be met in that they also require that the RCS minimum flow rate be greater than or equal to the limit specified in the COLR. The RCS minimum flow rate specified in the COLR is the minimum measured flow value. Therefore, the proposed change does not constitute a change to the current technical requirements.
The proposed change only affects the manner in which the specified limit on minimum measured flow is controlled, in that the COLR is controlled by I&M and submitted to the NRC at the start of each fuel cycle as required by TS 5.6.5.d. The proposed change is consistent with the current practice as recognized by the NRC in Reference
: 1. The proposed change will also make CNP Unit 2 TS 3.4.1 LCO and Surveillance Requirements for minimum RCS flow consistent with CNP Unit 1 TS 3.4.1 LCO and Surveillance Requirements for minimum RCS flow.4.2 Change to TS 3.5.2 I&M is proposing that Condition D be deleted from the TS 3.5.2 Actions, and reference to Condition D be deleted from Condition A and Condition C. Condition D allows the unit to be in Mode 1, 2 or 3, for an unlimited amount of time if an Sl system cross-tie valve is closed, provided that thermal power is reduced to less than or equal to a specified value. This allowance was originally added to the CNP Unit 2 TS by Reference
: 4. The new LBLOCA analysis proposed by this amendment does not address a condition in which an Sl cross-tie valve is closed. Therefore, the allowance provided by TS 3.5.2 Condition D should be deleted.Deletion of Condition D will render the CNP Unit 2 TS 3.5.2 Actions consistent with the CNP Unit 1 TS 3.5.2 Actions and the TS 3.5.2 Actions of the standard Westinghouse TS documented in NUREG 1431.
Enclosure 2 to AEP-NRC-2009-23 Page 5 4.3 Change to TS 5.6.5.b.4 I&M is proposing to replace the LBLOCA methodology listed in TS 5.6.5.b with a new LBLOCA methodology as described below. This change will result in TS 5.6.5.b.4 properly identifying the methodology used in the LBLOCA analysis of record upon implementation of this proposed amendment.
 
===4.4 Change===
to CNP Unit 2 LBLOCA Analysis Methodology I&M is requesting NRC approval to adopt a CNP Unit 2 plant-specific adaptation of the ASTRUM methodology documented in WCAP-16009-P-A.
The WCAP describes a realistic (or best-estimate)
ECCS evaluation model for demonstrating plant compliance with 10 CFR 50.46 for postulated plant-specific LBLOCA transients.
WCAP-16009-P-A uses a statistical approach in developing the peak cladding temperature (PCT), local maximum oxidation (LMO), and core wide oxidation (CWO) results at the 9 5 th percentile.
The ASTRUM methodology requires the execution of 124 transients to determine a bounding estimate of the 9 5 th percentile of the PCT, LMO, and CWO parameters with a 95% confidence level. These parameters are needed to satisfy 10 CFR 50.46 criteria.Westinghouse analyzed the CNP Unit 2 LBLOCA using a plant-specific adaptation of the ASTRUM methodology.
The analysis was performed in compliance with all of the conditions and limitations identified in the NRC SE approving ASTRUM. This proposed amendment is intended to meet the following limitation placed on WCAP-16009-P-A as stated in the NRC SE: The methodology described in WCAP-16009-P-A, Revision 0, is a separate and unique methodology.
