ML052230059

Request for Additional Information 1.38 Percent Measurement Uncertainty Recapture Power Uprate License Amendment Request

ML052230059
Person / Time
Site:	Calvert Cliffs
Issue date:	07/18/2005
From:	Vanderheyden G Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MC6210, TAC MC6211
Download: ML052230059 (55)

Text

George Vanderheyden Vice President Calvert Cliffs Nuclear Power Plant Constellation Generation Group, LLC 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410.495.4455 410.495.3500 Fax I3 Constellation Energy July 18, 2005 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

SUBJECT:

Document Control Desk Calvert Cliffs Nuclear Power Plant Unit Nos. 1 & 2; Docket Nos. 50-317 & 50-318 Request for Additional Information Re: 1.38 Percent Measurement Uncertainty Recapture Power Uprate License Amendment Request (TAC Nos. MC6210 and MC621 1)

REFERENCES:

(a)

Letter from Mr. G. Vanderheyden (CCNPP) to Document Control Desk (NRC),

dated January 31, 2005, License Amendment Request:

Appendix K Measurement Uncertainty Recapture -

Power Uprate Request (b)

Letter from Mr. R. V. Guzman (NRC) to Mr. G. Vanderheyden (CCNPP), dated June 16, 2005, same subject Reference (a) submitted a request to increase the licensed core power level for Calvert Cliffs Units I and 2 based on the use of more accurate feedwater flow measurement instrumentation.

The Nuclear Regulatory Commission staff has requested additional information concerning this license amendment request (Reference b).

Attachments (1) and (4) to this letter provide responses to the requested information. Attachment (2) provides corrected pages for Reference (a), as described in Attachments (1) and (4) to this letter. Attachment (5) contains a Westinghouse calculation as'described in Attachments (1) and (4). Some of the information contained in the attached responses is proprietary. Accordingly, it is requested that the proprietary information contained in Attachments (4) and (5) be withheld from public disclosure in accordance with the provisions of 10 CFR 2.390 and that this material be appropriately controlled. The reasons for the classification of this material as proprietary are delineated in the affidavit provided in Attachment (3). The non-proprietary version of the response is provided in Attachment (1).

As noted in the affidavit, there is no non-proprietary version of the attached proprietary calculation

[Attachment (5)].

The information contained in these responses supplements the information provided in Reference (a) and does not affect the No Significant Hazards Determination or the Environmental Consideration provided in that letter.

Document Control Desk July 18, 2005 Page 2 Should you have questions regarding this matter, please contact Mr. L. S.

at (410) 495-4922.

Very trul STATE OF MARYLAND COUNTY OF CALVERT

TO WIT:

I, George Vanderheyden, being duly sworn, state that I am Vice President - Calvert Cliffs Nuclear Power Plant, Inc. (CCNPP), and that I am duly authorized to execute and file this License Amendment Request on behalf of CCNPP. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other CCNPP employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subscribed and vsworn before me, a NotaryHiibli in and for the State of Maryland and County of 7Y,4 this /6~'<dayof

/JLy 2005.

WITN ESS my hai'd and Notarial Seal:

My Commission Expires:

Notary Publi It 0e200Z Date GV/PSF/bjd

Document Control Desk July 18, 2005 Page 3 Attachments:

(1)

Non-Proprietary Response to Request for Additional Information on the Measurement Uncertainty Recapture Power Uprate (2)

Summary of Calvert Cliffs Nuclear Power Plant Measurement Uncertainty Recapture Evaluation Corrected Pages (3)

Westinghouse Proprietary Affidavit (4)

Proprietary Response to Request for Additional Information on the Measurement Uncertainty Recapture Power Uprate (5)

Westinghouse Proprietary Calculation CN-PS-02-8, Revision 0 cc:

P. D. Milano, NRC (Attachments 1 and 2)

S. J. Collins, NRC Resident Inspector, NRC R. 1. McLean, DNR

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE Calvert Cliffs Nuclear Power Plant, Inc.

July 18,2005

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE NRC Ouestion i

1.

In Table 11-1 of the January 31, 2005, submittal, the licensee for the Calvert Cliffs Nuclear Power Plant, Units I and 2 (CCNPP I and 2), indicated that the power uprate conditions are bounded by the Updated Final Safety Analysis Report (UFSAR) Chapter 14 loss-of-coolant accident (LOCA) and non-LOCA transient analyses. Please provide the following information:

A.

Confirm that all UFSAR Chapter 14 events will remain bounded for operation at the new MUR [measurement uncertainty recapture] power uprate power level.

As part of the confirmation provide a list of references which demonstrates that each UFSAR Chapter 14 event remains bounded. Provide the specific details for each reference (i.e., flowpath) justifying how the referenced analyses bound the event. If multiple references are used to bound a single event, provide a detailed explanation of the relationship between the references.

For each UFSAR Chapter 14 event, provide a reference to the Nuclear Regulatory Commission (NRC) approval of the analysis. If there is not a specific reference to the NRC approval of the analysis, provide a detailed explanation and justification for operation at the measurement uncertainty recapture (MUR) power level.

CCNPP Response Summary of LOCA and Non-LOCA Safcty Analyses (UFSAR Chapter 14)

Bounded for UFSAR Operation at Reference Reference Discussion Topic Section MUR Power Last NRC 50.59/AOR Section Uprate Level Approval Control Element Assembly 14.2 Y

1 26, 24 A.2 Withdrawal Event Boron Dilution Event 14.3 Y

26 A.3 Excess Load Event 14.4 Y

2 26, 27 A.4 Loss of Load Event 14.5 Y

1 25, 28 A.5 Loss of Feedwater Flow 14.6 Y

3 25, 29 A.6 E v en t__

Excess Feedwater Heat 14.7 Y

2 26, 27 A.7 Removal Event Reactor Coolant System 14.8 Y

2 25, 30 A.8 Depressurization Event Loss of Coolant Flow Event 14.9 Y

1 26, 32, 33 A.9 Loss of Non-Emergency AC 14.10 Y

2 25,35 A.10 Power Event Control Element Assembly 14.11 Y

2 26,36 A.1 1 Drop Event Asymmetric Steam Generator 14.12 Y

2 23, 37 A.12 Event Control Element Assembly 14.13 Y

2 23, 38 A.13 Ejection Event Steam Line Break Event 14.14 Y

1 23, 26,39, A.14 4 0,4 1 Steam Generator Tube 14.15 Y

1 23, 42 A.15 Rupture Event I

I 1

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE Bounded for UFSAR Operation at Reference Reference Discussion Topic Section MUR Power Last NRC 50.59/AOR Section Uprate Level Approval

_.L(Y/N)

Seized Rotor Event 14.16 Y

2 26,45 A.16 Loss-of-Coolant Accident 14.17 Y

4 A.17 Large Break LOCA 14.17.2 Y

4 25, 26, 46, A.17.a Small Break LOCA 14.17.3 Y

4 25, 26, 46, A.17.b 47 Fuel Handling Incident 14.18 Y**

58, 59 A.18 Turbine-Generator Overspeed 14.19 Y

• t60, 61 A.19 Incident Containment Response 14.20 Y**

53 A.20 Hydrogen Accumulation in 14.21 Y

• • No Longer A.21 Containment Analyzed for Chapter 14 Waste Gas Incident 14.22 Y

50 A.22 Waste Processing System 14.23 Y

50 A.23 Incident Maximum Hypothetical 14.24 Y

62 A.24 Accident Excessive Charging Event 14.25 Y

52 A.25 Feedline Break Event 14.26 Y

1 26,56 A.26

• • Not applicable for reference to previous NRC review.

Table Discussion A.1.

Introduction Each Safety Analysis event listed in the above Table is discussed and treated separately in the following sub-sections with respect to its bounding nature and continued applicability relative to operation at the MUR power uprate power level of 2737 MWth. Included are discussions relating applicable events to respective references for codes, methodologies, limitations, and conditions that support the acceptability of the events and analyses presented with respect to approval of the NRC. All applicable Chapter 14 analyses have accounted for an assumed maximum initial power of 2754 MNVth including uncertainties.

Per the discussion presented in response to Request for Additional Informati6n 11.B (a), sufficient programmatic assurance exists under the process structures of both the approved vendors and Calvert Cliffs to ensure persistence of compliance with future analyses.

A.2 Control Element Assembly (CEA) Withdrawal Event A failure in either the CEA Drive Mechanism Control System or the Reactor Regulating System may initiate a sequential bank withdrawal, inserting positive reactivity and causing increases in reactor power, Reactor Coolant System (RCS) temperature and RCS pressure. The event is terminated by either the Variable High Power Trip, the High Pressurizer Pressure Trip, the Thermal Margin/Low Power Trip, or 2

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE the insertion of negative reactivity due to Doppler and negative Moderator Temperature Coefficient (MTC) feedbacks.

This event was reanalyzed for the Replacement Steam Generators (RSGs) as documented in Reference 25. The RCS flow rate was increased to match the revised minimum Technical Specification value. Plugged tube assumptions were changed to 10%. The analysis for the RSG (as with all applicable analyses discussed in Sections A.2 through A.26) assumes 5480F as the maximum indicated core inlet temperature. All results were found to be acceptable against approved criteria with implementation of the RSGs.

The current Analysis of Record (AOR) for the CEA Withdrawal Event is analyzed and documented in Reference 24.

In support of the MUR power uprate, this referenced analysis is performed with the assumption of a rated power of 2746 MWth, which bounds the MUR power uprate power level of 2737 MWth.

This re-analysis also implements the Asea Brown Boveri, Inc.-Turbo Vane (ABB-TV) correlation for critical heat flux (approved in Reference 7), and makes all appropriate input and assumption adjustments associated with both ABB-TV and the MUR power uprate. Included adjustments associated with the MUR power uprate are: permitted CEA insertions, inlet temperature program values, and affected Reactor Protective System (RPS) setpoints.

Approved methodologies and codes (References 1, 13, 14, and 15) were used, along with approved associated limits/constraints and acceptance criteria. As with all applicable Chapter 14 analyses, associated with implementation of the ABB-TV critical heat flux correlation was a change in the departure from nucleate boiling (DNB) specified acceptable fuel design limits (SAFDL) to a value of 1.24, determined by application of extended statistical combination of uncertainties (Reference 21). This value is acceptable in relation to the NRC-approved minimum departure from nucleate boiling ratio (DNBR) value of 1.13 associated with the approved methodologies of this analysis as described in the response to Question I.C.(b).

Assuming a rated full power level of 2746 MNYth and implementing ABB-TV, all acceptance criteria are met with respect to DNBR, peak linear heat generation rate (PLHGR), maximum primary and secondary pressure and radiological consequence.

The analysis for CEA Withdrawal is acceptable relative to applicable Safety Evaluation Reports (SERs) and bounds operation at the MUR power uprate power level of 2737 MWth.

A.3 Boron Dilution Event A Boron Dilution Event is defined as any event caused by a malfunction or an inadvertent operation of the Chemical and Volume Control System (CVCS) that results in a dilution of the active portion of the RCS. The analysis of this event covers all six modes of operation, each mode being associated with a required minimum time to lose required shutdown margin. This analysis, most recently documented in Reference 26 for the current operating conditions, is unaffected by the proposed MUR power uprate. The analysis is based on RCS and CVCS volumes, along with the boron concentration, to show that operator action within the required minimum time period will terminate the dilution prior to violating the assumed parameters for shutdown margin. The boron dilution event assumes boron concentration levels associated with operating modes, which will continue to bound the MUR power uprate power level of 2737 MWth.

A.4 Excess Load Event An Excess Load Event, as documented in the Calvert Cliffs UFSAR, is a rapid uncontrolled increase in steam generator steam flow not caused by a Steam Line Break (SLB). In the assumed presence of a negative MTC and Fuel Temperature Coefficient, positive reactivity addition leads to an increase in core power level, decreasing DNBR and linear heat rate margin. The transient continues until the Variable High Power Trip is reached on neutron flux or core temperature differential (AT), terminating the event.

3

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE The limiting scenario is most likely to be caused by a full opening of the turbine control valves, atmospheric dump valves (ADVs), or turbine bypass valves during steady-state operation. Limiting cases are determined at both Hot Full Power (HFP) and Hot Zero Power (HZP).

This event was re-analyzed for the RSG implementation as documented in Reference 25. The RCS flow rate was increased to match the revised Technical Specification minimum value. Some discretionary conservatism was removed from the assumed analytical value for the High Power Trip setpoint, however the new assumed analytical value remained above the field setpoint plus uncertainties. All criteria for acceptance were met via the use of NRC-approved codes and methodologies.

The current AOR, as documented in Reference 27, bounds operation at the MUR power uprate power level of 2737 MWth. That AOR also has been verified to use approved methodologies and codes, along with all associated limits and conditions as prescribed by associated SERs. The selected applicable SER references for this analysis are References 7, 14, and 15. The current AOR at HFP assumes an initial reactor thermal power of 2754.2 MWth, including uncertainties. This thermal power level bounds the proposed MUR power uprate power level of 2737 MWth. All criteria for acceptance are met with respect to DNBR, PLHGR, pressure limits and radiological consequence. The Excess Load Event AOR bounds operation at the MUR power uprate power level of 2737 MWth.

A.5 Loss of Load Event As defined in UFSAR Section 14.5, a Loss of Load Event is defined as any event that results in a reduction in the Steam Generators' heat removal capacity through a loss of secondary steam flow. Such an event could be caused by a closure of all main steam isolation valves (MSIVs), turbine stop valves, or turbine control valves along the steam flow path between the Steam Generators and the high pressure turbine. The most limiting Loss of Load Event is a turbine trip without concurrent reactor trip, or an inadvertent closure of the turbine stop valves at HFP.

This event was re-analyzed for RSG implementation as documented in Reference 25. The RCS flow rate was increased to match the revised Technical Specification minimum value. Plugged tube assumptions decreased to 10%. New instrument uncertainties resulted in a slightly lower minimum pressurizer pressure value of 2164 psia (vs. the previous 2165 psia). The range for the initial pressurizer liquid volume was expanded to 440 ft3 to maximize steam generator pressure. All acceptance criteria were met with the use of NRC-approved methodologies and associated limitations and constraints, as applicable.

Reference 28 provides a bounding AOR for both Calvert Cliffs Units Nos. 1 and 2. The assumed power level for transient initiation at HFP is 2771 MWth, which includes a 2.0% instrument uncertainty and a conservative assumption of an additional 17 MNVtl for reactor coolant pump (RCP) heat. This assumed power level in the analysis of 2771 MWth bounds the proposed operation at an MUR power uprate power level of 2737 MWth and the power measurement uncertainty.

All assumptions and methodologies associated with and documented in the AOR are consistent with previously approved analyses and associated SERs and limitations/conditions for application (Reference 15 for CESEC-ITI). All acceptance criteria were found to be met for the bounding analysis with respect to DNBR, fuel performance, peak pressures (RCS and secondary), and radiological consequence. This analysis, having been performed at HFP with a thermal power of 2771 MWth (including uncertainty and RCP heat), bounds operation at the proposed MUR power uprate power level of 2737 MWth plus uncertainties.

A.6 Loss of Feedwater Flow Event A Loss of Feedwater Flow Event is defined as a reduction or loss of feedwater to the Steam Generators without a corresponding reduction in steam flow from the Steam Generators. The most limiting Loss of 4

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE Feedwater initiating event is determined to be an inadvertent instantaneous closure of the feedwater regulating valves, which results in the largest steam and feedwvater flow mismatch and the most rapid reduction in Steam Generator inventory.

