Text

License Amendment Request #296, Revision 1, Measurement Uncertainty Recapture Uprate
	ML071830203
Person / Time
Site:	Crystal River
Issue date:	06/28/2007
From:	Young D; Progress Energy Florida
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	3F0607-05
	Download: ML071830203 (122)
	v • d • e

C2 Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10CFR50.90 June 28, 2007 3F0607-05 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - License Amendment Request #296, Revision 1, Measurement Uncertainty Recapture Uprate

References:

1. Regulatory Issue Summary 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," dated January 31, 2002

2. Amendment No. 205 to the Crystal River Unit 3 Facility Operating License (TAC NO. MB5289), dated December 4, 2002

3. FPC to NRC letter, 3F0407-10, LAR #296, Revision 0, Measurement Uncertainty Recapture Uprate, dated April 25, 2007

4. FPC to NRC letter, 3F0607-04, LAR #296, Revision 0, Measurement Uncertainty Recapture Uprate, Submittal Schedule Update, dated June 18, 2007

Dear Sir:

Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc., hereby submits License Amendment Request (LAR) #296, Revision 1. The proposed amendment will revise the maximum power limit in the Crystal River Unit 3 (CR-3) Facility Operating License, DPR-72, from 2568 MWt to 2609 MWt. The proposed uprate is characterized as a Measurement Uncertainty Recapture (MUR) using a Caldon Leading Edge Flowmeter (LEFM) CheckPlusTM system and other improvements to support an approximate 1.6 percent power increase. This submittal supersedes Reference 3 in its entirety. Except for minor editorial corrections, all substantial changes are contained in Attachment A, Attachment B and C, Attachment D (Sections 2.2.10, 2.2.14, 5.2.3), and deletion of Reference 3, Attachment F. Additionally, Figure D.2-2 was added to Attachment D.

Changes are requested to two definitions in Section 1.1 of the CR-3 Improved Technical Specifications (ITS), Effective Full Power Days and Rated Thermal Power, and License Condition 2.C.(1). The power uprate will permit a more economical operation of CR-3 and will not have a significant impact on the environment or the health and safety of the general public.

Regulatory Issue Summary (RIS) 2002-03, Reference 1, provided guidance for submittals requesting power uprates involving feedwater flow measurement uncertainty recapture. The RIS stated that a licensee's applications which follow this guidance will require less review time and could be approved in six months or less. The guidance of RIS 2002-03 was used to ensure all areas of concern to the NRC staff were addressed in this submittal. Therefore, FPC requests approval of this request by November 15, 2007 with a 60 day implementation period in order to support implementation following Refuel Outage 15 which is scheduled for Fall 2007.

Progress Energy Florida, Inc.

Crystal River Nuclear Plant 15760 W. Powerline Street Crystal River, FL 34428

U.S. Nuclear Regulatory Commission Page 2 of 3 3F0607-05 This letter provides the following information discussed in Reference 4. There are four specific questions being answered: grid reliability, a determination if critical operator actions are affected, clarifications to the description of proposed changes, and a determination if Attachment F of Reference 3 was still needed.

The grid reliability study information is included in Attachment D, Section 5.2.4. Critical operator actions are unaffected by the implementation of the MUR (the determination was made that there are no changes to critical operator actions resulting from MUR, Attachments A and D). Revisions to the change descriptions, which conflicted with the No Significant Hazards Consideration Determination, are included in Attachments A and D. Lastly, Attachment F of Reference 3 and ITS page 5.0-23A are being deleted, as the current mass and energy releases and maximum containment pressure bound the MUR Uprate.

Amendment 205 to the CR-3 Operating License, Reference 2, implemented a previous increase in power from 2544 to 2568 MWt. At that time, evaluations were performed at 102% of 2568 MWt to demonstrate compliance with applicable regulations and the capability of systems and components to operate safely.

Feedwater instrumentation will be supplemented with an ultrasonic flowmeter system that has significantly decreased measurement uncertainty. Existing feedwater flow instrumentation will continue to be used for protective and control functions. The majority of existing CR-3 accident analyses were performed at 2619 MWt or higher, therefore, few analytical changes are needed to support this power uprate.

The proposed amendment has been reviewed and recommended for approval by the Plant Nuclear Safety Committee.

Regulatory commitments are identified in Attachment G.

If you have any questions regarding this submittal, please contact Mr. Paul Infanger, Supervisor, Licensing and Regulatory Programs at (352) 563-4796.

Sincerely, Dale E. Young Vice President Crystal River Nuclear Plant DEY/par Attachments: A. Description of the Proposed Change, Background, Technical Analysis, Determination of No Significant Hazards Considerations, and the Environmental Assessment B. Proposed Operating License and Technical Specification Pages - Strikeout and Shadowed Text Format C. Proposed Operating License and Technical Specification Pages - Revision Bar Format.

D. Response to RIS 2002-03 Questions E. Uncertainty Calculation F. Anticipated Transient Without Scram (ATWS) Mitigation System Actuation Circuitry (AMSAC) Arming Setpoint Evaluation G. List of Regulatory Commitments xc: NRR Project Manager Regional Administrator, Region II Senior Resident Inspector State Contact

U.S. Nuclear Regulatory Commission . Page 3 of 3 3F0607-05 STATE OF FLORIDA COUNTY OF CITRUS Dale E. Young states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation, doing business as Progress Energy Florida, Inc.; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.

Dale E. Young Vice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this - day of 2007, by Dale E. Young.

Signature of Notary Publi* .... CM*M PSS State of Florida Stt flrdMY t1w..iz. COMMISSION FYPIRFS- JuIV#B.

D'E 2

(Print, type, or stamp Commissioned Name of Notary Public)

Personally Produced Known / -OR- Identification

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #296, REVISION 1 ATTACHMENT A DESCRIPTION OF THE PROPOSED CHANGE, BACKGROUND, TECHNICAL ANALYSIS, DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS, AND THE ENVIRONMENTAL ASSESSMENT

U.S. Nuclear Regulatory Commission Attachment A 3F0607-05 Page 1 of 7 DESCRIPTION OF THE PROPOSED CHANGE BACKGROUND, TECHNICAL ANALYSIS, DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS AND THE ENVIRONMENTAL ASSESSMENT

1.0 DESCRIPTION

OF PROPOSED CHANGE The proposed change would revise the Crystal River Unit 3 (CR-3) Operating License to read as follows:

2.C.(1) Maximum Power Level Florida Power Corporation is authorized to operate the facility at a steady state reactor core power level not in excess of 2609 Megawatts (100 percent of rated core power level).

The proposed change would revise ITS definitions to read as follows:

EFFECTIVE FULL POWER DAY (EFPD) - EFPD shall be the ratio of the number of hours of production of a given THERMAL POWER to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , multiplied by the ratio of the given THERMAL POWER to the RTP. One EFPD is equivalent to the thermal energy produced by operating the reactor core at RTP for one full day. (One EFPD is 2609 MWt times 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or 62616 MWhr.)

RATED THERMAL POWER (RTP) - RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2609 MWt.

2.0 BACKGROUND

CR-3 was initially licensed to operate at a maximum of 2452 Megawatts-Thermal (MWt). In Amendment 41, dated July 21, 1981, the NRC approved operation of CR-3 up to 2544 MWt. By letter dated June 5, 2002 (3F0602-05), CR-3 requested an increase in maximum Rated Thermal Power (RTP) to 2568 MWt. At that time, several transient and accident analyses (moderator dilution accident, letdown line failure, loss of feedwater event and small break loss-of-coolant accident) were reevaluated at 102% of 2568 MWt. The request was approved on December 4, 2002 as Amendment 205 to the CR-3 Facility Operating License. Since the evaluations in 2002 were performed, few changes to plant systems, structures or components (SSC) have occurred.

All analyses have been performed at 102% of 2568 MWt or higher except as otherwise discussed.

CR-3 has determined that implementation of the Caldon Leading Edge Flowmeter CheckPlusTM System and other changes are an effective way to obtain additional power from the plant without significantly changing current reactor core operations. This will permit the recovery of approximately 41 MWt by using more accurate instrumentation to calculate core power production. A spool piece will be installed in each of the two feedwater pipes, containing 16 ultrasonic, multi-path, transit time transducers. The CheckPlusTM system provides more accurate feedwater flow and temperature. New feedwater pressure and main steam temperature and pressure instrumentation will also be installed. The currently installed flow instruments will continue to provide inputs to other indication and control systems.

U.S. Nuclear Regulatory Commission Attachment A 3F0607-05 Page 2 of 7

3.0 TECHNICAL ANALYSIS

The licensed rated thermal power level for CR-3 is proposed to be increased from 2568 MWt to 2609 MWt. The uprate evaluation addressed the following categories: Nuclear Steam Supply System (NSSS) performance parameters, design transients, systems, components, accidents, and nuclear fuel, as well as interfaces between the NSSS and balance-of-plant (BOP) systems. Most of these evaluations are contained in Attachment D, Response to RIS 2002-03 Questions, which is based largely on AREVA Technical Report 51-9042409-000. No new analytical techniques were used to support the power uprate project. The uprate evaluation takes into consideration applicable licensing criteria and standards, including Regulatory Issue Summary 2002-03.

All current analyses were performed considering a maximum power output of 2568 MWt or higher. 10 CFR 50, Appendix K, analyses were done at 2619 MWt (102 percent of 2568 MWt) to account for the two percent uncertainty previously assumed in power measurement. This continues to be valid based on 100.4% of 2609 MWt. Some analyses were performed at higher power levels (generally 2772 MWt) because they were performed generically to bound all Babcock and Wilcox (BAW) plants. All of these analyses were reviewed to provide assurances that they remain the bounding or limiting analyses. Most of the systems, structures, and components (SSC) evaluated for the previous uprate, and discussed in the Safety Evaluation for Amendment 205, are virtually unchanged from the SSC that currently exist in the plant.