Any other version derived from this TR [topical report], such as designated by a new revision number, amendment number, addendum number or other equivalent designation, would constitute a definition of a new methodology requiring NRC review and acceptance prior to generic application and prior to any specific plant licensing application of a new methodology derived from ASTRUM.Preliminary results with the as-approved ASTRUM method applied to CNP Unit 1 (prior to submittal of the proposed CNP Unit 1 amendment approved by Reference
: 5) yielded unexpected results that have been attributed to overly conservative aspects of the model. A combination of CNP design features contribute to the as-approved WCOBRAJTRAC computer code predicting a significant loss of liquid inventory from the downcomer during reflood, due to extensive boiling by heat release from the vessel wall, core barrel, and thermal shield.Consequently, a plant-specific adaptation of ASTRUM has been used for the CNP Unit 2 analysis to better model the downcomer region by increasing the number of circumferential noding stacks from four to twelve. This finer nodalization has been assessed against experimental data, as described in Enclosure 3 to Reference
: 6. Upon approval and implementation, the CNP Unit 2 LBLOCA plant-specific adaptation of ASTRUM will be annotated in the COLR as "Plant-specific adaptation of WCAP-16009-P-A, Revision 1, as approved by Safety Evaluation dated [DATE]." The detailed radial noding of the vessel wall remains unchanged from the approved ASTRUM LBLOCA Evaluation Model (References 3 and 7) and therefore, does not change the historically Enclosure 2 to AEP-NRC-2009-23 Page 6 approved method for addressing downcomer boiling during reflood. The only difference from the previous noding method is that the vessel wall is partitioned into twelve segments connected to twelve downcomer channel "stacks" versus the four in the generic noding. This is the same twelve node stack downcomer model used for CNP Unit 1, as approved by Reference 5.Table 1 lists the major plantparameter assumptions in the best estimate (BE) LBLOCA analysis for CNP Unit 2 and Table 2 summarizes the results of the CNP Unit 2 ASTRUM analysis.Table 3 contains a sequence of events for the limiting PCT transient.
The scatter plot presented in Figure 1 shows the influence of the effective break area on the analysis PCT. The effective break area is calculated by multiplying the discharge coefficient with the sampled value of the break area, normalized to the cold-leg cross sectional area.Figure 1 is provided to illustrate that the break area is a significant contributor to the variation in PCT.From the 124 calculations performed as part of the ASTRUM analysis, the same case proved to be the limiting PCT, LMO, and CWO transient for CNP Unit 2. Figure 2 shows the predicted HOTSPOT cladding temperature transient at the PCT location for the limiting case. The HOTSPOT PCT plot includes local uncertainties applied to the Hot Rod. Figure 3 presents the WCOBRA/TRAC PCT transient predicted for the limiting case. This figure does not account for local uncertainties.
Figures 4 through 16 illustrate the key major response parameters for the limiting PCT and LMO transient.
The reference point for the lower plenum liquid level presented in Figure 11 is the bottom of the vessel (10.1 feet below the bottom of the active fuel). The reference point for the downcomer liquid level presented in Figure 12 is the bottom of the vessel. The reference point for the core collapsed liquid levels presented in Figure 13 is the bottom of the active fuel.The containment backpressure utilized for the LBLOCA analysis compared to the calculated containmentbackpressure is provided in Figure 17. The worst single failure for the LBLOCA analysis is the loss of one train of ECCS injection (consistent with the ASTRUM topical report);however, all containment systems which would reduce containment pressure are modeled for the LBLOCA containment backpressure calculation.
Figure 18 provides the CNP Unit 2 LBLOCA axial power distribution.
I&M and its analysis vendor (Westinghouse) have interface processes which identify plant configuration changes potentially impacting safety analyses.
These interface processes, along with vendor internal processes for assessing evaluation model changes and errors, are used to identify the need for LOCA analyses impact assessments.
Both I&M and Westinghouse have ongoing processes that assure that the ranges and values of the input parameters for the CNP Unit 2 LBLOCA analysis conservatively bound the ranges and values of the as-operated CNP Unit 2 parameters.
Implementation of the approved LBLOCA methodology will necessitate changes to TS 5.6.5.b to replace the reference to the current LBLOCA analysis-of-record with the plant-specific adaptation of WCAP-16009-P-A, as an approved LBLOCA analysis methodology for CNP Enclosure 2 to AEP-NRC-2009-23 Page 7 Unit 2. Using the new methodology does not result in any new operating limits requiring a change to the COLR.The core power level is one of the key input parameters included in Table 1. The CNP Unit 2 BELBLOCA analysis was performed modeling a core power of 3468 megawatts-thermal (MWt), with a maximum power level uncertainty of 0.34% being statistically sampled for all cases. The analysis maximum core power was therefore 3479.8 MWt.' The CNP Unit 2 licensed maximum core power level is 3468 MWt. SI delay time is another key parameter included in Table 1. A delay time of 54 seconds is assumed without offsite power available, to account for an emergency diesel generator (DG) start time of 30 seconds. Current TS Surveillance Requirements verify the DG starts within 10 seconds. I&M is not requesting any core power level changes or changes to the DG start time as part of this license amendment request.The WCOBRA/TRAC model used in the analysis assumes an RHR cross-tie valve modification has been installed.
