LIC-05-0023, License Amendment Request (LAR) Measurement Uncertainty Recapture Power Uprate

License Amendment Request (LAR) Measurement Uncertainty Recapture Power Uprate

ML050940389
Person / Time
Site:	Fort Calhoun
Issue date:	03/31/2005
From:	Ridenoure R Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-05-0023
Download: ML050940389 (61)

Text

Omaha Public ower District 444 South 16th Street Mall Omnaha NE 68102-2247 LIC-05-0023 March 31, 2005 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

REFERENCE:

Docket No. 50-285

SUBJECT:

Fort Calhoun Station Unit No. 1 License Amendment Request (LAR)

Measurement Uncertainty Recapture Power Uprate Pursuant to 10 CFR 50.90, Omaha Public Power District (OPPD) hereby requests the following amendment to the FCS Operating License and Technical Specifications:

Revise Paragraph 3.A. in Renewed Operating License DPR-40 to authorize operation at a steady state reactor core power level not in excess of 1522 megawatts thermal (MWt).

Revise the definition of RATED POWER in the Technical Specifications Definitions to reflect the increase from 1500 MWt to 1522 MWt.

Corresponding Technical Specifications Bases changes are also requested:

In the Basis of TS 2.1.6, Pressurizer and Main Steam Safety Valves, page 22, change all instances of "1500 MWt" to "RATED POWER."

1 In the Basis of TS 3.5, Containment Tests, page 3, replace "a reactor power level of 1500 MWt," with "RATED POWER."

The information provided in support of this request is based on NRC Regulatory Issue Summary (RIS) 2002-03, 'Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," January 31, 2002.

Appendix A of this letter is a listing of document references related to the previous LAR from OPPD for a Measurement Uncertainty Recapture (MUR) power uprate. On July 18, 2003, OPPD transmitted the original MUR power uprate LAR (Reference 1), which proposed a 1.67% increase in licensed power level from 1500 megawatts thermal Emlymn it qalOpotnty47 Employment with Equal Opportunity 4171

U. S. Nuclear Regulatory Commission LIC-05-0023 Page 2 (MWt) to 1525 MWt. This request was based on installation and use of the CROSSFLOW Ultrasonic Flow Measurement System supplied by Westinghouse/AMAG. In Reference 2, OPPD revised the MUR power uprate LAR to propose a 1.6% increase in licensed power level from 1500 MWT to 1524 MWt.

References 3 through 6 are the Nuclear Regulatory Commission (NRC) requests for additional information and OPPD's responses to those requests.

On January 16, 2004, the NRC issued Amendment No. 224 (Reference 7) to the FCS Renewed Facility Operating License to increase the licensed rated power by 1.6 percent from 1500 MWtto 1524 MWt. Subsequently, OPPD requested (Reference 8) and the NRC approved (Reference 9) an additional 120-day implementation period. As discussed in Reference 10, OPPD was unable to resolve problems encountered while testing the CROSSFLOW system associated with the MUR power uprate. As a result of these problems, OPPD submitted an exigent LAR (Reference 11) to return the FCS authorized power level to 1500 MWt, prior to implementation of Amendment 224. On May 14, 2004, the NRC issued Amendment 227 (Reference 12) restoring licensed thermal power to the previous value of 1500 MWt from 1524 MWt.

OPPD and Westinghouse/AMAG have resolved the problems encountered while testing the CROSSFLOW system. The causes of the problems and the corrective actions implemented to correct these problems are discussed in Reference 13.

This application updates information previously provided in References 2, 4, 5, 6, and 8, and specifically addresses the changes to these references as a result of the additional testing and analysis conducted by OPPD, and Westinghouse/AMAG. Thus, this is not an independent application, but it supplements and revises the application information previously reviewed and approved. The following is a summary of the information provided in this application and a comparison to the application provided in Reference 2.

• In Attachment 1, the page numbering for the requested Bases changes is updated to reflect the current page numbering format in the Technical Specifications.

Section 5.0, Technical Analysis, has been revised to reflect previous correspondence. Also, the list of precedents in Section 9.0 is updated to reflect the Reference 7 approval of the previous OPPD application. These changes are indicated by vertical bars in the margins. The remainder of Attachment 1 is unchanged from Attachment 1 of Reference 2.

• Attachment 2 addresses changes to the 'Summary of Measurement Uncertainty Recapture Evaluation Following Guidance Provided in Regulatory Issue Summary 2002-03," previously submitted as Attachment 2 of Reference 2. Updates include information provided by OPPD in response to the previous NRC requests for information, a global description of the reduction of the requested power uprate from

U. S. Nuclear Regulatory Commission LIC-05-0023 Page 3 1.6% to 1.5% (1524 MWt to 1522 MWt), and pagination changes in the FCS Technical Specifications. New information provided in this application is indicated by vertical bars in the margins. No changes were made to previously provided information in Reference 2 that addressed the remaining sections of RIS 2002-03.

• Attachments 3, 4, 5, and 6 provide an affidavit, a proprietary and non-proprietary discussion of the noise filtering methodology (Appendix A to Attachment 2), and a proprietary revised CROSSFLOW feedwater flow uncertainty calculation (Appendix B to Attachment 2). The entire calculation is considered proprietary, so a non-proprietary version does not exist.

• Proposed changes to the FCS Renewed Operating License and Technical Specifications in Attachments 7 and 8 correspond to Attachments 3 and 4 in Reference 2.

Pursuant to 10 CFR 2.790, OPPD requests that the proprietary information presented and discussed in Appendix A to Attachment 2, "CROSSFLOW Ultrasonic Flow Measurement System Correlated Noise Bias Remediation Through Signal Filtering,"

and Appendix B to Attachment 2, "CROSSFLOW feedwater flow uncertainty calculation," be withheld from public disclosure. This information is proprietary to Westinghouse, as justified in the supporting affidavit (Attachment 3). Attachment 4 is the signal filtering description with proprietary information enclosed in brackets. Attachment 5 is the non-proprietary version of Attachment 4 with the bracketed information deleted. is the wholly proprietary feedwater flow uncertainty.

The general content of each of the Attachments is as follows:

Attachment Content Description A description and assessment of the MUR power uprate including:

1 description, background, proposed OL and TS changes, technical assessment, a no significant hazards consideration and environmental considerations 2 Summary of the MUR power uprate evaluation following guidance provided in RIS 2002-03 3 Affidavit pursuant to 10 CFR 2.790 Appendix A to Attachment 2, "CROSSFLOW Ultrasonic Flow 4 Measurement System Correlated Noise Bias Remediation Through Signal Filtering," [Proprietary Version]

Appendix A to Attachment 2, "CROSSFLOW Ultrasonic Flow 5 Measurement System Correlated Noise Bias Remediation Through Signal Filtering," [Non-proprietary version]

U. S. Nuclear Regulatory Commission LIC-05-0023 Page 4 Appendix B to Attachment 2, "Feedwater Flow Measurement Using the 6 CROSSFLOW Ultrasonic Flowmeter at Fort Calhoun Station"

[Proprietary Calculation CN-PS-03-37, Rev. 1]

7 OL, TS, and TS bases pages marked up to show the proposed changes 8 Revised (clean) OL, TS, and TS bases pages List of regulatory commitments associated with this proposed amendment OPPD requests approval by June 30, 2005 and a 180-day implementation period.

In accordance with 10 CFR 50.91, a copy of this application, with non-proprietary attachments, is being provided to the designated state of Nebraska official.

I declare under penalty of perjury under the laws of the United States of America that I am authorized by Omaha Public Power District to make this request and that the foregoing is true and correct. (Executed on March 31, 2005)

If you have any questions or require information, please contact Thomas C. Matthews at 402-533-6938.

Sincerely, it I I RTR/tcm A

Attachments: See table above c: Division Administrator - Public Health Assurance, State of Nebraska B. S. Mallett, NRC Regional Administrator, Region IV A. B. Wang, NRC Project Manager J. D. Hanna, NRC Senior Resident Inspector

LIC-05-0023 Appendix A Reference Listing

1. Letter from OPPD (W. G. Gates) to NRC (Document Control Desk) dated July 18, 2003 (LIC-03-0067)

2. Letter from OPPD (S. K Gambhir) to NRC (Document Control Desk) dated August 28, 2003 (LIC-03-0122)

3. Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure) dated October 14, 2003, "Fort Calhoun Station Unit No. 1 - Measurement Uncertainty Recapture Power Uprate" (TAC No. MC0029) (NRC-03-198)

4. Letter from OPPD (S. K Gambhir) to NRC (Document Control Desk) dated October 31, 2003 Response to Request for Additional Information - Measurement Uncertainty Recapture Power Uprate (LIC-03-0148)

5. Letter from OPPD (S. K Gambhir) to NRC (Document Control Desk) dated December 15, 2003 Response to Additional Request for Information - Measurement Uncertainty Recapture Power Uprate (LIC-03-0164)

6. Letter from OPPD (S. K Gambhir) to NRC (Document Control Desk) dated December 22, 2003 Response to Additional Request for Information - Measurement Uncertainty Recapture Power Uprate (LIC-03-0166)

7. Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure) dated January 16, 2004, "Fort Calhoun Station Unit No. 1 - Issuance of Amendment" (TAC No. MC0029) (NRC-04-005)

8. Letter from OPPD (S. K. Gambhir) to NRC (Document Control Desk) dated February 6, 2004, Fort Calhoun Station Unit No. 1 License Amendment Request, "Extension of Implementation Period for License Amendment 224" (LIC-04-0017)

9. Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure) dated February 13, 2004, "Fort Calhoun Station Unit No. 1 - Issuance of Amendment" (TAC No. MC1949) (NRC-04-005)

10. Letter from OPPD (R. T. Ridenoure) to NRC (Document Control Desk) dated May 5, 2004, Fort Calhoun Station Unit No. 1 - Notification of Technical Issues Associated with License Amendment No. 225 (TAC No. MC1 949 and TAC No. MC0029) (LIC 0061)

11. Letter from OPPD (R. T. Ridenoure) to NRC (Document Control Desk) dated May 7, 2004, Fort Calhoun Station Unit No. 1 Exigent License Amendment Request, "Restoration of Previous Licensed Rated Power Limit" (LIC-04-0062)

12. Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure) dated May 14, 2004, "Fort Calhoun Station Unit No. 1 - Issuance of Amendment" (TAC No. MC3083) (NRC 017)

13. Letter from WOG (F. P. Schiffley) to NRC (C.l. Grimes) dated December 10, 2004, "Assurance that CROSSFLOW Technology, Installation and Operation Maintain Design and Licensing Basis" (WOG-04-62

LIC-05-0023 Page 1 Description of Change, Safety Evaluation, Significant Hazards Determination and Statement of Environmental Considerations 1.0 Introduction Omaha Public Power District (OPPD) proposes to amend the Facility Operating License (OL) DPR-40 and Technical Specifications (TS) to increase licensed rated power level for Fort Calhoun Station Unit No. 1 (FCS). FCS is currently licensed to operate at a maximum rated power of 1500 megawatts thermal (MWt). Approval is being requested to increase the licensed core rated power by 1.5% to 1522 MWt. This power increase will be accomplished by using more accurate main feedwater flow and temperature measurement to calculate the reactor thermal output. Increasing rated power by reducing measurement uncertainty is called a measurement uncertainty recapture (MUR) power uprate. OPPD has evaluated the impact of a 1.5% uprate to 1522 MWt for the applicable systems, structures, components, and safety analyses at FCS. The results of this evaluation and the new main feedwater flow measurement system are described in of this letter, "Summary of Measurement Uncertainty Recapture Power Uprate Evaluation Following Guidance Provided in NRC Regulatory Issue Summary (RIS) 2002-03," (Reference 10.1).

2.0 Description of License and Technical Specification Changes The proposed license amendment will revise the FCS OL and the TS to increase the licensed rated power by 1.5% from 1500 MWt to 1522 MWt. The proposed changes are described in detail below and are also indicated on the marked up and clean copy Operating License and TS pages in Attachments 6 and 7.

Revise paragraph 3.A of the operating license, DPR-40, to authorize operation at reactor core power levels not in excess of 1522 MWt.

Revise the definition of RATED POWER in the Technical Specifications Definitions to reflect the increase from 1500 MWt to 1522 MWt.

Corresponding TS Bases changes are also requested:

In the Basis of TS 2.1.6, Pressurizer and Main Steam Safety Valves, page 22, change all instances of "1500 MWt" to "RATED POWER."

In the Basis of TS 3.5, Containment Tests, page 3, replace "a reactor power level of 1500 MWt," with "RATED POWER."

LIC-05-0023 Page 2 3.0 Background The 1.5% power uprate for FCS is based on eliminating unnecessary analytical margin that is assumed in analyses to account for the measurement uncertainties associated with the calorimetric calculations. FCS current accident and transient analyses include a minimum 2% margin on rated power to account for power measurement uncertainty. This power measurement uncertainty was originally required by Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50), Appendix K, "ECCS Evaluation Models." The rule required a 2% power margin between the licensed power level and the power level assumed for the emergency core cooling system (ECCS) evaluations. In June 2000, the NRC amended 10 CFR 50, Appendix K to provide licensees the option of maintaining the 2% power margin or applying a reduced margin. For the latter case, the new assumed power level had to account for measurement uncertainties in the power level measurement instrumentation. The revised Appendix K rule had an effective date of July 31, 2000.

Uncertainty in the main feedwater flow measurement is one of the most significant contributors to power measurement uncertainty. Based on this fact and on the above Appendix K rule change, OPPD proposes a reduced power measurement uncertainty of 0.5%, thus allowing an increase in rated power of 1.5%. To accomplish this reduction in uncertainty and increase in power, OPPD has installed a CROSSFLOW Ultrasonic Flow Measurement System (CROSSFLOW system) for measuring the main feedwater flow at FCS. The CROSSFLOW system provides a more accurate measurement of feedwater flow than that assumed during the development of the original Appendix K requirements and that of the feedwater flow venturis currently used to calculate reactor power. As documented in Section 4.0 of the CROSSFLOW Safety Evaluation Report (Reference 10.3), the CROSSFLOW system is designed and tested to achieve a generic flow measurement uncertainty of 0.5% or better. The as-installed feedwater mass flow uncertainty at FCS was determined using plant-specific data to establish a bounding total power measurement uncertainty of 0.5%. Installation of new feedwater temperature RTDs provides more accurate temperature measurement than that assumed in the development of original Appendix K requirements. Based on this, OPPD proposes to reduce the power measurement uncertainty required by Appendix K to 0.5%. The improved power measurement uncertainty obviates the need for the 2% power margin originally required by Appendix K, thereby allowing an increase in the rated power available for electrical generation by 1.5%.

In addition to the proposal to increase the rated power to 1522 MWt, OPPD also proposes continued use of the topical reports identified in the OPPD proposal to implement a Core Operating Limits Report (COLR) at FCS (Letter LIC-02-0109, dated October 8, 2002). The topical reports describe the NRC-approved analytical methodologies used to determine the core operating limits for FCS. This includes the small and large break loss of coolant

LIC-05-0023 Page 3 accidents. In some of these topical reports, reference is made to the use of a 2% power measurement uncertainty being applied consistent with 10 CFR 50, Appendix K. OPPD requests that these topical reports be approved for use consistent with this MUR power uprate request (i.e., 0.5% power measurement uncertainty be assumed instead of 2%).

The proposed change was described in section 2.0 of this attachment. Additionally, the reduction of the power measurement uncertainty does not constitute a significant change as defined in1O CFR 50.46 (a) (3) (I) regarding ECCS evaluation models.

3.1 Licensing Methodologies for Uprate The proposed FCS MUR power uprate is consistent with topical report CENPD-397-P-A, Rev. 1, "Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology." The NRC has approved this topical report for referencing in MUR power uprate submittals. OPPD is specifically applying this topical report, and the criteria listed in the NRC SER for the CENPD-397-P-A, Rev. 1, for a requested 1.5% rated power increase.

In addition to the above methodology, OPPD has taken into account the specific guidance developed by the NRC for the content of MUR power uprate applications. This guidance was published on January 31, 2002, as NRC RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," (Reference 10.1). of this application provides an evaluation of the proposed MUR power uprate structured to be consistent with the NRC guidance. The NRC requests for additional information (RAI) for other licensee MUR power uprate requests were also reviewed and answers for applicable RAls have been incorporated into the text of .

3.2 Licensing Approach to Plant Safety, Component and System Analyses The reactor core power and the NSSS thermal power are used as inputs to most plant safety, component and system analyses. Generally, the FCS MUR power uprate analyses were evaluated as such:

• For safety analyses the power level was bounded at 153OMWt. 1

• For component analyses, reviews were conducted to verify original design basis limits were still applicable.

• For systems analyses, reviews were conducted for overall system performance to bounding uprate conditions. Some re-analyses were performed to ensure that parameters would be bound at the new power level.

Ill Note: some safety analyses are evaluated at zero percent power for most limiting conditions.

LIC-05-0023 Page 4 No new safety analysis techniques have been used to support this power uprate request.

3.3 Conclusion OPPD is requesting a 1.5% increase in core rated thermal power for FCS from 1500 MWt to 1522 MWt. This power increase will be accomplished by using a more accurate main feedwater flow measurement system to calculate the reactor power. This higher accuracy measurement is achieved with the use of a CROSSFLOW system and improved temperature instruments. This license amendment request has taken into account industry and NRC accepted methodologies and guidelines for power uprates.

This License Amendment Request (LAR) is made pursuant to 10 CFR 50.90 to modify the OL and the TS requirements associated with rated thermal power and the use of the power measurement uncertainty in safety analyses.

4.0 Regulatory Requirements & Guidance OPPD has evaluated the impact of the proposed power uprate on safety analyses, NSSS systems and components, and balance of plant (BOP) systems. Attachment 2 summarizes the results of the comprehensive engineering review performed to evaluate the increase in the licensed core rated power. Results of this evaluation are provided in a format consistent with the regulatory guidance provided in NRC RIS 2002-03 (reference 10.1). The results of OPPD's evaluation demonstrate that applicable acceptance criteria will continue to be met following the implementation of the proposed 1.5% MUR power uprate.

5.0 Technical Analysis addressesSection I of RIS 2002-03. No changes were made to previously provided information in Reference 2 that addressed the remaining sections of RIS 2002-

03. Section 1 has been updated to include information provided by OPPD in response to the NRC requests for information. Other sections have been updated as noted. New information provided in this application is indicated by vertical bars in the margins.

6.0 Regulatory Analysis Based on the detailed considerations discussed in Attachment 2 and the No Significant Hazards Determination, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be harmful to the common defense and security or to the health and safety of the public.

LIC-05-0023 Page 5 7.0 No Significant Hazards Determination OPPD has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or...

Response: No.