This section discusses the revised NSSS design thermal and hydraulic parameters that changed as a result of the power uprate and that serve as the basis for all of the NSSS analyses. A detailed assessment of the accident analyses performed for the steam generator tube rupture, Loss of Coolant Accident (LOCA), and non-LOCA areas was performed. The Loss of Coolant mass and energy release remain bounding. The radiological consequence evaluation is bounded by the current analysis since the radiological source term had not increased, due to the analytical limit of 2619 MWt not changing. The fuel was also evaluated for its ability to perform at the uprated power level. CR-3 concludes that the changes to the plant design basis and transient analyses are acceptable. Each of the NSSS systems and components were evaluated for the uprated conditions. The effects of the uprate on the BOP (secondary) systems, electrical power systems, control systems and instrumentation systems were also evaluated. The results of all of the analyses and evaluations performed demonstrate that all acceptance criteria continue to be met and that the plant requires minimal additional design changes other than setpoint adjustments to safely operate at the uprated conditions. A summary of these evaluations and assessments follows.

3.1 Nuclear Steam Supply System (NSSS) Design Parameter The NSSS parameters are the fundamental parameters which are used as input in all the NSSS analyses. They provide the Reactor Coolant System (RCS) and secondary system conditions (temperatures, pressures, flows) that are used as the basis for the design transient, system, component and accident evaluations. The parameters for design are established using conservative assumptions in order to provide bounding conditions to be used in the NSSS analyses.

U.S. Nuclear Regulatory Commission Attachment A 3F0607-05 Page 3 of 7 3.1.1 Input Parameters and Assumptions The total thermal power for the uprate analysis was set at 2609 MWt (core power). This is approximately 1.6 percent higher than the current core thermal power rating of 2568 MWt.

Feedwater and main steam pressure, feedwater and main steam temperature, feedwater flow instrument improvements and RCS cold leg temperature (Tcold), letdown flow, temperature, and make-up water temperature-uncertainty calculations were required to support this power uprate.

3.1.2 Discussion of Parametric Cases Table A. 1 provides the NSSS parameter cases which were generated and used as the basis for the uprate project. Uprated conditions were calculated at current and 20 percent once-through steam generator (OTSG) tube plugging (the CR-3 licensed tube plugging limit) to bound the range of RCS temperatures and steam conditions (flow rate and temperature) which could occur as part of the uprate. This Table provides the values used in the RCS functional specification as well as the calculated uprated conditions at current and 20 percent OTSG tube plugging. The parameters listed in Table 1 have been reviewed against the RCS functional specification. The RCS functional specification bounds the uprated conditions. For reactor coolant flow, the original functional specification design flow was not used for flow-induced vibration analysis.

As discussed in BAW-10051, Revision 1, "Design of Reactor Internals and Incore Nozzles for Flow-Induced Vibrations," conservative flow velocities were used. The very slight change in mass flow remains within the velocity used and thus has negligible impact on the components.

3.1.3 Conclusions Changes to plant operating conditions were determined for the 1.6 percent power uprate (values for current and 20 percent OTSG tube plugging are listed in Table A. 1). The new operating conditions were compared with original design conditions for the RCS. The power uprate will not result in operation outside the original design conditions. The change in operating conditions and increased power was used to evaluate systems, components, materials, fuel and safety analysis. It has been concluded that the cumulative effect of the evaluations for all systems, components and analyses support the power uprate.

U.S. Nuclear Regulatory Commission Attachment A 3F0607-05 Page 4 of 7 Table A. 1 Change in Operating Conditions For MIUR Parameter Case A Case B Case C Core Thermal Power (MWt) 2568 2609 2609 Other RCS Power MWt (RCP-LD)'- (QRCS) 16 16 16 Total RCS Power (MWt) 2584 2625 2625 Pressurizer Control Pressure (psig) 2155 2155 2155 SG A/B Tube Plugging (%) 2.4/5.7 2.4/5.7 20/20 Thot (TF) 601.9 602.2 602.9 Tcold (TF) 556.2 555.8 555.1 Tavg ('F) 579 579 579 RCS Mass Flow (E6 lbm/hr) 144.05 144.08 139.81 RCS Volumetric Flow (gpm) 386,873 386,729 374,896 Steam Temperature (TF) 591.0 590.5 580.7 Steam Superheat (TF) 54.9 54.4 44.6 Feedwater/Steam Flow Rate (E6 lbmjhr) 10.86 11.07 11.19 OTSG Steam Pressure (psia) (input) 931.7 931.7 931.7 Feedwater Temperature ('F) (input) 456.7 458.4 458.5 Case A - Existing tube plugging at 2568 MWt Case B - Existing tube plugging at 2609 MWt Case C - 20 percent tube plugging at 2609 MWt 4.0 NO SIGNIFICANT HAZARDS CONSIDERATION Florida Power Corporation (FPC) has evaluated the proposed License Amendment Request (LAR) against the criteria of 10 CFR 50.92 to determine if any significant hazards consideration is involved. FPC has concluded that this proposed LAR does not involve a significant hazards consideration. The following is a discussion of how each of the 10 CFR 50.92 (c) criteria is satisfied.

(1) Does not involve a significant increase in the probabilityor consequences of an accident previously evaluated.

The proposed change will increase the maximum core power level from 2568 MWt to 2609 MWt. This increase will only require adjustments and calibrations of existing plant instrumentation and control systems. The only equipment upgrades necessary for this uprate are spool pieces containing multiple ultrasonic flow instruments, which will be installed in each feedwater line, as well as more accurate instrumentation for feedwater pressure and steam pressure and temperature. Indication and control functions will continue to be performed by the currently installed feedwater instrumentation.

U.S. Nuclear Regulatory Commission Attachment A 3F0607-05 Page 5 of 7 Nuclear steam supply systems (NSSS) and balance-of-plant (BOP) systems and components that could be affected by the proposed change have been evaluated using revised NSSS design parameters based on a core power level of 2609 MWt. The results of these evaluations, which used well-defined analysis input assumptions/parameter values and currently approved analytical techniques, indicate that CR-3 systems and components will continue to function within their design parameters and remain capable of performing their required safety functions at 2609 MWt. Since the revised NSSS parameters remain within the design conditions of the Reactor Coolant System (RCS) functional specification, the proposed change will not result in any new design transients or adversely affect the current CR-3 design transient analyses.

The accidents analyzed in Chapter 14 of the CR-3 Final Safety Analysis Report (FSAR) have been reviewed for the impact of the uprate. Based on the power levels assumed in the current safety analyses, it has been determined that all FSAR and supporting analyses bound the uprate.

This includes the dose calculations for the design basis radiological accidents, which assume a power level of 2619 MWt (2568 MWt plus an assumed 2 percent measurement uncertainty).

Since the proposed change relies on less than 0.4 % uncertainty, the assumed power level of 100.4% of 2609 MWt remains 2619 MWt. Therefore, analyses performed at this power remain bounding.

(2) Does not create the possibility of a new or different kind of accidentfrom any accident previously evaluated.

As discussed above, the only equipment upgrades necessary for this uprate are spool pieces containing multiple ultrasonic flow instruments, which will be installed in each feedwater line, as well as more accurate instrumentation for feedwater pressure and steam pressure and temperature. All CR-3 systems and components will continue to function within their design parameters and remain capable of performing their required safety functions. The proposed change does not impact current CR-3 design transients or introduce any new transients.

Equipment failure modes are expected to be the same as for existing instruments. Protective and control functions will continue to be performed by the currently installed feedwater instrumentation. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

(3) Does not involve a significant reduction in a margin of safety Challenges to the fuel, RCS pressure boundary and containment were evaluated for uprate conditions. Core analyses show that the implementation of the power uprate will continue to meet the current nuclear design basis. Impacts to components associated with RCS pressure boundary structural integrity, and factors such as pressure/temperature limits, vessel fluence, and pressurized thermal shock (PTS) were determined to be bounded by current analyses.

As discussed above, all systems will continue to operate within their design parameters and remain capable of performing their intended safety functions following implementation of the proposed change. Finally, the current CR-3 safety analyses, including the design basis radiological accident dose calculations, bound the uprate. Therefore, this change does not involve a significant reduction in the margin of safety.

U.S. Nuclear Regulatory Commission Attachment A 3F0607-05 Page 6 of 7 5.0 ENVIRONMENTAL IMPACT EVALUATION 10 CFR 51.22(c)(9) provides criteria for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if the amendment changes a requirement with respect to use of a facility component within the restricted area provided that (i) the amendment involves no significant hazards consideration, (ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (iii) there is no significant increase in individual or cumulative occupational radiation exposure.

Florida Power Corporation (FPC) has reviewed this License Amendment Request (LAR) and has determined that it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22, no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the proposed license amendment. The basis for this determination is that this amendment does not significantly change feedwater flow instrumentation located inside the restricted area and:

(i) The proposed license amendment does not involve a significant hazards consideration, as described in the significant hazards evaluation.

(ii) As discussed in the Justification for the Request and the No Significant Hazards Evaluation, this change does not result in a significant change or significant increase in the release associated with any Design Basis Accident. The bounding accident involved, the Loss of Coolant Accident, has release rates not significantly affected by the increase in core power. Likewise, there will be no significant change in the types or a significant increase in the amounts of any effluents released offsite during normal operation. The specific activity of the primary and secondary coolant is expected to increase by no more than the percentage increase in power level. Therefore, the amount and specific activity of solid waste is not expected to increase significantly.