This modification was completed during the Unit 2 Cycle 17 outage (Fall 2007). The modification allows CNP Unit 2 to operate with RHR cross-tie valves open, providing four-loop injection during a postulated LBLOCA. CNP Unit 2 operates with RHR cross-tie valves open to meet analysis assumptions.
Prior to completion of the modification, the design required RHR cross-tie valves to be closed to address concerns communicated by NRC Bulletin 88-04, regarding pump-to-pump interaction during miniflow operation.
Based on the results presented in Table 2, I&M concludes that CNP Unit 2 continues to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.Table 1 Major Plant Parameter Assumptions Used in the CNP Unit 2 BELBLOCA Analysis Parameter Value Plant Physical Description
* Steam Generator Tube Plugging <10%Plant Initial Operating Conditions
* Maximum Reactor Power 3479.8 MWt F Q <2.335* Peaking Factors FAH _<1.644 Fluid Conditions
* Average Coolant Temperature (TAVG) 547.6 -5.6 OF -TAVG <578.1 + 4.1 OF 2100 -62.6 psia --<PRCS <2100 + 62.6 psia;* Pressurizer Pressure (PRcs) 2250 -62.6 psia --<PRCS _2250 + 62.6 psia* Reactor Coolant Flow 88,500 gpm per Loop (354,000 gpm total)* Accumulator Temperature (TAcc) 60 OF --<TAcc <120 OF* Accumulator Pressure (PAcC) 599.7 psia 5PACC <672.7 psia Enclosure 2 to AEP-NRC-2009-23 Page 8" Accumulator Water Volume (VAcC) 921 ft 3 <VACc 971 ft 3* Accumulator Boron Concentration 2228 ppm Accident Boundary Conditions*" Single Failure Assumptions Loss of one ECCS train* SI Flow Minimum* SI Temperature (Ts 1) 70 OF <Ts: 1 105 OF<27 sec (with offsite power)* SI Initiation Delay Time _54 sec (without offsite power)* Containment Pressure Bounded (minimum);
See Figure 17 Table 2 CNP Unit 2 Best-Estimate LBLOCA Results ASTRUM Result Value Criteria 95/95 PCT (OF) 2107 < 2,200 95/95 LMO (%) 9.7 < 17 95/95 CWO (%) 0.55 < 1 Table 3 CNP Unit 2 Best-Estimate Sequence of Events for the Limiting PCT Case Event Time (seconds)Break Initiation 0.0 Sl Signal 5.7 Accumulator Injection Begins 18.5 End of Blowdown 26.5 Bottom of Core Recovery 43.5 Accumulator Empty 52.2 SI Begins 59.7 PCT Occurs 315.9 End of Calculation
* 600.0* Quench is calculated to occur by the end of the calculation time.
Enclosure 2 to AEP-NRC-2009-23 Page 9 m A* PCT D DEC A PCT SPL 0 0 0 0 0 0 PCT DEGC [deg PCT SPL I T [deg:1 lln&#xfd;n/ /UU 2000-A A* A~n*...UA .* .....A 1800-1 AL A w F-C-)m AA k 1600-A m As A AA#A A A 'I 1400- 1..............