There are no changes as a result of the MUR power uprate to the design or operation of the plant that could affect system, component, or accident functions.

All systems and components function as designed and the performance requirements have been evaluated and found to be acceptable.

The reduction in power measurement uncertainty allows for safety analyses to continue to be used without modification. This is because those safety analyses were performed or evaluated at 102% of 1500 MWt (1530 MWt) or higher.

Analyses at these power levels support a core power level of 1522 MWt with a measurement uncertainty of 0.5%. Radiological consequences of USAR Chapter 14 accidents were assessed previously using the alternate source term methodology (Reference 10.2). These analyses were performed at 102% of 1500 MWt (1530 MWt) and continue to be bounding. Updated Safety Analysis Report (USAR) Chapter 14 analyses and accident analyses continue to demonstrate compliance with the relevant accident analyses' acceptance criteria. Therefore, there is no significant increase in the consequences of any accident previously evaluated.

The primary loop components (reactor vessel, reactor internals, control element drive mechanisms, loop piping and supports, reactor coolant pumps, steam generators, and pressurizer) were evaluated at an uprated core power level of 1524 MWt and continue to comply with their applicable structural limits. These analyses also demonstrate the components will continue to perform their intended design functions. Changing the heatup and cooldown curves is based on uprated fluence values. This does not have a significant effect on the reactor vessel integrity. Thus, there is no significant increase in the probability of a structural failure of the primary loop components. The LBB analysis conclusions remain valid and the breaks previously exempted from structural consideration remain unchanged.

All of the NSSS systems will continue to perform their intended design functions during normal and accident conditions. The auxiliary systems and components

LIC-05-0023 Page 6 continue to comply with the applicable structural limits and will continue to perform their intended functions. The NSSS/BOP interface systems were evaluated at 1522 MWt and will continue to perform their intended design functions. Plant electrical equipment was also evaluated and will continue to perform their intended functions.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or...

Response: No.

No new accident scenarios, failure mechanisms, or single failures are introduced as a result of the proposed change. All systems, structures, and components previously required for the mitigation of an event remain capable of fulfilling their intended design function at the uprated power level. The proposed change has no adverse effects on any safety related systems or component and does not challenge the performance or integrity of any safety related system. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

Response: No.

Operation at 1522 MWt core power does not involve a significant reduction in the margin of safety. The current accident analyses have been previously performed with a 2% power measurement uncertainty or at uprated core powers that exceed the MUR uprated core power. System and component analyses have been completed at the MUR uprated core power conditions. Analyses of the primary fission product barriers at uprated core powers have concluded that all relevant design basis criteria remain satisfied in regard to integrity and compliance with the regulatory acceptance criteria. As appropriate, all evaluations have been both reviewed and approved by the NRC, or are currently under review (the proposed Pressure-Temperature Limits Report). Therefore, the proposed change does not involve a significant reduction in margin of safety.

==

Conclusion:==

Operation of FCS in accordance with the proposed amendment will not result in a significant increase in the probability or consequences of any accident previously analyzed; will not result in a new or different kind of accident from any accident previously analyzed; and does not result in a significant reduction in a margin of safety.

LIC-05-0023 Page 7 Based on the above, OPPD concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92 (c), and, accordingly, a finding of "no significant hazards consideration" is justified.

8.0 Environmental Consideration The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22 (c) (9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

In accordance with RIS 2002-03, the environmental considerations pertaining to this license amendment request are addressed in detail in Attachment 2,Section VII, "Environmental Review."

9.0 Precedent Between October 10, 2002 and June 30, 2004, NRC Safety Evaluation Reports (SERs) were issued to the following stations for increased power level due to measurement uncertainty recapture:

1. Grand Gulf Nuclear Station, Issuance of Amendment, 1.7% Increase in Licensed Power Level (TAC No. MB3972), October 10, 2002.

2. H. B. Robinson Steam Electric Plant, UNIT NO. 2 (HBRSEP2) Issuance of Amendment Regarding a 1.7 Percent Power Uprate (TAC No. MB5106), November 5, 2002.

3. Peach Bottom Atomic Power Station, Units 2 and 3 Issuance of Amendment 1.62%

Increase in Licensed Power Level (TAC Nos. MB5192 and MB5193), November 22, 2002.

4. Indian Point Nuclear Generating Unit No. 3 Issuance of Amendment 1.4 Percent Power Uprate (TAC No. MB5297), November 26, 2002.

5. Point Beach Nuclear Plant, Units 1 and 2 Issuance of Amendments Measurement Uncertainty Recapture Power Uprate (TAC Nos. MB4956 and MB4957), November 29, 2002.

6. Donald C. Cook Nuclear Plant, Unit 1 Issuance of Amendment 273 Regarding Measurement Uncertainty Recapture Power Uprate (TAC No. MB5498), December 20, 2002.

7. River Bend Station, Issuance of Amendment 1.7 Percent Increase in Licensed Power Level (TAC No. MB5094) January 31, 2003.

LIC-05-0023 Page 8

8. Edwin I. Hatch Nuclear Plant, Units 1 And 2, Issuance of Amendments Regarding Appendix K Measurement Uncertainty Recovery (TAC Nos. MB7026 AND MB7027), September 23, 2003

9. Fort Calhoun Station Unit No. 1 - Issuance of Amendment [Regarding Measurement Uncertainty Recapture Power Uprate] (TAC No. MC0029),

January 16, 2004

10. Palisades Plant - Issuance of Amendment Regarding Measurement Uncertainty Recapture Power Uprate (TAC No. MB9469), June 23, 2004 Additional SERs have been issued to stations for measurement uncertainty recapture prior to October 2002.

10.0 References 10.1 NRC Regulatory Issue Summary 2002-03: "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," January 31, 2002.

10.2 NRC approved the FCS license amendment (T. S. Amendment 201) for implementation of R. G. 1.183 in NRC Letter dated December 5, 2001, "Fort Calhoun Station, Unit No. 1 Issuance of Amendment (TAC NO. MB1221)".

10.3 NRC Safety Evaluation Report: "Acceptance for Referencing of CENPD-397-P, Revision-01-P, "Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology" (TAC NO.

MA6452), March 20, 2000.

10.4 CENPD-397-P, Revision-01-P, "Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology," May, 2000.

LIC-05-0023 Page 1 Summary of the MUR Power Uprate Evaluation Following Guidance Provided in Regulatory Issue Summary 2002-03 Note: This attachment addresses changes to the "Summary of Measurement Uncertainty Recapture Evaluation Following Guidance Provided in Regulatory Issue Summary 2002-03," previously submitted as Attachment 2 of Reference 2. Specifically, portions of the Introduction,Section I, Section IV, and Section VIII have been updated to include information provided by OPPD in response to the previous NRC requests for information, and to reflect changes in the FCS Technical Specifications. New or revised information provided in this attachment is indicated by vertical bars in the margins. No changes were made to previously provided information in Attachment 2 of Reference 2 that addressed the remaining sections of RIS 2002-03.

References for Attachment 2

1. NRC Regulatory Issue Summary 2002-03: Guidance On The Content Of Measurement Uncertainty Recapture Power Uprate Applications, Nuclear Regulatory Commission Office Of Nuclear Reactor Regulation, January 31, 2002.

2. Letter from OPPD (S. K Gambhir) to NRC (Document Control Desk) dated August 28, 2003 (LIC-03-0122)

3. Letter from NRC (A. B. Wang) to OPPD (R. T. Ridenoure) dated January 16, 2004, "Fort Calhoun Station Unit No. 1 - Issuance of Amendment" (TAC No. MC0029) (NRC-04-005)

4. Letter from OPPD (S. K Gambhir) to NRC (Document Control Desk) dated October 31, 2003 Response to Request for Additional Information - Measurement Uncertainty Recapture Power Uprate (LIC-03-0148)

5. Letter from OPPD (S. K Gambhir) to NRC (Document Control Desk) dated December 15, 2003 Response to Additional Request for Information - Measurement Uncertainty Recapture Power Uprate (LIC-03-0164)

6. Letter from OPPD (S. K Gambhir) to NRC (Document Control Desk) dated December 22, 2003 Response to Additional Request for Information - Measurement Uncertainty Recapture Power Uprate (LIC-03-0166)

Introduction Omaha Public Power District (OPPD) proposes to amend the Operating License (OL)

DPR-40 and the Technical Specification (TS) for Fort Calhoun Station (FCS). FCS is presently licensed for a core power rating of 1500 MWt (Section 3.A). Through the use of more accurate feedwater flow measurement instrumentation, approval is sought to increase the licensed core power by 1.5%, to 1522 MWt. The proposed 1.5% power uprate is based on eliminating unnecessary analytical margin originally required of ECCS

LIC-05-0023 Page 2 evaluation models developed in accordance with the requirements set forth in 10 CFR 50, Appendix K "ECCS Evaluation Models."

In June 2000, the NRC approved a change to the 10 CFR 50, Appendix K, requirements to provide licensees with the option of maintaining the 2% power margin between the licensed core power level and the assumed core power level for ECCS evaluations, or apply a reduced margin to the ECCS evaluations. The proposed alternative to recapture margin for ECCS evaluation has been demonstrated to account for uncertainties due to a reduction in core power level measurement instrumentation error. OPPD has installed the Westinghouse CROSSFLOW system and improved feedwater temperature instrumentation with a calculated power measurement uncertainty of 0.5%. Based on the implementation of the CROSSFLOW system with improved feedwater temperature instrumentation and FCS-specific power calorimetric uncertainties, OPPD proposes to reduce the licensed core power uncertainty required by 10 CFR 50, Appendix K, to 0.5%

and to increase core power level by 1.5% using NRC-approved methodologies.

