CNS-15-058, Proposed Technical Specifications (TS) Amendments, TS 3.4.1, Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits

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Proposed Technical Specifications (TS) Amendments, TS 3.4.1, Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits
ML15168A009
Person / Time
Site: Catawba  Duke Energy icon.png
Issue date: 06/12/2015
From: Henderson K
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CNS-15-058
Download: ML15168A009 (58)
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Text

Kelvin Henderson DUKE Vice President Catawba Nuclear Station

ENERGY, Duke Energy CNOIVP 1 4800 Concord Road York, SC 29745 CNS-15-058 o: 803.701.4251 f: 803.701.3221 June 12, 2015 10 CFR 50.90 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Proposed Technical Specifications (TS) Amendments TS 3.4.1, Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits Pursuant to 10 CFR 50.90, Duke Energy is requesting amendments to the Catawba Units 1 and 2 Facility Operating Licenses (FOLs) NPF-35 and NPF-52, respectively, and the associated TS. This request is to modify the subject TS to allow lower minimum values of RCS flowrate. Specifically, Duke Energy proposes to modify TS Table 3.4.1-1, "RCS DNB Parameters", Parameter 3, "RCS Total Flow Rate", Limit as follows:

Unit 1: From '5 388,000 gpm and > the limit specified in the COLR (Unit 1)", to ">

384,000 gpm and > the limit specified in the COLR (Unit 1)"

Unit 2: From "> 390,000 gpm and > the limit specified in the COLR (Unit 2)", to ">

387,000 gpm and > the limit specified in the COLR (Unit 2)"

The contents of this amendment request package are as follows:

Attachment 1 provides the technical and regulatory evaluations associated with the proposed changes. Attachment 2 provides the marked-up TS page showing the proposed changes. The retyped (clean) TS page will be provided to the NRC immediately prior to issuance of the approved amendments. No changes to the corresponding TS Bases are required in conjunction with this amendment request.

Duke Energy is requesting NRC review and approval of this amendment request within one year of the date of submittal.

Duke Energy is requesting a 60-day implementation period in conjunction with this amendment request. Implementation of the approved amendments will require changes to the Updated Final Safety Analysis Report (UFSAR). Necessary UFSAR changes will be submitted to the NRC in accordance with 10 CFR 50.71(e), with approved exemptions.

www.duke-energy.com

I U.S. Nuclear Regulatory Commission Page 2 June 12, 2015 There are no regulatory commitments being made in conjunction with this amendment request.

In accordance with Duke Energy administrative procedures and the Quality Assurance Program Topical Report, this amendment request has been previously reviewed and approved by the Catawba Plant Operations Review Committee.

Pursuant to 10 CFR 50.91, a copy of this amendment request is being sent to the appropriate State of South Carolina official.

Inquiries on this matter should be directed to L.J. Rudy at (803) 701-3084.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on June 12, 2015.

Very truly yours, Kelvin Henderson Vice President, Catawba Nuclear Station LJPRs Attachments

A U.S. Nuclear Regulatory Commission Page 3 June 12, 2015 xc (with attachments):

V.M. McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 G.A. Hutto III Senior Resident Inspector (Catawba)

U.S. Nuclear Regulatory Commission Catawba Nuclear Station G.E. Miller (addressee only)

NRC Project Manager (Catawba)

U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 S.E. Jenkins Manager Radioactive & Infectious Waste Management Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull St.

Columbia, SC 29201

ATTACHMENT 1 TECHNICAL AND REGULATORY EVALUATIONS

-K

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Proposed Technical Specifications (TS) Amendments TS 3.4.1, Reactor Coolant System (RCS) Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits

1. DESCRIPTION
2. PROPOSED CHANGES
3. BACKGROUND
4. TECHNICAL EVALUATION
5. REGULATORY EVALUATION 5.1 Applicable Regulatory Requirements/Criteria 5.2 Precedent 5.3 No Significant Hazards Consideration 5.4 Conclusions
6. ENVIRONMENTAL CONSIDERATION
7. REFERENCES Attachment 1 Page 1
1. DESCRIPTION Pursuant to 10 CFR 50.90, Duke Energy is requesting amendments to the Catawba Units 1 and 2 Facility Operating Licenses (FOLs) NPF-35 and NPF-52, respectively, and the associated TS. This request is to modify the subject TS to allow lower minimum values of RCS flowrate. Specifically, Duke Energy proposes to modify TS Table 3.4.1-1, "RCS DNB Parameters", Parameter 3, "RCS Total Flow Rate", Limit as follows:

Unit 1: From "> 388,000 gpm and > the limit specified in the COLR (Unit 1)", to ">

384,000 gpm and > the limit specified in the COLR (Unit 1)"

Unit 2: From "> 390,000 gpm and > the limit specified in the COLR (Unit 2)", to '>

387,000 gpm and > the limit specified in the COLR (Unit 2)"

Attachment 1 Page 2

2. PROPOSED CHANGES The proposed changes reduce the required TS 3.4.1 minimum measured RCS flow rate from 388,000 gpm to 384,000 gpm for Catawba Unit 1 and from 390,000 gpm to 387,000 gpm for Catawba Unit 2. Although the methods of evaluation are similar for both units, each is evaluated separately below for clarity.

The safety and quality of operation of Catawba Nuclear Station will not be compromised by the implementation of these proposed amendments.

The methodologies used in the evaluations that follow are documented in the approved Core Operating Limits Report (COLR) methodology reports listed below:

1) DPC-NE-3001-P-A, "Multidimensional Reactor Transients and Safety Analysis Physics Parameter Methodology", Revision Oa
2) DPC-NE-3000-P-A, "Thermal-Hydraulic Transient Analysis Methodology", Revision 5a
3) DPC-NE-3002-A, "FSAR Chapter 15 System Transient Analysis Methodology",

Revision 4b

4) DPC-NE-2005-P-A, "Thermal Hydraulic Statistical Core Design Methodology",

Revision 4a Tables for all Updated Final Safety Analysis Report (UFSAR) transients discussed within each section below are included below. Tables are included for Unit 1 (Table 1) and for Unit 2 (Table 2) separately. These tables summarize the disposition of each UFSAR transient discussed below. The tables also include references for the various approved methodologies used for each analysis.

Attachment 1 Page 3

3. BACKGROUND Over a number of years, a decrease in RCS flow has been observed at Catawba Unit 1 and Catawba Unit 2. The actual amount of the flow decrease has been small, but nevertheless it has resulted in a loss of margin relative to the minimum values allowed by Catawba TS 3.4.1.

RCS flow data for the last eight fuel cycles spanning approximately eleven years is shown in the tables below. Data shown are the average values over the first month of full power operation for each fuel cycle.

Fuel design has remained constant throughout this period with the use of Westinghouse Robust Fuel Assembly (RFA) and RFA-2.

On Unit 1, the Loop A reactor coolant pump was replaced during the End of Cycle (EOC) 19 Refueling Outage (RFO). The new pump resulted in a significant increase in flow in Loop A. Loop A had experienced a significant decrease in flow in the previous fuel cycle, so part of the increase for Cycle 20 was likely attributed to the recovery of some of the previous flow decrease.

During Unit 2 Cycle 18, a Loop B flow transmitter was recalibrated which resulted in approximately a 500 gpm increase in Loop B and total flow rate. This is reflected in the Cycle 19 flow values.

Crud deposits on the fuel cladding and possibly the steam generator tubes appears to be a significant contributor to the observed changes in RCS flow. A Unit 1 reactor trip in Cycle 16 resulted in a step increase in RCS flow following the return to power, indicating that some crud had been removed from the fuel cladding and/or the steam generator tubes. A Unit 1 reactor trip in Cycle 20 did not have the same effect.

Zinc injection into the RCS began on Unit 1 during Cycle 17 and on Unit 2 during Cycle

15. Zinc is added to reduce outage dose rates for steam generator work by replacing activated material in the steam generator tubes with zinc. It was expected that some of this displaced material would adhere to the fuel cladding; ultrasonic fuel cleaning during outages was started before beginning zinc injection. Following zinc injection, RCS flow increased in the last thirty to sixty days of the fuel cycles (for Unit 1 Cycles 17, 18, and 19, and for Unit 2 Cycles 15, 16, 17, and 18). The flow increases along with radiochemistry analysis indicate that some crud is being released from the fuel cladding surface at the end of the fuel cycle. This phenomenon has not been observed for the most recent fuel cycles, however. The most recent fuel cycles also have not experienced significant decreases in RCS flow; in fact, RCS flow has increased from Unit 1 Cycle 19 and Unit 2 Cycle 18.

Attachment 1 Page 4

I Catawba Unit I Percent of TS Required Flow (Last Three RCS Flow Data, gpm Change from Previous Cycle, gpm Cycles)

Loop Loop Loop Loop Cycle Total Loop A Loop B Loop C Loop D Total A B C D C1C15 391,405 98,036 98,049 97,489 97,836 CLC16 390,974 97,846 97,909 97,485 97,741 -431 -190 -140 -4 -95 C1C17 390,224 97,844 97,734 97,232 97,446 -750 -2 -175 -253 -295 C1C18 389,948 97,840 97,564 97,059 97,491 -275 -3 -170 -173 45 CIC19 389,794 97,105 97,692 97,213 97,774 -155 -736 128 154 283 100.3 CLC20 390,731 98,110 97,730 97,240 .97,646 938 1,005 39 27 -127 100.6 CLC21 390,523 98,083 97,561 97,253 97,611 -209 -26 -169 13 -36 100.6 C1C22 390,860 98,221 97,626 97,350 97,653 337 138 66 97 42 Total Change: -545 185 -423 -138 -183 Catawba Unit 2 Percent of TS Required Flow (Last Three RCS Flow Data, gpm Change from Previous Cycle, gpm Cycles)

Loop Loop Loop Loop Cycle Total Loop A Loop B Loop C Loop D Total A B C D C2C13 394,685 97,285 98,696 99,704 99,009 C2C14 394,631 97,666 98,474 99,672 98,837 -54 380 -223 -32 -172 C2C15 394,563 97,604 98,336 99,678 98,941 -68 -62 -138 6 104 C2C16 394,396 97,535 98,243 99,691 98,939 -167 -69 -93 13 -2 C2C17 393,091 97,184 98,055 99,210 98,640 -1305 -351 -187 -480 -298 C2C18 391,846 97,169 97,480 98,591 98,618 -1245 -16 -575 -620 -22 100.5 C2C19 393,514 97,287 98,249 99,178 98,807 1668 119 769 587 188 100.8 C2C20 393,807 97,409 98,160 99,578 98,668 293 122 -89 400 -138 100.9 C2C21 393,761 97,437 98,287 99,237 98,792 -49 27 127 -342 123 Total Change: -925 152 -409 -468 -217 NOTE: Total flow does not equal the sum of the loop flows because of the method used to retrieve archived data.

Attachment 1 Page 5

In summary, the following insights were noted concerning the observed RCS flow decrease on Unit 1 and Unit 2:

  • The flow decrease is not due to drifting components or an instrumentation calibration issue.
  • The flow decrease could not be confirmed with an independent indication due to the small magnitude of the change as compared to the data resolution of other indications (e.g., RCS loop delta temperature and reactor coolant pump motor parameters).
  • The following potential causes were evaluated, and as noted, eliminated.

Reactor coolant boron concentration changes (eliminated)

Boron concentration effects in the elbow tap tubing runs (eliminated)

RCS chemistry transients (eliminated)

RCS chemistry pH and zinc addition programs (eliminated)

Shutdown crud burst pump configuration with hydrogen peroxide addition (eliminated)

Steam generator outage eddy current inspections (eliminated)

Core effects from bypass flow changes or temperature increases (eliminated)

Density effects from elbow tap tubing run temperature differences between the high and low taps (eliminated)

Nitrogen bubbles in the high tap tubing runs due to inadequate venting following nitrogen purge evolutions (eliminated)

Crud buildup in the tubing runs (eliminated)

RCS flow may eventually approach the minimum values allowed by Catawba TS 3.4.1.

The purpose of justifying a lower RCS flow for Unit 1 and Unit 2 is to accommodate any additional flow decrease that may occur during the course of future fuel cycles. The supporting analyses for this LAR submittal substantiate a reduction in the minimum RCS flow allowed by TS 3.4.1 to 384,000 gpm for Unit 1 and to 387,000 gpm for Unit 2.

Attachment 1 Page 6

I

4. TECHNICAL EVALUATION Basis for Proposed Changes (Catawba Unit 1)

The following summarizes the effect on UFSAR analyses of the requested reduction in Catawba Unit 1 required minimum measured RCS flow rate from 388,000 gpm to 384,000 gpm.

For the purposes of this evaluation, the UFSAR accident analyses are divided into three categories. For each category, the evaluation of the proposed RCS flow rate reduction is approached differently.