The transient causes an increase in primary and secondary pressures and is ultimately terminated by the High Pressurizer Pressure Trip or the Low Steam Generator Level Trip to ensure that all acceptance criteria are met.

The current AOR for this event was documented in References 25 and 29 and was reviewed and accepted by the NRC as documented in Reference 3. The 'current analysis includes accepted methodologies and evaluations of peak primary pressure, peak secondary pressure, and minimum steam generator inventory for both Calvert Cliffs Unit I and Unit 2 with the RSGs and all associated input and modeling changes.

The assumed initial power for the transient is 2771 MWth, which bounds operation at the MUR power uprate power level of 2737 MWth plus uncertainties. All acceptance criteria were met with respect to DNBR, PLHGR, peak pressures, steam generator inventory, and radiological consequences. The current AOR for Loss of Feedwater Flow bounds operation at the MUR rated thermal power (RTP) uprate power level of 2737 MWth.

A.7 Excess Feedwater Heat Removal Event The Excess Feedwater Heat Removal Event results from an extraction of excessive heat from the RCS through the Steam Generators caused by a reduction in steam generator feedwater temperature without a corresponding reduction in steam flow from the Steam Generators. The limiting circumstance of a loss of both high pressure feedwater heaters, coupled with the presence of a conservatively negative MTC and Fuel Temperature Coefficient, results in a core power increase due to the corresponding decrease in RCS temperature. This reactor power increase causes the system to approach the SAFDLs, and is ultimately mitigated by the Variable High Power Trip.

This analysis is documented as an Appendix to the Excess Load analysis of Reference 27 and discussed in Reference 26. This analysis is bounded by the inputs and results of the AOR for the Excess Load Event.

As the Excess Load Event has already been determined to bound operation at the proposed MUR power uprate power level of 2737 MNVth, the Excess Feedwater Heat Removal Event is also bounded by the current AOR. Bounding inputs with respect to initial reactor core power level (2754.2 MWth, including uncertainties), and associated methodologies, are identical to those discussed for the Excess Load Event.

As such, the current AOR for the Excess Feedwater Heat Removal Event bounds operation at the MUR RTP uprate power level of 2737 MWth.

A.8 Reactor Coolant System (RCS) Depressurization Event The RCS Depressurization Event is considered an Anticipated Operational Occurrence (AOO) for which action of the RPS is required to prevent SAFDL violation. The event is initiated by assuming the inadvertent opening of both power-operated relief valves, resulting in a rapid depressurization of the RCS.

The analysis shows that action of the RPS by way of Thermal Margin/Low Power trip prevents exceeding the associated SAFDLs, particularly DNBR.

The current AOR for this event was evaluated against the impacts associated with RSG installation as documented in Reference 30 and discussed with respect to reload and UFSAR implications in Reference 25. The previous AOR (Reference 31) was found to be bounding with consideration of the negligible impacts of the RSGs on the results of the analysis. As such, Reference 30 justifies the results of Reference 31 for the RSGs, and the results from the previous AOR were not updated in the UFSAR.

As also stated in the AOR, the assumed initial core power does not affect the results of the event.

However, the documented AOR (Reference 30, justifying results of Reference 31 with RSGs) is 5

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE performed with an assumed initial reactor core power level of 2771 MWth (including uncertainties and RCP energy), which bounds operation at the proposed MUR power uprate power level of 2737 MWth plus uncertainties.

A.9 Loss of Coolant Flow Event The Loss of Coolant Flow Event is classified as an AOO for which RPS trips and/or sufficient initial steady-state thermal margin, maintained by the applicable Limiting Conditions for Operation, are necessary to prevent exceeding acceptable limits. This transient event is initiated from a HFP condition and modeled to envelope the occurrence of two separate postulated scenarios for losing power to the RCPs: a complete loss of AC to the plant, and a failure of the fast transfer breakers to close following an assumed trip of the main generator. The intermediate system response to the RCP coast down is a rapid decrease in coolant mass flow rate through the reactor core, causing a rise in enthalpy across the core in the direction of coolant flow. A relatively slight power increase results due to the assumed presence of a positive MTC. The main concern with respect to SAFDLs for this event is DNBR, which is met in the analyses (References 32 and 33) by ensuring that initial steady-state margin is built into the DNB design operating limit such that, in conjunction with crediting of the low flow trip function, the DNBR SAFDL is not exceeded.

As documented in Reference 25, the Loss of Coolant Flow Event was re-analyzed in order to bound operation following RSG installation. For that analysis, RCS flow rate was increased to match the revised Technical Specification minimum value with the RSGs. All acceptance criteria were met.

Subsequent to the analysis for the RSGs, the Loss of Coolant Flow Event was reanalyzed to establish a new AOR (Reference 32) in order to credit the thermal margin gains associated with implementation of TURBO fuel. An additional intent of that analysis was to bound the proposed operation at an elevated power level associated with the MUR power uprate. For this analysis, the maximum core power with uncertainties applied was 2755 MWth, (2746 MWth plus uncertainties). The analyzed maximum power level of 2755 MWth bounds the proposed MUR power uprate power level of 2737 MWth, plus uncertainties. Methodologies associated with this analysis were verified to be consistent and within the limitations and conditions of associated SERs (References 14, 16, and 18) and previously NRC-approved analyses (Reference 1). Reported results for this analysis remain applicable to the current Unit 1 cycle.

The following paragraph describes a subsequent reanalysis (AOR), bounding and applicable to the current Unit 2 cycle for the implementation of Zirconium Diboride (ZrB2) Integral Fuel Burnable Absorber (IFBA) in conjunction with axial blankets. That particular fuel design has not yet been implemented in Unit 1, but is expected to be utilized for the next Unit I cycle. The current AOR, described below, will bound operation at the MUR power uprate power level of 2737 MWth for Unit 1.

Subsequent to the analysis for TURBO fuel implementation, the Loss of Coolant Flow was analyzed again in order to establish a new AOR for Calvert Cliffs cores with ZrB2 IFBA in conjunction with axial blankets. This analysis was performed with an assumed rated core power of 2746 MWth and an assumed maximum power level (including uncertainties) of 2755 MWth. Methodologies associated with this analysis were verified to be consistent and within the limitations and conditions of associated SERs (References 7, 9, 14, and 18) and previously NRC-approved analyses (Reference 1). All acceptance criteria were met with respect to DNBR, PLHGR, peak pressures, and radiological consequence. As the maximum analyzed power level of 2755 MWth (including uncertainties) and assumed RTP of 2746 MWth bound operation at the MUR power uprate power level of 2737 MWth, the Loss of Coolant Flow AOR bounds operation at the MUR power uprate power level of 2737 MWth.

6

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE A.10 Loss of Non-Emerzency AC Power Event The Loss of Non-Emergency AC Power Event involves a loss of electrical power to RCPs, resulting in an RCS flow coastdown that challenges SAFDLs and yields an increased steam release to the atmosphere via the main steam safety valves (MSSVs) and ADVs. With respect to DNBR and PLHGR, this event is bounded by the Loss of Coolant Flow Event described above and documented in References 25, 32, and

33. Loss of Coolant Flow has been verified to bound operation at the MUR power uprate power level, with use of applicable approved codes, methodologies and limitations/constraints.

The Loss of Non-Emergency AC Power was evaluated and documented as an AOR in Reference 35. An explicit analytical calculation was not performed for the reanalysis, but the documented AOR justifies the results of the previous AOR for operation with the RSGs. The analysis was performed at an initial power level of 2754 MWth, including uncertainties, which bounds the proposed MUR power uprate power level of 2737 MWth, plus uncertainties.

As previously stated, all SAFDL limits, including DNBR, are bounded by the Loss of Coolant Flow Event. Additionally, the peak pressures associated with Loss of Non-Emergency AC are bounded by the results of the Loss of Load Event (also discussed above).

Results of the referenced AOR meet all applicable criteria and are verified to be produced by NRC-approved methodologies in accordance with applicable SERs. Current analysis for the Loss of Non-Emergency AC Power bounds operation at the MUR power uprate power level of 2737 MWth.

A.11 Control Element Assembly (CEA) Drop Event The CEA Drop Event entails the drop of a single full length CEA into the core, reducing fission power in the vicinity of the dropped CEA and adding negative reactivity core-wide. A prompt drop in core power and heat flux results from the negative reactivity insertion, the magnitude of which depends on the reactivity worth of the dropped CEA. Assuming an inoperable turbine runback circuit, the resulting power mismatch between the primary and secondary systems leads to a cool-down of the RCS and a subsequent positive reactivity addition due to the effects of a negative MTC. Doppler reactivity and moderator feedbacks ultimately terminate the reactivity excursion, producing a re-stabilized core condition with an asymmetric power distribution and correspondingly higher peaking factors. Criteria with respect to DNB, PLHGR and radiological consequence must be shown analytically to be met.

An AOR was established to justify RSG operation within the bounds of the prior AOR results.

Reference 25 presents the impact of this analysis on operation following installation of the RSGs. In accordance with associated NRC-approved codes, methodologies and restrictions/limitations, the analysis of the CEA Drop Event for the RSGs shows that in all cases where the RSGs may affect the results of the CEA Drop Event, those effects are beneficial with respect to acceptance criteria for SAFDL results.

This event was reanalyzed and a new AOR was established for both Unit I and Unit 2 with References 26 and 36.

The reanalysis and currently bounding AOR were performed in order to implement the methodologies associated with TURBO fuel and to ensure bounding inputs and results for the anticipated MUR power uprate. Rated power for this event is assumed to be 2746 MWth, and the maximum initial power including uncertainties is 2754.2 MWth, which bounds operation at the proposed MUR power uprate power level of 2737 MWth plus uncertainties. The performance of this analysis has been verified by the vendor and Calvert Cliffs to have been done in accordance with all applicable SERs (References 7, 14, and 15) and limitations/conditions.

All results are shown to be acceptable with respect to the acceptance criteria for DNBR, PLHGR, peak pressures, and radiological consequence. As substantiated by the foregoing discussion, the current AOR for the CEA Drop Event bounds operation at the MUR power uprate power level of 2737 MWth.

7

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE A.12 Asymmetric Steam Generator Event The Asymmetric Steam Generator Event is classified as an AOO, described as a rapid imbalance in heat transfer between the two steam generators, initiated by one of the following: a loss of load to one steam generator, excessive increase in load to one steam generator, loss of feedwater to one steam generator, or excessive feedwater flow increase to one steam generator. The limiting cause evaluated for the current AOR at Calvert Cliffs is a loss of load to one steam generator, caused by instantaneous closure of one of two MSIVs. This circumstance produces the most rapid temperature tilt across the core, resulting in a limiting approach to the DNBR SAFDL for this analysis.

As discussed in Reference 25, the Asymmetric Steam Generator Event was reanalyzed to account for operation with the RSGs. The RCS flow rate was increased to match the revised Technical Specification minimum value.

Plugged tube assumptions decreased and allowed asymmetry increased with implementation of the RSGs in the analysis. The resultant calculated maximum steam generator pressure increased from 1001 psia to 1086 psia, but still remained below the 110% design pressure acceptance criterion. All acceptance criteria were met with respect to DNBR and PLHGR SAFDLs.

The current bounding AOR for Units I and 2 is documented in Reference 37. This revision to the AOR explicitly addresses the implementation of TURBO fuel, ABB-TV critical heat flux correlation, and the MUR power uprate. Rated power for this analysis is assumed to be 2746 MWth, and the maximum initial power, including uncertainties, for the analysis is assumed to be 2754.2 MWth. Methodologies and codes associated with this analysis are verified to be consistent and within the limitations and conditions of associated SERs (References 7, 14, and 15) and previously NRC-approved analyses (Reference 2). All acceptance criteria were met with respect to DNBR, PLHGR, peak pressures, and radiological consequence. As such, the current AOR for the Asymmetric Steam Generator Event bounds operation at the MUR power uprate power level.

A.13 Control Element Assembly (CEA) Election Event The CEA Ejection Event results from a postulated complete circumferential break of the control element drive mechanism housing or of the control element drive mechanism nozzle on the reactor vessel head.

The analysis is performed from postulated HFP and HZP initial conditions, each resulting in a rapid core power increase for a brief period of time. Doppler reactivity feedback inhibits the core reactivity and power rise, and the reactor is ultimately shut down by a high power level trip, thereby terminating the transient. The core is protected from fuel damage by CEA insertion limits associated with various power levels (Power-Dependent Insertion Limit of the Technical Specifications) and the high power trip. Being a postulated event, a small fraction of fuel failure is permitted in the analysis within the restrictions of criteria for acceptance placed on deposited energy limits and offsite radiological consequence.

The effect of the RSGs on this analysis was analyzed and documented as discussed in Reference 25. To evaluate the effects of the RSGs on the analysis, RCS flow rate was increased to match the revised Technical Specification minimum value and the effect of an increase in steam generator secondary side mass inventory was investigated. All applicable criteria were verified to be met within the applicability of results for the previously documented AOR.

The current bounding AOR for Units I and 2 at Calvert Cliffs is documented in Reference 38. This revision to the AOR explicitly addresses the implementation of ZrB2 fuel with axial blankets, as well as ZIRLOThI cladding and encompasses the MUR power uprate. Rated power for this analysis is assumed to be 2746 MWth, and the maximum initial power, including uncertainties, for the analysis is assumed to be 2754 MWth. Methodologies and codes associated with this analysis are verified to be consistent and within the limitations and conditions of associated SERs (References 5, 9, 17, and 19) and the NRC-8

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE approved analysis (Reference 2). All acceptance criteria were met with respect to fuel clad failure and radiological consequence. As such, the current bounding AOR for the CEA Ejection Event bounds operation at the MUR power uprate power level for both Calvert Cliffs units.

A.14 Steam Line Break (SLB) Event A SLB Event is defined as a breach in the Main Steam piping that carries steam from the Steam Generators to the turbine-generator and other equipment. That breach in the main steam piping produces an increase in heat extraction by the Steam Generators, causing a cooldown of the RCS. In the presence of an assumed negative MTC, that RCS cooldown leads to an addition of positive reactivity to the RCS.

The transient is terminated by reactor trip associated with the severe decrease in Steam Generator pressure, and the MSIVs in the main steam line close to isolate steam flow from the affected steam generator. The SLB Event is divided analytically into hvo separate phases, pre-trip and post-trip for separate safety concerns and associated evaluation against respective acceptance criteria. The primary concern in the pre-trip SLB analysis is the power excursion related to the RCS cooldown and the assumed negative MTC. A loss of power coincident with reactor trip is also assumed. A limiting combination of break size and MTC is determined parametrically for SLBs both inside and outside Containment during the pre-trip SLB analysis. The primary concern associated with the post-trip analysis is a return-to-power in the vicinity of an assumed stuck rod. Limiting scenarios with respect to DNBR are determined parametrically for HFP and HZP initial conditions, both with and without loss of power.

The critical heat flux correlation utilized in the SLB analysis is the MACBETH correlation, approved by the NRC in Reference 20. Associated with that documented SER is a minimum DNBR limit of 1.30 for the MACBETH critical heat flux correlation. Additional applicable SERs for the SLB and this discussion are References 6, 14, and 15.