OPPD provided the evaluated impact of the proposed power uprate on NSSS systems and components, BOP systems, safety analyses, and programs in Reference 2. The results of the NRC review of this evaluation are reported in Reference 3. The results of OPPD's analyses and evaluations, which demonstrate that applicable acceptance criteria will continue to be met, are summarized in this assessment. RIS 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications,"

(Reference 1) was used to establish the appropriate scope, structure, and level of detail presented in this assessment.

Overview of this Attachment A comprehensive engineering review program consistent with RIS 2002-03 was provided in Reference 2 to evaluate the increase in the licensed core power from 1500 MWt to 1522 MWt.

Reference 2 provided the evaluation of the accident and transient analyses that correspond to Items II and IlIl of RIS 2002-03. Reference 2 also provided the evaluations of the effect of the power uprate on the structural integrity of major plant components, electrical equipment and major plant systems that correspond to Items IV, V and VI of RIS 2002-03. Lastly, Reference 2 provided the evaluation that addresses Item VII of RIS 2002-03.

The results of the NRC review of the evaluations noted above were reported in Reference 3.

Revised Section I of this attachment describes the feedwater flow measurement technique and power measurement uncertainty that correspond to Item I of RIS 2002-03.

LIC-05-0023 Page 3 Included is a change to Section IV.1.1.2 concerning Reactor Vessel Pressure-Temperature Limits, reflecting the approved FCS Pressure-Temperature Limits Report.

Section VII.3 has been revised because both relief valves associated with feedwater heaters FW-16A & B (as well as the new feedwater temperature instrumentation (RTDs))

were replaced in the 2003 refueling outage.

Revised Section VIII updates the required changes to the FCS TS.

LIC-05-0023 Page 4

1. Feedwater Flow Measurement Technique and Power Measurement Uncertainty Instrumentation The feedwater flow measurement system installed at FCS is the CROSSFLOW ultrasonic flow measurement system (CROSSFLOW system). The installation of this system conforms to the requirements of the topical report cited next.

A. The referenced topical report for the CROSSFLOW system is CENPD-397-P-A, Revision 1, Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology, May, 2000. Updated to include safety evaluation reference.

B. The NRC approved CENPD-397-P, Revision-01-P "Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology" for referencing in power uprate license applications in a safety evaluation dated March 20, 2000 (TAC No. MA6452).

C. CROSSFLOW System The Westinghouse CROSSFLOW Ultrasonic Flow Measurement (UFM) system is used in conjunction with the plant process computer, to support the increase in reactor power. Reactor power is calculated using plant supplied inputs for feedwater temperature, steam generator pressure, steam generator moisture carryover, blowdown flow and feedwater venturi flow that has been corrected to improve its accuracy using the CROSSFLOW system.

Precision temperature instrumentation has been installed for feedwater temperature measurement. This instrumentation is used by the calorimetric calculation, XC105 (T1396 and T1399). The instrumentation consists of precision matched resistance temperature detectors (RTDs) and transmitters manufactured by Rosemount. The new RTDs and transmitters are installed in the same place as the previous equipment.

Consistent with Section 5.10 of CENPD-397-P-A, this higher accuracy feedwater temperature instrumentation lowers the total feedwater flow measurement uncertainty.

The new feedwater temperature instrumentation was installed to reduce the temperature measurement uncertainty. In addition to the new instrumentation, a reduction in the calibration tolerance for the instrument loop further reduces the instrument loop uncertainty. The new instrumentation and calibration tolerance in combination reduces the total loop uncertainty from approximately +/-4.80 F to less than

+/-0.80 F. The reduction in uncertainty is primarily accomplished through the use of transmitted sensor matching with the new RTD/transmitters. This greatly reduces the overall temperature measurement error and results in a reduction in the transmitted sensor uncertainty from approximately +/-4.7 0F (with the original RTD/transmitter combination) to less than +/-0.650 F (new RTD/transmitter combination). Entering the

LIC-05-0023 Page 5 temperature resistance profile specific to the RTD into the transmitter results in transmitter - sensor matching and eliminates the sensor interchangeability error that exists with the current sensor/transmitter combination. In accordance with the vendor information for the new RTD, the sensor interchangeability error (if transmitter - sensor are not matched) is 2.340 F at 3920 F and would result in an uncertainty value of 2.430F when inputted into the uncertainty equation. Matching of the sensor to the transmitter eliminates this error and results in the calculated transmitter sensor uncertainty value of less than +/-0.650F.

The CROSSFLOW system consists of ultrasonic sensors that are permanently mounted on the main feedwater common header, cables, signal conditioning unit, multiplexer, a data processing computer and associated proprietary software. The CROSSFLOW meter is mounted upstream of where the auxiliary feedwater enters the pipe. FCS is using two CROSSFLOW systems mounted in series which provide four redundant measurements of total feedwater flow. The four transducer sets (2 transmitters and 2 receivers in each set) are mounted on two metal support frames that attach, externally, to the feedwater pipe. The brackets are mounted on a straight pipe approximately 40 and 45 pipe diameters downstream of a 450 elbow in the same plane as the elbow. (Note: There was an uncorrected error in a previous response to a request for additional information contained in Reference 5 of the cover letter. The response should have indicated that the bracket was located approximately 45 pipe diameters downstream of the nearest elbow rather than 54 pipe diameters.) The upstream bracket has transducer sets mounted at the 1:30 (450) and 12:00 o'clock (0°)

positions and the downstream bracket has transducer sets mounted at the 4:30 (1350) and 3:00 o'clock (900) positions. Only one set of transducers is used to provide the total feedwater flow measurement. The other three sets provide redundant flow measurements. The four independent instruments are used to verify that the CROSSFLOW installation at FCS is equivalent to known calibration and plant configurations for the specific installation, including the propagation of flow profile effects at higher Reynolds numbers and can be utilized to provide the total feedwater flow measurement in the event of a failure in the primary transducer set.

Signals are passed from the ultrasonic transducers, through the multiplexer to the signal conditioning unit and data processing computers located in a non-harsh environment area. The functions of the signal conditioning unit and data processing computer are described in the Topical Report CENPD-397-P-A, Revision 1, Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology, May 2000.

The data processing computer receives values of feedwater flow, temperature and pressure for each loop from the plant process computer. The data processing computer then calculates a feedwater density and compensates for any thermal growth of the feedwater pipe due to a change in feedwater temperature. It also calculates an instantaneous correction factor for the venturi by measuring the feedwater flow with the CROSSFLOW meter and then dividing it by the corresponding sum of the venturi

LIC-05-0023 Page 6 readings for the same time period. The instantaneous correction factor is next added to a moving average of correction factors in order to smooth the data. The data processing computer then verifies the accuracy of the correction factor and passes the smoothed correction factor back to the plant computer along with a quality flag indicating that the correction factor meets the required accuracy for the Appendix K power uprate.

A plant computer interface has been developed for use with the CROSSFLOW system.

The CROSSFLOW / plant computer interface provides data between the ERFCS (plant computer) and the CROSSFLOW computer. This data link sends the required plant data from the plant computer to the CROSSFLOW computer (which generates a correction factor for feedwater flow), and returns the feedwater flow correction factor to the plant computer. New precision matched RTDs have been installed on each steam generator feedwater line for temperature measurement. An audible and visual alarm indicating that CROSSFLOW is inoperable is provided to alert plant operators when the UFM sensors are out of service and when the CROSSFLOW system diagnostics detect conditions outside of acceptable limits.

FCS has three installed electric motor driven main feedwater pumps; at full power, two of these pumps operate. During the commissioning of the CROSSFLOW system at FCS, the CROSSFLOW system self-identified that the venturi flow correction factor shifted as a function of which feedwater pumps were operating. Research determined that small pressure oscillations with frequencies that could be detected by the CROSSFLOW transducers were occurring in the feedwater system. This hydraulic noise changed as a function of which feedwater pumps were in operation. Frequency spectrum analysis determined that the cross-correlated signal was being contaminated by the noise. This noise contamination caused a shift in the time calculated by CROSSFLOW that it takes for the eddies to pass between the two ultrasonic beams.

Because of this noise contamination, OPPD and Westinghouse/AMAG have identified a modified approach for implementing the CROSSFLOW system at FCS. The modified approach employs signal filtering to correct for the effect of the noise. Use of filtering is discussed in CENPD-397-P-A, Revision 1, Section 3.3.4, wherein it notes that once the undesirable noise component is identified, a filter can be used to remove it. This filtering method has been successfully applied at FCS and validated in flow loop testing. In general, the approach is an off-line process that periodically evaluates the CROSSFLOW signal (including the adverse influence of the noise) and calculates the flow with and without noise contamination. Using this information a conservative noise correction factor for the continuously calculated venturi flow correction factor is determined, that effectively remediates the adverse impact of the hydraulic noise. This noise correction factor is entered into the plant computer to correct the venturi flow correction factor for the effects of noise. A detailed discussion of this approach is provided in Appendix A of this Attachment.

LIC-05-0023 Page 7 OPPD and Westinghouse/AMAG are developing an on-line version of the filter process described above that will continuously correct the CROSSFLOW signal and will eliminate the need for applying a noise correction factor to the venturi flow correction factor. Once this method is verified and validated OPPD intends to utilize it in the FCS CROSSFLOW system as an enhanced replacement for the off-line process.