Category 1: Transients bounded by current RCS flow assumption In the first category of analyses, the RCS flow rate assumed in the current analysis of record (AOR) is based on either the mechanical design flow of 420,000 gpm (where maximum RCS flows are conservative) or the thermal design flow of 382,000 gpm (where minimum RCS flows are conservative). Therefore, a change in the minimum RCS total flow rate limit from 388,000 gpm to 384,000 gpm would have no impact on the analysis.

1A. LOCA Blowdown Reactor Vessel and Loop Forces (UFSAR Section 3.6.4.1 and 3.9.1.5) lB. Containment Functional Design (UFSAR Section 6.2.1) 1C. Feedwater System Malfunction Causing an Increase in Feedwater Flow (zero power case) (UFSAR Section 15.1.2) 1D. Turbine Trip (UFSAR Section 15.2.3) 1E. Loss of Non-Emergency AC Power to the Station Auxiliaries (UFSAR Section 15.2.6) 1F. Anticipated Transients Without Trip (UFSAR Section 15.8)

Category 2: Inapplicable transients/transients bounded by another transient/transients insensitive to RCS flow In the second category of analyses, it is determined that the analysis itself is not applicable to Catawba (e.g., BWR transients) or is addressed in the UFSAR as being bounded by other analyzed transients. The proposed RCS flow reduction has no impact on these analyses.

In addition, many of the analyses are not sensitive to the reduction in RCS total flow, and a conclusion is reached that the 4000 gpm flow reduction does not affect the transient. Therefore, the following events are either not applicable, bounded by another accident, or are unaffected by the decrease of minimum RCS total flow rate to 384,000 gpm because RCS flow is not a factor in the event:

2A. Feedwater System Malfunctions that Result in a Reduction in Feedwater Temperature (UFSAR Section 15.1.1) The decrease in feedwater temperature transient is less severe than the increase in feedwater flow event (Section 15.1.2) or the increase in secondary steam flow event (Section 15.1.3). Based on results Attachment 1 Page 7

presented in Sections 15.1.2 and 15.1.3, the applicable acceptance criteria for the decrease in feedwater temperature event have been met.

2B. Inadvertent Opening of a Steam Generator Relief Valve or Safety Valve (UFSAR Section 15.1.4) The analyses performed assuming a rupture of a main steam line are given in Section 15.1.5. The main steam line break analysis bounds the inadvertent opening of a steam generator relief or safety valve analysis.

2C. Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow (UFSAR Section 15.2.1, not applicable) 2D. Loss of External Load (UFSAR Section 15.2.2) A loss of external load event results in a Nuclear Steam Supply System (NSSS) transient that is less severe than a turbine trip event (see Section 15.2.3).

2E. Inadvertent Closure of Main Steam Isolation Valves (UFSAR Section 15.2.4) The longer closing time of the MSIVs, relative to the turbine stop valve closure time, offsets the effects of the smaller steam piping volume, and therefore the MSIV closure event is less severe than a turbine trip event. Turbine trips are discussed in Section 15.2.3.

2F. Loss of Condenser Vacuum and Other Event Causing a Turbine Trip (UFSAR Section 15.2.5) Since steam dump is assumed not to be available in the turbine trip analysis, no additional adverse effects would result if the turbine trip were caused by loss of condenser vacuum. Therefore, the analysis results and conclusions contained in Section 15.2.3 apply to loss of condenser vacuum.

2G. Feedwater System Pipe Break (UFSAR Section 15.2.8) (short-term) The short-term results for the feedwater line break analysis are bounded by the analysis for a loss of normal feedwater flow (UFSAR Section 15.2.7).

2H. Reactor Coolant Pump Shaft Break (UFSAR Section 15.3.4) The consequences of a reactor coolant pump shaft break are similar to those calculated for the locked rotor incident (see UFSAR Section 15.3.3).

21. Dropped Rod Cluster Control Assembly (RCCA) Bank (UFSAR Section 15.4.3b)

The results for a dropped RCCA bank are bounded by the analysis presented for one or more dropped rods.

2J. Statically Misaligned RCCA (UFSAR Section 15.4.3c) The results for a statically misaligned RCCA are bounded by the analysis presented for one or more dropped rods (see UFSAR Section 15.4.3a) and for the Single RCCA Withdrawal (UFSAR Section 15.4.3d). DNBR does not fall below the limit value for the RCCA misalignment incident.

2K. BWR Transient (UFSAR Section 15.4.5, not applicable) 2L, Chemical Volume and Control System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant System (UFSAR Section 15.4.6, unaffected by change in RCS flow) 2M. Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (UFSAR Section 15.4.7, unaffected by change in RCS flow) 2N. BWR Transient (UFSAR Section 15.4.9, not applicable)

20. Inadvertent Operation of Emergency Core Cooling System (ECCS) During Power Operation (UFSAR Section 15.5.1) The DNB result of this transient is bounded by the inadvertent opening of a pressurizer safety or relief valve transient (see UFSAR Section 15.6.1).

2P. Chemical Volume and Control System Malfunction that Increases Reactor Coolant Inventory (UFSAR Section 15.5.2) An increase in reactor coolant inventory which results from the addition of cold, unborated water to the RCS is analyzed in Attachment 1 Page 8

Section 15.4.6. An increase in reactor coolant inventory which results from the injection of highly borated water into the RCS is analyzed in Section 15.5.1.

2Q. A Number of BWR Transients (UFSAR Section 15.5.3, not applicable) 2R. Break in Instrument Line or Other Lines from Reactor Coolant Pressure Boundary that Penetrate Containment (UFSAR Section 15.6.2) The limiting case for this accident is a break occurring in the Chemical Volume and Control System (CVCS).

The transient results would be unaffected by the proposed change in RCS flow.

2S. Spectrum of BWR Steam System Piping Failures Outside Containment (UFSAR Section 15.6.4, not applicable) 2T. A Number of BWR Transients (UFSAR Section 15.6.6, not applicable) 2U. All Radiological Accidents (UFSAR Sections 15.7, 15.9, and 15.10) These accidents are not impacted by changes in RCS flow.

Cateqory 3: Evaluated Transients The following summarizes the conclusions for the final category of safety analyses that have been re-evaluated with the reduced minimum RCS total flow rate of 384,000 gpm.

Each of the safety analyses was individually reviewed to determine if the results of the analyses are affected by a 4000 gpm reduction in RCS total flow.

In a subgroup of this category of analyses, the current AOR assumes a conservatively high core bypass flow of 8.5% of the total RCS flow. In the AOR for these transients, the flow passing through the core is 355,020 gpm (388,000 gpm minus the 8.5% bypass flow). For efficiency, each of these AORs was performed to bound all Duke Energy units with the Babcock and Wilcox feedring steam generator design (McGuire Units 1 and 2 and Catawba Unit 1). This 8.5% bypass flow assumption conservatively bounds the unit-specific core bypass flows for all three of these units. Therefore, each accident has a single AOR for this category of transients, which applies to all three of these units.

However, the unit-specific calculated core bypass flow for Catawba Unit 1 is 6.49%. At a total RCS flow rate of 384,000 gpm, the flow passing through the core is 359,078 gpm (384,000 gpm minus the 6.49% core bypass flow). These flow calculations are summarized below:

Catawba Unit 1 Total RCS Core Bypass Flow Net RCS Flow Flow Assumed Through Core

[gpm] [%] [gpm]

Existing AOR 388,000 8.5 355,020 With revised RCS 384,000 6.49 359,078 flow For the analyses listed below, core cooling is a concern and lower core flows are conservative. For these events, when the Catawba Unit 1 specific core bypass flow is considered, the existing AOR (with a net core flow of 355,020 gpm) remains bounding for Catawba Unit 1 with an RCS flow rate of 384,000 gpm (net core flow of 359,078 gpm). Therefore, no re-analysis was necessary since the existing AOR remains bounding:

3A. Feedwater Malfunction Causing an Increase in Feedwater Flow (full power case)

(UFSAR Section 15.1.2)

Attachment 1 Page 9

3B. Partial Loss of Coolant Flow (UFSAR Section 15.3.1) 3C. Complete Loss of Coolant Flow (UFSAR Section 15.3.2) 3D. Reactor Coolant Pump Shaft Seizure - Locked Rotor (DNB) (UFSAR Section 15.3.3) 3E. Uncontrolled RCCA Bank Withdrawal at Power (DNB) (UFSAR Section 15.4.2) 3F. Dropped RCCA Rod (UFSAR Section 15.4.3a) 3G. Single Rod Withdrawal (UFSAR Section 15.4.3d) 3H. Spectrum of Rod Cluster Control Assembly Ejection Accidents (UFSAR Section 15.4.8)

31. Inadvertent Opening of a Pressurizer Safety or Relief Valve (UFSAR Section 15.6.1) 3J. Steam Generator Tube Rupture (DNB) (UFSAR Section 15.6.3)

For the balance of the analyses, other evaluations were performed as described below:

3K. Excessive Increase in Secondary Steam Flow (UFSAR Section 15.1.3) 3L. Steam System Piping Failure (UFSAR Section 15.1.5) 3M. Loss of Normal Feedwater Flow (UFSAR Section 15.2.7) 3N. Feedwater System Pipe Break (long-term) (UFSAR Section 15.2.8)

30. Reactor Coolant Pump Shaft Seizure - Locked Rotor (peak RCS pressure)

(UFSAR Section 15.3.3) 3P. Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition (UFSAR Section 15.4.1) 3Q. Uncontrolled RCCA Bank Withdrawal at Power (peak RCS pressure) (UFSAR Section 15.4.2) 3R. Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (UFSAR Section 15.4.4) 3S. Steam Generator Tube Rupture (UFSAR Section 15.6.3) 3T. Loss of Coolant Accidents (UFSAR Section 15.6.5)

These events are summarized below:

3K. Excessive Increase in Secondary Steam Flow (UFSAR Section 15.1.3)

The increase in steam flow accident was originally evaluated at an RCS flow of 382,000 gpm, and it was concluded that approximately 6% margin to the overpower delta temperature trip existed during the event. Likewise, the cases analyzed at 388,000 gpm also concluded that at least 6% margin to the overpower delta temperature trip existed. Therefore, the analysis is not sensitive to RCS flow.

The proposed minimum RCS flow rate of 384,000 gpm is determined to be acceptable.

3L. Steam System Piping Failure (UFSAR 15.1.5)

This event was re-analyzed at a minimum RCS flow rate of 384,000 gpm to demonstrate DNB does not occur. The minimum DNBR was calculated to be 1.784, which is well above the W-3S Critical Heat Flux (CHF) correlation limit of 1.45.

3M. Loss of Normal Feedwater Flow (UFSAR Section 15.2.7)

The Loss of Normal Feedwater Flow short-term core cooling analysis was performed at the proposed RCS flow rate of 384,000 gpm with a conservatively Attachment 1 Page 10

high 8.5% core bypass flow assumption. The minimum DNBR was calculated to be 2.131, which is well above the WRB-2M CHF correlation limit of 1.45.

The long-term core cooling analyses have significant margin to the acceptance criteria, which is based on maintaining adequate decay heat removal. The long-term core cooling AOR assumes an RCS flow rate of 388,000 gpm, with auxiliary feedwater flows conservatively bounding the actual Catawba capacity by at least 10%. In the analysis, subcooling of at least 40 degrees F was maintained throughout the transient. Therefore, it is concluded that the proposed reduction in RCS flow rate to 384,000 gpm will have an inconsequential impact on this transient.

3N. Feedwater System Pipe Break (UFSAR Section 15.2.8)

The feedwater line break short-term analysis is bounded by the analysis documented in UFSAR Section 15.2.7, which was satisfactorily analyzed at 384,000 gpm.

The long-term core cooling analysis indicates adequate hot leg subcooling ensuring long-term core cooling capability. The key parameters in this analysis are the heat balance and the ability of auxiliary feedwater to maintain adequate hot leg subcooling. The proposed change in RCS flow does not alter either the heat sources (core decay heat and reactor coolant pump heat) or heat sink parameters (auxiliary feedwater flow and temperature). Therefore, it is concluded that the current AOR will not be significantly impacted by the proposed change in RCS flow rate.

30. Reactor Coolant Pump Shaft Seizure - Locked Rotor (peak RCS pressure)

(UFSAR Section 15.3.3)

This event was re-analyzed for a minimum RCS flow rate of 384,000 gpm to calculate the peak RCS pressure for this accident. The resulting peak RCS pressure of 2529 psig is well below the design value of 2735 psig.

3P. Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition (UFSAR Section 15.4.1)

This analysis addresses adequate core cooling (Case 1) and peak RCS pressure (Case 2) acceptance criteria. Case 1 assumes RCS flow with three reactor coolant pumps operational based on nominal flow of 388,000 gpm. The delivered flow per pump increases to between 104% and 107% with one pump not running.

The 1% reduction in nominal full flow to 384,000 gpm is therefore adequately accommodated by the three-pump assumption. Case 2 was re-analyzed with a total assumed RCS flow of 384,000 gpm. The results of the current AOR show about 130 psi of margin relative to the design value of 2735 psig.