Gaseous and liquid effluent releases are expected to increase from current values by no more than the percentage increase in power level. Offsite release concentrations and doses will continue to be maintained within the limits of 10 CFR 20 and 10 CFR 50, Appendix I, in accordance with the requirements of the CR-3 Offsite Dose Calculation Manual (ODCM). The ODCM contains methodologies and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, the methodologies and parameters used in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and controls for maintaining doses to the public from radioactive effluents As Low As Reasonably Achievable (ALARA). The proposed changes will not result in changes in the operation or design of the gaseous, liquid, or solid waste systems and will not create any new or different radiological release pathways.

Therefore, the proposed LAR will not result in a significant change in the types or increase in the amounts of any effluents that may be released offsite.

U.S. Nuclear Regulatory Commission Attachment A 3F0607-05 Page 7 of 7 (iii) The proposed LAR does not significantly increase core power and resultant dose rates in the Reactor Building and accessible areas of the plant. Individual worker exposures will be maintained within acceptable limits by the CR-3 ALARA Program. Therefore, the proposed LAR does not result in a significant increase to the individual or cumulative occupational radiation exposure.

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #296, REVISION 1 ATTACHMENT B Proposed Operating License and Technical Specification Pages Strikeout and Shadowed Text Format

Definitions 1.1 1.1 Definitions EFFECTIVE FULL POWER reactor core at RTP for one full day. (One EFPD is DAY (EFPD) 2568 ý-6Qq MWt times 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or 61632 MWhr.)

(continued)

EMERGENCY FEEDWATER The EFIC RESPONSE TIME shall be that time INITIATION AND CONTROL interval from when the monitored parameter (EFIC) RESPONSE TIME exceeds its EFIC actuation setpoint at the channel sensor until the emergency feedwater equipment is capable of performing its safety function (i.e.,

valves travel to their required positions, pump discharge pressures reach their required values, etc.) Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval FEATURE (ESF) RESPONSE from when the monitored parameter exceeds its ESF TIME actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing, that is captured and conducted to collection systems or a sump or collecting tank; or

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and quantified and known not to interfere with the operation of leakage detection systems and not to be pressure boundary LEAKAGE; or (continued)

Crystal River Unit 3 1.1-4 Amendment No. 2%

Definitions 1.1 1.1 Definitions PHYSICS TESTS These tests are:

(continued)

a. Described in Chapter 13, "Initial Tests and Operation" of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

PRESSURE AND The PTLR is the unit specific document that TEMPERATURE LIMITS provides the reactor vessel pressure and REPORT (PTLR) temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.2.19. Plant operation within these operating limits is addressed in LCO 3.4.3, "RCS Pressure and Temperature Limits."

QUADRANT POWER TILT QPT shall be defined by the following equation and (QPT) is expressed as a percentage.

QPT = 100 Power In Any Core Quadrant (Average Power of all Quadrants

-1)

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 2568 M MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until electrical power is interrupted at the control rod drive trip breakers. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or (continued)

Crystal River Unit 3 1.1-6 Amendment No. 2%

(7) Florida Power Corporation, pursuant to the Act and 10 CFR Parts 30 and 70, to receive and possess, but not separate, that by-product and special nuclear materials associated with four (4) fuel assemblies (B&W Identification Numbers 1A-01, 04, 05, and 36 which were previously irradiated in the Oconee Nuclear Station, Unit No. 1) acquired by Florida Power Corporation from Duke Power Company for use as reactor fuel in the facility.

C. This 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 or Part 50, 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:

2.C.(1) Maximum Power Level Florida Power Corporation is authorized to operate the facility at a steady state reactor core power level not in excess of 2568 9 Megawatts (100 percent of rated core power level).

2.C.(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 222, are hereby incorporated in the license. Florida Power Corporation shall operate the facility in accordance with the Technical Specifications.

The Surveillance Requirements contained in the Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment 149.

The Surveillance Requirements shall be successfully demonstrated prior to the time and condition specified below for each.

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #296, REVISION 1 ATTACHMENT C Proposed Operating License and Technical Specification Pages Revision Bar Format

Definitions 1.1 1.1 Definitions EFFECTIVE FULL POWER reactor core at RTP for one full day. (One EFPD is DAY (EFPD) 2609 MWt times 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or 62616 MWhr.)

(continued)

EMERGENCY FEEDWATER The EFIC RESPONSE TIME shall be that time INITIATION AND CONTROL interval from when the monitored parameter (EFIC) RESPONSE TIME exceeds its EFIC actuation setpoint at the channel sensor until the emergency feedwater equipment is capable of performing its safety function (i.e.,

valves travel to their required positions, pump discharge pressures reach their required values, etc.) Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval FEATURE (ESF) RESPONSE from when the monitored parameter exceeds its ESF TIME actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing, that is captured and conducted to collection systems or a sump or collecting tank; or

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and quantified and known not to interfere with the operation of leakage detection systems and not to be pressure boundary LEAKAGE; or (continued)

Crystal River Unit 3 1.1-4 Amendment No.

Definitions 1.1 1.1 Definitions PHYSICS TESTS These tests are:

(continued)

a. Described in Chapter 13, "Initial Tests and Operation" of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

PRESSURE AND The PTLR is the unit specific document that TEMPERATURE LIMITS provides the reactor vessel pressure and REPORT (PTLR) temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 5.6.2.19. Plant operation within these operating limits is addressed in LCO 3.4.3, "RCS Pressure and Temperature Limits."

QUADRANT POWER TILT QPT shall be defined by the following equation and (QPT) is expressed as a percentage.

QPT = 100 Power In Any Core Quadrant -1

ýAverage Power of all Quadrants )

RATED THERMAL POWER RTP shall be a total reactor core heat transfer (RTP) rate to the reactor coolant of 2609 MWt.

REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until electrical power is interrupted at the control rod drive trip breakers. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or (continued)

Crystal River Unit 3 1.1-6 Amendment No.

(7) Florida Power Corporation, pursuant to the Act and 10 CFR Parts 30 and 70, to receive and possess, but not separate, that by-product and special nuclear materials associated with four (4) fuel assemblies (B&W Identification Numbers 1A-01, 04, 05, and 36 which were previously irradiated in the Oconee Nuclear Station, Unit No. 1) acquired by Florida Power Corporation from Duke Power Company for use as reactor fuel in the facility.

C. This 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 or Part 50, 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:

2.C.(1) Maximum Power Level Florida Power Corporation is authorized to operate the facility at a steady state reactor core power level not in excess of 2609 I Megawatts (100 percent of rated core power level).

2.C.(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. , are hereby incorporated in the license. Florida Power Corporation shall operate the facility in accordance with the Technical Specifications.

The Surveillance Requirements contained in the Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment 149.

The Surveillance Requirements shall be successfully demonstrated prior to the time and condition specified below for each.

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #296, REVISION 1 ATTACHMENT D Response to RIS 2002-03 Questions

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 1 of 55 Response to RIS 2002-03 Questions 1.0 Feedwater Flow Measurement Technique and Power Measurement Uncertainty (RIS 2002-03 Section I Questions) 1.1 A detailed description of the plant-specific implementation of the feedwater flow measurement technique and the power increase gained as a result of implementing this technique. This description should include:

A. Identification (by document title, number, and date) of the approved topical report on the feedwater flow measurement technique B. A reference to the NRC's approval of the proposed feedwater flow measurement technique C. A discussion of the plant-specific implementation of the guidelines in the topical report and the staff's letter/safety evaluation approving the topical report for the feedwater flow measurement technique D. The dispositions of the criteria that the NRC staff stated should be addressed (i.e., the criteria included in the staff's approval of the technique) when implementing the feedwater flow measurement technique E. A calculation of the total power measurement uncertainty at the plant, explicitly identifying all parameters and their individual contribution to the power uncertainty F. Information to specifically address the following aspects of the calibration and maintenance procedures related to all instruments that affect the power calorimetric:

i. maintaining calibration ii. controlling software and hardware configuration iii. performing corrective actions iv. reporting deficiencies to the manufacturer

v. receiving and addressing manufacturer deficiency reports G. A proposed allowed outage time for the instrument, along with the technical basis for the time selected H. Proposed actions to reduce power level if the allowed outage time is exceeded, including a discussion of the technical basis for the proposed reduced power level.

1.2 Response to RIS 2002-03 Section I Questions Detailed description of the Crystal River Unit 3 (CR-3) Implementation of the Caldon LEFM CheckPlusTM Instrumentation and the 1.6% power increase.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 2 of 55 The feedwater flow measurement system installed at CR-3 is an LEFM CheckPlusTm ultrasonic, multi-path, transit time flowmeter. This equipment also provides a highly accurate feedwater temperature that will be input to the heat balance. The design of this advanced flow measurement system is addressed in detail by the manufacturer, Caldon, Inc., in Topical Reports ER-80P, Revision 0 (Reference 1.3.1), and ER-157P, Revision 5 (Reference 1.3.2). The current flow instruments will continue to measure main feedwater flow as well.

The LEFM ultrasonic flowmeter system consists of an electronic cabinet to be located in the Intermediate Building, Elevation 119 feet and two measurement section/spool pieces, also to be located in the Intermediate Building, in each of the two 18 inch main feedwater flow headers that feed each steam generator. The measurement sections are located between the existing feedwater flow nozzles and their respective upstream straightening vanes. The LEFM flowmeters will be calibrated at the Alden Research Laboratory, Inc. facility using the plant's current piping configuration and variations of the plant's configuration.

Each measurement section consists of sixteen (16) ultrasonic transducer housings, forming the pressure boundary. Each transducer may be removed at full-power conditions without disturbing the pressure boundary. The installation location of these flow elements conforms to the requirements in Topical Reports ER-80P and ER-157P (Reference 1.3.3 and 1.3.4).