Both pre-trip and post-trip portions of the SLB were reanalyzed prior for the impact of the RSGs (as documented in Reference 25), along with a newly analyzed end-of-cycle MTC limit at that time.

For the pre-trip SLB, RCS flow was increased to match the revised Technical Specification minimum value. A detailed parametric study was done on break size, break location, trip signal, and MTC, and yielded a different limiting break size for the RSGs. The assumed initial power level was changed to 2700 MoVth, and a 2% power measurement uncertainty was included in the overall analyses.

Additionally, the pre-trip SLB analysis was revised to include the effects of the RSG venturi steam nozzle flow restrictors. Results remained acceptable with respect to all associated criteria.

The post-trip SLB was also reanalyzed for the RSGs. The analysis included the presence of an integral steam nozzle flow restrictor, which limits break size. The RCS flow rate was increased to match the revised Technical Specification minimum value with the RSGs. All acceptance criteria continued to be met.

The current AOR for the pre-trip SLB was established in References 23 and 39. The maximum initial power level at event initiation assumed in that analysis is 2754.2 MWth, which bounds the proposed MUR power uprate power level of 2737 MWth (plus uncertainties) at Calvert Cliffs. The assumed rated power for that analysis is 2746 MWth, or 2754 MWth including instrument uncertainty. This AOR for the pre-trip SLB credits the thermal margin gains associated with TURBO fuel and the ABB-TV critical heat flux correlation, and bounds the proposed MUR power uprate operation. Acceptance criteria with respect to DNBR, PLHGR, peak pressures and radiological consequence are all met. Therefore, the current AOR for both Unit I and Unit 2 bounds operation at the proposed MUR power uprate power level of 2737 MWth.

9

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE The post-trip SLB Event is currently analyzed separately for each operating cycle to credit the cycle-specific physics input to the analysis. The current AORs for Unit I and Unit 2 are documented in References 40 and 41, respectively. Each of the respective AORs for the two operating units employs the MACBETH critical heat flux correlation (design DNBR 2 1.30, SER Reference 20). Each current AOR for the post-trip SLB is also performed with an assumed rated power level of 2746 MWth and a maximum initial total power of 2754 MWth including instrument uncertainties. This power level input assumption bounds the proposed operation at the MUR power uprate power level of 2737 MWth. As each AOR is specific to the core design and associated physics properties of the system, the AOR for Unit 2 (Reference 41) also bounds operation with ZrB2 IFBA in conjunction with axial blankets. In accordance with the foregoing discussion, the current separate AORs for each unit at Calvert Cliffs bound operation at the proposed MUR power uprate power level of 2737 MWth, and all applicable restrictions, limits and conditions associated with the respective methodologies, codes and correlations are met within the bounds of respective appropriate SER references.

A.15 Steam Generator Tule Rupture Event The Steam Generator Tube Rupture Event is a breach of the barrier between the RCS and the main steam system, resulting in mass transfer between the primary and secondary systems and, more consequentially, a radiological release to the environment through the MSSVs and the ADVs.

This analysis was reanalyzed for the RSGs as discussed in Reference 25. Most factors associated with the results of this analysis are affected favorably by implementation of the RSGs. The RCS flow rate was increased to match the revised Technical Specification minimum value with the RSGs. The analysis demonstrated that no design basis limits for fission product barriers would be exceeded or altered. Impact on calculated dose rates was insignificant and all results were shown to be bounded by previous analyses within applicable acceptance criteria.

The current bounding AOR for the Steam Generator Tube Rupture Event is documented in Reference 42.

As the primary concern associated with this analysis is radiological consequence, a reanalysis was not explicitly performed for TURBO fuel implementation. The AOR is documented as bounding in terms of affected neutronic parameters (e.g., Scram curves) for implementation of ZrB2 IFBAs in conjunction with axial blankets (Reference 9). The AOR is also supported by Reference 43 with regard to justifying parameter assumptions related to proportional and backup heater nominal heat rates, MSSV setpoints, and charging pump flow. The assumed maximum power level at initiation of the transient from HFP conditions in the AOR is 2754 MWth, which bounds operation at the MUR power uprate power level of 2737 MWth, plus uncertainties. The Steam Generator Tube Rupture event as documented in the current AOR bounds operation at the proposed MUR power uprate power level, and meets the requirements, limitations and conditions associated with all applicable SERs.

A.16 Seized Rotor Event The Seized Rotor Event is classified as a postulated event, for which a limited amount of fuel failure is permitted within the bounds of associated acceptance criteria. The transient event is caused by an instantaneous seizure of a RCP shaft, postulated to occur as a result of mechanical failure or a loss of component cooling to the RCP shaft seals. The flow rate rapidly reduces to a value corresponding to three RCPs, as opposed to four. The corresponding reduction in RCS flow rate causes a reactor trip on low RCS flow. The reduction of RCS flow rate results in a degradation of DNBR with respect to the SAFDL.

10

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE The effect of the RSGs on the Seized Rotor Event was documented in Reference 25. Establishing an AOR at the time, RCS flow rate was increased to match the revised Technical Specification minimum value. All acceptance criteria were met.

Reference 45 documents the current AOR, bounding operation of Calvert Cliffs Units 1 and 2, both before and after the MUR power uprate, for the Seized Rotor Event. The AOR credits the thermal margin benefits of TURBO fuel, as realized by application of the ABB-TV critical heat flux correlation in conjunction with CETOP-D (References 7 and 14). The effects of implementation of ZrB2 fuel in conjunction with axial blankets, as implemented for the current Unit 2 cycle, are evaluated in Reference 44. The assumed maximum power level for the currently bounding AOR is 2754 MWth, which bounds the proposed operation following MUR power uprate of 2737 MWth, plus uncertainties.

As all acceptance criteria with respect to DNBR, PLHGR, peak pressures, and radiological consequence are met within the restrictions, limitations and constraints of NRC-approved methodologies and codes, the Seized Rotor Event as currently analyzed bounds operation at the MUR power uprate power level of 2737 MWth.

A.17 Loss-of-Coolant Accident (LOCA)

The LOCA Analyses are performed in order to provide confirmation of the Emergency Core Cooling System (ECCS) performance within the criteria listed in 10 CFR 50.46. The following two subsections address the Analyses of Record for both large break and small break LOCA with respect to the projected MUR power uprate power level of 2737 MWth.

A.17.a Large Break Loss-of-Coolant Accident (LBLOCA)

The LBLOCA was reanalyzed for the RSG implementation as discussed in Reference 25. As part of the LBLOCA analysis for the RSGs, the RCS flow rate was increased to match the revised Technical Specification minimum flow value. The linear heat rate was increased to provide additional plant operating margin. Low Pressure Safety Injection flow was reduced to accommodate potential future pump degradation.

The current AORs, bounding operation for Unit 1 and Unit 2 are found in References 46 and 48, respectively.

The methodology was generically approved by the NRC and documented in Reference 4. Consequences of the LBLOCA continued to meet all acceptance criteria of 10 CFR 50.46.

Unit 1 Reference 46 documents the ECCS performance analysis (AOR) for the current operating cycle for Unit 1, demonstrating conformance with 10 CFR 50.46 for TURBO fuel in a mixed core analysis. The mixed core configuration is analyzed explicitly in Reference 49. The Unit I LBLOCA AOR also demonstrates conformance with 10 CFR 50.46, as applicable up to a power level of 2754 MWth. The limits of applicability for the Unit 1 AOR are:

• Rated thermal power (including measurement uncertainty) < 2754 MWth

• Peak Linear Heat Generation Rate of 14.3 wv/ft

• RSGs with tube plugging < 10% per Steam Generator ZIRLOT" and Zircaloy-4 clad Erbia and U0 2 Fuel Rods [both value added pellet (VAP) and previous standard fuel design]

• Fuel rods manufactured at both Hematite and Columbia manufacturing facilities

• Fuel assemblies with Turbo grids and with standard grids (mixed core, Reference 49)

The analysis for LBLOCA, as documented in Reference 46, is performed in accordance with the "1999 EM" Evaluation model, as approved by the NRC in Reference 10. Reference 5 provides the SER I1I

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE reference for limitations and conditions associated with ZIRLOT1I for the Unit I LBLOCA model. Unit I has not yet implemented ZrB2 IFBA fuel with axial blankets. This analysis and the explicit mixed core analysis of Reference 49 conform to all restrictions, limitations and conditions associated with applicable SERs.

The ECCS acceptance criteria of 10 CFR 50.46 are compared to the calculated results for the limiting break size in the LBLOCA analysis for the current Unit I cycle.

Parameter Criterion Result Peak Cladding Temperature

< 22000F 20260F Maximum Cladding Oxidation

• 17%

8.58%

Maximum Core-Wide Oxidation

< 1%

< 0.99%

Coolable Geometry Yes Yes These results are applicable to the current Unit 1 operation cycle with ZIRLOTI cladding in TURBO fuel assemblies, analyzed explicitly for a mixed core condition in Reference 49, and bound operation at the MUR power uprate power level of 2737 MWth.

Untit 2 The current bounding AOR for Unit 2 (and for Unit 1, once ZrB2 is implemented) is Reference 48. The results of that analysis are applicable to the following plant configuration conditions:

• Rated thermal power (including measurement uncertainty)

• 2754 MWth

• Maximum integrated radial peaking factor, Fr, Ms Core Operating Limits Report limit of 1.65 (full power, all rods out operation)

• Full core representation of the TURBO fuel assembly design

• VAP, ZIRLOTMI clad, ZrB2 IFBA, and U02 fuel rod designs operating at a peak linear heat generation rate of 14.5 kw/ft with 2x6-inch low-enriched axial blankets with annular pellets Once-burned VAP ZIRLOTNI clad Erbia fuel rod designs operating at 14.0 kW/ft PLHGR

• RSGs with < 10% tube plugging This bounding analysis employs the "1999 EM" version of Westinghouse's LBLOCA ECCS Performance Evaluation Model for Combustion Engineering designed Pressurized Water Reactors (PWRs), as documented in Reference 51 and approved by the NRC in Reference 10, conforming to the requirements associated with ZIRLOTNI (SER Reference 5) and ZrB2 (SER Reference 9).

The ECCS acceptance criteria of 10 CFR 50.46 are compared to the calculated results for the bounding LBLOCA analysis for any Calvert Cliffs operating cycle that meets the aforementioned applicability criteria.

Parameter Criterion Result Peak Cladding Temperature

< 22000F 20570F Maximum Cladding Oxidation

• 17%

9.95%

Maximum Core-Wide Oxidation

• 1%

< 0.99%

Coolable Geometry Yes Yes All results for the bounding LBLOCA analysis are acceptable with respect to acceptance criteria applied by 10 CFR 50.46. The LBLOCA, as evinced by the foregoing discussion, is performed according to all applicable SERs and bounds operation at the proposed MUR power uprate power level of 2737 MWth.

12

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE A.17.b Small Break Loss-of-Coolant Accident (SBLOCA)

The SBLOCA was reanalyzed for the RSG implementation as discussed in Reference 25. As part of the SBLOCA analysis for the RSGs, the plugged tube assumption was reduced to

• 10%. All consequences and results were within the acceptance criteria of 10 CFR 50.46.

Unit 1 As already mentioned, because Unit I has not yet implemented ZrB2, there is currently a bounding LOCA analysis, both LBLOCA and SBLOCA, for each operating unit (Unit I and Unit 2). The results of the SBLOCA analysis for Unit I are documented in Reference 46. This analysis is performed in accordance with SER References 5 and 11. The plant configuration conditions for SBLOCA are identical to those discussed for the LBLOCA for Unit 1. The results demonstrate conformance for the current Unit 1 cycle to the ECCS acceptance criteria of 10 CFR 50.46 as follows.

Parameter Criterion Result Peak Cladding Temperature

<22000F 1955 0F Maximum Cladding Oxidation

• 17%

9.03%

Maximum Core-Wide Oxidation

• 1%

< 0.81%

Coolable Geometry Yes Yes As the SBLOCA analysis is found to comply with all SER limitations and conditions, and all acceptance criteria for 10 CFR 50.46 are met, the current SBLOCA AOR for Unit I Cycle 17 bounds operation at the MUR power uprate power level of 2737 MWth.

Unit 2 The current AOR for SBLOCA applicable to Unit 2 and future applicable Calvert Cliffs operating cycles is documented in Reference 47 and discussed in Reference 26. The results of Reference 47 are applicable to the following plant configuration conditions:

• Rated thermal power (including measurement uncertainty)

• 2754 MWth.

• TURBO fuel assembly design.

• VAP, ZIRLOTNI and Zircaloy4 clad U02 fuel, with and without Erbia IFBA.

• VAP, ZIRLOT^' clad, ZrB2 IFBA, and U0 2 fuel rod designs with 2x6-inch low-enriched axial blankets with annular pellets.

• RSGs with

• 10% tube plugging.

• Peak linear heat generation rate of 15.0 kW/ft.

This SBLOCA ECCS performance analysis is performed with the NRC-accepted S2M version of the Westinghouse CE SBLOCA EM (Reference 11). As documented above for Unit 1 and for the LBLOCA for both units, the bounding AOR for SBLOCA complies with all limitations and conditions of applicable SERs, such as those associated with ZIRLOT^' and ZrB2.

13

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE The results demonstrate conformance for a bounding SBLOCA analysis (within the conditions of applicability) with respect to acceptance criteria of 10 CFR 50.46 as follows.

Parameter Criterion Result Peak Cladding Temperature

• 22000F 1815 0F Maximum Cladding Oxidation

• 17%

5.70%

Maximum Core-Wide Oxidation

• 1%

< 0.60%

Coolable Geometry Yes Yes As the bounding SBLOCA analysis (for current Unit 2 cycle and applicable future cycles) is found to comply with all SER limitations and conditions, and all acceptance criteria for 10 CFR 50.46 are met, the associated bounding LBLOCA AOR bounds operation at the proposed MUR power uprate of 2737 MWth.

A.18 Fuel Handline Incident The Fuel Handling Incident analysis assumes that a fuel assembly is dropped during fuel handling, either in the containment or in the Spent Fuel Pool. The results of this analysis are dependent upon the radionuclide inventory assumed for the dropped fuel assembly. The inventories associated with this analysis have been generated based on an assumption of core operating power of 2754 MWth, and source term values are based on the TID-14844 methodology in accordance with Regulatory Guide 1.25.

Therefore, the current analysis relating to the Fuel Handling Incident bounds operation at the MUR power uprate power level of 2737 MWth.

A.19 Turbine-Generator Overspeed Incident The Turbine-Generator Overspeed Incident is an analyzed event based on the failure of rotating elements of the steam turbines and generators.

This analysis is not a Design Basis Event or AOO and is documented in detail in UFSAR Section 5.3.1.2, not Chapter 14. The thermal power increase related to the MUR power uprate does not impact the results of this analysis. As such, the Turbine-Generator Overspeed Incident bounds operation at the MUR power uprate power level of 2737 MWth.

A.20 Containment Response The Containment Response is a Design Basis Event, the analysis of which verifies the integrity of the containment structure under the adverse pressure and temperature conditions resulting from a postulated LOCA or Main Steam Line Break (MSLB) Event. Parametric combinations of break size, break location, and power level are analyzed to determine the most limiting scenario with specific regard to containment response for both LOCA and MSLB.