Therefore, it can be concluded that the current AORs for this transient are not impacted by the proposed reduction in RCS flow.

3Q. Uncontrolled RCCA Bank Withdrawal at Power (peak RCS pressure) (UFSAR Section 15.4.2)

Attachment 1 Page 11

a a This event was re-analyzed for a minimum RCS flow rate of 384,000 gpm to calculate the peak RCS pressure for this accident. The resulting peak RCS pressure of 2708 psig is below the design value of 2735 psig.

3R. Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (UFSAR Section 15.4.4)

This analysis is initiated from an assumed initial power level of 50%. The analysis assumes RCS flow with three reactor coolant pumps initially operating. The current AOR is based on the calculated three-pump flow, starting from a nominal full power and four-pump flow of 388,000 gpm. A historical analysis with different fuel types was based on a nominal full power RCS flow of 382,000 gpm. Both cases showed similar core responses and considerable margin to the limiting DNBR in both cases. Based on the similarity of results between the two cases, it is concluded that the analysis results are insensitive to RCS flow. The proposed reduction in total RCS flow to 384,000 gpm will have no significant impact on the analysis results.

3S. Steam Generator Tube Rupture (UFSAR Section 15.6.3)

This event is examined for DNB, radiological consequences, and steam generator overfill. The overfill analysis determined the assumed RCS flow rate to be inconsequential. The DNB analysis was performed at 388,000 gpm and is evaluated above (item 3J). The dose input analysis was performed at 390,000 gpm plus uncertainty. Prior to trip, the marginal reduction in flow has an inconsequential impact on the analysis. Upon manual reactor trip, a loss of offsite power is assumed to trip the reactor coolant pumps and the impact of the small change in initial RCS flow has no impact on the balance of the transient.

3T. Loss of Coolant Accidents (UFSAR Section 15.6.5)

The Large Break LOCA (LBLOCA) analysis was evaluated by Westinghouse for the proposed reduction in RCS total flow rate. This analysis, which was applicable for both Catawba units, determined that the variations in the global model calculations are such that the 9 5 th percentile peak clad temperature is not impacted.

The significant factors in the Small Break LOCA (SBLOCA) analysis are decay heat, RCS mass, break flow, and ECCS delivery. Three of these variables are completely unrelated to initial RCS flow, and the fourth (RCS mass) is insignificantly affected. Changes in initial RCS flow may possibly change the thermal mass in the steam generators, and thus the operating pressure, which would in turn affect the time at which the safety valves would lift. However, this second order effect is considered insignificant to the SBLOCA outcome.

Refer to the summary Table 1 at the end of this attachment for a synopsis of the analysis results.

Attachment 1 Page 12

Basis for Proposed Changes (Catawba Unit 2)

The following summarizes the effect on UFSAR analyses of the requested reduction in Catawba Unit 2 required minimum measured RCS flow rate from 390,000 gpm to 387,000 gpm.

For the purposes of this evaluation, the UFSAR accident analyses are divided into three categories. For each category, the evaluation of the proposed RCS flow rate reduction is approached differently.

Category 1: Transients bounded by current RCS flow assumption In the first category of analyses, the RCS flow rate assumed in the current AOR is based on either the mechanical design flow of 420,000 gpm (where maximum RCS flows. are conservative) or the thermal design flow of 382,000 gpm (where minimum RCS flows are conservative). Therefore, a change in the minimum RCS total flow rate limit from 390,000 gpm to 387,000 gpm would have no impact on the analysis.

4A. LOCA Blowdown Reactor Vessel and Loop Forces (UFSAR Section 3.6.4.1 and 3.9.1.5) 4B. Containment Functional Design (UFSAR Section 6.2.1) 4C. Feedwater System Malfunction Causing an Increase in Feedwater Flow (zero power case) (UFSAR Section 15.1.2) 4D. Excessive Increase in Secondary Steam Flow (UFSAR Section 15.1.3) 4E. Turbine Trip (UFSAR Section 15.2.3) 4F. Loss of Non-Emergency AC Power to the Station Auxiliaries (UFSAR Section 15.2.6) 4G. Loss of Normal Feedwater Flow (long-term analysis) (UFSAR Section 15.2.7) 4H. Feedwater System Pipe Break (long-term analysis) (UFSAR Section 15.2.8)

41. Anticipated Transients Without Trip (UFSAR Section 15.8)

Category 2: Inapplicable transients/transients bounded by another transient/transients insensitive to RCS flow In the second category of analyses, it is determined that the analysis itself is not applicable to Catawba (e.g., BWR transients) or is addressed in the UFSAR as being bounded by other analyzed transients. The proposed RCS flow reduction has no impact on these analyses.

In addition, many of the analyses are not sensitive to the reduction in RCS total flow, and a conclusion is reached that the 3000 gpm flow reduction does not affect the transient. Therefore, the following events are either not applicable, bounded by another accident, or are unaffected by the decrease of minimum RCS total flow rate to 387,000 gpm because RCS flow is not a factor in the event:

5A. Feedwater System Malfunctions that Result in a Reduction in Feedwater Temperature (UFSAR Section 15.1.1) The decrease in feedwater temperature transient is less severe than the increase in feedwater flow event (Section 15.1.2) or the increase in secondary steam flow event (Section 15.1.3). Based on results Attachment 1 Page 13

presented in Sections 15.1.2 and 15.1.3, the applicable acceptance criteria for the decrease in feedwater temperature event have been met.

5B. Inadvertent Opening of a Steam Generator Relief Valve or Safety Valve (UFSAR Section 15.1.4) The analyses performed assuming a rupture of a main steam line are given in Section 15.1.5. The main steam line break analysis bounds the inadvertent opening of a steam generator relief or safety valve analysis.

5C. Steam Pressure Regulator Malfunction or Failure that Results in Decreasing Steam Flow (UFSAR Section 15.2.1, not applicable) 5D. Loss of External Load (UFSAR Section 15.2.2) A loss of external load event results in an NSSS transient that is less severe than a turbine trip event (see Section 15.2.3).

5E. Inadvertent Closure of Main Steam Isolation Valves (UFSAR Section 15.2.4) The longer closing time of the MSIVs, relative to the turbine stop valve closure time, offsets the effects of the smaller steam piping volume, and therefore the MSIV closure event is less severe than a turbine trip event. Turbine trips are discussed in Section 15.2.3.

5F. Loss of Condenser Vacuum and Other Event Causing a Turbine Trip (UFSAR Section 15.2.5) Since steam dump is assumed not to be available in the turbine trip analysis, no additional adverse effects would result if the turbine trip were caused by loss of condenser vacuum. Therefore, the analysis results and conclusions contained in Section 15.2.3 apply to loss of condenser vacuum.

5G. Feedwater System Pipe Break (UFSAR Section 15.2.8) (short-term) The short-term results for the feedwater line break analysis are bounded by the analysis for a loss of normal feedwater flow (UFSAR Section 15.2.7).

5H. Reactor Coolant Pump Shaft Break (UFSAR Section 15.3.4) The consequences of a reactor coolant pump shaft break are similar to those calculated for the locked rotor incident (see Section 15.3.3).

51. Dropped RCCA Bank (UFSAR Section 15.4.3b) The results for a dropped RCCA bank are bounded by the analysis presented for one or more dropped rods.

5J. Statically Misaligned RCCA (UFSAR Section 15.4.3c) The results for a statically misaligned RCCA are bounded by the analysis presented for one or more dropped rods (see UFSAR Section 15.4.3a) and for the Single RCCA Withdrawal (UFSAR Section 15.4.3d). DNBR does not fall below the limit value for the RCCA misalignment incident.

5K. BWR Transient (UFSAR Section 15.4.5, not applicable) 5L. Chemical Volume and Control System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant System (UFSAR Section 15.4.6, unaffected by change in RCS flow) 5M. Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (UFSAR Section 15.4.7, unaffected by change in RCS flow) 5N. BWR Transient (UFSAR Section 15.4.9, not applicable)

50. Inadvertent Operation of ECCS During Power Operation (UFSAR Section 15.5.1)

The DNB results of this transient are bounded by the inadvertent opening of a pressurizer safety or relief valve transient (see UFSAR Section 15.6.1).

5P. Chemical Volume and Control System Malfunction that Increases Reactor Coolant Inventory (UFSAR Section 15.5.2) An increase in reactor coolant inventory which results from the addition of cold, unborated water to the RCS is analyzed in Section 15.4.6. An increase in reactor coolant inventory which results from the injection of highly borated water into the RCS is analyzed in Section 15.5.1.

5Q. A Number of BWR Transients (UFSAR Section 15.5.3, not applicable)

Attachment 1 Page 14

5R. Break in Instrument Line or Other Lines from Reactor Coolant Pressure Boundary that Penetrate Containment (UFSAR Section 15.6.2) The limiting case for this accident is a break occurring in the CVCS. The transient results would be unaffected by the proposed change in RCS flow.

5S. Spectrum of BWR Steam System Piping Failures Outside Containment (UFSAR Section 15.6.4, not applicable) 5T. A Number of BWR Transients (UFSAR Section 15.6.6, not applicable) 5U. All Radiological Accidents (UFSAR Sections 15.7, 15.9, and 15.10) These accidents are not impacted by changes in RCS flow.

Categqory 3: Evaluated Transients The following summarizes the results of the final category of safety analyses that have been either re-evaluated or re-analyzed with the reduced minimum RCS total flow rate of 387,000 gpm. Each. of the safety analyses was individually reviewed to determine if the results of the analyses are affected by a 3000 gpm reduction in RCS total flow.

In a subgroup of this category of analyses, the current AOR for Catawba Unit 2 assumes a conservatively high core bypass flow of 7.5% of the total RCS flow. In the AOR for each of these transients, the flow passing through the core is 360,750 gpm (390,000 gpm minus the 7.5% bypass flow).

However, the unit-specific calculated core bypass flow for Catawba Unit 2 is 6.71%. At a total RCS flow rate of 387,000 gpm, the flow passing through the core is 361,032 gpm (387,000 gpm minus the 6.71% core bypass flow). These flow calculations are summarized below:

Catawba Unit 2 Total RCS Flow Core Bypass Flow Net RCS Flow

[gpm] Assumed Through Core

[%] [gpm]

Existing AOR 390,000 7.5 360,750 With revised RCS 387,000 6.71 361,032 flow For the analyses listed below, core cooling is a concern and lower core flows are conservative. For these events, when the Catawba Unit 2 specific core bypass flow is considered, the existing AOR (with a net core flow of 360,750 gpm) remains bounding for Catawba Unit 2 with an RCS flow rate of 387,000 gpm (net core flow of 361,032 gpm). Therefore, no re-analysis was necessary since the existing AOR remains bounding:

6A. Feedwater Malfunction Causing an Increase in Feedwater Flow (full power case)

(UFSAR Section 15.1.2) 6B. Loss of Normal Feedwater Flow (short-term analysis) (UFSAR Section 15.2.7) 6C. Partial Loss of Coolant Flow (UFSAR Section 15.3.1) 6D. Complete Loss of Coolant Flow (UFSAR Section 15.3.2) 6E. Reactor Coolant Pump Shaft Seizure - Locked Rotor (DNB) (UFSAR Section 15.3.3) 6F. Uncontrolled RCCA Bank Withdrawal at Power (DNB) (UFSAR Section 15.4.2) 6G. Dropped RCCA Rod (UFSAR Section 15.4.3a)

Attachment 1 Page 15

P 6H. Single Rod Withdrawal (UFSAR Section 15.4.3d)

61. Spectrum of Rod Cluster Control Assembly Ejection Accidents (UFSAR Section 15.4.8) 6J. Inadvertent Opening of a Pressurizer Safety or Relief Valve (UFSAR Section 15.6.1) 6K. Steam Generator Tube Rupture (DNB) (UFSAR 15.6.3)

For the balance of the analyses, other evaluations were performed as described below:

6L. Steam System Piping Failure (UFSAR Section 15.1.5) 6M. Reactor Coolant Pump Shaft Seizure - Locked Rotor (peak RCS pressure)

(UFSAR Section 15.3.3) 6N. Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition (UFSAR Section 15.4.1)

60. Uncontrolled RCCA Bank Withdrawal at Power (peak RCS pressure) (UFSAR Section 15.4.2) 6P. Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (UFSAR Section 15.4.4) 6Q. Steam Generator Tube Rupture (dose and overfill) (UFSAR Section 15.6.3) 6R. Loss of Coolant Accidents (UFSAR Section 15.6.5)

These events are summarized below:

6L. Steam System Piping Failure (UFSAR Section 15.1.5)

This event was re-analyzed at a minimum RCS flow rate of 387,000 gpm to demonstrate DNB does not occur. The minimum DNBR was calculated to be 1.987, which is well above the W-3S CHF correlation limit of 1.45.