The LEFM system measures the transit times of pulses of ultrasonic energy traveling along chordal acoustic paths through the flowing fluid. This technology provides significantly higher accuracy and reliability than the existing flow instruments, which use differential pressure measurements; and temperature instruments, which use conventional thermocouple or resistance thermometers. The sound will travel faster when the pulse traverses the pipe with the flow and will travel slower when the pulse traverses the pipe against the flow, due to the Doppler effect.

The LEFM uses these transit times and time differences between pulses to determine the fluid velocity. The LEFM also measures the speed of sound in water and uses this to determine the feedwater temperature.

The system's software employs the ultrasonic transit time method to measure the velocities at precise locations with respect to the pipe centerline. The system numerically integrates the measured velocities. The system's software has been developed and maintained under a verification and validation (V&V) program. The V&V Program has been applied to all system software and hardware, and includes a detailed code review. The mass flow rate and feedwater temperature are displayed on the electronic cabinet and transmitted via Ethernet to the automated unit load demand (AULD) and plant process computer (CP) for use in the calorimetric measurement of reactor thermal output based on an energy balance of the secondary system.

The improved accuracy of measurements of feedwater mass flow, pressure, and temperature as well as main steam temperature and pressure and updated instrument uncertainty calculations for other parameters results in a total uncertainty of less than 0.4 percent of reactor thermal power.

This is substantially more accurate than the nominal 2 percent rated thermal power (RTP) assumed in the accident analyses.

The LEFM indications of feedwater mass flow and temperature will be directly substituted for the existing feedwater flow instrumentation and the resistance temperature detector (RTD)

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 3 of 55 temperature inputs currently used in the plant calorimetric measurement calculations. The existing feedwater flow and RTD temperature will continue to be used for feedwater control and other functions that they currently fulfill.

The Caldon Panel has outputs for internally generated system trouble alarms, which will be wired into the plant process computer.

The AULD and the plant process computer (in the fixed incore detector monitoring system (FIDMS), each perform independent plant secondary heat balance calculations. The AULD heat balance is used in conjunction with the integrated control system to automatically control plant power at the operator selected Core Thermal Power (CTP) in megawatts thermal (MWt). The FIDMS heat balance is normally used by the plant operators to calibrate the nuclear instrumentation (NIs) and can be used by the plant operators to manually control reactor power upon loss of AULD. These two software routines are independent but receive identical inputs.

1.2.1 Caldon Topical Reports Applicable to the LEFM CheckPlusTM System (RIS 2002-03 Section 1.1.A)

The referenced Topical Reports are:

ER-80P, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM"/Tm System," Revision 0, dated March 1997 (Reference 1.3.1)

ER-157P, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM,/TM or LEFM CheckPlusTM System," Revision 5, dated October 2001 (Reference 1.3.2) 1.2.2 NRC Approval of Caldon LEFM CheckPlusTm System Topical Reports (RIS 2002-03 Section 1.1.B)

The NRC approved the subject Topical Reports referenced above on the following dates:

ER-80P, NRC SER dated March 8, 1999 (Reference 1.3.3)

ER-157P, NRC SER dated December 20, 2001 (Reference 1.3.4)

In addition, the NRC performed additional evaluations on the acceptability of the Caldon LEFMs and these are documented in, "Safety Evaluation by the Office of Nuclear Reactor Regulation The Hydraulic Aspects of the Caldon Leading Edge Flow Measurement (LEFM)

Check and CheckPlus"' Ultrasonic Flow Meters Caldon, Inc.," Project No. 1311, July 5, 2006 (TAC NO. MC6424) (Reference 1.3.9).

1.2.3 CR-3 Implementation of Guidelines and NRC SER for the Caldon LEFM CheckPlusTM System (RIS 2002-03 Section I.1.C)

The LEFM CheckPlusTM system will be installed at CR-3 in accordance with the requirements of Topical Reports ER-80P and ER-157P. This system will be used for continuous calorimetric power determination by serial link with the CP and incorporates self-verification features to ensure that hydraulic profile and signal processing requirements are met within its design basis uncertainty analysis.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 4 of 55 The CR-3 LEFM CheckPlus TM system will be calibrated in a site-specific model test at Alden Research Laboratories, with traceability to National Standards. The LEFM CheckPlusTM system will be installed and commissioned according to Caldon procedures, which include verification of ultrasonic signal quality and hydraulic velocity profiles as compared to those tested during site-specific model testing.

1.2.4 Disposition of NRC Criteria in the SER during Installation (RIS 2002-03 Section

.1.D)

In approving Caldon Topical Reports ER-80P and ER-157P, the NRC established four criteria to be addressed by each licensee. The four criteria and a discussion of how each will be satisfied for CR-3 follow:

Criterion 1 Discuss maintenance and calibration procedures that will be implemented with the incorporation of the LEEM, including processes and contingencies for unavailable LEFM instrumentation and the effect on thermal power measurements and plant operation.

Response to Criterion 1 Implementation of the power uprate license amendment will include developing the necessary procedures and documents required for operation, maintenance, calibration, testing, and training at the uprated power level with the new LEFM system. A preventative maintenance program will be developed for the LEFM using the vendor's maintenance and troubleshooting manual.

The preventative maintenance activities perform the following checks:

" General inspection of the terminal and cleanliness

" Power Supply inspection of magnitude and noise

Central Processing Unit inspection

Acoustic Processor Unit Checks of the 5 MHz clock and LED status

Analog Input checks of the A/D converter

Alarm Relay checks

Watchdog Timer checks that ensures the software is running

Transducer Cable checks

Calibration checks of each of the Feedwater pressure transmitters.

The preventative maintenance program and continuous monitoring of the LEFM ensures that the LEFM remains bounded by the analysis and assumptions set forth in the Topical Report ER-80P.

The incorporation of, and continued adherence to, these requirements will assure that the LEFM system is properly maintained and calibrated. Note that the LEFM provides both feedwater flow and temperature inputs to the core thermal power calculation.

Administrative controls will be implemented to provide guidance for plant control room operations staff in the event the LEFM system is unavailable. A requirement in plant compliance procedure CP-500, "Special Actions and Reporting Requirements," will state that if either LEFM or any low-uncertainty heat balance input parameters are inoperable, then reduce power to <

2568 MWt within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce the nuclear overpower - high setpoint to < 103.3% RTP

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 5 of 55 within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Logic will be programmed with the AULD to compare the improved calorimetric with the existing calorimetric and if the two deviate from one another by a pre-determined value, the AULD will be programmed to automatically transfer from Automatic to Manual in order to prevent any potential power excursions resulting from failed improved calorimetric sensors.

Administrative guidance will be developed to assist the Operator in determining whether to remain at 2609 MWt in AULD Automatic using the improved calorimetric or to reduce power to the previous rated thermal power of 2568 MWt (or lower) and transfer to AULD Automatic using the existing calorimetric. In addition to the above comparison logic, the AULD will also be programmed to detect an out-of-range condition for any of the calorimetric inputs. This condition will alert the Operator to investigate the validity of any suspect input.

These requirements ensure that an operable low uncertainty input shall be used whenever power is greater than the pre-uprate RTP level of 2568 MWt. With these requirements in place, the effect on plant operations is that power will be reduced and maintained to the pre-uprate level of 2568 MWt or lower, and that the existing flow nozzles and RTDs will be used for the calorimetric until the LEFM is returned to operable status. These requirements return the measurement techniques and maximum steady state power level to the currently licensed conditions.

Criterion 2 For plants that currently have LEFMs installed, provide an evaluation of the operational and maintenance history of the installed installation and confirmation that the installed instrumentation is representative of the LEFM system and bounds the analysis and assumptions set forth in Topical Report ER-80P.

Response to Criterion 2 Criterion 2 does not apply to CR-3.

Criterion 3 Confirm that the methodology used to calculate the uncertainty of the LEFM in comparison to the current feedwater instrumentation is based on accepted plant setpoint methodology (with regard to the development of instrument uncertainty). If an alternative approach is used, the application should be justified and applied to both Venturi and ultrasonic flow measurement instrumentation installations for comparison.

Response to Criterion 3 The LEFM uncertainty calculation is based on the ASME PTC 19.1 methodology (Reference 1.3.6) and Alden Research Laboratory Inc. calibration tests. This ASME PTC 19.1 methodology was reviewed by the NRC as part of the Seabrook MUR application and Safety Evaluation Report (Reference 1.3.7). The feedwater flow and temperature uncertainties are then combined with other plant measurement uncertainties (steam temperature, steam pressure, feedwater pressure) to calculate the overall heat balance uncertainty.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 6 of 55 This LEFM uncertainty calculation method is consistent with the current heat balance uncertainty calculation that uses the feedwater flow nozzles and feedwater RTDs. The current calculation is based on a square-root-sum-squares calculation, as described in ASME PTC 19.1.

FPC will provide the results of the Alden Research Laboratory calibration and testing to the Staff by September 1, 2007.

Criterion 4 For plants where the ultrasonic meter (including LEFM) was not installed and flow elements calibrated to a site-specific piping configuration (flow profiles and meter factors not representative of the plant specific installation), additional justification should be provided for its use. The 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 calibrations and plant configurations for the specific installation including the propagation of flow profile effects at higher Reynolds numbers. Additionally, for previously installed calibrated elements, confirm that the piping configuration remains bounding for the original LEFM installation and calibration assumptions.

Response to Criterion 4 A bounding uncertainty for the LEFM has been provided for use in the uncertainty calculation described below (Reference 1.3.8). The acceptability of the bounding calibration factor for the CR-3 spool pieces will be established by tests of these spools at Alden Research Laboratory.

These include tests of a full-scale model of the CR-3 hydraulic geometry and tests in a straight pipe. An Alden data report for these tests and a Caldon engineering report evaluating the test data will be prepared. The calibration factor used for the LEFM CheckPlus Tm at CR-3 will be verified as acceptable against these reports. The site-specific uncertainty analysis (Attachment E) documents these analyses and will be maintained as a CR-3 design basis calculation.