Design and acceptance criteria are placed on the limiting temperature and pressure results, which ensure the integrity of the containment structure under the conditions of the analyzed events.

Reference 53 is the current bounding AOR for containment response, applicable to plant conditions with and without the RSGs, and valid beyond a rated power level of 2737 MWth (MUR power uprate). In support of the RSG installation, the bounding AOR (Reference 53) was established.

Reference 54 provides the qualification of the GOTHIC computer code for modeling containment response at Calvert Cliffs. This methodology was implemented at Calvert Cliffs in accordance with the 10 CFR 50.59 process, as documented in Reference 55.

Limiting mass and energy releases are determined parametrically, and include power levels of 2754 MWth. Decay heat values following the modeled plant trip are calculated based on the NRC Branch Technical Position ASB 9-2 for LOCA and American Nuclear Society 1973 for the MSLB analysis. The MSLB results bound those of LOCA with respect to both peak pressure and peak temperature in containment during the course of the analyzed limiting 14

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE events. The limiting initial power level for the MSLB event is 75% RTP, however a power level of 2754 MWth, plus pump heat, was analyzed parametrically with various break sizes to determine the limiting contribution of mass and energy to the containment atmosphere through the break.

All documented bounding results in Reference 53 (AOR) bound operation at the MUR power uprate power level of 2737 MWth, plus uncertainties, and are found to be in compliance with the applicable qualification restraints of Reference 54. Therefore, the current AOR for the Containment Response Analysis is appropriately applicable to, and bounds, operation at the MUR power uprate power level of 2737 MWth.

A.21 Hydrozen Accumulation in Containment This analysis has been deleted from the UFSAR per License Amendment Nos. 262/239. Reanalysis is not required to verify that the analysis bounds operation at MUR power uprate power level of 2737 MWth.

A.22 Waste Gas Incident The limiting Waste Gas Incident analyzed for UFSAR Chapter 14 is an uncontrolled and unexpected release to the atmosphere of radioactive xenon and krypton fission gases stored in one waste decay tank.

The assumed maximum activity, in accordance with Reference 50, is determined based on conditions in the waste gas decay tank shortly after plant heatup and startup after cold shutdown conditions near the end of a 24-month operating cycle. Associated limiting activity levels are calculated in Reference 50 with the assumption of constant full-power operation at 2754 MWth. Radiological consequence limits are met.

The Waste Gas Incident bounds operation at the proposed MUR power uprate power level of 2737 MWth.

A.23 Waste Processine Svstem Incident The Waste Processing System Incident assumes a seismically-induced failure of the reactor coolant Waste Processing System whereby the contents of the system are released.

Reference 50, as discussed in Section A.22, contains the analysis for this event. As previously mentioned, the depletion calculations for generating radio-isotopic inventories for these analyses is performed at a core thermal power level of 2754 MWth. Therefore, the Waste Processing System Incident analysis bounds operation at the proposed MUR power uprate power level of 2737 MWth.

A.24 Maximum Hypothetical Accident The results of this analysis demonstrate bounding compliance with the guidelines of 10 CFR Part 100. As stated in UFSAR Section 14.24, the pre-accident thermal power for the Maximum Hypothetical Accident is 2754 MWth. All methodologies and results are consistent with approved methodologies and previously submitted analyses. The documented Maximum Hypothetical Accident bounds operation at the proposed MUR power uprate power level of 2737 MWth.

A.25 Excessive Chargin2 Event The Excessive Charging Event is analyzed to verify compliance with the limits of Technical Specification 3.4.4, and to provide the basis for associated alarm setpoints.

Specifically, the AOR, Reference 52, verifies that operator action no sooner than 15 minutes following receipt of pressurizer high level alarm suffices to terminate the event without violating limits on pressurizer level. The associated analysis is based on RCS volumes and CVCS flow rates (letdown and charging). Reactor power level does not affect the results. The current AOR is bounding and acceptable with respect to the plant configuration (charging pump flows, installed pressurizer level setpoints, etc.), and remains valid for the power level associated with the MUR power uprate.

15

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NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE A.26 Feedline Break Event The Feedline Break Event is a postulated accident whereby a piping failure occurs downstream of the check valves between the Steam Generator and Containment. The affected Steam Generator empties, causing elevated temperatures in that Steam Generator and the RCS. A reactor trip occurs on either loss of Steam Generator Level or High Pressurizer Pressure, terminating the pressure transient in combination with the opening action of the pressurizer safety valves and MSSVs.

The Feedwater Line Break Event was reanalyzed for the RSGs as described in Reference 25. Reactor Coolant System flow rate was increased to match the revised Technical Specification minimum value.

The minimum initial pressurizer pressure was decreased to account for uncertainties.

The analyses demonstrated no design basis limits for fission product barriers will be exceeded or altered. For the RSGs, the impact of the increase in Steam Generator secondary side mass inventory on calculated offsite doses was found to be insignificant.

The site boundary doses for the Seized Rotor Event remained bounding with respect to the results of the Feedline Break. All results were acceptable with respect to acceptance criteria.

The AOR for the Feedline Break Event is contained in Reference 56, and described in Reference 44, and bounds operation under current and proposed MUR power uprate power levels. The maximum core power level assumed in the analysis as an input condition is 2771 MWth, including rated power plus uncertainties and RCP energy. All acceptance criteria for the event with regard to DNBR, peak RCS and secondary pressure limits, radiological consequence, and long-term cooling capability are verified to have been met. Additionally, all methodologies and code implementation are consistent with the most recent NRC-reviewed analysis documented in Reference 1. Compliance with applicable SERs is verified for use of CESEC-III (Reference 15). As all results and methodologies are acceptable, all results meet associated acceptance criteria, and the maximum initial power level exceeds the proposed MUR power uprate plus uncertainties, the results of the current Feedline Break Event AOR bound the proposed MUR power uprate power level of 2737 MWth.

Table References

1.

A. W. Dromerick (NRC) to C. H. Cruse (BGE), "Issuance of Amendments for Calvert Cliffs Nuclear Power Plant Unit No. I (TAC No. M97855 and Unit No. 2 (TAC No. M97856)," Docket Nos. 50-317 and 50-318, May23, 1998

2.

D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BGE), "Amendment No. 71 to Facility Operating License No. DPR-53 for Calvert Cliffs Nuclear Power Plant, Unit No. I," June 24, 1982 D. H. Jaffe (NRC) to A. E. Lundvall, Jr. (BGE), "Amendment No. 61 to Facility Operating License No. DPR-69 for Calvert Cliffs Nuclear Power Plant, Unit No. 2," January 10, 1983

3.

D. M. Skay (NRC) to C. H. Cruse (Calvert Cliffs), " Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2-Amendment RE: Reanalysis of Loss of Feedwater Event (TAC Nos. MB3442 and MB3443)," Docket Nos. 50-317 and 50-318, February 26, 2002

4.

D. G. McDonald (NRC) to G. C. Creel (BGE), "Issuance of Amendment for Calvert Cliffs Nuclear Power Plant Unit No. I (TAC No. M82277)," May 26, 1992.

5.

LTR-ESI-01-224, "Limitation/Constraint Identification in the NRC SER for CENPD-404-P, Rev. 0, 'Implementation of ZIRLOThI Cladding Material in CE Nuclear Power Fuel Assembly Designs,"' C. M. Molnar, December 7, 2001 16

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE

6.

C. 0. Thomas (NRC) to A. E. Scherer (CE), "Acceptance for Reference of Licensing Topical Report CENPD-207(P/NP), C-E Critical Heat Flux Correlation for C-E Fuel Assemblies with Standard Spacer Grids, Non-Uniform Axial Power Distribution," November 2, 1984

7.

S. A. Richards (NRC) to I. C. Rickard (ABB-CE), "Acceptance for Referencing of CENPD-387-P, Revision-00-P, 'ABB Critical Heat Flux Correlations for PWR Fuel' (TAC NO. MA6109),"

March 16, 2000

8.

Not Used.

9.

H. N. Berkow (NRC) to J. A. Gresham (WEC), "Final Safety Evaluation for Topical Report WCAP-16072-P, Revision 00, 'Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs,"' May 6, 2004

10.

S. A. Richards (NRC) to P. W. Richardson (WEC), "Safety Evaluation of Topical Report CENPD-132, Supplement 4 Revision 1, 'Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model' (TAC No. MA5660)," December 15, 2000

11.

T. H. Essig (NRC) to 1. C. Rickard (WEC), "Acceptance for Referencing of the Topical Report CENPD-137(P), Supplement 2, 'Calculative Methods for the C-E Small Break LOCA Evaluation Mode' (TAC NO. M95687)," December 16, 1997

12.

Not Used.

13.

K. Kniel (NRC) to A. E. Scherer (CE), "Evaluation of Topical Report CENPD-161-P ["TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core"]," September 14, 1976

14.

J. R. Miller (NRC) to J. W. Williams (FPL), "Amendment No. 8 to NPF-16 (SLU 2), with Safety Evaluation (SE)," November 9, 1984 [CETOP-D SER]

15.

C. 0. Thomas (NRC) to A. E. Scherer (CE), "Combustion Engineering Thermal-Hydraulic Computer Program CESEC-IIl," April 3, 1984

16.

CENPD-183-A (inc. Amendment I-P), "Loss of Flow - C-E Methods for Loss of Flow Analysis,"

June 1984 H. Bernard (NRC) to A. E. Scherer (CE), "Acceptance for Referencing of Licensing Topical Report CENPD-1 83," May 12, 1976

17.

CENPD-190-A, "CEA Ejection, C-E Method for Control Element Assembly Ejection," July 1976

0. D. Parr (NRC) to A. E. Scherer (CE), June 10, 1976

18.

CENPD-188-A, "HERMITE:

A Multi-Dimensional Space-Time Kinetics Code for PWR Transients," March 1976

0. D. Parr (NRC) to A. E. Scherer (CE), June 10, 1976

19.

R. L. Baer (NRC) to A. E. Scherer (CE), "Evaluation of Topical Report CENPD-135 Supplement 5," September 6, 1978

20.

R. C. Clark (NRC) to A. E. Lundvall, Jr. (BGE), No Title, July 15, 1983

21.

S. A McNeil, Jr. (NRC) to J. A. Tiernan (BGE), "Safety Evaluation of Topical Report CEN-348(B)-P, 'Extended Statistical Combination of Uncertainties,"' October 21, 1987

22.

Not Used.

23.

SE00492, Revision 0000, "Unit I Cycle 17 Reload Physics and Transients Safety Evaluation,"

April 23, 2004 17

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE

24.

CA06346, Revision 0001, "Calvert Cliffs Units I & 2 Control Element Assembly Withdrawal Event," December 14, 2004

25.

SE00471, Revision 0001, "Unit I Cycle 16 Reload Physics and Transients Safety Evaluation,"

November 21, 2002

26.

SE00495, Revision 0003, "Unit 2 Cycle 16 Core Reload (2005 RFO)," March 11, 2005

27.

CA063 89, Revision 0000, "Calvert Cliffs Units I & 2 Excess Load Event," April 13, 2004

28.

CA05745, Revision 0000, "Calvert Cliffs Units 1 & 2 Loss of Load Transient Analysis,"

February 1, 2002

29.

CA05733, Revision 0001, "Calvert Cliffs Units I & 2 Loss of Feedwater Flow Event," February 22, 2002

30.

CA03352-0001, "Revision Completed to Support UIC16 Reload Including Replacement Steam Generators," February 22, 2002.

[Calvert Cliffs Owner Acceptance Review of Westinghouse Calculation A-CC-FE-0029, Revision 02, "Calvert Cliffs Units I and 2 RCS Depressurization Event Analysis"]

31.

CA03552, Revision 01, "Calvert Cliffs Units I and 2 RCS Depressurization Event Analysis for the Low Flow Reduction Project," January 20, 1997

32.

CA06381, Revision 0000, "Calvert Cliffs Units I & 2 Loss of Coolant Flow Event," March 30, 2004

33.

CA06509, Revision 0000, "Calvert Cliffs Units 1 & 2 Loss of Coolant Flow Event," February 23, 2005

34.

Not Used.

35.

CA3553-001, Calculation Change Notice for "BGE Calvert Cliffs Units 1 and 2 Loss of Non-Emergency AC Power Evaluation for Reduced Flow and 2500 Plugged Tubes," Addressing Revision Completed to Support UIC16 Reload Including Replacement Steam Generators, February 22, 2002

36.

CA06385, Revision 0000, "Calvert Cliffs Units I & 2 Control Element Assembly Drop Event,"

March 30, 2004

37.

CA06388, Revision 0000, "Calvert Cliffs Unit I & 2 Asymmetric Steam Generator Event,"

March 15,2004

38.

CA06508, Revision 0000, "Calvert Cliffs Units I & 2 Control Element Assembly Ejection Event,"

February 3, 2005

39.

CA06383, Revision 0000, "Calvert Cliffs Units I & 2 Pre-Trip Steam Line Break Event,"

March 15, 2004

40.

CA06382, Revision 0000, "Calvert Cliffs Unit 1 Cycle 17 Post-Trip Steam Line Break Event,"

March 22, 2004

41.

CA06510, Revision 0000, "Calvert Cliffs Unit 2 Cycle 16 Post-Trip Steam Line Break Event,"

February 3, 2005

42.

A-CC-FE-0067, Revision 07, "Calvert Cliffs SGTR Event with EOP-Based Operator Actions and Isolated ADVs," December 15, 2003

43.

A-CC2-FE-0097, Revision 003, "Calvert Cliffs Unit 2 Cycle 14:

Evaluation of Non-LOCA Transient Analyses and Summary of Setpoint Analysis Inputs," January 12, 2001 1 8

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE

44.

LTR-TAS-04-100, Revision 0, "Assessment of Scram Curve Changes on Various Non-LOCA Safety Analyses for Calvert Cliffs Unit 2 Cycle 16," September 21, 2004

45.

CA063 84, Revision 0000, "Calvert Cliffs Units I & 2 Seized Rotor Event," March 30, 2004

46.

CA06380, Revision 0000, "Calvert Cliffs Unit I Cycle 17 ECCS Performance Analysis and Appendix K Uprate Evaluation," March 17, 2004

47.

CA06551, Revision 0000, "Calvert Cliffs Units I and 2 SBLOCA ECCS Performance Analysis for Implementation of ZrB2 IFBA and Axial Blankets," January 15, 2005

48.

CA06550, Revision 0000, "Calvert Cliffs Units I and 2 1999 EM LBLOCA ECCS Performance Analysis for Implementation of ZrB2/Axial Blankets," January 15, 2005

49.

A-CC-FE-01 14, Rev. 000, "Calvert Cliffs Units 1 and 2 Mixed Core ECCS Performance Analysis with Turbo Fuel," December 17, 2001

50.

CA05994, Revision 0000, "RC Waste Processing System Incident and Waste Gas Incident Dose Analysis," October 18, 2002

51.

CENPD-132, Supplement 4-P-A, "Calculative Methods for the C-E Nuclear Power Large Break LOCA Evaluation Model," March 2001

52.

CA03746, Revision 0000, "Excess Charging Event," August 28, 1998

53.

CA05892, Revision 0001, "Containment Response to OSG and RSG DBA for USAR," May 15, 2002

54.

CA03559, Revision 0000, "Topical Report, GOTHIC Code Containment Response Analysis Model Qualification," April 9, 1998

55.