D. Compliance with NRC SER The installation and commissioning of the CROSSFLOW flow measurement system at FCS are consistent with topical report CENPD-397-P-A, Rev. 1, Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology, May, 2000 (updated to include safety evaluation reference). The NRC approved CENPD-397-P, Revision-01-P Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology" for referencing in power uprate license applications in a safety evaluation dated March 20, 2000 (TAC No. MA6452). Although multiple meters have been installed, the feedwater flow uncertainty credits only one set of ultrasonic transducers. The NRC identified the following four criteria that must be addressed by licensees requesting a license amendment based on the Topical Report. FCS will be consistent with the four criteria as described below.

Criterion 1 Discuss the development of maintenance and calibration procedures that will be implemented with the CROSSFLOW UFM installation. These procedures should include process and contingencies for an inoperable CROSSFLOW UFM and the effect on thermal power measurement and plant operation.

Response to Criterion 1 Maintenance and calibration will be performed using FCS maintenance and calibration procedures, which will be developed from vendor information and FCS specific experience, or will be performed by a combination of vendor and FCS procedures. The site procedures will be developed using the CROSSFLOW technical manuals. All work will be performed in accordance with site work control procedures.

Verification of proper CROSSFLOW system operation is provided by onboard system diagnostics. CROSSFLOW operation will be monitored on a periodic basis using an internal time delay check. The onboard system diagnostics enable verification that the signal conditioning unit, computer, and software remain within the stated accuracy.

The FCS-specific flow measurement uncertainty for the CROSSFLOW system is 0.51% of full power flow. This number is based on a detailed analysis of the installed CROSSFLOW instrumentation by Westinghouse. A one sided confidence interval methodology was utilized to determine the plant specific calorimetric measurement uncertainty. This methodology is discussed in Reference 2 and the results of the NRC review of this evaluation are reported in Reference 3. The thermal

LIC-05-0023 Page 8 power measurement uncertainty analysis method previously submitted in Reference 4 utilized a feedwater flow measurement uncertainty of 0.3922%. The thermal power measurement uncertainty was recalculated using this NRC reviewed analysis methodology utilizing the FCS-specific flow measurement uncertainty for the installed CROSSFLOW system. The accuracies of all other process input parameters to the core thermal power calculation reflect either vendor or operating data uncertainties for one full fuel cycle. Hence, there is only one condition that requires an explanation for the proposed actions - an inoperable CROSSFLOW system. If only one of the four individual flow measurements is affected, FCS will switch to one of the three remaining redundant measurements. The following paragraphs describe the proposed actions for complete CROSSFLOW UFM failure (i.e., all of the four redundant measurements).

Complete CROSSFLOW UFM failure will be detected and transmitted to the Plant Computer and will cause an audible alarm in the control room. The CROSSFLOW system does not perform any safety function and is not used to directly control any plant systems. However, adjustments to RPS power indication based on CROSSFLOW are considered important to safety. Operations will enter an operating procedure that will contain a step for complete CROSSFLOW UFM failure.

Plant operations may remain at an Rated Power of 1522 MWt, while continuing to use the last valid CROSSFLOW UFM correction factor in the heat balance calculation. If the CROSSFLOW system is not returned to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, power will be reduced and maintained at the appropriate power levels until at least one of the four redundant CROSSFLOW UFMs is returned to service. The power level is described below.

Condition Measurement Uncertainty Power Level With CROSSFLOW RTP Uncertainty with UFM 1522 MWt Without CROSSFLOW RTP Uncertainty without UFM 1500 MWt As part of the approved amendment implementation process, OPPD will revise appropriate operations procedures to reflect the above responses to the unavailability of the CROSSFLOW system, and will include this information in the operator training program.

Criterion 2 For plants that currently have the CROSSFLOW UFM installed, the licensee should provide an evaluation of the operational and maintenance history of the installed UFM and confirm that the instrumentation is representative of the CROSSFLOW UFM and is bounded by the requirements set forth In Topical Report CENPD-397 -P.

Response to Criterion 2 The CROSSFLOW system installed at FCS is representative of the CROSSFLOW UFM and is bounded by the requirements set forth in Topical Report CENPD-397-P-A, Rev. 1. Four transducer sets are mounted in two Mounting/Transducer Support

LIC-05-0023 Page 9 Frames (M/TSF) upstream of where the auxiliary feedwater enters the main feedwater pipe. These meters provide four redundant measurements of total feedwater flow.

Each M/TSF attaches externally to the feedwater pipe. The M/TSFs are mounted on a straight pipe approximately 40 and 45 pipe diameters downstream of a 450 elbow. The upstream M/TSF has transducer sets mounted at the 1:30 (45°) and 12:00 o'clock (0°)

positions and the downstream M/TSF has transducer sets mounted at the 4:30 (1350) and 3:00 o'clock (90°) positions. The CROSSFLOW UFM sets have operated successfully since they were installed at these locations in August 2004. Only one CROSSFLOW UFM is used to provide the total feedwater flow measurement for use in the calorimetric power calculation.

FCS has three installed electric motor driven main feedwater pumps; at full power, two of these pumps operate. During the commissioning of the CROSSFLOW system at FCS, the CROSSFLOW system self-identified that the venturi correction factor shifted as a function of which feedwater pumps were operating. Research determined that small pressure oscillations with frequencies that could be detected by the CROSSFLOW transducers were occurring in the feedwater system. This hydraulic noise changed as a function of which feedwater pumps were in operation. Frequency spectrum analysis determined that the cross-correlated signal was being contaminated by the noise. This noise contamination caused a shift in the time calculated by CROSSFLOW that it takes for the eddies to pass between the two ultrasonic beams.

Because of this noise contamination, OPPD and Westinghouse/AMAG have identified an approach for implementing the CROSSFLOW system at FCS that employs signal filtering to compensate for the effect of the noise. Use of filtering is discussed in CENPD-397-P-A, Revision 1, Section 3.3.4, which notes that once the undesirable noise component is identified, a filter can be used to remove it. A filtering method has been successfully applied at FCS and validated in flow loop testing. In general, the approach periodically evaluates the CROSSFLOW signal (including the adverse influence of the noise) and calculates the flow with and without noise contamination.

Using this information, a conservative noise correction factor for the continuously calculated venturi flow correction factor is determined that effectively remediates the impact of the hydraulic noise. This noise correction factor is entered into the plant computer to correct the venturi flow correction factor for the effects of noise. A detailed discussion of this approach is provided in Appendix A of this attachment.

OPPD and Westinghouse/AMAG are developing an on-line version of the above filter that will continuously correct the CROSSFLOW signal and will eliminate the need for applying a noise correction factor to the venturi flow correction factor. Once this method is verified and validated, OPPD intends to utilize it in the FCS CROSSFLOW system.

LIC-05-0023 Page 10 Criterion 3 Confirm that the methodology used to calculate the uncertainty of the CROSSFLOW UFM in comparison to the current feedwater flow instrumentation is based on accepted plant setpoint methodology (with regard to the development of instrument uncertainty).

If an alterative methodology is used, the application should be justified and applied to both the venturi and the CROSSFLOW UFM for comparison.

Response to Criterion 3 The methodology used to calculate the uncertainty of the CROSSFLOW UFM in comparison to the current feedwater flow instrumentation is based on accepted plant setpoint methodology using instrument uncertainty guidance of Regulatory Guide 1.105 and ISA S67.04, as described in the Topical Report. An alternative methodology is not used. Further discussion of OPPD's implementation of ISA S67.04.02 is included in References 5 and 6. The NRC review of this implementation is included in Reference 3.

Criterion 4 For a plant at which the installed CROSSFLOW UFM was not calibrated to a site-specific piping configuration (flow profile and meter factors not representative of the plant-specific Installation) should submit additional justification. This justification should show that the meter installation is either independent of the plant-specific flow profile for the stated accuracy or that the installation can be shown to be equivalent to known calibration and plant configurations for the specific installation, including the propagation of flow profile effects at higher Reynolds numbers. Additionally, for previously installed and calibrated CROSSFLOW UFM, the licensee should confirm that the plant-specific Installation follows the guidelines in the CROSSFLOW UFM topical report.

Response to Criterion 4 In-situ testing using both a temporary CROSSFLOW system and the permanent CROSSFLOW system, computational fluid dynamic calculations and scale model tests of the FCS specific feedwater piping configuration have confirmed that the FCS installation is equivalent to known calibration and piping configurations discussed in the topical report for the CROSSFLOW system CENPD-397-P-A, Revision 1, including the propagation of flow profile effects at higher Reynolds numbers. Because the FCS installation is equivalent to known calibration and piping configurations the FCS flow profile and meter factors are consistent with those discussed in the topical report for the CROSSFLOW system CENPD-397-P-A, Revision 1. The FCS installation follows the guidelines in the CROSSFLOW topical report.

LIC-05-0023 Attachment 2 Page 11 E. The following table summarizes the core thermal power measurement uncertainty at FCS:

Table 1-1 FCS Process Parameter Inputs to Reactor Thermal Power Independent Term Uncertainty Sensitivity Variable Feedwater Flow UWFW 0.51 % 1.0107 I Feedwater UTFW 0.690 F 0.4136 TemperatureI SG Pressure UPSG 14.68 psia 0.0144 SG Moisture UMCO A/B 0.11%/0.05% 0.0011/0.0008 C arryover__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

SG Blowdown Flow UWBD 1873 Ibm/hr 0.0034 SG Blowdown UTBD 2.940 F 0.0038 Temperature Reference 4 provided the detailed proprietary calorimetric uncertainty calculation previously performed for FCS. The results of the NRC review of this evaluation are reported in Reference 3. The identical methodology was used to recalculate the calorimetric uncertainty using the FCS specific flow measurement uncertainty.