* A'AWN* l 1200-A AL A A A....................AU A A I I ItI I I I I I I fln-*I UUU C 0.5 1 1.5 CD
* Abreak/ACL 2 2.5 Figure 1 -HOTSPOT PCT versus Effective Break Area Scatter Plot (CD = Discharge Coefficient, Abreak = Break Area, ACL = Cold Leg Area)
Enclosure 2 to AEP-NRC-2009-23 HOTSPOT ROD 1 PEAK CLADDING TEMPERATURE 2500-2000 ............20 0 .........................1-' 1500.Il_ ..E a .an ..1~.Page 10 0 100 200 300 400 500 600 Time After Break (s)700 823161850 Figure 2 -HOTSPOT Cladding Temperature Transient for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 WC/T ROD 1 PEAK CLADDING TEMPERATURE Page 11 C-E C, 0 100 200 300 400 Time After Break (s)500 600 700 23161M80 Figure 3 -WCOBRA/TRAC Cladding Temperature-Transient for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 PRESSURIZER PRESSURE 2500 2000 .............1500-ci---s C,, Page 12 0 50 100 150 200 Time After Break (s)250 823161 850 Figure 4 -Pressurizer Pressure for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 BREAK FLOW 60000 50000.40000 -. ........ .E-o30000.0 U.- 20000-o Page 13 0 50 100 150 200 Time After Break (s) C 250 1123161MG Figure 5 -Break Flow for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 Intact Loop--- --Broken Loop Page 14 C-0 w, 0_0 50 100 150 200 Time After Break (s)250 a23161B00 Figure 6 -Void Fraction in Pumps for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 VAPOR FLOW RATE IN CORE HOT ASSEMBLY CHANNEL Page 15 C', E-o 0 0 0~0 0 100 200 300 Time After Break (s)400 500 a23161&50 Figure 7 -Vapor Flow at Top of Core Hot Assembly Channel for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 VAPOR FLOW RATE IN CORE HOT ASSEMBLY CHANNEL Page 16 Cl)E t--o 0 0~0 100 200 300 Time After Break (s)400 500 823161850 Figure 8 -Vapor Flow at Bottom of Core Hot Assembly Channel for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 INTACT LOOP 2 ACCUMULATOR MASS FLOW RATE Page 17 E 0 C/)0 50 100 150 200 Time After Break (s)250 8231 61850 Figure 9 -Accumulator Injection Flow for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 Page 18 INTACT LOOP 2 CHARGI NG SI MASS FLOW RATE 14 121-10--U)E C, 0., U)............................................................I I I........................................................................................................................................
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...................
41 A I I I I 823161850I I I E .I I I ...U 0 50 I I 100 150 Time After Break (s)2 200 250 82318150 Figure 1OA -Charging Safety Injection (SI) Flow for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 Page 19 INTACT LOOP 2 RHR + HHSI MASS FLOW RATE E C', W 0 50 100 150 200 Time After Break (s)250 823161850 Figure 1 OB -Residual Heat Removal (RHR) + High Head Safety Injection (HHSI) Flow for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 Page 20 LOWER PLENUM COLLAPSED LIQUID LEVEL 12 10 ... .... ...........Cn,-2 2 -8 .. ...........................................0-0 100 200 300 400 500 Time After Break (s)823161850 Figure 11 -Lower Plenum Collapsed Liquid Level for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 LIQUID LEVEL IN INTACT LOOP 2 DOWNCOMER Page 21 27 C,, C-.)0 100 200 300 400 Time After Break (s)500 82161850 Figure 12 -Downcomer Collapsed Liquid Level for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 COLLAPSED LIQUID LEVEL IN CORE AVERAGE CHANNEL Page 22 a,-J-o 0~~0'I, U)0~0 C-, 0 100 200 300 400 Time After Break (s)500 Figure 13 -Core Average Channel Collapsed Liquid Level for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 VESSEL LIQUID MASS Page 23 cn-Ej-0 100 200 300 400 Time After Break (s)500 823161850 Figure 14 -Vessel Fluid Mass for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 Page 24 Hot Rod Hot Assembly Guide Tubes Open Holes/Support Low Power Columns/Flow Mixers U-cJ)a, 0~E a.)0 100 200 300 4W 500 600 700 Time After Break (s)823161850 Figure 15 -WCOBRA/TRAC Peak Cladding Temperature for all 5 Rod Groups for the Limiting Case Enclosure 2 to AEP-NRC-2009-23 HOT ROD PCT ELEVATION Page 25 w 0 100 200 300 400 Time After Break (s)500 823181850 Figure 16 -Peak Cladding Temperature Elevation for the Hot Rod for the Limiting Case**The PCT location is based on the core noding (approximately one node for every 1.9 inches of core elevation).