SE00040, 10 CFR 50.59 Safety Evaluation, "Updated FSAR 14.20 Containment Response Safety Evaluation," December 20, 1995

56.

CA0621 1, Revision 0000, "Calvert Cliffs Units I & 2 Feedwater Line Break Event," March 7, 2003

57.

Not Used.

58.

NEU 94-030, Revision 0, "Offsite Doses at the Exclusion Area Boundary with a Fuel Handling Incident in the Spent Fuel Pool Area," December 22, 1994

59.

000-DA-9302, Revision 1, "Re-evaluation of Fuel Handling Accident Supporting Both Personnel Air Lock Doors Open During Fuel Movement (Open Door Policy)," October 13, 1993

60.

C. E. Rossi (NRC) to J. A. Martin (Westinghouse Electric Corporation), "Safety Evaluation Report, dated February 2, 1987, Approval for Referencing of Licensing Topical Reports: March.1974 Report; WSTG-2-P, May 1981; and WSTG-3-P," July 1984

61.

Turbine Missile Analysis Statement, Constellation Nuclear, Calvert Cliffs Unit 1, TB.# 170X413, Mark S. Page, GE Energy Services, July 3, 2003

62.

M-89-583, Revision 3, "Offsite and Control Room Doses Following a LOCA," July 9, 1991 19

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE NRC Ouestion B.

Provide justification that plant operation at the MUR power level will be acceptable for all UFSAR Chapter 14 events, as a minimum large and small break LOCAs that were not reviewed and approved by the NRC. The discussion should include:

(a) a program that supports acceptability of the analyses in terms of input data used in the NRC-approved computer codes, compliance with safety evaluation (SE) restrictions on the NRC-approved methods, and quality assurance of the analyses and results; and (b) reference to licensing submittals that support adequacy of the analyses including effects of plant operating condition changes, the use of new correlations, and installation of the replacement steam generators (SGs) and new fuel design with mixing vane grids.

CCNPP Response L.B(a) All analyses and evaluations that support the license basis and AOR documentation with respect to UFSAR Chapter 14 are performed in accordance with all requirements and criteria of 10 CFR Part 50, Appendix B, as applied to quality assurance programs within the programmatic structures of approved vendors and the internal engineering processes at Calvert Cliffs. Specifically, all Safety Analysis calculations supplied by approved vendors are subject to an Appendix B program for quality assurance. Calvert Cliffs' internal processes for accepting and/or generating and reviewing associated engineering products within the scope of the Safety Analysis also rely on, and implement, the guidance of 10 CFR Part 50, Appendix B. Additionally, Calvert Cliffs employs a programmatically and procedurally controlled process of vendor oversight to ensure quality and appropriate use of input data within the SER restrictions placed upon all associated NRC-approved codes and methodologies. Each calculation and analysis for UFSAR Chapter 14 is assessed explicitly and separately within the distinct contexts of Appendix B quality assurance programs at both the vendor and Calvert Cliffs with specific respect to its compliance with NRC-approved codes and methodologies within all applicable limits and restrictions.

I.B(b) The response to this question is contained within our answer to Question I.A above.

NRC Question C.

Provide information for the following items:

(a)

List the NRC-approved values of the design safety departure from nucleate boiling ratio (DNBR) limits with the associated SEs approving the values for the CE-1, ABB-NV [Asea Brown Boveri, Inc.-Non-Mixing Vane] and ABB-turbine event critical heat flux correlations with use of the extended statistical combination of uncertainties.

(b)

Confirm that the values of the DNBR limits used in UFSAR Chapters 3.5 and 14 analyses are consistent with the values specified in Section IV-5-2 of the January 31, 2005, submittal and are acceptable, or revise the UFSAR (such as the information for DNBRs in Table 14.9-2) as needed.

(c)

Provide a commitment to reference the correlations that were used for DNBR calculations applicable for current and future cycles in appropriate documents, such as UFSAR Chapters 3.5 and 14, and the core operating limits report.

20

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE (d)

Specify the NRC-approved value of the DNBR penalty factor for the mixing core DNBR analysis with turbo and non-turbo fuel assemblies, or provide ajustification for cases where the core mixing factor was not considered.

(e)

Confirm that the references discussed in each section of Attachment 2 to the January 31, 2005, submittal are adequate and consistent with the references listed in Section IX, Reference Section of Attachment 2.

CCNPP Response 1.C.(a) Minimum DNBR limits listed in Chapter 14 may differ from (but must be greater than) the NRC-approved minimum DNBR limit associated with use of the referenced methodologies with the referenced codes. For instance, the design minimum DNBR for Unit 2 Cycle 16 and Unit 1 Cycle 17, given the application of approved Extended Statistical Combination of Uncertainties methodology, is 1.24.

I.C.(b) Chapters 3 and 14 correctly refer to 95/95 minimum DNBR values for respective codes and methodologies. The statistically derived values (such as 1.24 for Unit 2 Cycle 16 and Unit 1 Cycle 17) are also correctly referenced to the applicable analyses in UFSAR Chapter 14. For instance, Table 14.9-2 correctly references the statistically-derived DNBR limit for that analysis, applicable to the use of the ABB-NV and ABB-TV correlations in TORC and CETOP-D codes (minimum DNBR 2 1.24).

For Unit 2 Cycle 15, which represented the first introduction of TURBO fuel in Unit 2 (approximately 1/3 of the core), the statistical combination of uncertainties resulted in a design limit for DNBR of 1.21. The answer to question 1.C.(c) immediately below addresses the issue of providing clear distinction in the applicable sections of the UFSAR.

1.C.(c) Calvert Cliffs commits to clarifying throughout the UFSAR (e.g., Section 3.5 and Chapter 14) and Core Operating Limits Report, as appropriate, the difference between NRC-approved minimum DNBR limits for a given code or methodology, and the statistically derived design DNBR limits for the safety analyses.

I.C.(d) The safety analyses for Calvert Cliffs that address the implementation of TURBO and non-TURBO fuel assembly designs are not subject to application of a DNBR penalty factor for the "mixed core" condition during the phase-in implementation of TURBO fuel. In lieu of applying a DNBR penalty factor, the TURBO and non-TURBO fuel assemblies are treated explicitly in the safety analysis via the combined application of ABB-NV and ABB-TV critical heat flux correlations. As the mixed core condition is treated explicitly with respect to DNBR, there is no penalty factor required.

I.C.(e) The January 31, 2005 submittal contained misaligned reference citations in Attachment (2),

Section IV.5.3 and Section XI. The affected references are the Section IV references. Revised pages for Section IV.5.3 and a list of references for Section IV are attached. Please replace the affected pages of the January 31, 2005 submittal with the revised pages in Attachment (2) of this letter.

21

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE NRC Ouestion D.

Discuss the impact of the proposed power uprate on the analyses of record for:

(a) natural circulation cooldown.

(b) station blackout (SBO).

(c) capacity of the relief valves for the high temperature overpressure protection (UFSAR Section 4.2.1).

(d) the relief valve capacity for low-temperature overpressure protection (UFSAR Section 4.2.2).

CCNPP Response I.D.(a) The AOR for natural circulation corresponds to full-power operation at a thermal power of 102%

(2754 MWth) and bounds the proposed uprated conditions. As such, the proposed uprate does not impact the documented analysis for verifying natural circulation with the RSGs.

1.D.(b) See answer to Question No. 3 below.

l.D.(c) The limiting design basis accident (DBA) for maximum RCS pressure is a Loss of Load without simultaneous reactor trip while at full power. The Loss of Load scenario is described under UFSAR Section 14.5. The inputs to this accident analysis include the reactor being at 2754 MWth (does not include an additional 17 MWth of pump heat added to core power).

With the 2754 MWth input, the pressurizer safety relief valves are of acceptable relieving capacity to limit the RCS pressure to 2686 psia. Since the uprated conditions under Appendix K are within the bounds of the analyzed Loss of Load event, the pressurizer safety relief valves are verified to be of adequate capacity for power uprate. The pressurizer safety relief valves therefore have enough relieving capacity to protect the RCS against high temperature overpressurization.

The relieving capacity at high temperature is not impacted by MUR power uprate.

I.D.(d) Low temperature overpressurization analyses were performed for the worst-case scenarios for mass addition and energy addition.

The RCS mass addition events were postulated as an inadvertent high pressure safety injection (HPSI) pump start; inadvertent HPSI and charging pump start; and inadvertent mismatch of charging and letdown flow. Since the MUR power uprate does not affect charging, HPSI flow, or letdown flow, it therefore does not affect the RCS mass addition analysis.

The RCS energy addition analyses were postulated to have RCS expansion following loss of shutdown cooling, including steam generator heat addition, inadvertent pressurizer heater actuation; and energy addition from the. steam generator secondary side to the RCS due to a start of an RCP when the SGs are at a higher temperature than the reactor vessel inventory. The inputs to the analysis for the power-operated relief valve flow model uses a decay heat value taken at 102%

power (2754 MWth). This analysis therefore bounds the MUR power uprate conditions.

The relieving capacity for low temperature overpressure protection is unaffected by MUR power uprate.

22

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE NRC Question

2.

Provide information for the following items:

A.

Identify the nature and quantity of megavolt ampere reactive (MVAR) support necessary to maintain post-trip loads and minimum voltage levels as a result of the power uprate.

B.

Identify what MVAR contributions CCNPP I and 2 are credited for providing to the grid following implementation of the power uprate.

C.

After the power uprate, identify any changes in MVAR associated with Items A and B above.

D.

Address the compensatory measures that the licensee would take to compensate for the depletion of the nuclear unit MVAR capability on a grid-wide basis following implementation of the power uprate.

E.

Evaluate the impact of any MVAR shortfall listed in Item D above on the ability of the offsite power system to maintain minimum post-trip voltage levels and to supply power to safety buses during peak electrical demand periods. The subject evaluation should document information exchanges with the transmission system operator.

CCNPP Response 2.A. The responsible transmission operator (RTO) requires that generators connected to the grid maintain a composite power delivery at continuous power output at the generators terminals at a power factor of at least 0.95 leading to 0.90 lagging.

This establishes the minimum reactive capability for the manufacturer's reactive capability curves at the prescribed uprated megawatt (MW) capacity. This data was provided to the RTO along with plant post-trip load data so that the proper stability and loadflow/voltage studies could be run as part of an interconnection impact study. These studies establish that grid minimum voltages (both pre and post uprated unit trip) satisfy the RTO's published voltage ranges for the 500 kV system. These values, for the Calvert Cliffs grid connection, are used as inputs to the plant loadflow/voltage studies.

2.B. Both units' manufacturer's reactive capability curves along with uprated MW ratings were provided to the RTO so that the units can be properly modeled for use in their planning and stability studies.

2.C. Changes to the units' reactive capability is prescribed by the reactive capability curves and the uprated MW values for each unit.

Unit I was uprated +35 MW to:

• Unit I Gross Output 908 MW at 921F Summer Ambient Temperature,

• 367 MVAR Lagging,

• -50 MVAR leading, and

• 35 MW post unit trip load.

Unit 2 was uprated +20 MW to:

• Unit 2 Gross Output 902 MW at 921F Summer Ambient Temperature,

• 350 MVAR Lagging,

• -50 MVAR leading, and

• 35 MW post unit trip load.

23

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE These updated values envelope the expected MUR power uprate values on both Units while still satisfying the RTO's reactive capability requirements.

2.D. No compensatory measures were required since the uprated output of both units met the minimum reactive requirements specified in 2.a above and the resulting stability and voltage studies demonstrated that grid stability and voltage (pre and post trip) requirements were met.

2.E.

A grid stability analysis was performed as part of the Generator Interconnection Impact Study by the RTO with the Calvert Cliffs power uprate and MVAR shortfalls were not identified.

NRC Ouestion

3.

Provide in detail the effect of the power uprate on the SBO coping capability in accordance with Title 10 of the Code of Federal Regulations Section 50.63, "Loss of all alternating current power." The SBO coping analysis should address the condensate inventory for decay heat removal, Class IE battery capacity, compressed air, effect of loss of ventilation, containment isolation, etc.

CCNPP Response The initial evaluation of a SBO event for Calvert Cliffs was performed in accordance with the requirements of Regulatory Guide 1.155, "Station Blackout" and approved by the NRC (References I and 1I). That evaluation included the effects on condensate inventory, class IE battery capacity, compressed air, loss of ventilation, containment isolation, and reactor coolant inventory.

A description of the evaluation is contained in UFSAR Section 1.8.

We have evaluated the effects of the proposed MUR power uprate on the plant response to a SBO and have determined that the plant will continue to meet the requirements of 10 CFR 50.63 following the implementation of the proposed MUR power uprate with no change to the coping capability.

Specifically, the initial assumption for the SBO analyses described in the UFSAR was 102% RTP. Since the proposed MUR power uprate does not infringe upon the 102% RTP limit, there is no impact to this initial condition. Therefore, we believe that the proposed MUR power uprate is acceptable with respect to the plant response to a SBO.

NRC Ouestion

4.

Provide in detail the uprated ratings and the effect of the power uprate on the following equipment:

A.

Main generator B.

Isophase bus C.

Main power transformer D.

Start-up transformer E.

Station power transformer 24

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE CCNPP Response 4.A. Main Generator The current main generator ratings are:

Unit I Main Generator Rating: 1020 MVA at 25 kV, 60 Hz when operated at 0.9 lagging power factor,' and hydrogen pressure of 60 psig (918.0 MW). The operating point (MVAR - MW) is determined as shown in the reactive capability curve contained in the operating procedure.

Unit 2 Main Generator Rating: 1012 MVA at 22 kV, 60 Hz when operated at 0.9 lagging power factor, and hydrogen pressure of 75 psig (910.8 MWV). The operating point (MVAR - MW) is determined as shown in the reactive capability curve contained in the operating procedure.

The ratings of the main generator as described in the UFSAR will not be changed by this activity.

The new operating point corresponding to a 1.6-1.7% power uprate is within the design rating of both machines. The generators are operated to produce power output within the limits of their associated reactive capability curves as described in the associated operating procedure.

If required, the MWAR output of the generators can be adjusted such that the total MVA rating is not exceeded. No modification to auxiliary or support equipment is required due to the MUR power uprate.

To illustrate the effect of MUR power uprate, the following comparison shows the calculated change in generator output MW and MVAR at maximum generation in summer conditions:

MVA MWe MVAR PF Unit I Base 1020 876.0 522.5 0.859 Unit I MUR 1020 890.4 497.6 0.873 Unit 2 Base 1012 891.5 478.9 0.881 Unit 2 MUR 1012 906.3 450.3 0.896 4.B. Isophase Bus The isolated phase bus and switchyard equipment associated with the main generators is rated for maximum current flow from the generators, thus no modification to this equipment is required.

However, the existing unit limitations for conditions when either unit's isophase bus duct cooling is not available still remains.

Unit I and Unit 2 Isophase Bus Duct Ratings (Main Bus)

Unit I Unit 2 General Ratings 25 kV 22 kV 150kVBIL IlOkVBIL 650C Rise 651C Rise Continuous Current Rating 25,000A - Forced Cooled 28,OOOA -Forced Cooled 16,OOOA - Self Cooled 18,OOOA - Self Cooled Momentary Current Rating

'320,OOOA (ASY) 360,OOOA (ASY)

The maximum possible calculated current through the isophase main bus before and after the MUR power uprate is 23,556A and 26,558A for Unit I and Unit 2, respectively which is below the continuous current ratings of 25,OOOA and 28,OOOA.