1.2.5 Total Power Measurement Uncertainty at CR-3 (RIS 2002-03 Section 1.1.E)

The total power uncertainty using the LEFM CheckPlus TM at CR-3 is 0.4%. This calculation is provided in AREVA NP Calculation 32-9042687-000 (Reference 1.3.5) and is included as Attachment E. The parameters, their uncertainty, and relative contributions are shown in Table D. 1-1. The ASME Performance Test Code Methodology was used to calculate the expected core thermal power uncertainty to be achieved using the Caldon CheckPlusTM System ultrasonic flow meter. The analysis concluded that using the following instrument uncertainty values, the core thermal power uncertainty would be 0.394% of 2609 MWt, thus allowing a power uprate of 1.6% to be pursued. The feedwater flow and temperature measurement is the bulk of this uncertainty (0.34% absolute and -84% of the total uncertainty). The new steam temperature/pressure instrumentation results in -4% of the total uncertainty, while the steam pressure measurement uncertainty is -1% of the total. The Reactor Coolant (RC) pump energy uncertainty, ambient loss uncertainty and an atmospheric pressure correction uncertainty were chosen to be treated as a bias (algebraically added and not square root sum of the squares (SRSS)) and they are -11% of the total uncertainty. After the final feedwater flow/temperature uncertainty is determined for the CR-3 specific equipment (post fabrication testing), the total uncertainty may be reduced.

Table D. 1-1 below summarizes the core thermal power measurement uncertainty.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 7 of 55 TABLE D. 1 HEAT BALANCE PARAMETER UNCERTAINTY CONTRIBUTIONS Uncertainty Contribution Symbol Description Absolute Relative Relative Percent (Btu/hr)2 (SRSS) (Total) of Total Power WFW/TFW Feedwater Flow/Temp 9.166E+14 94.2733% 83.7722% 0.3301%

TS Steam Temperature 4.432E+13 4.5582% 4.0504% 0.0160%

PS Steam Pressure 8.573E+12 0.8818% 0.7836% 0.0031%

PFW Feedwater Pressure 6.140E+08 0.0001% 0.0001% 0.0000%

TLD Letdown Temperature 2.147E+06 0.0000% 0.0000% 0.0000%

PLD Letdown Pressure 3.117E+10 0.0032% 0.0028% 0.0000%

WLD Letdown Flow Rate 2.726E+12 0.2804% 0.2492% 0.0010%

TMU Makeup Temperature 2.995E+10 0.0031% 0.0027% 0.0000%

TOTAL 9.723E+14 100%

Bias Corrections (Btu/hr)

QRCP RCP Power 9.215E+04 NA 2.6261% 0.0103%

QLOSS Ambient Heat Loss 2.560E+06 NA 7.2948% 0.0287%

PATMOS Pqaqe to Pabsolute 4.274E+05 NA 1.2181% 0.0048%

Totals 100% 0.394%

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 8 of 55 1.2.6 Calibration and Maintenance Procedures of All Instruments Affecting the Power Calorimetric (RIS 2002-03 Section 1.1.F)

Information to specifically address the following aspects of the calibration and maintenance procedures related to all instruments that affect the power calorimetric:

1.2.6.1 Maintaining Calibration Calibration of the LEFM will be ensured by preventative maintenance activities previously described in Section 1.2.4, Response to Criterion 1.

New instruments that contribute to the power calorimetric will be maintained according to required calibration and maintenance procedures. The other instruments that contribute to the power calorimetric were unaffected by the addition of the LEFM and will be maintained according to existing calibration and maintenance procedures.

1.2.6.2 Controlling Hardware and Software Configuration Hardware configuration will be controlled in accordance with Progress Energy procedures, including EGR-NGGC-0005, "Engineering Change," and EGR-NGGC-0012, "Equipment Data Base".

LEFM software will be properly classified in accordance with Progress Energy procedure CSP-NGGC-2507, "Software Documentation and Testing". AULD software will be classified, developed, tested, and controlled in accordance with EGR-NGGC-0157, "Engineering of Plant Digital Systems and Components". Implementation of the AULD software will be performed under the design control process governed by EGR-NGGC-0005, "Engineering Change".

Software control will be in accordance with EGR-NGGC-0157, "Engineering of Plant Digital Systems and Components". Software control will be in accordance with EGR-NGGC-0157, "Engineering of Plant Digital Systems and Components".

Instruments that affect the power calorimetric, including the Caldon LEFM CheckPlusTM System inputs, are monitored by CR-3 personnel. Equipment problems for plant systems, including the Caldon LEFM CheckPlusTM System equipment, fall under site work control processes.

Conditions that are adverse to quality are documented under the corrective action program.

Corrective action procedures, which ensure compliance with the requirements of 10 CFR 50, Appendix B, include instructions for notification of deficiencies and error reporting.

1.2.6.3 Performing Corrective Actions Corrective actions will be monitored and performed in accordance with Progress Energy procedures CAP-NGGC-0200, "Corrective Action Program," and ADM-NGGC-0104, "Work Management Process".

1.2.6.4 Reporting Deficiencies to the Manufacturer Reporting deficiencies to the manufacturer will be performed in accordance with Progress Energy procedure CAP-NGGC-0200, "Corrective Action Program".

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 9 of 55 1.2.6.5 Receiving and Addressing Manufacturer Deficiency Reports Manufacturer deficiency reports will be received and addressed in accordance with Progress Energy procedure REG-NGGC-0013, "Evaluating and Reporting of Defects and Noncompliance in Accordance with 10 CFR 21".

1.2.7 Allowed Outage Time and Technical Basis (RIS 2002-03 Section I.I.G)

The Nuclear Instrumentation (NI) indicated power is compared against heat balance power on a daily basis. In the event that the LEFM or other low uncertainty heat balance input parameters becomes unavailable, it must be restored to operable status or the plant power will be reduced to 98.4% RTP (< 2568 MWt) within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (see Item 1.2.8 below). The justification for the allowed outage time of the LEFM is that the NIs were compared to the last known good heat balance calculation using the LEFM measurement, do not routinely require adjustments and thus can continue to be relied upon for power measurement until the next daily comparison. The time period is reasonable based on the functions and the capability of performing a calorimetric calculation without using the LEFM.

1.2.8 Actions for Exceeding Allowed Outage Time and Technical Basis (RIS Section 1.1.H)

Administrative controls will be placed in CR-3 procedure CP-500 to address LEFM or any low-uncertainty heat balance input parameter unavailability. Should the LEFM system become unavailable, the current flow nozzle-based feedwater flow and RTD feedwater temperature instrumentation will be used as input to the core power calorimetric, and the core power will be limited to the current licensed power level of 2568 MWt. The reactor operators will be provided with procedural guidance for those occasions when the LEFM CheckPlusTM or any low-uncertainty heat balance input parameter is not available.

1.3 SECTION

1.0 REFERENCES

1.3.1 ER-80P, Revision 0, "Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFM,"TM System," Caldon, Inc., dated March 1997.

1.3.2 ER-157P, Revision 5, "Supplement to Topical Report ER-80P: Basis for a Power Uprate with the LEFM,/VT or LEFM CheckPlusTM System," Caldon, Inc., dated October 2001.

1.3.3 Letter from Project Directorate IV-1, Division of Licensing Project Management, Office of Nuclear Reactor Regulation, to C.L. Terry, TU Electric, Comanche Peak Steam Electric Station, Units 1 and 2 - Review of Caldon Engineering Topical Report ER 80P, Improving Thermal Power Accuracy and Plant Safety while Increasing Power Level Using the LEFM System (TAC Nos. MA2298 and 2299), dated March 8, 1999.

1.3.4 Letter from S. A. Richards, NRC, to M. A. Krupa, Entergy, Waterford Steam Electric Station, Unit 3; River Bend Station; and Grand Gulf Nuclear Station - Review of Caldon, Inc. Engineering Report ER-157P (TAC Nos. MB2397, MB 2399 and MB2468), dated December 20, 2001.

1.3.5 32-9042687-000, "CR-3 Heat Balance Uncertainty for CR-3 MUR."

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 10 of 55 1.3.6 ASME PTC 19.1-1998, Test Uncertainty, Instruments and Apparatus, American Society of Mechanical Engineers, NY, NY, 1998.

1.3.7 NRC Letter, G. Edward Miller to Gene F. St. Pierre,

Subject:

Seabrook Station, Unit No.

I - Issuance of Amendment RE: Measurement Uncertainty Recapture Power Uprate (TAC No. MC8434), May 2006.

1.3.8 ER-579 Rev. 0, "Bounding Uncertainty Analysis for Thermal Power Determination at Crystal River Unit 3 Using the LEFM,/+ System," dated February 2007.

1.3.9 NRC Letter, B. E. Thomas to E. M. Hauser, "Evaluation of the Hydraulic Aspects of the Caldon Leading Edge Flow Measurement (LEFM) Check AND CheckPluSTM Ultrasonic Flow Meters (UFMs)" (TAC NO. MC6424) July 5, 2006.

2.0 Accidents and Transients for Which the Existing Analyses of Record Bound Plant Operation at the Proposed Uprated Power Level (RIS 2002-03 Section II Questions) 2.1 A matrix that includes information for each analysis in this category and addresses the transients and accidents included in the plant's updated final safety analysis report (UFSAR) (typically Chapter 14 or 15) and other analyses that licensees are required to perform to support licensing of their plants (i.e., radiological consequences, natural circulation cooldown, containment performance, anticipated transient without scram, station blackout, analyses to determine environmental qualification parameters, safe shutdown fire analysis, spent fuel pool cooling, flooding):

A. Identify the transient or accident that is the subject of the analysis B. Confirm and explicitly state that

i. the requested uprate in power level continues to be bounded by the existing analyses of record for the plant ii. the analyses of record either have been previously approved by the NRC or were conducted using methods or processes that were previously approved by the NRC iii. the analyses of record are not changed by the requested power uprate C. Confirm that bounding event determinations continue to be valid D. Provide a reference to the NRC's previous approvals discussed in Item B. above.