F. The following information addresses specific aspects of calibration and maintenance procedures addressing the CROSSFLOW system.

i. Calibration and maintenance will be performed by OPPD personnel using site procedures. Westinghouse/AMAG will initially assist OPPD personnel in applying the noise correction factor discussed above. The site procedures will be developed using the CROSSFLOW technical manuals. All work will be performed in accordance with site work control procedures. Routine preventive maintenance activities will include physical inspections, and power supply checks. I&C maintenance personnel will be trained in the operation of the equipment prior to performing any system calibration.

ii. The CROSSFLOW system is designed and manufactured in accordance with Westinghouse's quality assurance program (class 4, considered important to safety) and in accordance with the topical report CENPD-397-P-A, Rev. 1.

iii. Corrective actions involving maintenance will be performed by OPPD l&C maintenance personnel, qualified in accordance with FCS training program.

iv. Reliability of the CROSSFLOW system will be monitored by OPPD reliability engineering personnel. Equipment problems for all plant systems will be under site work control processes. Corrective Action procedures will be maintained that include instructions for notification of deficiencies and error reporting.

LIC-05-0023 Page 12 G/H. If the primary flow measurement is unavailable, FCS may switch to one of the three remaining redundant flow measurements. If the CROSSFLOW system becomes completely unavailable (i.e., all four flow measurements are unavailable),

steady state plant operations at a core thermal output up to rated power may continue for a maximum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the last valid UFM correction factor was used in the calorimetric calculation for use in the daily nuclear power range surveillance. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is based on the minimum frequency for the calibration of the power range channels found in FCS Technical Specifications (TS). Per TS 3.1, Table 3-1, the power range channels are adjusted daily against a calorimetric balance standard (channel adjustment to agree with heat balance calculation). Since the nuclear power range channels will have been adjusted using the heat balance calculated with a valid CROSSFLOW UFM correction factor, the nuclear power range channel adjustment will be acceptable until the next performance of the surveillance.

The control room operators will receive a computer alarm if the CROSSFLOW UFM system becomes inoperable. The operators will then enter an operating procedure, which will direct them through the actions for a CROSSFLOW failure. The procedure will require that a power range nuclear instrumentation channel adjustment surveillance test be performed within one hour of the failure, using the last good correction factor. The CROSSFLOW system must then be returned to service prior to the next power range channel surveillance (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of last good correction factor). If the CROSSFLOW system cannot be returned to service prior to the next surveillance time, reactor power will be reduced to the current maximum allowed thermal power level of 1500 MWt. This maximum power level is based on an overall calorimetric uncertainty of 2% and is conservative the actual calorimetric uncertainty utilizing the high precision RTDs.

LIC-05-0023 Page 13 In the following sections of Attachment 2 to Reference 2, there were many instances where the text noted an uprate of 1.6%, corresponding to a power level of 1524 MWt, as requested by Reference 2. OPPD has determined that in all cases, the justifications and discussions in Reference 2 are bounding for the revised uprate of 1.5% (1522 MWt) described in this letter. This is because the discussions in Reference 2 assumed bounding analyses done at 1.67% or 1.7% uprates (with corresponding power uncertainties and MWt power levels), or at 102.0% power. Therefore, there is no need for global revisions of the uprate and uncertainty values noted in the following sections of Reference 2:

II. Accidents and Transients for which the Existing Analyses of Record Bound Plant Operation at the Proposed Uprated Power Level Ill. Accidents and Transients for which the Existing Analyses of Record do not Bound Plant Operation at the Proposed Uprated Power Level IV. Mechanical/Structural/Material Component Integrity and Design V. Electrical Equipment Design VI. System Design VII. Other Due to the approval of a License Amendment and the completion of a modification subsequent to the Reference 2 request, the Sections IV. 1.1.2 and V1.3 of Attachment 2 to Reference 2 are revised:

IV.1.1.2 Reactor Vessel Pressure-Temperature Limits Per 10 CFR 50 Appendix G, "Fracture Toughness Requirements," the P-T limit curve must be updated prior to the end of the applicable fluence period. The current P-T limit curves are valid to 40 EFPY (Reference IV.1.3) of reactor operation, which is based on the following:

Neutron fluence accumulation of 2.15 x 1019 n/cm 2 .

Limiting weld is the 3-410 axial weld consisting of weld wire heat number 12008/13253.

The fluence value used in the P-T limit curve was based on a "best estimate" fluence analysis (Reference IV.1.4) that must be adjusted by the increase in fast neutron flux by 1.67% which bounds the increase due to the MUR power uprate. This adjustment has been completed by the evaluation "Reactor Vessel Fluence Assessment for Measurement Uncertainty Recovery Uprate" in Reference IV.1.5. Based on this evaluation, the fluence value of 2.15 x 1019 n/cm2 for the limiting 3-410 weld is expected to occur at 39.9 EFPY, which reduces the P-T limit curve applicability to 39.9 EFPY versus 40.0 EFPY. Figure 5-1, "FORT CALHOUN STATION UNIT 1 COMPOSITE P/T

LIC-05-0023 Page 14 LIMITS, 40 EFPY" in the FCS Core Operating Limits Report will be revised prior to the reactor vessel reaching 39.9 EFPYs of operation.

V1.3 Power Uprate Modifications As demonstrated in Sections II through VI, the current plant analyses, design, and operation ensure that the applicable acceptance criteria are met for the MUR Uprate Program. No changes to the RCS or NSSS systems are required to support the MUR Uprate Program.

The changes in flowrates, pressures, and other operating parameters can be accommodated by all existing equipment in the condensate or feedwater systems.

Therefore, no plant changes/modifications are required to the condensate or feedwater systems to implement the MUR Uprate Program other than the installation of the CROSSFLOW flow instrumentation itself and change out of FW-16 A and B relief valves (completed in the 2003 RFO).

As the impacts of the MUR Uprate Program are bounded by the current design and operation of the AFW system, no modifications are required to this system for implementation of the MUR Uprate Program.

Feedwater Heater FW15-A/B was replaced in 2003 due to component reliability requirements. No plant changes/modifications are required to the feedwater heaters and drains for implementation of the MUR Uprate Program.

The Environmental Review has been revised to reflect the 1.5% uprate request:

VII.5 Environmental Review OPPD has evaluated this license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. OPPD has determined that this license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9).

This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:

(i) The amendment involves no significant hazards consideration.

As demonstrated in Attachment 1 of LAR, Section 5.1, No Significant Hazards Consideration, this proposed amendment does not involve a significant hazards consideration.

LIC-05-0023 Page 15 (ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

The MUR Uprate Program thermal power increase will not alter or increase the inventory of radionuclides in the RCS above the current analysis of record. This change will not alter the fuel cladding in a way that affects its mechanical and structural integrity or affects its leakage characteristics. This power uprate will not alter or increase the primary pressure, so there is no additional challenge to the RCS or other fission product barriers. Additionally, increasing core thermal power by 1.5% will not affect or increase water production or inventory use in any way that will affect effluent volume or production. Finally, the 1.5% uprated plant heat discharge will remain below the site FCS limit. The 1.5% power uprate is bounded by the previously evaluated thermal effluent limits. Therefore, this change will not result in a significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The MUR Uprate Program thermal power increase will not alter or increase the analysis of record inventory of radionuclides in the RCS. The radionuclide source terms applicable to personnel dose determination were calculated assuming a core thermal power of 1530 MWt, which bounds the uprated core power of 1522 MWt. This change will not alter the fuel cladding in a way that affects its mechanical and structural integrity or affects its leakage characteristics; therefore, there is no additional challenge to the RCS or other fission product barriers. Finally, no new effluents or effluent release paths are created by the MUR Uprate Program. Therefore, this change will not result in an increase in individual or cumulative occupational radiation exposures.

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Vill. Changes to Technical Specifications, Protection System Settings, and Emergency System Settings The proposed license amendment would revise the FCS OL and TS to increase licensed power level to 1522 MWt, or 1.5% greater than the current level of 1500 MWt.

The proposed changes are described below:

Revise Paragraph 3.A. in OL DPR-40 to authorize operation at a steady state reactor core power level not in excess of 1522 MWt.

LIC-05-0023 Page 16 Revise the definition of RATED POWER in TS to reflect the increase from 1500 MWt to 1522 MWt.

Corresponding TS Bases changes are also proposed:

In the Basis of TS 2.1.6, Pressurizer and Main Steam Safety Valves, page 22, change all instances of "1500 MWt" to "RATED POWER."

In the Basis of TS 3.5, Containment Tests, page 3, replace "a reactor power level of 1500 MWt," with "RATED POWER."

LIC-05-0023 Attachment 3 Affidavit Pursuant to 10 CFR 2.790

LIC-05-0023 Attachment 5 Appendix A to Attachment 2, "CROSSFLOW Ultrasonic Flow Measurement System Correlated Noise Bias Remediation Through Signal Filtering," [Non-Proprietary Version]

LIC-05-0023 NON-PROPRIETARY Appendix A CROSSFLOW Ultrasonic Flow Measurement System Correlated Noise Bias Remediation Using Time Domain Analysis

Background

The presence of correlated noise (e.g., due to standing pressure waves) of sufficient intensity in a feedwater system has the potential to produce a bias in the CROSSFLOW Ultrasonic Flow Measurement System (CROSSFLOW) time delay measurement and, consequently, the calculated fluid velocity. Noisy signals (or contamination) have been observed in the CROSSFLOW signal at some nuclear power plants. The noise was identified using the CROSSFLOW frequency spectrum analyzer software known as DIAGNOSE. Typically, noise is characterized by a periodic anomaly in the frequency spectrum of the demodulated CROSSFLOW signal with marked peaks on the spectrum at one or more frequencies.