Enclosure 2 to AEP-NRC-2009-23 Page 26 DC Cook Unit 2 Containment Backpressure Comparison WCOBRA/TRAC Containment Backpressure LOTIC2 Calculated Containment Backpressure (n cf)0 100 200 300 4W0 Time (s)500 367082624 Figure 17 -Analyzed Versus Calculated Containment Backpressure Enclosure 2 to AEP-NRC-2009-23 Page 27 0.40 0.3, 0.38 0.35, 0.38 0.38 _ 0 0 0.36 0.34 0.32 0.3, 0.3 0.30 S0.28 0.26 0.24 0.22 0.38, 0.235 0.43, 0.235 0.20 0.25 0.30 0.35 0.40 0.45 PMID Figure 18 -CNP Unit 2 BELBLOCA Analysis Axial Power Shape Operating Space Envelope PBOT = integrated power fraction in the bottom third of the core PMID = integrated power fraction in the middle third of the core Enclosure 2 to AEP-NRC-2009-23 Page 28 5.0 REGULATORY SAFETY ANALYSIS 5.1 No SiQnificant Hazards Consideration Indiana Michigan Power Company (I&M) has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: 1. Does the proposed change involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated?
Response:
No The current minimum departure from nucleate boiling ratio Technical Specifications (TS)specify a minimum measured flow value in the Reactor Coolant System (RCS) total flow requirement.
I&M is proposing to replace this minimum measured flow value with a thermal design flow value. The current minimum departure from nucleate boiling ratio TS also require that RCS total flow meet the requirements in the Core Operating Limits Report (COLR). The COLR specifies the minimum measured flow value. Consequently, the minimum measured flow value will continue to be met. This proposed change does not alter any system or actual flow value.I&M is proposing to delete a TS provision that allows the unit to operate for an unlimited amount of time with a Safety Injection (SI) system cross tie valve closed, provided that thermal power is reduced. As discussed below, I&M is proposing to adopt a new large break loss-of-coolant accident (LBLOCA) analysis.
The new analysis does not evaluate plant operation with an SI system cross-tie valve closed. The position of the SI system cross connect valve does not affect the likelihood of an accident.
This proposed change will assure the plant will be operated within the new LBLOCA analysis.I&M is proposing to modify the TS such that it identifies the new LBLQCA analysis methodology rather than the analysis methodology being replaced.
This TS change is administrative in that it will identify the new methodology following approval of the new methodology by the Nuclear Regulatory Commission (NRC).I&M is proposing to adopt a new LBLOCA analysis which uses a plant-specific adaptation of a best-estimate methodology using the automated statistical treatment of uncertainty methodology (ASTRUM).
The analysis is based on the current plant configuration and the plant will be operated within the assumptions of the analysis.
The analysis demonstrates that the current emergency core cooling system design performance conforms to the criteria contained in 10 CFR 50.46.b. An LBLOCA is the only accident involved in this change. No changes are being made to any reactor protection system or engineered safeguards features actuation system setpoints.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
Enclosure 2 to AEP-NRC-2009-23 Page 29 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response:
No The proposed changes to the TS will not result in the operation of any structure, system, or component in a new or different manner. Adoption of a plant-specific adaptation of the ASTRUM methodology will not create any new failure modes that could lead to a different kind of accident.Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?Response:
No It has been shown that the analytical technique used in the analysis realistically describes the expected behavior of the Donald C. Cook Nuclear Plant Unit 2 reactor system during a postulated LBLOCA. Uncertainties have been accounted for as required by 10 CFR 50.46.A sufficient number of loss-of-coolant-accidents (LOCAs) with different break sizes, different locations, and other variations in properties have been analyzed to provide assurance that the most severe postulated LOCAs were analyzed.
WCOBRA/TRAC validation with the revised downcomer noding has been found acceptable for application of the ASTRUM methodology, with no changes to the uncertainty treatment.
The analysis has demonstrated that all acceptance criteria contained in 10 CFR 50.46, Paragraph b, continue to be satisfied.