2.2 Response to RIS 2002-03 Section II Questions In order to support the CR-3 MUR Power Level Uprate, with respect to the accident analyses, a review of the Final Safety Analysis Report (FSAR), Reference 4.3.1, Chapters 6 and 14 and other related sub-sections was performed. Evaluations were performed on other analyses as well and it was determined there was no impact from the MUR. The purpose of the review was to confirm that the analysis results, as currently presented in the FSAR, were performed conservatively and bound the proposed power uprate. All of the analyses that are included in the FSAR have been performed using NRC-approved tools and methods. If any event was determined to be not bounded by the current FSAR analyses, then a new analysis was performed as discussed in Section 3.0.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 11 of 55 For the Babcock and Wilcox (B&W)-designed plants, the heat balance uncertainty is accounted for in the initial core power level that is modeled in the accident analyses and included in the determination of the nuclear overpower reactor trip setpoint. For the implementation of the MUR, the intent is to use a higher accuracy feedwater flow measurement device to reduce the secondary side heat balance uncertainty from 2% to 0.4%. The reduced uncertainty would then be used to increase the rated thermal power level of the plant from 2568 MWt to 2609 MWt.

Ideally, no new accident analyses would be required as long as the new error-adjusted power level does not exceed what was modeled in the analyses (i.e., 102% of 2568 MWt).

As stated above, the heat balance uncertainty is also used to determine the overpower reactor trip setpoint specified in the technical specifications for Crystal River Unit 3. The current overpower reactor trip setpoint modeled in the accident analyses is 112% of 2568 MWt (or -2876 MWt).

The methodology used to derive the technical specification trip setpoint is described in Section 7 of BAW-10179P-A, Reference 2.3.5. The uncertainty associated with power measurement, including the secondary side heat balance uncertainty, is applied to the analysis value to derive the current plant technical specification setpoint allowable value of 104.9% of 2568 MWt (or

-2694 MWt). Therefore, in order to ensure that the over power limit modeled in the accident analyses, in terms of absolute megawatts, is preserved, the overpower analysis setpoint must be reduced proportionally to 110.2% (or 2876 MWt / 2609 MWt

100%). When the power measurement uncertainties with the reduced heat balance uncertainty are applied to 110.2%

analysis value, the technical specification allowable value will be 104.9% of 2609 MWt.

Although the allowable value will remain at the same percentage, the actual reactor trip will occur at a higher absolute power (104.9% of 2609 MWt rather than 104.9% of 2568 MWt). The analysis determined that the 104.9% allowable value still supports the assumption of a reactor trip prior to 2876 MWt. This allowable value remains valid as long as the higher accuracy feedwater flow measurement system is operable.

Another way of looking at this is the overpower reactor trip limit of 2876 MWt is not changing.

At the current licensed power limit of 2568 MWt, this is equal to 112% full power. At the proposed power limit of 2609 MWt, this equals 110.2% full power. The overpower trip limit is the accident analysis value that protects against exceeding the departure from nucleate boiling limits.

Working backwards from the overpower limit, with the current licensed power limit, there is a 4% bias and a 0.7% bistable error, as well as the current measurement uncertainty limit of 2%

which brings the value down to 105.3%. The current plant Technical Specifications Nuclear Overpower Trip Setpoint is 104.9% or 2694 MWt. This setpoint was selected to retain some analytical margin for future use. Working backward from the overpower limit with the proposed power limit of 2609 MWt, there is a 4% bias, a 0.7% bistable error, and 0.4% instrument uncertainty. This brings the value down for the proposed power limit to 105.1% of 2609 MWt.

The setpoint of 104.9% of 2609 MWt (2737 MWt) was selected because it remains conservative and eliminates the need to change the plant Technical Specifications or plant implementing documents. Therefore, the proposed power limit of 2609 MWt will use the same Nuclear Overpower High Setpoint (104.9%) as currently used. Even though the actual MWt value for the reactor trip is greater than the current setpoint, the reduced instrumentation uncertainty will provide assurance that the overpower trip limit of 2876 MWt will not be exceeded. This allowable value remains valid as long as the higher accuracy feedwater measurement system is operable. See Table D. 2-2 for a graphical representation.

If the higher accuracy feedwater flow measurement system becomes inoperable, the secondary

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 12 of 55 heat balance uncertainty will return to the 2% value associated with the main feedwater flow nozzles. Accordingly, the plant power level and reactor overpower trip setpoints in terms of absolute megawatts must be returned to the pre-MUR values. That is, core power must be reduced to 2568 MWt and the Nuclear Overpower High Flux trip setpoint will be reduced to 103.3% of RTP (2694 MWt, 104.9% of 2568 MWt.).

A summary of each accident is provided below and is summarized in Table 2-1 (All information is taken from Reference 2.3.2).

2.2.1 Uncompensated Operating Reactivity Changes (FSAR Section 14.1.2.1)

During normal operation of the reactor, the overall reactivity of the core changes because of fuel depletion, burnable poison depletion, and changes in fission product poison concentration.

These reactivity changes, if left uncompensated, can cause the operating limits to be exceeded.

In all cases, the Reactor Protection System (RPS) setpoints are placed to prevent the safety limits from being exceeded. No damage occurs from these conditions.

There are two acceptance criteria for this accident. First, the rate of reactivity will be much less than the rate at which the operator can compensate for the addition. Second, the rate of temperature change will be much less than the rate at which the automatic control system can compensate for the change.

The plant and control system response to reactivity changes resulting from fuel depletion, burnable poison depletion, and changes in fission product poison concentration are not significantly affected by the initial core power level. As a result, the change in the magnitude of reactivity changes caused by fuel depletion, burnable poison depletions, and/or changes in fission product poison concentration will be negligible. The analysis was initiated at 2575 MWt and is insensitive to initial core power. An increase in the analyzed power to 102% of 2568 MWt will not result in any appreciable change in the accident as previously analyzed.

The analysis of record for this accident was accepted by the NRC as part of the approval of the original CR-3 FSAR. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205. The individual accidents are discussed below.

2.2.2 Startup Accident (FSAR Section 14.1.2.2)

The startup accident is a moderate frequency event that results from a spurious control rod withdrawal from hot zero power conditions. The acceptance criteria for the event are that the peak reactor coolant system (RCS) pressure does not exceed 2750 psig, and the maximum allowed core power does not exceed 112% of rated thermal power. Therefore, the primary reactor protection system (RPS) trip functions that are credited for this event are the high RCS pressure and core over-power. This reactivity addition event is considered a heat-up transient that results in the pressurization of the RCS. The startup accident is the limiting event in ensuring overpressure protection of the RCS as outlined in the FSAR. The transient is initiated from hot zero power conditions and as a result, the MvUR power uprate has no effect on the initial conditions within the RCS.

The startup accident credits the reactor trip on high neutron flux. In analytical space, the high neutron flux set point is presently defined as 112% of 2568 MWt (2876.16 MWt). For MUR

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 13 of 55 conditions, the absolute power of 2876.16 MWt will remain the analytical limit for the high neutron flux setpoint. The setpoint expressed as a percent of the rated power condition will be reduced to 110.2% of rated power at MUR conditions. Using the same absolute power for the setpoint ensures the same protection at MUR conditions that currently exist for the rated power condition at 2568 MWt.

The analysis of record for this accident is reflected in the CR-3 FSAR, and remains acceptable for the MUR power uprate. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

2.2.3 Rod Withdrawal at Rated Power Operation Accident (FSAR Section 14.1.2.3)

The rod withdrawal accident is a moderate frequency event that results from a spurious control rod withdrawal from rated power conditions. The acceptance criteria for the event are that the peak RCS pressure does not exceed 2750 psig and the maximum allowed core power does not exceed 112% of rated power. Therefore, the primary RPS trip functions that are credited for this event are the high RCS pressure and core over-power. This reactivity addition event is considered a heat-up transient that results in pressurization of the RCS.

The initial core power level for the current rod withdrawal at power accident analysis is 100% of 2568 MWt. With the MUR power uprate, the RCS average temperature and initial pressurizer level will not change. The steam space in the pressurizer will also not be affected.

A spectrum of reactivity insertion rates (RIRs) is simulated to demonstrate compliance with the event acceptance criteria. For slow RIRs, the neutron and thermal power increase at nearly the same rate. The RCS temperature, and hence reactor pressure, increases rapidly to the high pressure trip setpoint. A different RIR will become limiting, but the MUR power uprate will still be bounded by the current plant FSAR analysis from the peak pressure perspective.

For fast RIRs, reactor protection is provided by the over-power trip setpoint. The transient response will be governed by the power difference between the initial core power and the over-power trip setpoint. The larger the difference between these values will result in a more severe transient.

In the current FSAR analysis, the power difference (or the net energy added) is 114% of 2568 MWt (Reference 2.3.1, Section 14.1.2.3.2). As discussed in the FSAR Section 14.1.2.3.2, the over-power trip setpoint was reduced to the current analytical setpoint of 112% of 2568 MWt due to fuel densification issues. This change took place after the original FSAR analysis and no new analyses were performed.

In order to provide the same over-power protection under MUR conditions, the over-power setpoint used in the analysis will be reduced to 110.2% of the MUR power level of 2609 MWt, or approximately 2876 MWt, which is the same net value as 112% of 2568 MWt. Since the initial core power level would be higher with the MUR, the net energy added to the RCS before the reactor trip setpoint is reached would be less than as in the current FSAR analysis.