6M. Reactor Coolant Pump Shaft Seizure - Locked Rotor (peak RCS pressure)

(UFSAR Section 15.3.3)

This event was re-analyzed for a minimum RCS flow rate of 387,000 gpm to calculate the peak RCS pressure for this accident. The resulting peak RCS pressure of 2536 psig is well below the design value of 2735 psig.

6N. Uncontrolled RCCA Bank Withdrawal from a Subcritical or Low Power Startup Condition (UFSAR Section 15.4.1)

This analysis addresses adequate core cooling (Case 1) and peak RCS pressure (Case 2) acceptance criteria. Both Case 1 and Case 2 have sufficient margin to their respective acceptance criteria. The minimum DNBR for Case 1 for the accident analysis at zero power is 3.395, which is well above the acceptance criteria of 1.45. The results of Case 2 showed about 150 psi of margin relative to the design value of 2735 psig.

It is concluded that a reduction in RCS flow from 390,000 gpm to 387,000 gpm would not have an appreciable impact on the results in either case due to the amount of margin between the existing AOR results and the acceptance criteria.

60. Uncontrolled RCCA Bank Withdrawal at Power (peak RCS pressure) (UFSAR Section 15.4.2)

Attachment 1 Page 16

This event was re-analyzed for a minimum RCS flow rate of 387,000 gpm to calculate the peak RCS pressure for this accident. The resulting peak RCS pressure of 2709 psig is below the design value of 2735 psig.

6P. Startup of an Inactive Reactor Coolant Pump at an Incorrect Temperature (UFSAR Section 15.4.4)

This analysis is initiated from an assumed initial power level of 50%. The analysis assumes RCS flow with three reactor coolant pumps initially operating. The current AOR is based on the calculated three-pump flow, starting from a nominal full power and four-pump flow of 388,000 gpm. A historical analysis with different fuel types was based on a nominal full power RCS flow of 382,000 gpm. Both cases showed similar core responses and considerable margin to the limiting DNBR in both cases. Based on the similarity of results between the two cases, it is concluded that the analysis results are insensitive to RCS flow. The proposed reduction in total RCS flow to 387,000 gpm will have no significant impact on the analysis results.

6Q. Steam Generator Tube Rupture (UFSAR Section 15.6.3)

This event is examined for DNB, radiological consequences, and steam generator overfill. The overfill analysis determined the assumed RCS flow rate to be inconsequential. The DNB analysis was performed at 388,000 gpm and is evaluated above (item 6K). The dose input analysis was performed at 390,000 gpm plus uncertainty. Prior to trip, the marginal reduction in flow has an inconsequential impact on the analysis. Upon manual reactor trip, a loss of offsite power is assumed to trip the reactor coolant pumps and the impact of the small change in initial RCS flow has no impact on the balance of the transient.

6R. Loss of Coolant Accidents (UFSAR Section 15.6.5)

The LBLOCA analysis was evaluated by Westinghouse for the proposed reduction in RCS total flow rate. (The Catawba Unit 2 proposed RCS flow reduction is only 3000 gpm, so the Westinghouse evaluation is applicable for both Catawba units.)

It was determined that the variations in the global model calculations are such that the 9 5 th percentile peak clad temperature is not impacted by this proposed RCS flow reduction.

The significant factors in the SBLOCA analysis are decay heat, RCS mass, break flow, and ECCS delivery. Three of these variables are completely unrelated to initial RCS flow, and the fourth (RCS mass) is insignificantly affected. Changes in initial RCS flow may possibly change the thermal mass in the steam generators, and thus the operating pressure, which would in turn affect the time at which the safety valves would lift. However, this second order effect is considered insignificant to the SBLOCA outcome.

Refer to the summary Table 2 at the end of this attachment for a synopsis of the analysis results.

Other UFSAR analyses with common evaluations for Units 1 and 2 Best Estimate/Thermal/Mechanical Desiqn Flows (UFSAR Section 5.1)

Attachment 1 Page 17

The RCS Best Estimate Design Flow, Thermal Design Flow, and Mechanical Design Flow values for Catawba Units 1 and 2 are discussed in Section 5.1 of the UFSAR.

(Unit-specific values are given in Table 5-1.) None of these design flows are impacted by the proposed reduction in the TS 3.4.1 minimum RCS flows discussed here.

Operation at 98% thermal power/99% RCS total flow rate (TS 3.4.1, Required Action B.1)

Catawba is allowed by TS 3.4.1 to operate at RCS flows between 99% and 100% of the limit specified in the COLR, as long as Reactor Thermal Power is reduced to < 98% of the Rated Thermal Power level. The proposed reduction in TS 3.4.1 minimum flows would therefore allow operation at RCS flows as low as 380,160 gpm (Unit 1) and 383,130 gpm (Unit 2) as long as the required reduction in Reactor Thermal Power occurs.

This "stair step" flow reduction was considered as part of the evaluation of the entire list of accident analyses described above for Catawba Units 1 and 2. The only accidents where operation with the 2%/1 % "stair step" reduction in RCS flow/Reactor Thermal Power may represent a bounding condition have been evaluated for the additional reduction in RCS minimum flow proposed here. There is no impact on the TS 3.4.1 Required Action B.1 requirement due to the proposed reduction in RCS minimum flow.

Conclusion The UFSAR transient and accident analyses have been reviewed for a decrease in the TS RCS minimum measured flow rate from 388,000 gpm to 384,000 gpm for Catawba Unit 1, and for a decrease from 390,000 gpm to 387,000 gpm for Catawba Unit 2.

Analysis revisions were performed for those events sensitive to a decrease in RCS flow.

The acceptance criteria were met for all events. Therefore, the reductions in RCS minimum measured flow described above are justified.

Attachment 1 Page 18

5. REGULATORY EVALUATION 5.1 Applicable Regulatory Requirements/Criteria 10 CFR 50.36(b) states that the TS will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to 10 CFR 50.34. The minimum allowable RCS flow rates for Unit 1 and Unit 2 that are located in the TS are supported by existing safety analyses that demonstrate that the units can meet safety analysis acceptance criteria at the revised flow rates. All applicable safety analyses either are not affected by the proposed changes or have been reanalyzed at the proposed new values. 10 CFR 50, Appendix A, General Design Criterion 10 (Reactor design) states that the reactor core and associated coolant, control, and protection systems shall be designed with appropriate margin to assure that specified acceptable fuel design limits are not exceeded during any condition of normal operation, including the effects of anticipated operational occurrences. 10 CFR 50, Appendix A, General Design Criterion 15 (Reactor coolant system design) states that the reactor coolant system and associated auxiliary, control, and protection systems shall be designed with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not exceeded during any condition of normal operation, including anticipated operational occurrences. The reanalyzed safety analyses demonstrate that these two criteria will continue to be met with the proposed changes.

5.2 Precedent On December 19, 2003 (ADAMS Accession Numbers ML033570127 and ML033580393), the NRC approved Amendments 210 and 204 for Catawba Units 1 and 2, respectively. These amendments changed the TS for both units to relocate certain RCS cycle-specific parameter limits from the TS to the Core Operating Limits Report (COLR), and revised the minimum allowable RCS flow rate for Unit 1 from 390,000 gpm to 388,000 gpm. The approved amendments reference all of the Duke Energy correspondence associated with this license amendment request. This correspondence included the following:

Initial submittal dated March 24, 2003 (ADAMS Accession Number ML030920393)

Supplemental submittal dated June 25, 2003 (ADAMS Accession Number ML032130497)

Supplemental submittal dated October 15, 2003 (ADAMS Accession Number ML033100103) 5.3 No Significant Hazards Consideration The following discussion is a summary of the evaluation of the changes contained in this proposed amendment against the 10 CFR 50.92(c) requirements to demonstrate that all three standards are satisfied. A no Attachment 1 Page 19

significant hazards consideration is indicated if operation of the facility in accordance with the proposed amendment would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated, or
2. Create the possibility of a new or different kind of accident from any accident previously evaluated, or
3. Involve a significant reduction in a margin of safety.

First Standard Does operation of the facility in accordance with the proposed amendment involve a significant increase in the probabilityor consequences of an accident previously evaluated?

Response: No.

The reduction in Catawba Unit 1 Reactor Coolant System (RCS) minimum measured flow from 388,000 gpm to 384,000 gpm and the reduction in Catawba Unit 2 RCS minimum measured flow from 390,000 gpm to 387,000 gpm will not change the probability of actuation of any Engineered Safeguard Feature or any other device. The consequences of previously analyzed accidents have been found to be insignificantly different when these reduced flow rates are assumed.

The system transient response is not affected by the initial RCS flow assumption unless the initial assumption is so low as to impair the steady state core cooling capability or the steam generator heat transfer capability. This is clearly not the case with the small proposed reductions in RCS flow. The proposed changes will not result in the modification of any system interface that would increase the likelihood of an accident since these events are independent of the proposed changes. The proposed amendments will not change, degrade, or prevent actions, or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the Updated Final Safety Analysis Report (UFSAR). Therefore, the proposed amendments do not result in the increase in the probability or consequences of an accident previously evaluated.

Second Standard Does operation of the facility in accordancewith the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

These changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. No new accident causal mechanisms are created as a result of NRC approval of this amendment request. No changes are being made to the facility which would introduce any new accident Attachment 1 Page 20

causal mechanisms. This amendment request does not impact any plant systems that are accident initiators.

Third Standard Does operation of the facility in accordance with the proposed amendment involve a significant reduction in the margin of safety?

Response: No.

Implementation of these amendments would not involve a significant reduction in the margin of safety. The decreases in Catawba Unit 1 and Unit 2 RCS minimum measured flow have been analyzed and found to have an insignificant effect on the applicable transient analyses as described in the UFSAR. Margin of safety is related to the confidence of the fission product barriers being able to perform their accident mitigating functions. These fission product barriers include the fuel cladding, the RCS, and the containment. The proposed amendments will have no impact upon the ability of these barriers to function as designed. Consequently, no safety margins will be impacted.

Based upon the preceding discussion, Duke Energy has concluded that the proposed amendments do not involve a significant hazards consideration.

5.4 Conclusions 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 Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Attachment 1 Page 21

6. ENVIRONMENTAL CONSIDERATION Pursuant to 10 CFR 51.22(b), an evaluation of this license amendment request has been performed to determine whether or not it meets the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) of the regulations.

Implementation of these amendments will have no adverse impact upon the Catawba units; neither will it contribute to any additional quantity or type of effluent being available for adverse environmental impact or personnel exposure.

It has been determined there is:

1. No significant hazards consideration,
2. No significant change in the types, or significant increase in the amounts, of any effluents that may be released offsite, and
3. No significant increase in individual or cumulative occupational radiation exposures involved.

Therefore, these amendments to the Catawba TS meet the criteria of 10 CFR 51.22(c)(9) for categorical exclusion from an environmental impact statement.

Attachment 1 Page 22

7. REFERENCES 7.1 Letter from Sean Peters, NRC to D.M. Jamil, Duke Energy, "CATAWBA NUCLEAR STATION, UNITS 1 AND 2 RE: ISSUANCE OF AMENDMENTS (TAC NOS. MB8359 AND MB8360)", December 19, 2003, ADAMS Accession Numbers ML033570127 and ML033580393.