25

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE 4.C.

Main Power Transformer The main generator step-up transformer ratings are:

• 25/500 kV (Unit 1)

• 22/500 kV (Unit 2)

• 2-810 MVA (Parallel) per unit (1620 MVA total per unit)

Two main power transformers are connected via the isolated phase bus, to the output of each main generator, and are designed to carry the maximum generator output and transform generator output voltage to the transmission system voltage.

Each of the paralleled transformers is rated for 810 MVA at 650C rise. The maximum MVA rating of the Unit 1 and Unit 2 generators remain at 1020 and 1012 MVA, which is well within the rating of the paralleled transformers, thus no changes are required to the generator step-up transformers due to the MUR power uprate.

4.D. Start-Up Transformer Calvert Cliffs does not utilize start-up transformers. Plant loads are connected to the P-13000 station service transformers which are fed from the 500 kV switchyard.

4.E.

Station Power Transformer The ratings of the station service transformers (P-13000-1 and P-13000-2) are:

• 500/14 kV

• 60 MVA (0A) /80 MVA (FA) / 100 MVA (FOA)

• Z = 10.5 % 60 MVA Base The most limiting operating condition for the station service transformers is when one of the two transformers is out-of-service and the other transformer is supplying both units' load. Historically, the combined load requirement of both units is approximately 89.7 MVA. The calculated worst case load flow data with both units in Mode I is 97.4 MVA. The rating of a station power transformer is 100 MVA with forced air and oil cooling. The anticipated load increase due to the MUR power uprate is approximately 0.86 MVA. Therefore the total worst case calculated loading on a station power transformer after the MUR power uprate is 98.3 MVA, which is within the design rating of the station power transformer (100 MVA).

NRC Ouestion

5.

Provide a list of loads affected by the power uprate.

CCNPP Response The only calculated discemable electrical load increase resulting from the MUR power uprate is due to an increase in the Brake Horsepower of the following loads:

• 2000 HP Condensate Booster Pumps 11, 12, 13, 21,22,23

• 1250 HP Condensate Pumps 11, 12, 13, 21, 22, 23

• 1250HPHeaterDrainPumps 11, 12, 21, 22 The resulting maximum total calculated electrical system load increase from these pumps due to the MUR power uprate is approximately 0.86 MVA for both units.

26

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE NRC Ouestion

6.

The proposed 1.38 percent power uprate (2737 MWth) is based on the assumed power measurement uncertainty of 0.62 percent. Response to Criterion 1 shows an uncertainty of 0.56 percent with an operable Crossflow whereas Table 1-1 shows a calculated uncertainty of 0.41 percent for CCNPP I and 0.46 percent for CCNPP 2. Explain these values.

CCNPP Response Table I-1 presents both the random and bias components of uncertainty. The random component of uncertainty summarized in Table 1-1 is 0.41% for Unit 1 and 0.46% for Unit 2.

However, total uncertainty must also include the 0.08% bias component summarized in Table I-1.

Thus, the total uncertainty for Unit 1, using the values summarized in Table I-1, is 0.41% + 0.08% = 0.49%; for Unit 2, 0.46% + 0.08% = 0.54% RTP. With respect to the uprate, the Unit 2 uncertainty is more limiting and is used as the basis for the uprate.

The 0.02% difference between the 0.56% uncertainty described in the response to Criterion I and 0.54%

uncertainty using the values summarized in Table I-1 results from the summarization of the data in Table I-I. For simplicity, the values summarized in the table do not account for the dependency between feedwater flow uncertainty and temperature measurement uncertainty, which results in a slight increase in uncertainty.

The 0.56% uncertainty is the more precise value based on the detailed calculation of Enclosure (1) to Reference III and has been rounded up to the nearest 0.01%. The 0.02% value is also caused by round-off error.

Each value in Table I-1 was rounded off to the nearest 0.01%.

In Enclosure (1) to Reference III, uncertainty was not rounded to the nearest 0.0 1% until the final step.

As described in Enclosure (1) to Reference III, Unit 2 calorimetric uncertainty, expressed in megawatts, is 15.1 MW. Thus, the new licensed power limit must be equal to or less than (1.02)(2700 MWth) -

15.1 MWth = 2738.9 MWth. However, the requested uprate was limited to 2737 MWth, resulting in an additional margin.

Note that the response to Criterion 1 does not present a value of 0.62% RTP, however, this value can be derived from the maximum thermal power limit of 2754 MWth and the proposed uprate of 2737 MWth.

Again the difference between the 0.62% derived value and the 0.56% calculated uncertainty represents an additional, conservative margin of 0.06%.

NRC Ouestion

7.

Response to Criterion I and 3 refers to a plant-specific feedwater mass flow measurement uncertainty calculation performed by Westinghouse for each CCNPP unit (References 6.3.1 and 6.3.2). Please submit a calculation for NRC staff review.

CCNPP Response A copy of the referenced Westinghouse proprietary calculation is attached (Attachment 5). Note that per the response to Criterion 1, the uncertainty values presented in the calculation will be confirmed-after final system set up and prior to an increase in plant power.

27

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE NRC Ouestion

8.

The table in Section 4.4.1.6 of Enclosure I (Calculation CA06494) lists the Crossflow ultrasonic flow meter (UFM) uncertainties for each of the four headers. These values are different from those listed in Table I-I of Attachment 2 for CCNPP I and 2. Please explain how the contents of Table I-1 were calculated and how the random and bias components were used to calculate total calorimetric uncertainty. Section 3 of Enclosure I states that this calculation uses methodology established in ES-028. Was ES-028, "Instrument Loop Uncertainty and Setpoint Methodology" (Reference 6.1) reviewed and approved by the NRC staff?

CCNPP Response The table in Section 4.4.1.6 of Enclosure (1) to Reference III, lists the CROSSFLONV'm uncertainties as a percentage of measured flow. Table 1-1 presents the contribution of CROSSFLOW\\m to calorimetric uncertainty as a percentage of RTP.

The values in Table I-1 were calculated using the val'ues provided in Enclosure (1) to Reference III. For simplicity, the values summarized in the table do not account for the dependency between feedwater flow uncertainty and temperature measurement uncertainty. The dependency between flow and temperature has been accounted for in Enclosure (1).

In Enclosure (1) to Reference III, the contribution of each instrument uncertainty to calorimetric uncertainty was determined by calculating calorimetric power using nominal values for each input to the calorimetric calculation. Then, each input was individually and separately varied by its uncertainty, and calorimetric power recalculated.

For each input, the difference between the calculated calorimetric powers represents the contribution of that input's uncertainty to calorimetric uncertainty.

Thus, a flow uncertainty of approximately 0.56% flow (average of Unit 2's UFM uncertainties) corresponds to a calorimetric uncertainty of approximately 0.39% (not accounting for the dependency between temperature measurement and flow).

Random uncertainties consist of independent and dependent terms and are combined using the square root sum of squares methodology. Dependent random terms are algebraically summed prior to performing the square root sum of squares methodology determination.

Bias terms are algebraically summed.

Total uncertainty is the sum of the combined random uncertainty and the bias uncertainty.

Engineering Standards ES-028, "Instrument Loop Uncertainty and Setpoint Methodology," is a site specific procedure used to develop uncertainty calculations. This procedure has not been reviewed by the NRC. Engineering Standard ES-028 is based upon Instrument Society of America (ISA) standard, S67.04 (1987) "Setpoints for Nuclear Safety-Related Instrumentation," and ISA recommended practice RP67.04, Part II (1994), "Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation."

Table I-1 should only be considered a simplified presentation of calorimetric uncertainty. Please refer to Enclosure (1) to Reference III for the complete calculation.

28

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE NRC Ouestion

9.

A flow profile "fully developed" term is used in the 1st, 3rd, and 4th paragraphs of response to Criterion 4, while a flow profile "stable" term is used in the 3rd paragraphs of response to Criterion 2 and 4. Crossflow Topical Report CENPD-397-P-A used a "fully developed" term for locating the UFM. Westinghouse has subsequently replaced this term with a "stable flow in an isothermal and subcooled" liquid condition.

Replacement of this new terminology and the Crossflow in-situ calibration is explained in the X-Beam Topical Report WCAP-16163-P, Revision I (Supplement 1-P to CENPD-397-P, Revision 1), which is not yet approved by the NRC staff.

Explain and justify using the new terminology and maintain consistency.

CCNPP Response The response to this question has been modified to include information requested in a teleconference with the NRC staff on July 13, 2005.

'PROPRIETARY TEXT' NRC Question

10.

It is proposed to install a new X-Beam UFM farther down-stream from the flow straightener to address the observed increase in the venturi correction factor on three of the four Crossflow meter locations at steady state power levels between 85 percent and 100 percent rated thermal power. Is this exercise to confirm the upstream installed, in-situ calibrated, loops 11, 21, and 22 Crossflow UFM readings by comparing it with a higher accuracy X-Beam UFM? Will the X-Beam UFMs be removed after the proposed testing on both units? Please note that X-Beam topical report is not yet approved by the NRC staff.

29

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE CCNPP Response X-Beam UFMs were temporarily installed on Unit I to determine if the observed power dependency could be eliminated at the new location. If power dependency was eliminated, the X-Beam transducer brackets would remain installed and used to correct venturi flow. However, the X-Beam technology would be applied as a standard CROSSFLONVm technology approved in accordance with CENPD-397-P-A.

Each set of transducers would be treated individually and no credit taken for the improved accuracy, which results by averaging the flows from both transducers sets.

One set of transducers would be used to correct venturi flow. The second set of transducers would essentially function as an installed spare. Consistent with the response to Criteria 1, the uncertainty values would be confirmed after final system set up and prior to an increase in plant power.

As noted above (Response 9) the X-Beam meters do show a power dependency in the venturi flow correction factor similar in nature to the dependency observed in the original CROSSFLONVWm meters.

Since the power dependency was not reduced, the use of X-Beam UFMs are not required to support the requested uprate and they will be removed now that the testing is complete.

NRC Ouestion

11.

Section 1.4 of Attachment 2 addresses various aspects of calibration and maintenance procedures only for Crossflow UFM. Section I-F of RIS 2002-03 guidance discusses addressing various aspects of calibration and maintenance of "all instruments that affect the power calorimetric."

Please submit complete information.

CCNPP Response The remaining instruments used for calorimetric power calculation are periodically calibrated to ensure the instrument is performing within the limits established by the instrument's uncertainty calculations.

Calibration and maintenance is performed by Calvert Cliffs maintenance personnel using site procedures.

All work is performed in accordance with site work control procedures.

Reliability of the calorimetric instrumentation is monitored by Calvert Cliffs Systems Engineering personnel. Equipment deficiencies are documented under the site's corrective action process, which includes requirements for the resolution of identified deficiencies.

Equipment manufacturers are contacted as required to correct the deficiency.

If a deficiency report is received from a vendor, that deficiency report would be considered Operating Experience and reviewed for applicability.

Those deficiencies applicable to Calvert Cliffs are documented under the site's corrective action process.

NRC Question

12.

Section 3.3.6 of Topical Report CENPD-397-P-A states that each rejected data point is highlighted and if the build up of rejected data point is higher than normal (plant-specific), the operator shuts down the Crossflow meter. Please identify CCNPP 1 and 2's normal number of rejected data points.

Additionally, Section 3.2.4.6 of the topical report indicates that the standard design recommends averaging 4 seconds of data in one time delay. The operator of a power plant has a choice to use multiple time delays for averaging data. Please identify if CCNPP Units 1 and 2 are using multiple time delays for Crossflow data averaging, and also provide reasons for this choice.

30

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THlE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE CCNPP Response The response to this question has been modified to include information requested in a teleconference with the NRC staff on July 13, 2005.

'PROPRIETARY TEXT' Additionally, other indications are available to the operator to alert of a potential change in venturi performance, such as changes in first stage pressure, feedwater temperature, primary calorimetric power and steam flow. Therefore, the operator has multiple pieces of information on which to react. Based on this information, the operator would not adjust reactor power, but would attempt to determine why venturi performance had changed.

NRC Ouestion

13.

Surveillance Requirement 3.3.1.2 in the CCNPP 1 and 2 technical specifications (TSs) requires adjusting excore power range and delta T power channels to agree with calorimetric calculation if the absolute difference is 2 1.5 percent. The TSs Bases do not explain the reason for choosing the 31

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWNER UPRATE 1.5 percent value. Was this value chosen to remain within the required Appendix K margin of 2.0 percent to allow for instrumentation error? Please explain and justify retaining the same value of the absolute difference with the proposed 1.38 percent MUR power uprate.

CCNPP Response The response to this question has been modified to include information requested in a teleconference with the NRC staff on July 13, 2005.

The 1.5% absolute difference between the secondary calometric calculation and the excore power range and delta T power channels ensures that the excore power range and delta T power channels do not excessively deviate from the secondary calometric calculation. The excore power range and delta T power channels support the Power Level - High RPS trip function. This trip function is credited in the Chapter 14 safety analysis as described in the UFSAR.

The Power Level - High function has an allowable value of 107% RTP (Technical Specification Table 3.3.1-1). The analytical limit assumed in the analyses for the affected events is 110.3% thermal power. These limits are much higher than the 1.5%

difference between the secondary calometric calculation and the excore power range and delta T power channels at the MUR power uprate power level. Therefore, the MUR power uprate does not impact the 1.5% absolute difference between the secondary calometric calculation and the excore power range and delta T power channels in Surveillance Requirement 3.3.1.2.

NRC Ouestion

14.

The last sentence in the first paragraph in Section 1.5 of Attachment 2 states that the random component of uncertainty could be substantially reduced permitting an even greater (-0.24 percent) variation in correction factor over the 72-hour interval. Please explain this statement.

CCNPP Response This statement should be disregarded and is withdrawn. The statement pertains to draft calculations of calorimetric uncertainty which did not credit a reduction in uncertainty for single side of interest as permitted by ISA recommended practice RP67.04. Since the approved uncertainty calculation makes use of the single side of interest reduction, no credit can be taken for any additional reduction in uncertainty.

The basis for the use of a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> interval is that correction factor trends over that interval have shown very little variation as stated in the submittal.

NRC Question

15.

How many effective full power years (EFPYs) of operation is the projected end of license neutron fluence based on for both CCNPP units?

CCNPP Response The projected end of license neutron fluence is based on 48 EFPY for both Units 1 and 2.

NRC Ouestion

16.

What is the core thermal power level for the projected end of license neutron fluence for both CCNPP units?

32

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE CCNPP Response The core thermal power level for the projected end of license neutron fluence for both Units 1 and 2 is 2700 MWth. The projected thermal power for calculating neutron fluence, in accordance with Regulatory Guide 1.99, has not been updated. A comprehensive reactor surveillance program for Calvert Cliffs continues to monitor the limits of the vessel analysis assumptions.

NRC Ouestion

17.

For the evaluation of the upper shelf energy (USE), the beltline of the CCNPP 1 and 2 reactor vessels is limited by the USE drop that is projected to occur in what material heat number in the CCNPP I and 2 reactor vessels? What is the projected end of life USE value for these materials based on for the uprated number of EFPYs and 1/4 T neutron fluence. Please indicate what the 1/4 T neutron fluence is for both units.