Therefore, the rod withdrawal at power accident analyses described in the FSAR will remain applicable for the MUR power uprate from the peak power perspective.

The analyses of record for this accident are reflected in the CR-3 FSAR and remain acceptable

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 14 of 55 for the MIUR power uprate. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

2.2.4 Moderator Dilution Accident (FSAR Section 14.1.2.4)

The moderator dilution accident (MDA) is a moderate frequency event and results from an uncontrolled dilution of the primary coolant. The dilution of the moderator will result in a positive reactivity addition to the core and a corresponding heatup and pressurization of the RCS.

The acceptance criteria for this accident relate to peak RCS pressure, maximum allowed power, and minimum subcritical margin.

The transient progression is determined by the combinations of the dilution flow rate and the cycle-specific reactivity parameters. Conservative reactivity parameters and dilution flow rates are modeled to ensure a bounding analysis. These cycle-specific parameters are validated during each reload analysis to ensure the bounding analyses remain conservative. The analysis was performed at 102% of 2568 MWt, and complies with the acceptance criteria that peak power not exceed 112% of rated thermal power and peak RCS pressure not exceed 110% of design pressure.

Also for the dilution event, a minimum shutdown margin must be maintained during refueling conditions. Compliance to the shutdown margin requirement is demonstrated as part of the cycle-specific reload calculations because no system level transient is simulated and the results are largely unaffected by the MUR power uprate.

The CR-3 FSAR discusses an unterminated dilution event through the decay heat removal system. A plant modification was performed to prevent the possibility of dilution by sodium hydroxide (NaOH) addition. Therefore, this event is no longer possible.

The moderator dilution accident credits the reactor trip on high neutron flux. In analytical space, the high neutron flux setpoint is presently defined as 112% of 2568 MWt (2876.16 MWt). For MUR conditions, the absolute power of 2876.16 MWt will remain the analytical limit for the high neutron flux setpoint. The setpoint expressed as a percent of the rated power condition will be updated to 110.2% of rated thermal power at MUR conditions. Using the same absolute power for the setpoint ensures the same protection at MIUR conditions that currently exist for the rated power condition at 2568 MWt.

The analysis of record for this accident is reflected in the CR-3 FSAR and remains acceptable for the MUR power uprate. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

2.2.5 Cold Water Accident (FSAR Section 14.1.2.5)

This transient results from the startup of an idle loop while the plant is operating at reduced power. The cold water accident (CWA) is a moderate frequency event. The acceptance criteria for the event are that the peak RCS pressure does not exceed 2750 psig and the maximum departure from nucleate boiling ratio (MDNBR) does not decrease below 1.30.

The analysis assumed that the plant was operating with one reactor coolant pump in each loop at 50% of rated power when the remaining two pumps were started. The increase in primary

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 15 of 55 coolant flow and negative reactivity coefficients results in a positive reactivity insertion and subsequent increase in core power. The increase in core power limits the primary coolant temperature decrease and the plant reaches equilibrium at a new power level of approximately 65% which is still less than the rated power. No RPS trip setpoints are challenged. The increase in coolant flow combined with an increase in power to 65% (thermal) does not result in an unacceptable minimum DNBR. The RCS pressure increases approximately 137 psi and remains well below the high pressure reactor trip setpoint. The MUR will not impact the results of this analysis.

The analysis of record for this accident is reflected in the CR-3 FSAR and remains acceptable for the MUR power uprate. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

2.2.6 Loss-of-Coolant-Flow Accident (FSAR Section 14.1.2.6)

The loss of coolant flow (LOCF) accidents result from either loss of power or mechanical failure of one or more of the RCPs. The LOCF accidents are comprised of three different transients.

The simultaneous coastdown of all four RCPs is considered an infrequent event. The single locked pump rotor is considered a limiting fault transient. Although the four pump coastdown is considered an infrequent event, it is typically analyzed to the more restrictive criteria of the moderate frequency event category. For the locked rotor transient, no fuel cladding failure is allowed. These events are evaluated for each new fuel reload. The acceptance criteria for these events relate to the minimum allowed DNBR based on the applicable critical heat flux correlation for the fuel design being analyzed. These events were analyzed at 102% of 2568 MWt and include a 2% power measurement uncertainty in the calculations. In addition, the DNBR calculations are verified for each new core design.

The analysis of record for this accident is reflected in the CR-3 FSAR and remains acceptable for the MUR power uprate. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

2.2.7 Stuck-Out, Stuck-In, or Dropped Control Rod Accident (FSAR Section 14.1.2.7)

The dropped rod accident is the limiting event in the group of transients identified with misaligned control rods. A misaligned Control Rod Assembly (CRA) is defined as the deviation of a CRA from its group reference position by more than nine inches (indicated). This definition encompasses both the action of having a single CRA stick while moving its associated group or dropping a single CRA. With respect to stuck CRAs, core design requirements ensure that a Shutdown Margin (SDM) of a least 1.0% AK/K exists with the greatest worth CRA full withdrawn from the core. On the other hand, should a CRA stick while pulling its associated group, control systems will function to sound an alarm, inhibiting all CRAs out-movement. The consequences of stuck CRA accidents are therefore limited in severity because of the restrictions associated with rod movement and core design. Thus, the dropped CRA accident, which has restrictions for rod insertion, is the limiting CRA misalignment event.

The dropped control rod accident is a moderate frequency event and the acceptance criteria for this event relate to peak RCS pressure and MDNBR.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 16 of 55 The FSAR analysis of record is based on a core power level of 2772 MWt and a core design with steady-state peaking factors allowed by implementation of the statistical core design. The power level bounds the MUR. A cycle specific DNBR evaluation is addressed in the maneuvering analysis during the standard reload process.

2.2.8 Load Rejection Accident (Turbine Trip) (FSAR Section 14.1.2.8)

The plant was originally design to withstand the effects of a load rejection transient without a reactor or turbine trip. The reactor power would automatically be runback to the power level corresponding to the steam generator low level limit. The power operated relief valve (PORV) was available to relieve pressure to prevent a reactor trip. The acceptance criteria for this event are that fuel damage would not occur and that the RCS pressure would not exceed the core pressure limit of 110% of the design pressure. Fuel is not expected to fail during the load rejection analysis, and therefore the dose consequences are bounded by the Main Steam Line failure accident. The Analysis of Record (AOR) for the original load rejection accident is discussed in the FSAR.

The current plant response to a load rejection is different than the description presented above because the PORV lift setpoint has been raised above the high reactor coolant pressure reactor trip setpoint. A load rejection from 100% power with the higher PORV lift setpoint would result in a reactor trip on high reactor coolant pressure. The plant response to a load rejection under this configuration is similar to a turbine trip, but is less severe because the closure of the turbine stop valves during a turbine trip causes a more rapid pressurization. It is noted that the Anticipatory Reactor Trip System (ARTS) is not credited in the turbine trip analysis of record.

The turbine trip accident from full power bounds the load rejection accident. This analysis was evaluated over a range of power levels up to 112% of 2568 MWt. The analysis concluded that a 3% maximum tolerance for the main steam safety valves (MSSV) for one inoperable MSSV on each once-through steam generator (OTSG) was sufficient for power levels up to 112% of 2568 MWt, which is equivalent to the maximum allowed power for the MUR (110% of 2609 MWt).

The analyses of record for this accident are reflected in the CR-3 FSAR and remain acceptable for the MUR power uprate. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

2.2.9 Station Blackout Accident (FSAR Section 14.1.2.9)

The original Station Blackout Accident (SBO) analysis was evaluated as a Loss of AC Power (LOAC) event. This analysis was performed to show that the plant would transition to a stable condition in which decay heat would be removed by the steam generators via natural circulation.

During this event, the loss of AC power was assumed to initiate a reactor trip, RC pump trip, a turbine trip and close the turbine stop valves (TSV). As a result, the secondary side pressure will increase to the main steam safety valve setting which limits the secondary heat removal capacity.

This causes an initial reactor coolant heatup and pressure increase. This analysis is historical and has been superseded by a calculation that was prepared in response to the recommendations of the Nuclear Management and Resources Council (NUMARC) to determine the capability of a nuclear plant to cope and to recover from a SBO event for four hours.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 17 of 55 The acceptance criteria for a SBO event dictate that fuel damage shall not occur, the reactor coolant system shall not exceed core pressure limits, and the accident doses shall be within the 10 CFR 50.67 limits. The original LOAC power event, evaluated at 100% of 2568 MWt, is discussed in the FSAR. The NUMARC analyses are documented in the FSAR and were evaluated at 100% of 2772 MWt and therefore bound the MIUR power uprate conditions.

2.2.10 Steam Line Failure Accident (FSAR Section 14.2.2.1)

The steam line break is a rupture in the steam lines between the steam generators and the turbine.

The rapid depressurization causes an increase in the main feedwater flow rate. The increase in steam flow to the break and the turbine results in a large overcooling of RCS. The steam line break accident is the most severe overcooling transient. The acceptance criteria relates to effective core cooling, offsite dose release, reactor coolant system integrity, and containment vessel integrity.

For the core response, as documented in the FSAR, the core power was evaluated at 100% of 2568 MWt, to minimize the heat input to the reactor coolant system. The heat balance uncertainty of 2% was accounted for in the steam generator mass inventories. Therefore, the FSAR analysis bounds the MUR power uprate for the core and reactor coolant system. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

The dose release calculations are evaluated on a reload basis at a power level of 102% of 2568 MWt.

For the containment response, the steam line break event has been evaluated at 102% of 2568 MWt. The analyses of record for this accident are reflected in the CR-3 FSAR and remain acceptable for the MUR power uprate.