Attachment 1 Page 23

TABLE 1 Summary of UFSAR Transient Information for Use in Proposed Catawba Unit I RCS Flow Reduction (384,000 gpm)

Evaluated Cases (see Note 3)

Approved by NRC, or Conducted LAR Designation Using Methods Core Flow in and Evaluation or Processes High RCS Flow/ Core Flow in Evaluation with UFSAR Type Approved by Reference for Low RCS Flow Current AOR Reduced RCS Section Analysis Title (see Note 1) NRC NRC Approval Limiting (see Note 2) Flow Comment (1) Feedwater System 2A Reference 1 References 2, Low Transient results are 15.1.1 Malfunction that Bounded by 3, 4, and 5 less severe than Results in a Reduction other transient 15.1.2 and 15.1.3 -

in Feedwater results not analyzed since Temperature transient is bounded (2) Feedwater System 1C (zero power) Reference 1 References 2, Low 353,350 gpm NA Zero power case -

15.1.2 Malfunction Causing Bounded by 3, 4, and 5 (zero power 382,000 gpm total an Increase in current AOR case) RCS flow rate -

Feedwater Flow bounds proposed RCS flow rate of 384,000 gpm total flow 3A (full power) Reference 1 References 2, Low 355,020 gpm 359,078 gpm (full Full power case -

Evaluated 3, 4, and 5 (full power power case) Evaluated case)

(3) Excessive Increase in 3K Reference 1 References 2, Low 355,020 gpm 359,078 gpm Evaluated 15.1.3 Secondary Steam Evaluated 3, 4, and 5 Flow (4) Inadvertent Opening 2B Reference 1 References 2, Low Results are 15.1.4 of a Steam Generator Bounded by 3, 4, and 5 bounded by those in Relief or Safety Valve other transient Main Steam Line results Break Analysis

... ___. .. _"(15.1.5)

Attachment 1 Table 1 Page 1

Evaluated Cases (see Note 3)

Approved by NRC, or Conducted

-AR Designation Using Methods Core Flow in and Evaluation or Processes High RCS Flow/ Core Flow in Evaluation with UFSAR Type Approved by Reference for Low RCS Flow Current AOR Reduced RCS Section Analysis Title (see Note 1) NRC NRC Approval Limiting (see Note 2) Flow Comment (5) Steam System Piping 3L References 6 References 7 Low 337,590 gpm 337,590 gpm Reanalyzed with 15.1.5 Failure Evaluated and 10 and 10 RCS total flow of 384,000 gpm minus flow uncertainty, stairstep flow reduction, and bounding high core bypass flow (6) Steam Pressure 2C NA NA NA NA 15.2.1 Regulator Not applicable to Malfunction or Failure Catawba that Results in Decreasing Steam Flow (7) Loss of External Load 2D Reference 8 Reference 9 Low Results are 15.2.2 Bounded by . .  : bounded by those in other transient Turbine Trip (15.2.3) results (8) Turbine Trip 1D Reference 1 References 2, Low (Pk primary) (Peak NA Peak primary 15.2.3 Bounded by 3, 4, and 5 primary pressure: 382,000 current AOR pressure) gpm (total) 345,576 gpm High (Pk (Peak NA Peak secondary secondary) secondary pressure: 420,000 pressure) gpm (total) 388,500 gpm (9) Inadvertent Closure 2E NA NA Low I Results are 15.2.4 of Main Steam Bounded by bounded by those in Isolation Valves other transient Turbine Trip (15.2.3) results Attachment 1 Table 1 Page 2

Evaluated Cases (see Note 3)

Approved by NRC, or Conducted

_AR Designation Using Methods Core Flow in and Evaluation or Processes High RCS Flow/ Core Flow in Evaluation with UFSAR Type Approved by Reference for Low RCS Flow Current AOR Reduced RCS Section Analysis Title (see Note 1) NRC NRC Approval Limiting (see Note 2) Flow Comment (10) Loss of Condenser 2F NA NA Low Results are 15.2.5 Vacuum and Other Bounded by , -. ,.;::bounded by those in Events Causing a other transient Turbine Trip (15.2.3)

Turbine Trip results __*_____________

(11) Loss of Non- 1E Reference 1 References 2, Low 345,576 gpm NA Current AOR 15.2.6 Emergency AC Bounded by 3, 4, and 5 assumes Power to the Station current AOR conservatively low Auxiliaries total RCS flow of 1_ 382,000 gpm (12) Loss of Normal 3M Reference 1 References 2, (Short-term) (Short-term) (Short-term) (Short-term) 15.2.7 Feedwater Flow Evaluated 3, 4, and 5 Low 351,360 gpm 351,360 gpm Reanalyzed with 384,000 gpm total RCS flow (Long-term) (Long-term) (Long-term) (Long-term)

Low 343,415 gpm 343,415 gpm Current AOR assumes 388,000 gpm total RCS flow, but has substantial margin to success criteria (13) Feedwater System 2G Reference 1 References 2, (Short-term) (Short-term) 15.2.8 Pipe Break (Short-term) 3, 4, and5 Low I Results are Bounded by bounded by those in other transient Loss of Normal results Feedwater Flow

_(15.2.7)

Attachment 1 Table 1 Page 3

Evaluated Cases (see Note 3)

Approved by NRC, or Conducted

.AR Designation Using Methods Core Flow in and Evaluation or Processes High RCS Flow/ Core Flow in Evaluation with UFSAR Type Approved by Reference for Low RCS Flow Current AOR Reduced RCS Section Analysis Title (see Note 1) NRC NRC Approval Limiting (see Note 2) Flow Comment 3N Reference 1 References 2, (Long-term) (Long-term) (Long-term) (Long-term)

(Long-term) 3, 4, and 5 Low 352,813 gpm 352,813 gpm Current AOR Evaluated assumes 390,000 gpm total RCS flow, but results are insensitive to assumed RCS flow (14) Partial Loss of 3B Reference 1 References 2, Low 355,020 gpm 359,078 gpm Evaluated 15.3.1 Forced Reactor Evaluated 3, 4, and 5 Coolant Flow (15) Complete Loss of 3C Reference 1 References 2, Low 355,020 gpm 359,078 gpm Evaluated 15.3.2 Forced Reactor Evaluated 3, 4, and 5 Coolant Flow (16) Reactor Coolant 3D T&H: T&H: Low 355,020 gpm 359,078 gpm (DNB) 15.3.3 Pump Shaft Seizure (DNB) Reference 1 References 2, Evaluated (Locked Rotor) Evaluated 3, 4, and 5 Dose:

Reference 10 Dose:

Reference 10 30 Reference 1 Reference 2, Low 341,001 gpm NA (peak RCS (peak RCS 3, 4, and 5 pressure) pressure) Reanalyzed with Evaluated 384,000 gpm total RCS flow (17) Reactor Coolant 2H NA NA Low Results are similar 15.3.4 Pump Shaft Break Bounded by to those calculated other transient in Locked Rotor results . (15.3.3)

Attachment 1 Table 1 Page 4

Evaluated Cases (see Note 3)

Approved by NRC, or Conducted

-AR Designation Using Methods Core Flow in and Evaluation or Processes High RCS Flow/ Core Flow in Evaluation with UFSAR Type Approved by Reference for Low RCS Flow Current AOR Reduced RCS Section Analysis Title (see Note 1) NRC NRC Approval Limiting (see Note 2) Flow Comment (18) Uncontrolled Rod 3P Reference 1 References 2, Low (3 RCP) (3 RCP) (3 RCP - core 15.4.1 Cluster Control (DNB) 3, 4, and 5 274,146 gpm 274,146 gpm cooling) - AOR has Assembly Bank Evaluated total RCS flow rate Withdrawal From a based on 388,000 Subcritical or Low gpm and 3 pump Power Startup operation at HZP; Condition proposed decrease in RCS flow has no impact on 3 RCP case 3P Reference 1 References 2, Low (Peak RCS (Peak RCS (Peak RCS (Peak RCS 3, 4, and 5 pressure) pressure) pressure) -

pressure) 337,591 gpm 337,591 gpm Reanalyzed with Evaluated RCS total flow of 384,000 gpm (19) Uncontrolled Rod 3E Reference 1 References 2, Low 355,020 gpm 359,078 gpm Evaluated 15.4.2 Cluster Control (DNB) 3, 4, and 5 Assembly Bank Evaluated Withdrawal at Power 3Q Reference 1 References 2, Low 337,590 gpm NA (Peak RCS (Peak RCS 3, 4, and 5 pressure)-

pressure) Reanalyzed with Evaluated RCS total flow of 384,000 gpm (20) Rod Cluster Control 15.4.3a: 3F Reference 6 Reference 7 Low 355,020 gpm 359,078 gpm 15.4.3a: (One or 15.4.3 Assembly Evaluated more dropped Misoperation (System RCCA rods):

Malfunction or Evaluated Attachment 1 Table 1 Page 5

Evaluated Cases (see Note 3)

Approved by NRC, or Conducted

-AR Designation Using Methods Core Flow in and Evaluation or Processes High RCS Flow/ Core Flow in Evaluation with UFSAR Type Approved by Reference for Low RCS Flow Current AOR Reduced RCS Section Analysis Title (see Note 1) NRC NRC Approval Limiting (see Note 2) Flow Comment Operator Error) 15.4.3b: 21 Reference 1 References 2, Low 15.4.3b: (Dropped Bounded by 3, 4, and 5 "RCCA bank):

other transient

  • Results are results bounded by case with one or more dropped rods (15.4.3a and 15.4.3c) 15.4.3c: 2J Reference 1 References 2, Low 15.4.3c: (Statically Bounded by 3, 4, and 5 misaligned RCCA):

other transient Results are results bounded by the analysis of one or more dropped rods, or by single RCCA

__.__

  • withdrawal (SUCR) 15.4.3d: 3G Reference 1 References 2, Low 358,900 gpm 359,078 gpm 15.4.3d: (SUCR):

Evaluated 3, 4, and 5 Flow through the core: Evaluated (21) Startup of an Inactive 3R Reference 1 References 2, Low Based on Based on 339,980 AOR has total RCS 15.4.4 Reactor Coolant Evaluated 3, 4, and 5 339,980 gpm gpm initial core flow rate based on Pump at an Incorrect initial core flow 388,000 gpm; Temperature flow proposed decrease in RCS flow has no impact for analysis with 3 RCP initial condition (22) A Malfunction or 2K NA NA NA NA 15.4.5 Failure of the Flow (BWR Transient)

Controller in a BWR Loop that Results in an Increased Reactor Attachment 1 Table 1 Page 6

Evaluated Cases (see Note 3)

Approved by NRC, or Conducted

-AR Designation Using Methods Core Flow in and Evaluation or Processes High RCS Flow/ Core Flow in Evaluation with UFSAR Type Approved by Reference for Low RCS Flow Current AOR Reduced RCS Section Analysis Title (see Note 1) NRC NRC Approval Limiting (see Note 2) Flow Comment Coolant Flow Rate _

(23) Chemical and 2L Reference 1 References 2, NA NA 15.4.6 Volume Control (Unaffected) 3, 4, and 5 System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant (24) Inadvertent Loading 2M Reference 1 References 2, NA NA 15.4.7 and Operation of a (Unaffected) 3, 4, and 5 Fuel Assembly in an Improper Position (25) Spectrum of Rod 3H T&H: T&H: Low 343,621 gpm 344,655 gpm Evaluated 15.4.8 Cluster Control Evaluated Reference 6 Reference 7 Assembly Ejection Accidents Dose: Dose:

Reference 10 Reference 10 (26) Spectrum of Rod Drop 2N NA NA NA N..A 15.4.9 Accidents (BWR) (BWR Transient)

(27) Inadvertent Operation 20 Reference 1 References 2, NA NA 15.5.1 of Emergency Core (Bounded by 3, 4, and 5 Cooling System other transients)

During Power Operation (28) Chemical and 2P NA NA NA NA 15.5.2 Volume Control (Bounded by System Malfunction other transients) that Increases Reactor Coolant Inventory _ _ _ __ _ _ _

(29) A Number of BWR 2Q NA NA NA NA 15.5.3 Transients (BWR Transient) ,,

Attachment 1 Table 1 Page 7

Evaluated Cases (see Note 3)

Approved by NRC, or Conducted LAR Designation Using Methods Core Flow in and Evaluation or Processes High RCS Flow/ Core Flow in Evaluation with UFSAR Type Approved by Reference for Low RCS Flow Current AOR Reduced RCS Section Analysis Title (see Note 1) NRC NRC Approval Limiting (see Note 2) Flow Comment (30) Inadvertent Opening of 31 Reference 1 References 2, Low 355,020 gpm 359,078 gpm Evaluated 15.6.1 a Pressurizer Safety or Evaluated 3, 4, and 5 Relief Valve (31) Break in Instrument 2R Reference 10 Reference 10 NA . ;NA 15.6.2 Line or Other Lines (Unaffected)  ;

From Reactor Coolant Pressure Boundary that Penetrate Containment (32) Steam Generator 3J (DNB) T&H: T&H: Low 355,020 gpm 359,078 gpm Evaluated 15.6,3 Tube Failure Evaluated Reference 1 References 2, 3, 4, and 5 3S (overfill, Dose: Dose: Low 358,900 gpm NA For the SG overfill dose) Reference 10 Reference 10 (overfill) and dose Evaluated consequences of 343,414 gpm SGTR, the analysis (dose) results are not sensitive to the assumed RCS flow (33) Spectrum of BWR 2S NA NA NA . . . . NA 15.6.4 Steam System Piping (BWR Transient)

Failures Outside Containment (34) Loss-of-Coolant 3T T&H: T&H: Low LBLOCA/ LBLOCA/SBLOCA: LBLOCA/SBLOCA 15.6.5 Accidents Evaluated References 11 References 18 SBLOCA: Based on 384,000 consequences are and 12 and 19 Based on gpm total RCS flow insensitive to the 390,000 gpm rate proposed change in Dose: Dose: total RCS RCS flow References 13, Reference 20 flow rate 14, 15,16, and 17 Attachment 1 Table 1 Page 8

Evaluated Cases (see Note 3)

Approved by NRC, or Conducted LAR Designation Using Methods Core Flow in and Evaluation or Processes High RCS Flow/ Core Flow in Evaluation with UFSAR Type Approved by Reference for Low RCS Flow Current AOR Reduced RCS Section Analysis Title (see Note 1) NRC NRC Approval Limiting (see Note 2) Flow Comment (35) A Number of BWR 2T NA NA NA NA 15.6.6 Transients (BWR Transient)

(36) Radioactive Gas 2U Reference 21 Reference 20 NA NA 15.7.1 Waste System Leak (Unaffected) or Failure (37) Radioactive Liquid 2U Reference 10 Reference 10 NA NA 15.7.2 Waste System Leak (Unaffected) or Failure (38) Postulated 2U Reference 10 Reference 10 NA NA 15.7.3 Radioactive Releases (Unaffected)