Depending on the specific noise structure, its influence can result in either a conservative or non-conservative bias in the time delay measurement.

Since all piping systems have noise to some degree, the potential for noise contamination was not unexpected. [

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Ia,c References

2. CENPD-397-P-A, Rev. 1, "Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology", May 2000

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LIC-05-0023 NON-PROPRIETARY Figure Ia Clean Frequency Spectrum for Both the Upstream (A) and the Downstream (B) Demodulated Signals With No Correlated Noise igure lb Correlated Noise Contaminated Frequency Spectrum for Both the Upstream (A) and the Downstream (B) Demodulated Signals at a Frequency of Approximately 10 Hz a, c

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LIC-05-0023 NON-PROPRIETARY Figure 2 Cross-Correlation Curves for Contaminated Signal With Noise and Clean Signal + Noise Autocorrelation Curve

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LIC-05-0023 NON-PROPRIETARY Figure 3 Installed Transducer Location a, C IN K

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LIC-05-0023 NON-PROPRIETARY Figure 4 Sample of Power Spectrum for S2 (Correlated Noise) a, c k

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LIC-05-0023 NON-PROPRIETARY Figure 5 Sample Power Spectrum for S4 (Correlated Noise) a, c

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LIC-05-0023 NON-PROPRIETARY Figure 6 Sample Power Spectrum for S3 (Noise free)

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LIC-05-0023 NON-PROPRIETARY Figure 7 Sample Power Spectrum for S5 (Noise free) r

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LIC-05-0023 NON-PROPRIETARY Attachment 5 Figure 8 ARL Piping Configuration With Installed Four Transducer Locations

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LIC-05-0023 Attachment 7 Facility Operating License, TS, and TS Bases pages marked up to show the proposed changes

(5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. Maximum Power Level Omaha Public Power District is authorized to operate the Fort Calhoun Station, Unit 1, at steady state reactor core power levels not in excess of 4500 1522 megawatts thermal (rated power).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.

228, are hereby incorporated in the license. Omaha Public Power District shall operate the facility in accordance with the Technical Specifications.

C. Security and Safeguards Contingency Plans The Omaha Public Power District shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Fort Calhoun Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan,"

submitted by letter dated October 18, 2004.

Renewed Operating License No. DPR-40 Revised by letter dated October 28, 2004

TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATIONS DEFINITIONS The following terms are defined for uniform interpretation of these Specifications.

REACTOR OPERATING CONDITIONS Rated Power A steady state reactor core output of 4500 1522 MWt.

Reactor Critical The reactor is considered critical for purposes of administrative control when the neutron flux logarithmic range channel instrumentation indicates greater than 104% of rated power.

Power Operation Condition (Operating Mode 1)

The reactor is in the power operation condition when it is critical and the neutron flux power range instrumentation indicates greater than 2% of rated power.

Hot Standby Condition (Operating Mode 2)

The reactor is considered to be in a hot standby condition if the average temperature of the reactor coolant (Tavg) is greater than 51 50F, the reactor is critical, and the neutron flux power range instrumentation indicates less than 2% of rated power.

Hot Shutdown Condition (Operating Mode 3)

The reactor is in a hot shutdown condition if the average temperature of the reactor coolant (Tavg) is greater than 515SF and the reactor is subcritical by at least the amount defined in Paragraph 2.10.2.

Definitions - Page 1 Amendment No. 32,50,224, 227

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued)

The power-operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

Action statements (5)b. and c. include the removal of power from a closed block valve to preclude any inadvertent opening of the block valve at a time the PORV may not be closed due to maintenance. However, the applicability requirements of the LCO to operate with the block valve(s) closed with power maintained to the block valve(s) are only intended to permit operation of the plant for a limited period of time not to exceed the next refueling shutdown (Mode 5), so that maintenance can be performed on the PORV(s) to eliminate the seat leakage condition.

To determine the maximum steam flow, the only other pressure relieving system assumed operational is the main steam safety valves. Conservative values for all systems parameters, delay times and core moderator coefficients are assumed. Overpressure protection is provided to portions of the reactor coolant system which are at the highest pressure considering pump head, flow pressure drops and elevation heads.

If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve lift pressure would be less than half of the capacity of one safety valve. This specification, therefore, provides adequate defense against overpressurization when the reactor is subcritical.

Performance of certain calibration and maintenance procedures on safety valves requires removal from the pressurizer. Should a safety valve be removed, either operability of the other safety valve or maintenance of at least one nozzle open to atmosphere will assure that sufficient relief capacity is available. Use of plastic or other similar material to prevent the entry of foreign material into the open nozzle will not be construed to violate the "open to atmosphere" provision, since the presence of this material would not significantly restrict the discharge of reactor coolant.

The total relief capacity of the ten main steam safety valves is 6.606 x 106 lblhr. If, following testing, the as found setpoints are outside +/-1% of nominal nameplate values, the valves are set to within the +/-1% tolerance. The main steam safety valves were analyzed for a total loss of main feedwater flow while operating at 4500MW-t RATED POWER( to ensure that the peak secondary pressure was less than 1100 psia, the ASME Section III upset pressure limit of 10% greater than the design pressure. At the powel Of 10Wt RATED POWER, sufficient relief valve capacity is available to prevent overpressurization of the steam system on loss-of-load conditions.(4) These analyses are based on a minimum of four-of-five operable main steam safety valves on each main steam header.

The power-operated relief valve low setpoint will be adjusted to provide sufficient margin, when used in conjunction with Technical Specification Sections 2.1.1 and 2.3, to prevent the design basis pressure transients from causing an overpressurization incident. Limitation of this requirement to scheduled cooldown ensures that, should emergency conditions dictate rapid cooldown of the reactor coolant system, inoperability of the low temperature overpressure protection system would not prove to be an inhibiting factor. The effective full flow area of an open PORV is 0.94

.2 in.

Removal of the reactor vessel head provides sufficient expansion volume to limit any of the design basis pressure transients. Thus, no additional relief capacity is required.

References (1) Article 9 of the 1968 ASME Boiler and Pressure Vessel Code,Section III (2) USAR, Section 14.9 (3) USAR, Section 14.10 (4) USAR, Sections 4.3.4, 4.3.9.5 2.1 - Page 22 Amendment No. 3 9 ,47 ,5 ,l1 6 ,1 6 ,I 8

9. 2 4 ,2 2 7

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (Continued)

Basis The containment is designed for an accident pressure of 60 psig.( 2) While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a maximum temperature of about 1200F. With these initial conditions the temperature of the steam-air mixture at the peak accident pressure of 60 psig is 2880 F.

Prior to initial operation, the containment was strength-tested at 69 psig and then was leak tested. The design objective of the pre-operational leakage rate test has been established as 0.1% by weight for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig. This leakage rate is consistent with the construction of the containment, which is equipped with independent leak-testable penetrations and contains channels over all inaccessible containment liner welds, which were independently leak-tested during construction.

Safety analyses have been performed on the basis of a leakage rate of 0.1% of the free volume per day of the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the maximum hypothetical accident. With this leakage rate, atreactor power level of 1500 MM RATED POWER, and with minimum containment engineered safety systems for iodine removal in operation (one air cooling and filtering unit), the public exposure would be within 10 CFR Part 50.67 values in the event of the maximum hypothetical accident.(3) The performance of an integrated leakage rate test and performance of local leak rate testing of individual penetrations at periodic intervals during plant life provides a current assessment of potential leakage from the containment.

The reduced pressure (5 psig) test on the PAL is a conservative method of testing and provides adequate indication of any potential containment leakage path. The test is conducted by pressurizing between two resilient seals on each door. The test pressure tends to unseat the resilient seals which is opposite to the accident pressure that tends to seat the resilient seals. A periodic test ensures the overall PAL integrity at 60 psig.

The integrated leakage rate test (Type A test) can only be performed during refueling shutdowns.

3.5 - Page 3 Amendment No. 68,97,139,151,185, 216,224, 227

LIC-03-0122 Attachment 8 Revised (clean) Facility Operating License, TS, and TS Bases pages

(5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by operation of the facility.

3. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter 1: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. Maximum Power Level Omaha Public Power District is authorized to operate the Fort Calhoun Station, Unit 1, at steady state reactor core power levels not in excess of 1522 megawatts thermal (rated power).

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.

228, are hereby incorporated in the license. Omaha Public Power District shall operate the facility in accordance with the Technical Specifications.

C. Security and Safeguards Contingency Plans The Omaha Public Power District shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Fort Calhoun Station Security Plan, Training and Qualification Plan, Safeguards Contingency Plan,"

submitted by letter dated October 18, 2004.

Renewed Operating License No. DPR-40 Revised by letter dated October 28, 2004

TECHNICAL SPECIFICATIONS TECHNICAL SPECIFICATIONS DEFINITIONS The following terms are defined for uniform interpretation of these Specifications.