CCNPP Response

-Calvert Cliffs Unit 1 and Unit 2 Reactor Vessel Beltline USE.and Fluence (Plate M aterial)

Material w/ Limiting EOL USE 1/4T Fluence USE Drop

(@48 EFPY)

((ai)48 EFP)

Unit 1 C4420-1/D-7207-1 55 3.0 x10' Unit 2 A4463-2/D-8906-3 52 3.4 x10'9 Calvert Cliffs Unit -and Unit 2 Reactor Vessel Beltline USE and Fluence (W eld M aterial)

Weld Code iv/

EOL USE 1/4T Fluence Limiting USE Drop

((@48 EFPY)

(P,48 EFTY)

Unit 1 2-203-A,B,C 59 3.0 x10 Unit 2 2-203-A,B,C 50 3.4 x10' 9 The projected end-of-life (EOL) values have not been updated. A comprehensive reactor surveillance program for Calvert Cliffs continues to monitor the limits of the vessel analysis assumptions, and end-of-life values will be re-analyzed at the time of the next test capsule withdrawal.

NRC Ouestion

18.

The licensee stated in its application that the blowdown system will continue to operate normally with no change at a continuous rate of up to 200 gallons per minute per SG.

A.

Since the MUR power uprate will require the blowdown system to operate beyond its current design time, explain whether the additional operating time will make any components in the system more susceptible to flow accelerated corrosion (FAC).

B.

Does the licensee's current evaluation of the blowdown system under uprate conditions consider the effect of a potential increase of impurities in the SG water?

C.

Provide a technical justification as to why the blowdown system is to operate normally without any changes during uprate conditions.

33

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE CCNPP Response 18.A. The steam generator blowdown system currently operates continuously. The system is only taken out-of-service to perform maintenance or repairs. The MUR power uprate does not change the operating time or design time for the steam generator blowdown system. The steam generator blowdown system will continue to operate continuously under the MUR power uprate conditions.

Since the system is controlled manually, the flow rate is controlled. Therefore, the FAC of the steam generator blowdown lines should continue with a similar trend as currently experienced.

18.B. The MUR power uprate may increase the level of impurities in the steam generator and condensate systems. The steam generator blowdown system helps remove some of the impurities from the steam generator water by removing an amount of water (approximately 1%) from the steam generator regions. This water is then passed through ion exchangers or discharged. With the increase in impurity levels, the resin from the ion exchangers may need to be replaced more often.

This does not result in an operational change or a change to the system performance. Any increase in blowdown flow required to remove potentially higher levels of impurities is within the blowdown system flow capacity and throttle valve adjustment capacity.

18.C. The steam generator blowdown system flow is manually controlled using throttle valves. These throttle valves are set to the desired flow rate, based on system chemistry, and engineering and operational needs. The MUR power uprate will not change the operation of this system. The flow rate will continue to be set as needed by manual manipulation of the throttle valves.

NRC Question

19.

Since the inlet pressure to the SG blowdown system varies proportionally with operating steam pressure, the blowdown flow control valves must be designed to handle a corresponding range of inlet pressures. Confirm that any change to the inlet pressure of the SG blowdown system is still inside the revised range of operating parameters for the uprate.

CCNPP Response The effect of the MUR power uprate on secondary side plant systems was analyzed and determined to create an expected 1.5 psi drop in steam pressure and a 1.7% increase in steam flow in the steam generator. This small change will also drop the steam generator blowdown inlet pressure by 1.5 psi. The result of this pressure drop will be adjusted for by the throttle control valves. These globe valves are designed to drop the high pressure of the steam generator blowdown inlet flow to a much lower pressure prior to entering the steam generator blowdown tank. These design parameters more than meet the change in system conditions due to the MUR power uprate, since the inlet pressure is lower and the flow rate is unchanged.

NRC Ouestion

20.

The licensee stated that the evaluation resulted in long-term FAC program impacts and no short-term impacts. Describe what the long-term FAC program impacts identified in your evaluation are and explain how these will be addressed under your corrective action program implementation.

CCNPP Response The system most susceptible to long term impacts is the feedwater system. This system is experiencing the highest wear rates with the lowest margins between nominal wall thickness and minimum wall 34

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE thickness. Since there is a projected change in flow velocity due to the MUR power uprate, the FAd computer model (Chec-Works) will be updated to account for the increased flow rate. Based on the revised model results, areas identified as high wear locations are scheduled for as-is inspections and follow-up inspections to evaluate the actual wear rate once the MUR power uprate has been implemented.

NRC Question

21.

The licensee stated that the MUR power uprate will result in a slight increase in inspection scope for some specific systems and possibly some additional replacement scope prior to the end-of-plant life expectancy.

A.

Describe the criteria used in the FAC program for selecting components for inspection after the MUR power uprate.

CCNPP Response 21.A. The Chec-Works model results are used for selection of inspection locations. Components are sorted by wear rate and time until reaching code minimum wall thickness. Based on these results, the highest wear-ranked components and components with the shortest time to reach minimum wall thickness will be selected for inspection.

Once components are determine to be experiencing FAC wear, components are trended and evaluated for time until reaching code minimum, replacements are then scheduled at an outage prior to reaching this minimum value.

NRC Question

22.

What are the significant changes in FAC rates anticipated as a result of the MUR power uprate conditions.

CCNPP Response We believe that the FAC wear rates resulting from the MUR power uprate will be very slight and probably not measurable using existing measurement methods.

NRC Question

23.

Identify the systems that are expected to experience the greatest increase in wear as a result of the uprate. In addition, for the five components most susceptible to FAC, provide numerical data that show changes in (1) velocity, (2) temperature, and (3) predictive wear rate result from the uprate.

For the same components, provide numerical data comparing the measured FAC rate with the as found condition.

CCNPP Response The feedwater system is expected to experience the greatest increase in wear rates due to the MUR power uprate. Below are listed the five most susceptible components to FAC wear. It should be noted that the increases in velocity, temperature and wear are calculated for a Chec-Works model upgrade, not the MUR power uprate. The Chec-Works model upgrade accounts for changes not related to the MUR power uprate (such as guarantee vs. maximum power output, etc.). As noted above, the wear rate due the MUR 35

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE power uprate is expected to be very slight and probably not measurable using existing measurement methods.

Unit 1 16" feedwater system Velocity Temperature Pred. Increase Increase Rise Wear DB-01-1010-16FE 3.384 ft/s 4.5 0F 0.383 mils/yr DB-01-1009-18FE 3.384 fl/s 4.5 0F 0.383 mils/yr DB-01-1018-04E 2.017 fl/s 4.5 0F 0.176 mils/yr DB-01-1019-02E 2.020 ft/s 4.50F 0.217 mils/yr DB-01-1019-04P 2.020 ft/s 4.50F 0.118 mils/yr Below is the table that compares the current measured wear with the predicted measured wear for the same components as above. Again note that these predicted numbers are calculated for a Chec-Works model upgrade, not the MUR power uprate. In addition, the majority of the difference between the actual wear and the predicted wear is due to the absence of chrome alloy data from the Check-Works model.

Inclusion of the chrome alloy data, will greatly improve the accuracy of the model. Future Chec-Works models will include both the chrome alloy data and the MUR power uprate flow changes Point to Point Chec-Works Model Component Current Total Ce-ok oe Meturredt Wtar Current Total Predicted Wear Measured Wear DB-01-1010-16FE*

20 mils 34 mils DB-01-1009-18FE*

12 mils 57 mils DB-01-1018-04E 288 mils 218 mils DB-01-1019-02E 359 mils 233 mils DB-01-1019-04P 184 mils 150 mils

• Components replaced in 1992 NRC Question

24.

Confirm that the coating qualification temperature and pressure profile used to qualify the original maintenance Service Level I coatings continues to bound the design basis accident temperature and pressure profile under power uprate conditions.

CCNPP Resnonse The Service Level 1 coatings were originally qualified to temperature-pressure curves that bound CCNPP's temperature-pressure curves for the DBA. The coatings are therefore still qualified and their testing still bounds the containment design of 50 psig at 200 seconds and 276°F (surface temperature) at 250 seconds. See the below list of qualified coatings within Containment, and their associated test parameters.

36

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE Coating Qualification Test and Critical Parameters Ameron Dimetcote D6N with Ameron Protective Coatings Division DBA Test Report No.

topcoat of Ameron Amercoat 66 1781-Al, 1842, 2113, 2032, 2178, 2534, 1862, Southwest epoxy Research Inst. Dated January 1978. The curve utilized consisted of a 53 psig maximum pressure and a 300'F maximum temperature starting at 20 seconds and continuing to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Note that Unit 1 was identified in the past with thicker than allowed coatings; this thicker coating was tested and requalified for this as-installed configuration, Technical Report No. 3428/2004.

Mobilzinc 7 inorganic zinc primer Mobil Chemical Company DBA Test Report dated December, with a Mobil 89 Series epoxy 1976. The curve utilized consisted of a 70 psig maximum pressure topcoat (Now sold as Valspar 13-and a 3401F maximum temperature starting at 20 seconds and F-12 and Valspar 89 Series, continuing to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

respectively)

Ameron I IOAA sealer with a Ameron Protective Coatings Division DBA Test Report LSR Ameron Amercoat 66 epoxy 2052, 1862 dated June 1976. The curve utilized consisted of a topcoat 70 psig maximum pressure starting at 10 second and continuing to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the temperature ramps up from 290'F starting at 20 seconds, to 3001F at 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Ameron Amercoat 71 epoxy primer Ameron Protective Coating Division DBA Test Report No.

and Ameron Amercoat 66 epoxy LSR 2178, 1840, 2204, 2197, 1862, 1994, dated November, 1983.

topcoat The curve utilized consisted of a 70 psig maximum pressure starting at 10 seconds and a 3401F maximum temperature starting at 20 seconds and both continuing for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Ameron Amercoat 90 epoxy, two Ameron Protective Coating Division DBA Test Report No. LSR coat system 2206, 2207, 2052, 1538, 2178, 1862, 2408 A-1, 2452 dated January 1993. The curve utilized consisted of a 70 psig maximum pressure starting at 10 seconds and a 340'F maximum temperature starting at 20 seconds and both continuing for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Carboline Carboguard 890N epoxy Carboline DBA Test Report No. 02927.

The curve utilized consisted of a maximum pressure of 50 psig from 10 seconds to 20 minutes, with a corresponding maximum temperature of 320'F at 10-180 seconds, and ramping down to 270'F from 180-300 seconds.

Carboline Starglaze 2011 Ssurfacer Carboline DBA Test Report No. 02658.

The curve utilized and Carboline Carboguard 890N consists of a maximum pressure of 60 psig and temperature of 307'F from 42 seconds, and continuing for I hour and 8 minutes.

NRC Ouestion

25.

Under the power uprate conditions, the differential pressure across the tubes, the temperature of the primary and secondary water, and the flow through the SG will change. Confirm that the SG will still satisfy all original design criteria (e.g., American Society of Mechanical Engineers Boiler and Vessel Pressure Code) under these conditions. In addition, confirm that your analysis addresses the current condition of the CCNPP 1&2 SGs (e.g., the plugs, any repairs the as-built configuration, loose parts, etc.) and addressed flow-induced vibration.

37

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON TIHE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE CCNPP Response The steam generators at CCNPP, replaced in Refueling Outage 2002 (Unit 1) and Refueling Outage 2003 (Unit 2), are referred to as the RSGs. Babcock & Wilcox Canada, the RSG manufacturer, was contacted to review the RSG with the MUR power uprate condition (Note, they conservatively used 1.7%) and verify its acceptability.

Their response included a thermo-hydraulic and structural evaluation of a 101.7% power uprate.

The uprated 101.7% full power conditions were analyzed using the steady-state thermal-hydraulic code CIRC and compared with the 100% power cases in the existing reports (see Table IV-1 in Reference III).

Both beginning-and end-of-life conditions were analyzed. The steam/water inventory was calculated for start-up conditions.

The pressure across the tubes of the RSG between the primary and the secondary side increased by approximately 0.313% at the uprated conditions. This new differential pressure is bounded by the original analysis per the RSG design specification; which utilized pressures of 2500 psia maximum for the primary side (unchanged for MUR power uprate) and 850 psia minimum for the secondary side (860.3 psia at EOL per MUR power uprate estimate). The temperature of the RCS, Th.t, is increased approximately 0.80F to a maximum expected value of 595.91F. This is bounded by the original RSG specification design temperature of 6501F and a maximum actual temperature of 600'F.

The steam/water inventory was calculated for the start-up conditions. The total inventory of steam/water decreased 0.36%. The total steam generator weight decreased 0.04% due to the inventory change. An analysis of all pressure boundary components for level A, B, C, D and test condition loads was carried out. Due to the negligible changes in temperature and pressure of the primary side and secondary side, for the power uprate scenario, the following was found: stress range and usage factor calculation would stay the same, external loads including LOCA remain unchanged, and the MSLB load would actually decrease due to lower secondary side temperature.

Babcock & Wilcox Canada also performed a flow induced vibration (FIV) analysis for the CCNPP RSGs at 1.7% uprated power, end-of-life conditions. The results of this analysis confirm that the 1.7% power uprate will have an insignificant impact on the FIV response of the critical U-Bend and bundle entrance tubes.

This result is true for all three FIV mechanisms, including fluidelastic instability, random turbulence excitation and vortex shedding, as well as the tube support fretting rate. Furthermore, due to the minimal increase in the bundle entrance and riser flow velocity as a result of the 1.7% power uprate, the existing FIV calculation for other internal components will not be affected by operation at the MUR power uprate conditions. Therefore, the results of the FIV analysis are valid for the 1.7% power uprate.

The current conditions of the RSGs were reviewed; very few tubes are currently plugged, due to the young age of the RSGs. The current plug and tube stabilizer design parameters bound the MUR power uprate conditions. Therefore, the plugs and stabilizers remain an effective means of maintaining the pressure boundary. Loose parts in the RSG were reviewed and the MUR power uprate was determined to make no difference due to location of parts, and that the tubes in the area were preemptively plugged and stabilized to ensure RSG pressure boundary for the unit.

NRC Ouestion

26.

Confirm that CCNPP I and 2's plugging limit is still appropriate for power uprate conditions, given the guidance in Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes."

38

ATTACHMENT (1)

NON-PROPRIETARY RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ON THE MEASUREMENT UNCERTAINTY RECAPTURE POWER UPRATE CCNPP Response The tube plugging margin required by Unit I and 2 Technical Specifications is 10%. The Regulatory Guide 1.121 analysis was reviewed and found to still be appropriate for the MUR power uprate conditions, with no changes. The analysis utilized a combination of 1400 psi differential for normal operating conditions, which bounds the MUR power uprate conditions of 1390 psi differential maximum.

REFERENCES I.

Letter from D. G. McDonald (NRC) to G. C. Creel (BGE), dated February 12, 1991, Response to Station Blackout Rule - Calvert Cliffs Nuclear Power Plant, Units I and 2 II.