Due to Liquid Tank Failures (39) Fuel Handling 2U References 22, References 25 NA NA 15.7.4 Accidents in the (Unaffected) 23, 24, and 25 and 26 Containment and Spent Fuel Storage Buildings ________________

(40) Anticipated 1F Reference 27 Reference 28 NA Based on NA Current AOR 15.8 Transients Without Bounded by nominal total assumes Trip current AOR RCS flow of conservatively low 377,600 gpm total RCS flow of 377,600 gpm (41) Formerly Chapter 15 2U NA NA NA NA 15.9 Appendix A - Models (Unaffected)

Used for Calculation of Accident Doses (42) Formerly Chapter 15 2U References 29, References 9 NA NA 15.10 Appendix B - (Unaffected) 30, and 31 and 32 Supplementary (ac),n1d Radiological Analyses Attachment 1 Table 1 Page 9

Evaluated Cases (see Note 3)

Approved by NRC, or Conducted LAR Designation Using Methods Core Flow in and Evaluation or Processes High RCS Flow/ Core Flow in Evaluation with UFSAR Type Approved by Reference for Low RCS Flow Current AOR Reduced RCS Section Analysis Title (see Note 1) NRC NRC Approval Limiting (see Note 2) Flow Comment (43) Containment 1B Reference 33 References 34 High 388,500 gpm NA Current AOR 6.2.1.1.3.1 Performance Bounded by and 35 assume 6.2.1.1.3.3 Analyses current AOR conservatively high 6.2.1.1.3.4 RCS flow of 420,000 6.2.1.3.2 gpm 6.2.1.4 (44) Postulated 1B References 33 References High 388,500 gpm NA Current AOR 6.2.1 Secondary System Bounded by and 36 34, 35, and 37 assumes Pipe Rupture Outside current AOR conservatively high Containment RCS flow of 420,000

_ _ gpm (45) LOCA Blowdown 1A NA NA High Based on NA Current AOR 3.6.2.2.1 Reactor Vessel and Bounded by total RCS assume 3.9.1.4 Loop Forces current AOR flow rate of conservatively high 398,000 gpm RCS flow of 398,000 gpm NOTE 1: LAR Designations listed by Category pages B-1 to B-5.

NOTE 2: "Core Flow" represents net flow through core in analysis, defined as Total RCS flow minus Core Bypass Flow.

NOTE 3: Transients which are bounded by other transient results, not applicable for Catawba Unit 1, or unaffected by the proposed RCS flow reduction have shaded fields in this column.

REFERENCES:

1. DPC-NE-3002-A, Revision 4b, "McGuire and Catawba Nuclear Station UFSAR Chapter 15 System Transient Analysis Methodology", September 2010
2. Letter from Tim Reed (NRC) to H. B. Tucker (Duke) dated November 15, 1991, "Safety Evaluation on Topical Report DPC-NE-3002, "FSAR Chapter 15 System Transient Analysis Methodology," (TAC No. 66850)"
3. Letter from Robert Martin (NRC) to M. S. Tuckman (Duke) dated December 28, 1995, "Safety Evaluation for Revision 1 to Topical Report DPC-NE-3002-A, "FSAR Chapter 15 System Transient Analysis Methodology" McGuire Nuclear Station, Units 1 and 2; and Catawba Nuclear Station, Units I and 2 (TAC Nos. M89944, M89945, and M89946)"

Attachment 1 Table 1 Page 10

4. Letter from Herbert Berkow (NRC) to M. S. Tuckman (Duke) dated April 26, 1996, "Safety Evaluation on Change to Topical Report DPC-NE-3002-A on Opening Characteristics of Safety Valves - McGuire Nuclear Station, Units 1 and 2; and Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M94405, M94406, M94407, and M94408)"
5. Letter from Chandu Patel (NRC) to G. R. Peterson (Duke) dated April 6, 2001, "Catawba Nuclear Station, Units 1 and 2 RE: Revision 4 to the Duke Energy Corporation Topical Report DPC-NE-3002-A, "UFSAR Chapter 15 Transient Analysis Methodology" (TAC Nos. MA8928 and MA8929)"
6. DPC-NE-3001-PA, Revision Oa, "McGuire and Catawba Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology", May 2009
7. Letter from Timothy Reed (NRC) to H. B. Tucker (Duke) dated November 15, 1991, "Safety Evaluation on Topical Report DPC-NE-3001, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters" (TAC Nos. 75954/75955/75956/75957)"
8. Technical Information Document (TID) 14844, Calculation of Distance Factors for Power and Test Reactors, U.S. Atomic Energy Commission, 1962
9. 10 CFR Part 100, Section 100.11
10. Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluation Design Basis Accidents at Nuclear Power Reactors"
11. WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best-Estimate LOCA Analysis," March 1998.
12. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.
13. Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Power Company LLC d/b/a Duke Energy Carolinas, LLC McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 License Amendment Request for Full Scope Implementation of the Alternative Source Term," March 20, 2008.
14. Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas: LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 License Amendment Request for Full Scope Implementation of the Alternative Source Term Revision to Control Room Atmospheric Dispersion Factors," March 20, 2008.
15. Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 License Amendment Request for Full Scope Implementation of the Alternative Source Term. Response to Request for Additional Information," October 6, 2008.
16. Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Response to Request for Additional Information related to the License Amendment Request for Implementation of Alternative Source Term," December 17, 2008.
17. Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Response to Request for Additional Information related to the License Amendment Request (LAR) for Implementation of Alternative Source Term (AST)," February 12, 2009.
18. Letter from R. C. Jones (NRC) to N. J. Liparulo (Westinghouse) dated June 28, 1996, "Acceptance for Referencing of the Topical Report WCAP-12945 (P), Westinghouse Code Qualification Document for Best Estimate Loss-of-Coolant Analysis".

Attachment 1 Table 1 Page 11

19. Letter from R. C. Jones (NRC) to E. P. Rahe (Westinghouse) dated May 21, 1985, "Acceptance for Referencing of Licensing Topical Report WCAP-1 0054 (P), Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code".
20. Letter from John Stang (USNRC) to B.H. Hamilton (Duke), "McGuire Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Adoption of the Alternative Source Term Radiological Analysis Methodology (TAC Nos. MD8400 and MD8401)," March 30, 2009.
21. Regulatory Guide (RG) 1.109, "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"
22. Letter from G. R. Peterson (Duke) to U.S. NRC dated December 20, 2005, "License Amendment Request for Selective Implementation of the Alternative Source Term and Revision to Technical Specification 3.9.4, Containment Penetrations"
23. Letter from G. R. Peterson (Duke) to U.S. NRC dated May 4, 2006, "License Amendment Request for Selective Implementation of the Alternative Source Term and Revision to Technical Specification 3.9.4, Containment Penetrations. Response to Request for Additional Information"
24. Letter from G. R. Peterson (Duke) to U.S. NRC dated August 31, 2006, "License Amendment Request for Selective Implementation of the Alternative Source Term and Revision to Technical Specification 3.9.4, Containment Penetrations. Additional Commitments Regarding Containment Closure Administrative Controls."
25. ISG-5, Revision 1 - "Confinement Evaluation", Spent Fuel Project Office, NRC
26. Letter from John Stang (NRC) to G. R. Peterson (Duke) dated December 22, 2006, "McGuire Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Implementation of Alternative Source Term Methodology (TAC Nos. MC9746 AND MC9747)"
27. Letter from T. M. Anderson (Westinghouse) to S. H. Hanauer (NRC) dated December 30, 1979, "NS-TMA-2182, ATWS Submittal"
28. Letter from Darl S. Hood (NRC) to H. B. Tucker (Duke) dated November 6, 1987, "ATWS Rule (10 CFR 50.62) for McGuire and Catawba Nuclear Stations, Units 1 and 2 (TACs 59081/59111/59112/64535)"
29. Letter from Hal Tucker (Duke) to Harold Denton (NRC) dated February 17, 1984, "McGuire Nuclear Station Docket Nos. 50-369, 50-370", Spent Fuel Pool Re-rack LAR
30. Letter from Hal Tucker (Duke) to Harold Denton (NRC) dated March 20, 1984, "McGuire Nuclear Station Docket Nos. 50-369, 50-370", Safety & Environmental Analysis for Spent Fuel Pool Re-rack
31. Regulatory Guide (RG) 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (Safety Guide 25)"
32. Letter from Elinor Adensam (NRC) to Hal Tucker (Duke) dated September 24, 1984, "Issuance of Amendment No.35 to Facility Operating License NPF-9 and Amendment No.

16 to Facility Operating License NPF McGuire Nuclear Station, Units 1 and 2"

33. DPC-NE-3004-PA, Revision 1, McGuire and Catawba Mass and Energy Release and Containment Response Methodology, December 2000
34. Letter from NRC to M. S. Tuckman (Duke) dated September 6, 1995, "Safety Evaluation for Topical Report DPC-NE-3004-P, "Mass and Energy Release and Containment Response Methodology", McGuire Nuclear Station, Units 1 and 2; and Catawba Nuclear Station , Units 1 and 2 (TAC Nos. M90646, M90647, and M90648)"

Attachment 1 Table 1 Page 12

35. Letter from NRC to H. B. Barron (Duke) dated February 29, 2000, "McGuire Nuclear Station and Catawba Nuclear Station RE: Review of Topical Report DPC-NE-3004-PA, Rev. 1, Regarding Proposed Finer Nodalization of Ice Condenser (TAC Nos. MA551 1, MA5512, MA5517, and MA5518)"
36. Letter from M. S. Tuckman (Duke) to U. S. NRC dated March 15, 1996, "Catawba Nuclear Station, Units 1 and 2, Docket Nos. 50-413 and 414; McGuire Nuclear Station, Units 1 and 2, Docket Nos. 50-369 and 370; Response to Request for Additional Information"
37. Letter from Victor Nerses (NRC) to H. B. Barron (Duke) dated May 5, 1997, "Issuance of Amendments - McGuire Nuclear Station, Units 1 and 2 (TAC Nos. M90590 and M90591 )"

Attachment 1 Table 1 Page 13

TABLE 2 Summary of UFSAR Transient Information for Use in Proposed Catawba Unit 2 RCS Flow Reduction (387,000 gpm)

Evaluated Cases (see Note 3)

Approved by NRC or Conducted LAR Designation Using Methods Core Flow in Core Flow in and Evaluation or Processes Reference High RCS Flow/ Current AOR Evaluation with UFSAR Type Approved by for NRC Low RCS Flow (see Note 2) Reduced RCS Section Analysis Title (see Note 1) NRC Approval Limiting Flow Comment (1) Feedwater System 5A Reference 1 References Low Transient results are 15.1.1 Malfunction that Bounded by other 2, 3, 4, and less severe than Results in a Reduction transient results 5 15.1.2 and 15.1.3 -

in Feedwater ":not analyzed since Temperature transient is bounded (2) Feedwater System 4C (zero power) Reference 1 References Low 353,350 gpm NA Zero power case -

15.1.2 Malfunction Causing Bounded by 2, 3, 4, and (zero power 382,000 gpm total an Increase in current AOR 5 case) RCS flow rate -

Feedwater Flow bounds proposed RCS flow rate of 387,000 gpm total flow 6A (full power) Reference 1 References Low 358,800 gpm 361,032 gpm (full Full power case -

Evaluated 2, 3, 4, and (full power power case) Evaluated 5 case)

(3) Excessive Increase in 4D Reference 1 References Low 356,125 gpm 361,032 gpm Current AOR 15.1.3 Secondary Steam Bounded by 2, 3, 4, and assumes RCS total Flow current AOR 5 flow rate of 385,000 gpm (4) Inadvertent Opening 5B Reference 1 References Low Results are 15.1.4 of a Steam Generator Bounded by other 2, 3, 4, and < bounded by those in Relief or Safety Valve transient results 5 Main Steam Line Break Analysis

____._(15.1.5)

Attachment 1 Table 2 Page 1

Evaluated Cases (see Note 3)

Approved by NRC or Conducted LAR Designation Using Methods Core Flow in Core Flow in and Evaluation or Processes Reference High RCS Flow/ Current AOR Evaluation with UFSAR Type Approved by for NRC Low RCS Flow (see Note 2) Reduced RCS Section Analysis Title (see Note 1) NRC Approval Limiting Flow Comment (5) Steam System Piping 6L References 6 References Low 353,343 gpm 353,343 gpm Reanalyzed with 15.1.5 Failure Evaluated and 10 7 and 10 RCS flow of 387,000 gpm minus flow uncertainty, stairstep flow reduction, and bounding core bypass flow (6) Steam Pressure 5C NA NA NA  : NA 15.2.1 Regulator Not applicable to Malfunction or Failure Catawba that Results in Decreasing Steam Flow .