Therefore, the proposed change does not involve a significant reduction in the margin of safety.Based on the above, I&M concludes that the proposed amendment presents a no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
 
===5.2 Applicable===
 
Regulatory Requirements/Criteria In accordance with 10 CFR 50.46, the conclusions of the new LBLOCA analysis show that CNP Unit 2 continues to maintain a margin of safety to the limits prescribed by the following criteria: 1. The calculated maximum fuel element cladding temperature (i.e., peak cladding temperature) will not exceed 2,200 0 F.2. The calculated total oxidation of the cladding (i.e., maximum cladding oxidation) will nowhere exceed 0.17 times the total cladding thickness before oxidation.
: 3. The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam (i.e., maximum hydrogen generation) will not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding Enclosure 2 to AEP-NRC-2009-23 Page 30 cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.4. The calculated changes in core geometry are such that the core remains amenable to cooling.5. After successful initial operation of the emergency core cooling system, the core temperature will be maintained at an acceptably low value and decay heat will be removed for the extended period of time required by the long-lived radioactivity remaining in the core.In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the NRC's regulations, and (3)the issuance of the amendment will not be inimical to the common defense and security or to the health or safety of the public.6.0 ENVIRONMENTAL CONSIDERATIONS A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.
However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
 
==7.0 REFERENCES==
: 1. Letter from P. C. Wen, NRC, to A. Drake, Westinghouse Electric Corporation, "Request for Additional Information for Westinghouse Topical Report WCAP-14483, "Generic Methodology for Expanded Core Operating Limits Report," dated September 2, 1998.2. Letter from S. A. Greenlee, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Units 1 and 2, Annual Report of Loss-of-Coolant Accident Evaluation Model Changes," C0801-19, dated August 31, 2001 (ADAMS Accession Number ML012490022).
: 3. Letter from H. N. Berkow, NRC, to J. A. Gresham, Westinghouse Electric Company,"Final Safety Evaluation for WCAP-16009-P, Revision 0, 'Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM)' (TAC NO. MB9483)," dated November 5, 2004 (ADAMS Accession Number ML043100073).
: 4. Letter from J. B. Hickman, NRC, to E. E. Fitzpatrick, I&M, "Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2 -Issuance of Amendment Re: Increase in Main Steam Safety Enclosure 2 to AEP-NRC-2009-23 Page 31 Valve Setpoint Tolerances (TAC Nos. M84979 and M84980)," dated September 9, 1994 (ADAMS Accession Number ML021060010).
: 5. Letter from T. A. Beltz, NRC, to M. W. Rencheck, I&M, "Donald C. Cook Nuclear Plant, Unit 1 -Issuance of Amendment to Renewed Facility Operating License Regarding Use of The Westinghouse ASTRUM Large Break Loss-Of-Coolant Accident Analysis Methodology (TAC No. MD7556)," dated October 17, 2008 (ADAMS Accession Number ML082670351).
: 6. Letter from J. N. Jensen, I&M, to NRC Document Control Desk, "Donald C. Cook Nuclear Plant Unit 1, Docket No. 50-315, License Amendment Request Regarding Large Break Loss-of-Coolant Accident Analysis Methodology," dated December 27, 2007 (ADAMS Accession Number ML080090268).
: 7. WCAP-12945-P-A, "Westinghouse Code Qualification Document for Best Estimate Loss of Coolant Analysis," dated March 1998.8.0 PRECEDENTS As documented in Reference 5, the NRC has approved a plant-specific adaptation of the ASTRUM methodology for CNP Unit 1.The NRC has approved use of the ASTRUM methodology for the nuclear power plants listed below. The CNP Unit 2 license amendment request differs from these amendments in that I&M is requesting approval of a plant-specific adaptation of the NRC-approved ASTRUM methodology.