Letter from D. G. McDonald (NRC) to R. E. Denton (BGE), dated September 22, 1993, Supplemental Safety Evaluation of Response to the Station Blackout Rule, 10 CFR 50.63 - Calvert Cliffs Nuclear Power Plant Unit I and Unit 2 III. Letter from G. Vanderheyden (CCNPP) to Document Control Desk (NRC), dated January 31, 2005, License Amendment Request: Appendix K Measurement Uncertainty Recapture - Power Uprate Request 39

ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION CORRECTED PAGES Calvcrt Cliffs Nuclear Power Plant, Inc.

July 18, 2005

ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION IV.5.1 Fuel Core Design The impact of a bounding 1.7% uprate condition on the fuel core design was evaluated against the data used in the current Calvert Cliffs safety AOR. Since the MUR power uprate is relatively small, the range of parameters used in the current safety AOR are adequate to accommodate the range of parameters expected for future cores that have implemented the MUR power uprate. The core analyses for specific uprate cycles has shown that the implementation of the MUR power uprate does not result in significant changes to the current nuclear design basis for the safety analysis documented in the UFSAR. The impact of the MUR power uprate on peaking factors, rod worths, reactivity coefficients, shutdown margin, and kinetics parameters is either well within normal cycle-to-cycle variation of these values or controlled by the core design and will be addressed on a cycle-specific basis consistent with reload methodology.

The methods and core models used in the MUR power uprate analyses are consistent with those presented in the Calvert Cliffs UFSAR. No changes to the nuclear design philosophy, methods, or models are necessary due to the MUR power uprate. The current range of required cycle specific analysis will be sufficient to verify the applicability of these parameters for future cycles.

IV.5.2 Core Thermal-Hydraulic Design The core thermal-hydraulic design and methodology were evaluated for the MUR power uprate.

The thermal hydraulic design is based on the TORC computer code described in Reference IV-2, the ABB-TV [Asea Brown Boveri, Inc.-Turbo vane] and ABB-NV [non-mixing vane] critical heat flux correlations described in Reference IV-3, the simplified TORC modeling methods described in Reference IV-4, and the CETOP-D code described in Reference IV-5. In addition, the departure from nucleate boiling ratio (DNBR) analysis uses the methodology for determining the limiting fuel assembly or assemblies.

The Extended Statistical Combination of Uncertainties presented in Reference IV-6 and approved in Reference IV-7 was applied to validate the design limit of 1.24 on the ABB-TV and ABB-NV minimum DNBR. This DNBR limit includes the following allowances:

1.

NRC specified allowances for TORC code uncertainty.

2.

Rod bow penalty equivalent to 0.6% on minimum DNBR as discussed in Reference IV-8.

The core thermal-hydraulic design and methodology remain applicable for Calvert Cliffs with the MUR power uprate.

IV.5.3 Fuel Rod Design The thermal performance of erbia and U02 fuel rods for Calvert Cliffs Unit 1 Cycle 17 core with the MUR power uprate were evaluated using the FATES3B version of the fuel EM (References IV-9, IV-1 0, and IV-1 1), the erbia burnable absorber methodology described in Reference IV-12, the maximum pressure methodology described in Reference IV-13, and the ZIRLO' fuel rod cladding methodology described in Reference IV-14. This evaluation included a power history that enveloped the power and burnup levels expected for the peak fuel rod at each burnup interval, from beginning-of-cycle to end-of-cycle burnups, including a reduction in maximum permitted Fr, consistent with implementation of the power uprate. The maximum predicted fuel rod internal pressure for the uprated core remains below the critical pressure for No-Clad-Lift-Off (Reference IV-13).

ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION The thermal performance of erbia and U02 fuel rods for Calvert Cliffs Unit 1 (subsequent to Cycle 17) and Unit 2 (subsequent to Cycle 15) cores with the MUR power uprate is evaluated using the FATES3B version of the fuel EM (References IV-9, IV-10, and IV-1 1), the erbia burnable absorber methodology described in Reference IV-12, the maximum pressure methodology described in Reference IV-13, and the ZIRLOTM fuel rod cladding methodology described in Reference IV-14. These evaluations include a power history that envelopes the power and burnup levels expected for the peak fuel rod at each burnup interval, from beginning-of-cycle to end-of-cycle burnups. The maximum predicted fuel rod internal pressure for the uprated conditions will be shown to remain below the critical pressure for No-Clad-Lift-Off (Reference IV-13).

The expected fuel rod corrosion performance for Calvert Cliffs Units 1 and 2 cores with MUR power uprate was evaluated and found acceptable. This evaluation was conducted consistent with requirements of the NRC SER on the high burnup topical report for 14x14 CE design fuel of Reference IV-15. This evaluation also considered the impact of recent high duty corrosion observations for OPTINTM clad fuel that may be resident in Calvert Cliffs Units 1 and 2 (see also Reference IV-16). The fuel rod corrosion performance of OPTINTm and ZIRLOTM clad fuel specifically for the MUR power uprated Calvert Cliffs Unit 1 Cycle 17 core were evaluated (References IV-14, IV-15, and IV-16) and found to be acceptable. The fuel rod corrosion performance for MUR power uprated Calvert Cliffs Unit 1 (subsequent to Cycle 17) and Unit 2 (subsequent to Cycle 15) cores is evaluated using this same methodology.

As noted in previous sections (e.g., 11.3) Calvert Cliffs Unit 1 Cycle 17 was targeted as the lead unit for implementation of the MUR power uprate. Application of the MUR power uprate analyses and evaluations to Unit 2 would be made as part of the normal reload process.

Subsequent to performance of these Unit 1 analyses and evaluations, a fuel design change that implements the use of Zirconium Diboride (ZrB2) integral burnable absorbers was submitted to the NRC (Reference IV-1 7). Unit 2 Cycle 16 is the lead unit for the fuel design change. The analyses that are being performed to support the transition to ZrB 2 have already included the MUR power uprate as an initial condition. No additional analyses to support both ZrB2 and the MUR power uprate are required.

IV.6 BALANCE OF PLANT PIPING The balance of plant (BOP) piping systems impacted by the uprate (main steam, feedwater, extraction steam, moisture separator drains, reheater drains, condensate, and heater drain piping) have been evaluated by comparing the conditions for the proposed power uprate with the current operating conditions. The design temperatures and pressures used in the analyses continue to bound the uprate conditions. The maximum operating temperatures with the MUR power uprate are within 1% of the existing maximum operating temperatures.

The BOP piping systems remain acceptable for operation at the MUR power uprate conditions, and the proposed 1.38% power uprate will not have adverse effects on the BOP piping.

ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION 11-11 CENPD-132P, "Calculative Methods for the C-E Large Break LOCA Evaluation Model,"

August 1974 11-12 Letter from A. W. Dromerick (NRC) to C. H. Cruse (BGE), dated October 2, 1997, "Issuance of Amendments for Calvert Cliffs Nuclear Power Plant Unit 1 (TAC No.

M95181) and Unit No. 2 (TAC No. M95182)"

11-13 Letter from S. A. McNeil (NRC) to J. A. Tiernan (BGE), dated November 2, 1988, "Safety Evaluation Concerning Conformance to the ATWS Rule (TACs 59079 and 59080)"

11-14 Letter from C. H. Cruse (BGE) to Document Control Desk (NRC), dated July 31, 1997,

'Response to Request for Additional Information Regarding the Technical Specification Change to the Moderator Temperature Coefficient (TAC Nos. M95181 and M95182)"

Section III None.

Section IV IV-1 CEN-387-P, Revision 1-P-A, "Pressurizer Surge Line Flow Stratification Evaluation,"

May 1994 IV-2 CENPD-161-P-A, 'TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," April 1986 IV-3 CENPD-387-P-A, "ABB Critical Heat Flux Correlations for PWR Fuel," May 2000 IV-4 CENPD-206-P-A, 'TORC Code:

Verification and Simplified Modeling Methods,"

June 1981 IV-5 CEN-191(B)-P, "CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2," December 1981 IV-6 CEN-348(B)-P-A, Supplement 1-P-A, "Extended Statistical Combination of Uncertainties," January 1997 IV-7 Letter from G. M. Holahan (NRC) to S. A. Toelle (ABB), dated August 31, 1994, "Generic Approval of CEN-348(B)-P-A,

'Extended Statistical Combination of Uncertainties' (TAC No. M90019)"

IV-8 CENPD-225-P-A, "Fuel and Poison Rod Bowing," June 1983 IV-9 CEN-161(B)-P Supplement 1-P-A, "Improvements to Fuel Evaluation Model,"

January 1992 IV-10 CENPD-139-P-A, "Fuel Evaluation Model," July 1974 IV-11 CEN-161(B)-P-A, 'Improvements to Fuel Evaluation Model," August 1989 IV-12 CENPD-382-P-A, "Methodology for Core Designs Containing Erbium Burnable Absorbers," August 1993 IV-13 CEN-372-P-A, "Fuel Rod Maximum Allowable Gas Pressure," May 1990 IV-14 CENPD-404-P-A, Revision 0, "Implementation of ZIRLOTm Cladding Material in CE Nuclear Power Fuel Assembly Designs," November 2001 IV-15 CEN-382(B)-P-A, 'Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 14x14 PWR Fuel," August 1993 I

ATTACHMENT (2)

SUMMARY

OF CALVERT CLIFFS NUCLEAR POWER PLANT MEASUREMENT UNCERTAINTY RECAPTURE EVALUATION IV-16 CENPD-384-P, 'Report on the Continued Applicability of 60 MWD/kgU for ABB Combustion Engineering PWR Fuel," September 1995 IV-17 Letter from B. S. Montgomery (CCNPP) to document Control Desk (NRC), dated July 15, 2004, 'License Amendment Request: Incorporate Methodology References for the Implementation of PHOENIX-P, ANC, PARAGON, and Zirconium Diboride into the Technical Specifications"Section V None.

Section VI None.

Section Vil None.

Section VilI VI11-1 CENPD-199-P, Revision 1-P-A, 'C-E Local Power Density and DNB LSSS and LCO Setpoint Methodology for Analog Protection Systems," January 1986 V11-2 CEN-124(B)-P, "Statistical Combination of Uncertainties, Part 1," December 1979 V11-3 CEN-348(B)-P-A Supplement 1-P-A, "Extended Statistical Combination of Uncertainties," January 1997 V1I1-4 CEN-124(B)-P, "Statistical Combination of Uncertainties Methodology Part 3,"

December 1979 ATTACHMENT (3)

WESTINGHOUSE PROPRIETARY AFFIDAVIT Calvert Cliffs Nuclear Power Plant, Inc.

July 18,2005

e) Westinghouse U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 Westinghouse Electric Company Nudear Services P.O. Box355 Pittsburgh, Pennsylvania 15230-0355 USA Directtel: 412-374-4643 Direct fax: 412-374-4211 e-mail: greshajaewestinghouse.com Our ref: CAW-05-2025 July 14, 2005 APPLICATION FOR WITHHOLDING PROPRIETARY INTORMATION FROM PUBLIC DISCLOSURE

Reference:

CCNPI Letter, "Request for Additional Information Re: 1.38 Percent Measurement Uncertainty Recapture Power Uprate License Amendment Request (TAC Nos. MC6210 and MC621 1)", dated July 18, 2005 The proprietary information for which withholding is being requested is contained in the above referenced Calvert Cliffs Nuclear Power Plant, Inc. (CCNPPI) letter in Attachment (4),

"Proprietary Response to Request for Additional Information on the Measurement Uncertainty Recapture Power Uprate" and in proprietary Attachment (5), Calculation CN-PS-02-8, Rev. 0, "Feedwater Flow Measurement Using the CROSSFLOW Ultrasonic Flowmeter at Constellation Nuclear Calvert Cliffs Unit 2", as identified in Affidavit CAW-05-2025 signed by the owner of the proprietary information, Westinghouse Electric Company LLC (Westinghouse). The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations. Westinghouse also affirms that a non-proprietary version of Attachment (5) does not exist and, due to the extent of proprietary information contained, a non-proprietary version of the calculation would be meaningless.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Calvert Cliffs Nuclear Power Plant, Inc.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-05-2025, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC, P.O. Box 355, Pittsburgh, Pennsylvania 15230-0355.

Manager Regulatory Compliance and Plant Licensing

Enclosure:

As stated A BNFL Group Company

CAW-05-2025 Page I CAW-05-2025 Page 1 AFFIDAVIT STATE OF CONNECTICUT

)

) ss: WINDSOR, CT COUNTY OF HARTFORD

)

Before me, the undersigned authority, personally appeared Ian C. Rickard, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC ("Westinghouse"), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

z72)

Ian R

rpd, ingProject Manager Systems and Safety Analysis, Nuclear Services Westinghouse Electric Company, LLC Sworn to and subscribed before me this 14'h day of July 2005.

I

CAW-05-2025 Page 2 (1)

I, Ian C. Rickard, depose and say that I am the Licensing Project Manager, Systems and Safety Analysis, in Nuclear Services, Westinghouse Electric Company LLC ("Westinghouse"), and as such I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Electric Company LLC.

(2)

I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.

(3)

I have personal knowledge of the criteria and procedures utilized by the Westinghouse Electric Company LLC in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4)

Pursuant to the provisions of paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i)

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence.

The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a)

The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b)

It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c)

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

CAW-05-2025 Page 3 CAW-OS -2025 Page 3 (f)

It contains patentable ideas, for which patent protection may be desirable.

(iii) There are sound policy reasons behind the Westinghouse system for classification of proprietary information, which include the following:

(a)

The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b)

It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c)

Use of this information by our competitors would put Westinghouse at a competitive disadvantage by reducing their expenditure of resources at our expense.

(d)

Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e)

Unrestricted disclosure of this information would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f)

The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390; it is to be received in confidence by the Commission.

(v) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(vi) The proprietary information sought to be withheld by this submittal is that which is contained in proprietary Attachment (4), "Proprietary Response to Request for Additional Information on the Measurement Uncertainty Recapture Power Uprate" and in proprietary Attachment (5), Calculation CN-PS-02-8, Rev. 0, "Feedwater Flow Measurement Using the CROSSFLOW Ultrasonic Flowmeter at Constellation Nuclear Calvert Cliffs Unit 2",

transmitted by Calvert Cliffs Nuclear Power Plant, Inc. letter, "Request for Additional Information Re: 1.38 Percent Measurement Uncertainty Recapture Power Uprate License Amendment Request (TAC Nos. MC6210 and MC621 1)", dated July 18, 2005, for submittal to the Commission, being transmitted herewith and Application for Withholding Proprietary Information from Public Disclosure, to the NRC Document Control Desk. The proprietary information as submitted for use by Westinghouse is expected to be applicable in other licensee submittals in response to certain NRC requirements for justification of the application of the CROSSFLOW Ultrasonic Flow Measurement System performance within its approved accuracy limit.

CAWV-05-2025 Page 4 CAW-05-2025 Page 4 This information is part of that which will enable Westinghouse to:

(a) Validate CROSSFLOW Ultrasonic Flow Measurement System performance.

(b) Support licensees in implementing the CROSSFLOW Ultrasonic Flow Measurement System within its approved accuracy limit.

Further this information has substantial commercial value as follows:

(a)

Westinghouse plans to sell the use of similar information to its customers for purposes of meeting NRC requirements for licensing documentation.

(b) Westinghouse can sell support and defense of the CROSSFLOW Ultrasonic Flow Measurement System performance within its approved accuracy limit.

(c) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar advanced nuclear power plant designs and to provide licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

CAW-05-2025 Page 5 PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval. In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection wvith generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.