(7) Loss of External Load 5D Reference 8 Reference 9 Low .* ° Results are 15.2.2 Bounded by other . bounded by those in transient results Turbine Trip (15.2.3)

(8) Turbine Trip 4E Reference 1 References Low (Peak (Peak NA Peak primary 15.2.3 Bounded by 2, 3, 4, and primary) primary pressure: 382,000 current AOR 5 pressure) gpm (total) 345,576 gpm High (Peak (Peak NA Peak secondary secondary) secondary pressure: 420,000 pressure) gpm (total) 388,500 gpm (9) Inadvertent Closure 5E NA NA Low " .... Results are 15.2.4 of Main Steam Bounded by other . bounded by those in Isolation Valves transient results ___.. ."_..._....Turbine Trip (15.2.3)

(10) Loss of Condenser 5F NA NA Low Results are 15.2.5 Vacuum and Other Bounded by other . . bounded by those in Events Causing a transient results...... i Turbine Trip (15.2.3)

Turbine Trip Attachment 1 Table 2 Page 2

Evaluated Cases (see Note 3)

Approved by NRC or Conducted LAR Designation Using Methods Core Flow in Core Flow in and Evaluation or Processes Reference High RCS Flow/ Current AOR Evaluation with UFSAR Type Approved by for NRC Low RCS Flow (see Note 2) Reduced RCS Section Analysis Title (see Note 1) NRC Approval Limiting Flow Comment (11) Loss of Non- 4F Reference 1 References Low 345,576 gpm NA Current AOR 15.2.6 Emergency AC Bounded by 2, 3, 4, and assumes Power to the Station current AOR 5 conservatively low Auxiliaries RCS flow of 382,000 gpm (12) Loss of Normal (Short-term) Reference 1 References (Short-term) (Short-term) (Short-term) (Short-term) 15.2.7 Feedwater Flow 6B 2, 3, 4, and Low 360,750 gpm 361,032 gpm Evaluated Evaluated 5 (Long-term) Reference 1 References (Long-term) (Long-term) (Long-term) (Long-term) 4G 2, 3, 4, and Low 348,290 gpm NA Current AOR Bounded by 5 assumes 385,000 current AOR gpm (13) Feedwater System 5G Reference 1 References (Short-term) (Short-term) 15.2.8 Pipe Break (Short-term) 2, 3, 4, and Low . .... Results are Bounded by other 5  : bounded by those in transient results Loss of Normal Feedwater Flow

__ _ _ (15.2.7) 4H Reference 1 References (Long-term) (Long-term) (Long-term) (Long-term)

(Long-term) 2, 3, 4, and Low 339,972 gpm NA Current AOR Bounded by 5 assumes 382,000 current AOR gpm (14) Partial Loss of 6C Reference 1 References Low 358,800 gpm 361,032 gpm Evaluated 15.3.1 Forced Reactor Evaluated 2, 3, 4, and Coolant Flow 5 (15) Complete Loss of 6D Reference 1 References Low 358,800 gpm 361,032 gpm Evaluated 15.3.2 Forced Reactor Evaluated 2, 3, 4, and Coolant Flow 5 Attachment 1 Table 2 Page 3

Evaluated Cases (see Note 3)

Approved by NRC or Conducted LAR Designation Using Methods Core Flow in Core Flow in and Evaluation or Processes Reference High RCS Flow/ Current AOR Evaluation with UFSAR Type Approved by for NRC Low RCS Flow (see Note 2) Reduced RCS Section Analysis Title (see Note 1) NRC Approval Limiting Flow Comment (16) Reactor Coolant 6E T&H: T&H: Low 360,750 gpm 361,032 gpm Evaluated 15.3.3 Pump Shaft Seizure (DNB) Reference 1 References (Locked Rotor) Evaluated 2, 3, 4, and Dose: 5 Reference 10 Dose:

Reference 10 6M Reference 1 References Low 350,099 gpm NA (Peak RCS (peak RCS 2, 3, 4, and pressure) -

pressure) 5 Reanalyzed with Evaluated RCS total flow of 387,000 gpm (17) Reactor Coolant 5H NA NA Low Results are similar 15.3.4 Pump Shaft Break Bounded by other to those calculated transient results in Locked Rotor (15.3.3)

(18) Uncontrolled Rod 6N Reference 1 References Low (3 RCP) NA (3 RCP - core 15.4.1 Cluster Control (DNB, peak RCS 2, 3, 4, and 357,142 gpm cooling)- AOR has Assembly Bank pressure) 5 total RCS flow rate Withdrawal From a Evaluated based on 390,000 Subcritical or Low gpm; proposed Power Startup decrease in RCS Condition flow has no impact on 3 RCP case Attachment 1 Table 2 Page 4

r 7 I r Evaluated Cases (see Note 3)

Approved by NRC or Conducted LAR Designation Using Methods Core Flow in Core Flow in and Evaluation or Processes Reference High RCS Flow/ Current AOR Evaluation with UFSAR Type Approved by for NRC Low RCS Flow (see Note 2) Reduced RCS Section Analysis Title (see Note 1) NRC Approval Limiting Flow Comment

+/- I.

(Peak RCS NA (Peak RCS pressure) pressure) -

based on AOR has total RCS 349,285 gpm flow rate based on 390,000 gpm; proposed decrease in RCS flow has no impact on 3 RCP case (19) Uncontrolled Rod 6F Reference 1 References Low 357,142 gpm 357,421 gpm Evaluated 15.4.2 Cluster Control (DNB) 2, 3, 4, and Assembly Bank Evaluated 5 Withdrawal at Power 60 Reference 1 References Low 346,597 gpm NA Reanalyzed with (Peak RCS 2, 3, 4, and RCS total flow of pressure) 5 387,000 gpm Evaluated (20) Rod Cluster Control 15.4.3a: 6G 15.4.3a: 15.4.3a: (All cases) 358,800 gpm 361,032 gpm 15.4.3a: (One or 15.4.3 Assembly Evaluated Reference 6 Reference 7 Low more dropped Misoperation (System RCCA rods):

Malfunction or Dose: Dose: Evaluated Operator Error) Reference 10 Reference 10 15.4.3b: 51 Reference 1 References Low 15.4.3b: (Dropped Bounded by other 2, 3, 4, and . RCCA bank):

transient results 5 Results are bounded by case with one or more dropped rods (15.4.3a and

_ _*_ _ _ _ _ _ _ _ _. _ __  :.. 15.4.3c)

Attachment 1 Table 2 Page 5

Evaluated Cases (see Note 3)

Approved by NRC or Conducted LAR Designation Using Methods Core Flow in Core Flow in and Evaluation or Processes Reference High RCS Flow/ Current AOR Evaluation with UFSAR Type Approved by for NRC Low RCS Flow (see Note 2) Reduced RCS Section Analysis Title (see Note 1) NRC Approval Limiting Flow Comment 15.4.3c: 5J Reference 1 References Low . . . 15.4.3c: (Statically Bounded by other 2, 3, 4, and misaligned RCCA):

transient results 5 Results are bounded by cases with one or more dropped rods, or by

single RCCA
_____________________j__________=__withdrawal (SUCR) 15.4.3d: 6H Reference 1 References Low 360,750 gpm 361,032 gpm 15.4.3d: (SUCR):

Evaluated 2, 3, 4, and Flow through the 5 core: Evaluated (21) Startup of an Inactive 6P Reference 1 References Low Based on Based on 339,980 (3 RCP - core 15.4.4 Reactor Coolant Evaluated 2, 3, 4, and 339,980 gpm gpm initial core cooling) - AOR has Pump at an Incorrect 5 initial core flow total RCS flow rate Temperature flow based on 388,000 gpm; proposed decrease in RCS flow has no impact on 3 RCP case (22) A Malfunction or 5K NA NA NA NA 15.4.5 Failure of the Flow (BWR Transient)

Controller in a BWR Loop that Results in an Increased Reactor Coolant Flow Rate (23) Chemical and 5L Reference 1 References NA NA 15.4.6 Volume Control (Unaffected) 2, 3, 4, and System Malfunction 5 that Results in a Decrease in Boron Concentration in the Reactor Coolant Attachment 1 Table 2 Page 6

Evaluated Cases (see Note 3)

Approved by NRC or Conducted LAR Designation Using Methods Core Flow in Core Flow in and Evaluation or Processes Reference High RCS Flow/ Current AOR Evaluation with UFSAR Type Approved by for NRC Low RCS Flow (see Note 2) Reduced RCS Section Analysis Title (see Note 1) NRC Approval Limiting Flow Comment (24) Inadvertent Loading 5M Reference 1 References NA NA 15.4.7 and Operation of a (Unaffected) 2, 3, 4, and Fuel Assembly in an 5 Improper Position (25) Spectrum of Rod 61 T&H: T&H: Low 343,621 gpm 349,558 gpm Evaluated 15.4.8 Cluster Control Evaluated Reference 6 Reference 7 Assembly Ejection Accidents Dose: Dose:

Reference 10 Reference 10 (26) Spectrum of Rod 5N NA NA NA NA 15.4.9 Drop Accidents (BWR Transient)

(BWR)

(27) Inadvertent Operation 50 Reference 1 References NA NA 15.5.1 of Emergency Core (Unaffected) 2, 3, 4, and Cooling System 5 During Power Operation _ ___ ,__:__ ,__ :____

(28) Chemical and 5P NA NA NA NA 15.5.2 Volume Control (Bounded by System Malfunction other transients) that Increases Reactor Coolant I Inventory _ ________"__ __ _ __ _ ___,

(29) A Number of BWR 5Q NA NA NA NA 15.5.3 Transients (BWR Transient) . . .. . ____ . ...

(30) Inadvertent Opening 6J Reference 1 References Low 355,020 gpm 361,032 gpm Evaluated 15.6.1 of a Pressurizer Evaluated 2, 3, 4, and Safety or Relief Valve , 1 1_5 Attachment 1 Table 2 Page 7

Evaluated Cases (see Note 3)

Approved by NRC or Conducted LAR Designation Using Methods Core Flow in Core Flow in and Evaluation or Processes Reference High RCS Flow/ Current AOR Evaluation with UFSAR Type Approved by for NRC Low RCS Flow (see Note 2) Reduced RCS Section Analysis Title (see Note 1) NRC Approval Limiting Flow Comment (31) Break in Instrument 5R Reference 10 Reference NA NA 15.6.2 Line or Other Lines (Unaffected) 10 From Reactor Coolant Pressure  :

Boundary that Penetrate Containment ......

(32) Steam Generator 6K T&H: T&H: Low 360,750 gpm 361,032 gpm Evaluated 15.6.3 Tube Failure (DNB) Reference 1 References Evaluated 2, 3, 4, and 5

6Q Dose: Dose: Low 345,576 gpm NA For the SG overfill (overfill, dose) Reference 10 Reference (dose) and dose Evaluated 10 consequences of 356,125 gpm SGTR, the analysis (overfill) results are not sensitive to the assumed RCS flow (33) Spectrum of BWR 5S NA NA NA NA 15.6.4 Steam System Piping (BWR Transient)

Failures Outside Containment (34) Loss-of-Coolant 6R T&H: T&H: Low LBLOCA/ LBLOCAISBLOCA: LBLOCA/SBLOCA 15.6.5 Accidents Evaluated References 11 References SBLOCA: Based on 387,000 consequences are and 12 18 and 19 Based on gpm total RCS flow insensitive to the 390,000 gpm rate proposed change in Dose: Dose: total RCS RCS flow References 13, Reference flow rate 14, 15, 16, and 20

__ _ 1_ _ _ __ _ _ _ 1__ _ __ _ _ 1__ 171 (35) A Number of BWR 5T NA NA NA NA NA 15.6.6 Transients (BWR Transient)s: ... ... T i WA  :

Attachment 1 Table 2 Page 8

Evaluated Cases (see Note 3)

Approved by NRC or Conducted LAR Designation Using Methods Core Flow in Core Flow in and Evaluation or Processes Reference High RCS Flow/ Current AOR Evaluation with UFSAR Type Approved by for NRC Low RCS Flow (see Note 2) Reduced RCS Section Analysis Title (see Note 1) NRC Approval Limiting Flow Comment (36) Radioactive Gas 5U Reference 21 Reference NA NA 15.7.1 Waste System Leak (Unaffected) 21 or Failure (37) Radioactive Liquid 5U Reference 10 Reference NA NA 15.7.2 Waste System Leak (Unaffected) 10 "

  • or Failure (38) Postulated 5U Reference 10 Reference NA *:... . . NA 15.7.3 Radioactive Releases (Unaffected) 10 Due to Liquid Tank Failures (39) Fuel Handling 5U References 22, References NA . NA 15.7.4 Accidents in the (Unaffected) 23, 24, and 25 25 and 26 Containment and . .