Joseph M. Farley, Units 1 and 2 (ADAMS Accession No. ML061810306)
Indian Point Nuclear Generating Plant, Unit 2 (ADAMS Accession No. ML061710291)
Diablo Canyon Power Plant, Unit 2 (ADAMS Accession No. ML063380020)
Prairie Island Nuclear Generating Plant, Units 1 and 2 (ADAMS Accession No. ML071230789)
Attachment to AEP-NRC-2009-23 DONALD C. COOK NUCLEAR PLANT UNIT 2 TECHNICAL SPECIFICATION PAGES MARKED TO SHOW CHANGES 3.4.1-1 3.4.1-2 3.5.2-1 5.6-3 RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS)3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below: a. Pressurizer pressure is greater than or equal to the limit specified in the COLR;b. RCS average temperature is less than or equal to the limit specified in the COLR; and c. RCS total flow rate is greater than or equal to the limit specified in the COLR. The minimum RCS total flow rate shall be>366,1034,000 gpm.APPLICABILITY:
MODE 1.-----------------------------
NOTE -----------------------
Pressurizer pressure limit does not apply during: a. THERMAL POWER ramp > 5% RTP per minute; or b. THERMAL POWER step> 10% RTP.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more RCS DNB A.1 Restore RCS DNB 2 hours parameters not within parameter(s) to within limit.limits.B. Required Action and B.1 Be in MODE 2. 6 hours associated Completion Time not met.Cook Nuclear Plant Unit 2 3.4.1-1 Amendment No. 269 RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is greater than or equal 12 hours to the limit specified in the COLR.SR 3.4.1.2 Verify RCS average temperature is less than or 12 hours equal to the limit specified in the COLR.SR 3.4.1.3 Verify RCS total flow rate is>'366,4An 4 gpm 12 hours and greater than or equal to the limit specified in the COLR.SR 3.4.1.4 ----------------------
NOTE ----------------
Not required to be performed until 24 hours after 90% RTP.Verify by precision heat balance that RCS total flow 24 months rate is.366,04. gpm and greater than or equal to the limit specified in the COLR.Cook Nuclear Plant Unit 2 3.4.1-2 Amendment No. 269 ECCS -Operating ,3.5.2 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)3.5.2 ECCS -Operating LCO 3.5.2 APPLICABILITY:
Two ECCS trains shall be OPERABLE.MODES 1, 2, and 3.ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more trains A.1 Restore train(s) to 72 hours inoperable-fer reasees OPERABLE status.o-the-r th-An Condition D).B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. AND B.2 Be in MODE 4. 12 hours C. Less than 100% of the C.1 Enter LCO 3.0.3. Immediately ECCS flow equivalent to a single OPERABLE ECCS train available-f-F resn thcr than D. O-ne or mo-re Safety D.1 Reduce THERMAL -heu.Injection (SI) System POWER tf330, MWt.cro)Ss tie valves closed.Cook Nuclear Plant Unit 2 3.5.2-1 Amendment No. 269 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
: 5. LCO 3.1.6, "Control Bank Insertion Limits";6. LCO 3.2.1, "Heat Flux Hot Channel Factor (FQ(Z))";7. LCO 3.2.2, "Nuclear Enthalpy Rise Hot Channel Factor (FNH)";8. LCO 3.2.3, "AXIAL FLUX DIFFERENCE (AFD)";9. LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation," Functions 6 and 7 (Overtemperature AT and Overpower AT, respectively)
Allowable Value parameter values;10. LCO 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits"; and 11. LCO 3.9.1, "Boron Concentration." b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
: 1. WCAP-9272-P-A, "Westinghouse Reload Safety Evaluation Methodology," (Westinghouse Proprietary);
: 2. WCAP-8385, "Power Distribution Control and Load Following Procedures
-Topical Report," (Westinghouse Proprietary);
: 3. WCAP-1 0216-P-A, "Relaxation of Constant Axial Offset Control/FQ Surveillance Technical Specification," (Westinghouse Proprietary);
: 4. WGAP-1 0266 P A, "The 1981 VeriGRn of Westing'hou.
,-, U BAS, H Code,",,, (^ , ..... , , -- ...... dPlant-specifi adaptation of WCAP-1 6009-P-A, "Realistic Large-Break LOCA&#xfd;Evaluation Methodology Using the Automated Statistical Treatment ol Uncertainty Method (ASTRUM)," (Westinghouse Proprietary);
: 5. WCAP-1261 0-P-A, "VANTAGE+
Fuel Assembly Reference Core Report," (Westinghouse Proprietary);
: 6. WCAP-8745-P-A, "Design Bases for the Thermal OverpowerAT and Thermal OvertemperatureAT Trip Functions," (Westinghouse Proprietary);
and 7. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," (Westinghouse Proprietary).
Cook Nuclear Plant Unit 2 5.6-3 Amendment No. 26-9, 270}}

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