REACTOR OPERATING CONDITIONS Rated Power A steady state reactor core output of 1522 MWt. I Reactor Critical The reactor is considered critical for purposes of administrative control when the neutron flux logarithmic range channel instrumentation indicates greater than 104% of rated power.

Power Operation Condition (Operating Mode 1)

The reactor is in the power operation condition when it is critical and the neutron flux power range instrumentation indicates greater than 2% of rated power.

Hot Standby Condition (Operating Mode 2)

The reactor is considered to be in a hot standby condition if the average temperature of the reactor coolant (Tavg) is greater than 51 50F, the reactor is critical, and the neutron flux power range instrumentation indicates less than 2% of rated power.

Hot Shutdown Condition (Operating Mode 3)

The reactor is in a hot shutdown condition if the average temperature of the reactor coolant (Tavg) is greater than 5150 F and the reactor is subcritical by at least the amount defined in Paragraph 2.10.2.

Definitions - Page 1 Amendment No. 32,50,224, 227

TECHNICAL SPECIFICATIONS 2.0 LIMITING CONDITIONS FOR OPERATION 2.1 Reactor Coolant System (continued) 2.1.6 Pressurizer and Main Steam Safety Valves (continued)

The power-operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

Action statements (5)b. and c. include the removal of power from a closed block valve to preclude any inadvertent opening of the block valve at a time the PORV may not be closed due to maintenance. However, the applicability requirements of the LCO to operate with the block valve(s) closed with power maintained to the block valve(s) are only intended to permit operation of the plant for a limited period of time not to exceed the next refueling shutdown (Mode 5), so that maintenance can be performed on the PORV(s) to eliminate the seat leakage condition.

To determine the maximum steam flow, the only other pressure relieving system assumed operational is the main steam safety valves. Conservative values for all systems parameters, delay times and core moderator coefficients are assumed. Overpressure protection is provided to portions of the reactor coolant system which are at the highest pressure considering pump head, flow pressure drops and elevation heads.

If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve lift pressure would be less than half of the capacity of one safety valve. This specification, therefore, provides adequate defense against overpressurization when the reactor is subcritical.

Performance of certain calibration and maintenance procedures on safety valves requires removal from the pressurizer. Should a safety valve be removed, either operability of the other safety valve or maintenance of at least one nozzle open to atmosphere will assure that sufficient relief capacity is available. Use of plastic or other similar material to prevent the entry of foreign material into the open nozzle will not be construed to violate the "open to atmosphere" provision, since the presence of this material would not significantly restrict the discharge of reactor coolant.

The total relief capacity of the ten main steam safety valves is 6.606 x 106 lb/hr. If, following testing, the as found setpoints are outside +/-1% of nominal nameplate values, the valves are set to within the +/-1% tolerance. The main steam safety valves were analyzed for a total loss of main feedwater flow while operating at RATED POWER 3) to ensure that the peak secondary pressure was less than 1100 psia, the ASME Section III upset pressure limit of 10% greater than the design pressure. At RATED POWER, sufficient relief valve capacity is available to prevent overpressurization of the steam system on loss-of-load conditions.(4) These analyses are based on a minimum of four-of-five operable main steam safety valves on each main steam header.

The power-operated relief valve low setpoint will be adjusted to provide sufficient margin, when used in conjunction with Technical Specification Sections 2.1.1 and 2.3, to prevent the design basis pressure transients from causing an overpressurization incident. Limitation of this requirement to scheduled cooldown ensures that, should emergency conditions dictate rapid cooldown of the reactor coolant system, inoperability of the low temperature overpressure protection system would not prove to be an inhibiting factor. The effective full flow area of an open PORV is 0.94

.2 in.

Removal of the reactor vessel head provides sufficient expansion volume to limit any of the design basis pressure transients. Thus, no additional relief capacity is required.

References (1) Article 9 of the 1968 ASME Boiler and Pressure Vessel Code,Section III (2) USAR, Section 14.9 (3) USAR, Section 14.10 (4) USAR, Sections 4.3.4, 4.3.9.5 2.1 - Page 22 Amendment No. 39,47,54,146,161,189,224,227

TECHNICAL SPECIFICATIONS 3.0 SURVEILLANCE REQUIREMENTS 3.5 Containment Tests (Continued)

Basis The containment is designed for an accident pressure of 60 psig.(2) While the reactor is operating, the internal environment of the containment will be air at approximately atmospheric pressure and a maximum temperature of about 120 0F. With these initial conditions the temperature of the steam-air mixture at the peak accident pressure of 60 psig is 288 0F.

Prior to initial operation, the containment was strength-tested at 69 psig and then was leak tested. The design objective of the pre-operational leakage rate test has been established as 0.1% by weight for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 60 psig. This leakage rate is consistent with the construction of the containment, which is equipped with independent leak-testable penetrations and contains channels over all inaccessible containment liner welds, which were independently leak-tested during construction.

Safety analyses have been performed on the basis of a leakage rate of 0.1% of the free volume per day of the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the maximum hypothetical accident. With this leakage rate, at RATED POWER, and with minimum containment engineered safety systems for iodine removal in operation (one air cooling and filtering unit), the public exposure would be within 10 CFR Part 50.67 values in the event of the maximum hypothetical accident.(3) The performance of an integrated leakage rate test and performance of local leak rate testing of individual penetrations at periodic intervals during plant life provides a current assessment of potential leakage from the containment.

The reduced pressure (5 psig) test on the PAL is a conservative method of testing and provides adequate indication of any potential containment leakage path. The test is conducted by pressurizing between two resilient seals on each door. The test pressure tends to unseat the resilient seals which is opposite to the accident pressure that tends to seat the resilient seals. A periodic test ensures the overall PAL integrity at 60 psig.

The integrated leakage rate test (Type A test) can only be performed during refueling shutdowns.

3.5 - Page 3 Amendment No. 68,97,139,151,185, 216,224, 227

LIC-05-0023 Attachment 9 List of Regulatory Commitments The following table identifies those actions committed to by OPPD in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

COMMITMENT Due Date/Event Figure 5-1, UFORT CALHOUN STATION Prior to reactor vessel reaching 39.9 EFPYs UNIT 1 COMPOSITE P/T LIMITS, 40 of operation.

EFPY" in the FCS Core Operating Limits Report will be revised prior to the reactor vessel reaching 39.9 EFPYs of operation.

Westinghouse Electric Company Nuclear Services Westin houseP.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 USA U.S. Nuclear Regulatory Commission Direct tel: 412/374-4643 ATTN: Document Control Desk Direct fax: 412/3744011 Washington, DC 20555 e-mail: greshaja@westinghouse.com Project No.: 700 Our ref: CAW-05-1969 March 21,2005 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Transmittal of Application For Withholding Proprietary Information Supporting the Ft. Calhoun Appendix K Measurement Uncertainty Recapture Power Uprate License Amendment Request

Reference:

OPPD Letter, "Fort Calhoun Station Unit No. 1, License Amendment Request (LAR),

Measurement Uncertainty Recapture Power Uprate", LIC-05-0023 The proprietary information for which withholding is being requested is contained in the above referenced Omaha Public Power District License Amendment Request in Appendix A, "CROSSFLOW Ultrasonic Flow Measurement System Correlated Noise Bias Remediation Using Time Domain Analysis" and Appendix B, calculation CN-PS-03-37, Rev. 1, "Feedwater Flow Measurement Using the CROSSFLOW Flowmeter at Fort Calhoun Nuclear Station" as identified in Affidavit CAW-05-1969 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 Appendix B does not exist and, due to the extent of proprietary information contained, a non-proprietary version of such calculation would be meaningless.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Omaha Public Power District.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-05-1969, 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.

Very truly V urs, A. Gresham, Manager Regulatory Compliance and Plant Licensing cc: J. K. Gasper (OPPD)

R. A. Gramm (NRC)

T. C. Matthews (OPPD)

A. B. Wang (NRC)

CAW-05-1969 AFFIDAVIT STATE OF CONNECTICUT:

ss: TOWN OF WINDSOR COUNTY OF HARTFORD:

Before me, the undersigned authority, personally appeared J. A. Gresham, 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, a Delaware limited liability company ("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:

. A. Gresham, Manager Regulatory Compliance and Plant Licensing Sworn to and subscribed before me this ___ 21 "t day of March , 2005 N~otaryublic..

My commission expires:

,f

2 CAW-05-1969 (1) I, J. A. Gresham, am the Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), a Delaware limited liability company 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 rulemaking proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(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 Westinghouse 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 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 constitutes Westinghouse policy and provides 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.

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

There are sound policy reasons behind the Westinghouse system 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 which 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 by our competitor would put Westinghouse at a competitive disadvantage by reducing his 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.

3 CAW-05-1969 (e) Unrestricted disclosure 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.

(iii) 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.

(iv) 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.

(v) The proprietary information sought to be withheld by this submittal is that which is contained in proprietary Appendix A, "CROSSFLOW Ultrasonic Flow Measurement System Correlated Noise Bias Remediation Using Time Domain Analysis" and in proprietary Appendix B, calculation CN-PS-03-37, Rev. 1, "Feedwater Flow Measurement Using the CROSSFLOW Flowmeter at Fort Calhoun Nuclear Station", transmitted by Omaha Public Power District Letter, "Fort Calhoun Station Unit No. 1, License Amendment Request (LAR), Measurement Uncertainty Recapture Power Uprate", LIC-05-0023, 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.

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 technical evaluation justifications and 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 for developing the enclosed improved core thermal performance methodology.

Further the deponent sayeth not.

CAW-05-1969 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). Thejustification 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 with 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.