Spent Fuel Storage Buildings (40) Anticipated 41 Reference 27 Reference NA Based on NA Current AOR 15.8 Transients Without (Bounded by 28 nominal total assumes Trip current AOR) RCS flow of conservatively low 377,600 gpm RCS flow of 377,600 gpm (41) Formerly Chapter 15 5U NA NA NA.. NA 15.9 Appendix A - Models (Unaffected)

Used for Calculation of Accident Doses (42) Formerly Chapter 15 5U References 29, References NA NA 15.10 Appendix B - (Unaffected) 30, and 31 9 and 32 Supplementary Analyses _ _ _. "_.

Attachment 1 Table 2 Page 9

0 Evaluated Cases (see Note 3)

Approved by NRC or Conducted LAR Designation Using Methods Core Flow in Core Flow in and Evaluation or Processes Reference High RCS Flow/ Current AOR Evaluation with UFSAR Type Approved by for NRC Low RCS Flow (see Note 2) Reduced RCS Section Analysis Title (see Note 1) NRC Approval Limiting Flow Comment (43) Containment 4B Reference 33 References High 388,500 gpm NA Current AOR 6.2.1.1.3.1 Performance (Bounded by 34 and 35 assume 6.2.1.1.3.3 Analyses current AOR) conservatively high 6.2.1.1.3.4 RCS flow of 420,000 6.2.1.3.2 6.2.1.4 gpm (Unit 1 analysis is used since results bound those for Unit 2)

(44) Postulated 4B References 33 References High 388,500 gpm NA Current AOR 6.2.1 Secondary System (Bounded by and 36 34, 35, and assumes Pipe Rupture Outside current AOR) 37 conservatively high Containment RCS flow of 420,000 1 gpm (45) LOCA Blowdown 4A NA NA High Based on NA Current AOR 3.6.2.2.1 Reactor Vessel and (Evaluated) total RCS assume 3.9.1.4 Loop Forces flow rate of conservatively high 398,000 gpm RCS flow of 398,000 gpm NOTE 1: LAR Designations listed by Category pages B-8 to B-1 1.

NOTE 2: "Core Flow" represents net flow through core in analysis, defined as Total RCS flow minus Core Bypass Flow.

NOTE 3: Transients which are bounded by other transient results, not applicable for Catawba Unit 2, or unaffected by the proposed RCS flow reduction have shaded fields in this column.

REFERENCES:

1. DPC-NE-3002-A, Revision 4b, "McGuire and Catawba Nuclear Station UFSAR Chapter 15 System Transient Analysis Methodology", September 2010
2. Letter from Tim Reed (NRC) to H. B. Tucker (Duke) dated November 15, 1991, "Safety Evaluation on Topical Report DPC-NE-3002, "FSAR Chapter 15 System Transient Analysis Methodology," (TAC No. 66850)"

Attachment 1 Table 2 Page 10

3. Letter from Robert Martin (NRC) to M. S. Tuckman (Duke) dated December 28, 1995, "Safety Evaluation for Revision 1 to Topical Report DPC-NE-3002-A, "FSAR Chapter 15 System Transient Analysis Methodology" McGuire Nuclear Station, Units 1 and 2; and Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M89944, M89945, and M89946)"
4. Letter from Herbert Berkow (NRC) to M. S. Tuckman (Duke) dated April 26, 1996, "Safety Evaluation on Change to Topical Report DPC-NE-3002-A on Opening Characteristics of Safety Valves - McGuire Nuclear Station, Units 1 and 2; and Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M94405, M94406, M94407, and M94408)"
5. Letter from Chandu Patel (NRC) to G. R. Peterson (Duke) dated April 6, 2001, "Catawba Nuclear Station, Units 1 and 2 RE: Revision 4 to the Duke Energy Corporation Topical Report DPC-NE-3002-A, "UFSAR Chapter 15 Transient Analysis Methodology" (TAC Nos. MA8928 and MA8929)"
6. DPC-NE-3001-PA, Revision Oa, "McGuire and Catawba Multidimensional Reactor Transients and Safety Analysis Physics Parameters Methodology", May 2009
7. Letter from Timothy Reed (NRC) to H. B. Tucker (Duke) dated November 15, 1991, "Safety Evaluation on Topical Report DPC-NE-3001, "Multidimensional Reactor Transients and Safety Analysis Physics Parameters" (TAC Nos. 75954/75955/75956/75957)"
8. Technical Information Document (TID) 14844, Calculation of Distance Factors for Power and Test Reactors, U.S. Atomic Energy Commission, 1962
9. 10 CFR Part 100, Section 100.11
10. Regulatory Guide (RG) 1.77, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors"
11. WCAP-12945-P-A, Volume 1, Revision 2, and Volumes 2 through 5, Revision 1, "Code Qualification Document for Best-Estimate LOCA Analysis," March 1998.
12. WCAP-10054-P-A, 'Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985.
13. Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Power Company LLC d/b/a Duke Energy Carolinas, LLC McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 License Amendment Request for Full Scope Implementation of the Alternative Source Term," March 20, 2008.
14. Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas: LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 License Amendment Request for Full Scope Implementation of the Alternative Source Term Revision to Control Room Atmospheric Dispersion Factors," March 20, 2008.
15. Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 License Amendment Request for Full Scope Implementation of the Alternative Source Term. Response to Request for Additional Information," October 6, 2008.
16. Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Response to Request for Additional Information related to the License Amendment Request for Implementation of Alternative Source Term," December 17, 2008.
17. Letter from B.H. Hamilton (Duke) to U.S. Nuclear Regulatory Commission, "Duke Energy Carolinas, LLC (Duke) McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Response to Request for Additional Information related to the License Amendment Request (LAR) for Implementation of Alternative Source Term (AST)," February 12,2009.
18. Letter from R. C. Jones (NRC) to N. J. Liparulo (Westinghouse) dated June 28, 1996, "Acceptance for Referencing of the Topical Report WCAP-12945 (P), Westinghouse Code Qualification Document for Best Estimate Loss-of-Coolant Analysis".

Attachment 1 Table 2 Page 11

19. Letter from R. C. Jones (NRC) to E. P. Rahe (Westinghouse) dated May 21, 1985, "Acceptance for Referencing of Licensing Topical Report WCAP-10054 (P), Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code".
20. Letter from John Stang (USNRC) to B.H. Hamilton (Duke), "McGuire Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Adoption of the Alternative Source Term Radiological Analysis Methodology (TAC Nos. MD8400 and MD8401)," March 30, 2009.
21. Regulatory Guide (RG) 1.109, "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"
22. Letter from G. R. Peterson (Duke) to U.S. NRC dated December 20, 2005, "License Amendment Request for Selective Implementation of the Alternative Source Term and Revision to Technical Specification 3.9.4, Containment Penetrations"
23. Letter from G. R. Peterson (Duke) to U.S. NRC dated May 4, 2006, "License Amendment Request for Selective Implementation of the Alternative Source Term and Revision to Technical Specification 3.9.4, Containment Penetrations. Response to Request for Additional Information"
24. Letter from G. R. Peterson (Duke) to U.S. NRC dated August 31, 2006, "License Amendment Request for Selective Implementation of the Alternative Source Term and Revision to Technical Specification 3.9.4, Containment Penetrations. Additional Commitments Regarding Containment Closure Administrative Controls."
25. ISG-5, Revision 1 - "Confinement Evaluation", Spent Fuel Project Office, NRC
26. Letter from John Stang (NRC) to G. R. Peterson (Duke) dated December 22, 2006, "McGuire Nuclear Station, Units 1 and 2, Issuance of Amendments Regarding Implementation of Alternative Source Term Methodology (TAC Nos. MC9746 AND MC9747)"
27. Letter from T. M. Anderson (Westinghouse) to S. H. Hanauer (NRC) dated December 30, 1979, "NS-TMA-2182, ATWS Submittal"
28. Letter from Darl S. Hood (NRC) to H. B. Tucker (Duke) dated November 6, 1987, "ATWS Rule (10 CFR 50.62) for McGuire and Catawba Nuclear Stations, Units 1 and 2 (TACs 59081/59111/59112/64535)"
29. Letter from Hal Tucker (Duke) to Harold Denton (NRC) dated February 17, 1984, "McGuire Nuclear Station Docket Nos. 50-369, 50-370", Spent Fuel Pool Re-rack LAR
30. Letter from Hal Tucker (Duke) to Harold Denton (NRC) dated March 20, 1984, "McGuire Nuclear Station Docket Nos. 50-369, 50-370", Safety & Environmental Analysis for Spent Fuel Pool Re-rack
31. Regulatory Guide (RG) 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors (Safety Guide 25)"
32. Letter from Elinor Adensam (NRC) to Hal Tucker (Duke) dated September 24, 1984, "Issuance of Amendment No.35 to Facility Operating License NPF-9 and Amendment No.

16 to Facility Operating License NPF McGuire Nuclear Station, Units 1 and 2"

33. DPC-NE-3004-PA, Revision 1, McGuire and Catawba Mass and Energy Release and Containment Response Methodology, December 2000 Attachment 1 Table 2 Page 12
34. Letter from NRC to M. S. Tuckman (Duke) dated September 6, 1995, "Safety Evaluation for Topical Report DPC-NE-3004-P, "Mass and Energy Release and Containment Response Methodology", McGuire Nuclear Station, Units 1 and 2; and Catawba Nuclear Station, Units 1 and 2 (TAC Nos. M90646, M90647, and M90648)"
35. Letter from NRC to H. B. Barron (Duke) dated February 29, 2000, "McGuire Nuclear Station and Catawba Nuclear Station RE: Review of Topical Report DPC-NE-3004-PA, Rev. 1, Regarding Proposed Finer Nodalization of Ice Condenser (TAC Nos. MA5511, MA5512, MA5517, and MA5518)"
36. Letter from M. S. Tuckman (Duke) to U. S. NRC dated March 15, 1996, "Catawba Nuclear Station, Units 1 and 2, Docket Nos. 50-413 and 414; McGuire Nuclear Station, Units 1 and 2, Docket Nos. 50-369 and 370; Response to Request for Additional Information"
37. Letter from Victor Nerses (NRC) to H. B. Barron (Duke) dated May 5, 1997, "Issuance of Amendments - McGuire Nuclear Station, Units 1 and 2 (TAC Nos. M90590 and M90591)"

Attachment 1 Table 2 Page 13

ATTACHMENT 2 MARKED UP TS PAGE

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 F,00R "IJ110 -ýG JLY 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 in Table 3.4.1-1.

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. Pressurizer pressure or A.1 Restore DNB parameter(s) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> RCS average to within limit.

temperature DNB parameters not within limits.

B. RCS total flow rate > B.1 Reduce THERMAL 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 99%, but < 100% of the POWER to < 98% RTP.

limit specified in the COLR. AND B.2 Reduce the Power Range 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Neutron Flux - High Trip Setpoint below the nominal setpoint by 2% RTP.

(continued)

Catawba Units 1 and 2 3.4.1-1 Amendment Nos.210 & 204

KC0 CW'"'IES T1h!S PACE.

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. RCS total flow rate < C.1 Restore RCS total flow rate 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 99% of the value to > 99% of the value specified in the COLR. specified in the COLR.

OR C.2.1 Reduce THERMAL 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> POWER to < 50% RTP.

AND C.2.2 Reduce the Power Range 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Neutron Flux - High Trip Setpoint to < 55% RTP.

AND C.2.3 Restore RCS total flow rate 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to > 99% of the value specified in the COLR.

-jb D. Required Action and D.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

Catawba Units 1 and 2 3.4.1-2 Amendment Nos.210 & 204

- t A.

~3R'~S TIS PACE.

FGJ1 l2C TlOl,~OL RCS Pressure, Temperature, and Flow DNB Limits 3-4.1

()

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is within limits. In accordance with the Surveillance Frequency Control Program SR 3.4.1.2 Verify RCS average temperature is within limits. In accordance with the Surveillance Frequency Control Program SR 3.4.1.3 Verify RCS total flow rate is within limits. In accordance with the Surveillance Frequency Control Program SR 3.4.1.4 Perform CHANNEL CALIBRATION for each RCS total In accordance with flow indicator, the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.1-3 Amendment Nos. 263, 259

A * , ft RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 Table 3.4.1-1 (page 1 of 1)

RCS DNB Parameters No. OPERABLE PARAMETER INDICATION CHANNELS LIMITS

1. Indicated RCS Average meter 4 < the value specified Temperature - Unit 1 in the COLR meter 3 < the value specified in the COLR computer 4 < the value specified in the COLR computer 3 < the value specified in the COLR Indicated RCS Average meter 4 < the value specified Temperature - Unit 2 in the COLR meter 3 < the value specified in the COLR computer 4 < the value specified in the .COLR computer 3 < the value specified in the COLR
2. Indicated Pressurizer meter 4 > the value specified Pressure in the COLR meter 3 > the value specified in the COLR computer 4 > the value specified in the COLR computer 3 > the value specified in the COLR
3. RCS Total Flow Rate A38 0 gpm and>

/ the limit specified in the COLR (Unit 1);

~ý 9 00 gpm and >

ecified in the COLR (Unit 2)

Catawba Units 1 and 2 3.4.1-4 Amendment Nos67YD

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