2.2.11 Steam Generator Tube Rupture Accident (FSAR Section 14.2.2.2)

The Steam Generator Tube Rupture (SGTR) is a postulated double-ended rupture of a steam generator tube with unrestricted discharge from both ends of the tube. The acceptance criteria are related to offsite dose and further degradation of the primary-to secondary boundary beyond the affected tube.

The SGTR is a breach of the reactor coolant pressure boundary and results in a transfer of primary coolant to the secondary system. The core protection aspects of a SGTR are bounded by a small break LOCA. Therefore, the SGTR event is analyzed to determine the offsite doses resulting from the release of contaminated primary coolant into the steam generator and to the atmosphere via the main steam safety valves.

The system response for the SGTR analysis of record is based on a constant leak rate. The leak flow rate is based on critical flow from each end of the ruptured tube. The leak rate was assumed to be constant until the plant was cooled down to the decay heat removal cut-in temperature.

This is conservative because it does not credit the decrease in the leakage rate with RCS depressurization or the secondary side pressurization following the reactor trip and turbine trip.

The SGTR calculation is independent of power level based on the analytical method used.

Therefore, there is no impact on the system response due to uprate.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 18 of 55 The acceptance criteria for the evaluation of this accident are public radiological doses must not exceed the allowable limits prescribed by 10 CFR 50.67 and Regulatory Guide 1.183 (2.5 rem Total Effective Dose Equivalent (TEDE) for a coincident iodine spike). Additionally, the event must not result in additional tube failures and further degradation of the integrity of the reactor coolant pressure boundary caused by the effects of temperature gradients.

The analyses of record for this accident was evaluated at a power level of 102% of 2568 MWt as reflected in the CR-3 FSAR, and remains acceptable for the MUR power uprate. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

2.2.12 Fuel Handling Accident (FSAR Section 14.2.2.3)

Mechanical damage to a fuel assembly is postulated during refueling operations. The analyses for this accident consider an accident inside containment and outside containment. The core power level is used to determine the activity levels in the fuel-to-clad region prior to the accident.

The acceptance criteria for the Fuel Handling Accident are based on the requirements of 10 CFR 50.67 and Regulatory Guide 1.183. The analyses of record for this accident was evaluated at a power level of 102% of 2568 MWt as reflected in the CR-3 FSAR, and remains acceptable for the MUR power uprate. This analysis was also reviewed and accepted by the Staff during the review for Amendment.205.

2.2.13 Rod Ejection Accident (FSAR Section 14.2.2.4)

The rod ejection event is a postulated event involving a physical failure of a pressure barrier component in the Control Rod Drive assembly and subsequent ejection of the control rod. The event is classified as an infrequent event. The acceptance criteria for the Rod Ejection from full power event relate to peak RCS pressure and peak fuel enthalpy.

The ejection of a control rod with the reactor at full power causes a rapid positive reactivity insertion. Core power and fuel temperatures increase rapidly. The rapid fuel temperature rise produces negative Doppler reactivity feedback that terminates the power excursion. A reactor trip occurs on over-power and the reactor is returned subcritical by control rod insertion. The primary safety valves provide steam relief to limit the peak RCS pressure to less than the acceptance criterion. Limiting the reactivity worth of a given rod in the fuel design and the initial fuel enthalpy at full power will ensure that the peak fuel enthalpy does not exceed the maximum allowable limit.

At hot zero power conditions, a rod ejection accident initiated from zero power is not directly impacted by the MUR power uprate.

The rod ejection at hot full power conditions were originally evaluated at 100% of 2568 MWt as documented in the FSAR. The neutron power response during a control rod ejection accident is not sensitive to the initial power conditions. Due to the rapid ejection time of 0.15 seconds, the transient is defined by the ejected rod worth and the kinetics parameters. The MUR will cause a slight increase in the fuel heat up of approximately 2 calories per gram (cal/g). There is approximately 80 cal/g margin to the fuel enthalpy limit for an ejected rod worth of 0.7%AK/K.

The reload core design ensures that maximum ejected rod worth will not exceed 0.65%AK/K

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 19 of 55 including a 15% uncertainty. Therefore, the current fuel enthalpy margin more than compensates for the small power increase.

The radiological analyses for the rod ejection accident assumed that the fuel gap activity for 14%

of the fuel rods is completely released. The dose release calculations are evaluated on a reload basis at a power level of 102% of 2568 MWt. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

2.2.14 Loss of Coolant Accident (FSAR Section 14.2.2.5)

A spectrum of break sizes and break locations is postulated in the primary coolant piping. The LOCAs are considered limiting fault transients, events that are not expected to occur, but are postulated because of the potential for large releases of radiation. The acceptance criteria relate to ensuring adequate core cooling for the short and long term post-LOCA, containment vessel pressure and temperature, and offsite dose consequences.

For compliance to adequate core cooling, the large and small break loss of coolant accident analyses were evaluated at 102% of 2568 MWt as documented in the FSAR.

For compliance to offsite dose consequences, the loss of coolant accidents were evaluated at radioactive nuclide inventories consistent with 102% of 2568 MWt. These analyses address the maximum hypothetical accident discussed in Section 14.2.2.7 of the FSAR.

In addition, post-LOCA boron control management analyses were performed as discussed in the FSAR. These analyses were evaluated at 102% of 2568 MWt. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

For compliance with containment integrity, the mass and energy releases were performed at 102% of 2568 MWt. The analyses of record for this accident are reflected in the CR-3 FSAR and remain acceptable for the MIUR power uprate.

2.2.15 Makeup System Letdown Line Failure Accident (FSAR Section 14.2.2.6)

Regulatory Guide 1.70, Table 15-1, indicates breaks in lines connected to the reactor coolant system that carry reactor coolant outside containment should be evaluated for dose consequences. The most severe piping rupture for which radioactivity release is postulated during normal plant operation is in the letdown line of the Makeup and Purification System. The acceptance criteria for this accident are described in 10 CFR 50.67.

The reactor is operating at 102% of 2568 MWt. The rupture is modeled as a complete severance of the 21/2 inch nominal diameter letdown line at a location downstream of the outboard isolation valve. A single emergency diesel generator is assumed to fail, and no credit is taken for the operators to increase the steam generator levels. Operators are assumed to isolate the letdown line at 10 minutes after the hot leg reaches saturated conditions.

The analyses of record for this accident was evaluated at a power level of 102% of 2568 MWt as reflected in the CR-3 FSAR and remains acceptable for the MUR power uprate. Radiological consequences are assessed during the standard reload process. This analysis was also reviewed

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 20 of 55 and accepted by the Staff during the review for Amendment 205.

2.2.16 Waste Gas Decay Tank Rupture Accident (FSAR Section 14.2.2.8)

The waste gas decay tank is used in the radioactive waste disposal system to store radioactive gaseous waste from the station until such time that the radioactive decay renders the gas safe for release to the site environment. Rupture of a waste gas tank would result in the premature release of its radioactive contents to the station ventilation system and to the atmosphere through the station vent.

The acceptance criteria for the Waste Gas Decay Tank Rupture Accident (WGDTRA) are based on the requirements of 10 CFR 50.67 and Regulatory Guide 1.183. The analysis of record for this accident is reflected in the CR-3 FSAR. The analysis conservatively assumes that all three available waste gas decay tanks rupture. Each tank is assumed to contain the maximum curie inventory allowed by the Offsite Dose Calculation Manual. The WGDTRA would release more radioactivity to the atmosphere than any other credible radwaste system accident. The dose assessment for the WGDTRA is based on WGDT inventories of radioactive nuclides and are independent of power level. Therefore, the WGDTRA analysis is not affected by operation of CR-3 at 2609 MWt.

2.2.17 Loss of Feedwater and Main Feedwater Line Break Accident (FSAR Section 14.2.2.9)

A loss of feedwater accident results from either a reduction in or the complete loss of secondary feedwater to the steam generators. The loss of feedwater may be caused by pump failure, valve closure, or a feedwater line break. The acceptance criteria are that fuel failure shall not occur, the peak RCS pressure will not exceed code pressure limits of 110% of the design pressure, and offsite dose consequences remain less than the limits specified in 10 CFR 50.67. The loss of feedwater and feedwater line break accidents were evaluated at 102% of 2568 MWt. The loss of feedwater accident is also used to establish the minimum required emergency feedwater (EFW) flow rate of 550 gpm. The feedwater line break accident is considered a limiting fault event.

However, the analysis is analyzed with an imposed minimum DNBR limit to prevent fuel failures.

The analyses of record for these accidents are reflected in the CR-3 FSAR and remain acceptable for the MUR power uprate. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

2.2.18 ATWS Transients (DSS, AMSAC) (FSAR Section 7.5)

The Anticipated Transients Without Scram (ATWS) events are evaluated in compliance to 10 CFR 50.62. An ATWS event is an anticipated operational occurrence followed by the failure of the reactor trip portion of the reactor protection system.

For compliance to 10 CFR 50.62 criteria, CR-3 has installed a Diverse Scram System (DSS) and an ATWS Mitigating System Actuating Circuitry (AMSAC) system. DSS provides an interruption of power to the control systems on high reactor pressure, and AMSAC provides an actuation of emergency feedwater and trips the turbine at power levels above approximately 50%

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 21 of 55 of rated and feedwater flow below 17% of rated. Both systems are independent of the reactor protection system (RPS), and both are operable during a loss of offsite power.

The design basis transient for the DSS is the loss of main feedwater (LOFW) with a failure of the RPS reactor trip. DSS actuates on a high RCS pressure of 2450 psig (FSAR Section 7.5). The LOFW transient was evaluated generically at 2772 MWt and ensured the peak RCS pressure remained below 3250 psia. Therefore, the current analysis of record for full power operation bounds the conditions for MUR power uprate. This analysis was also reviewed and accepted by the Staff during the review for Amendment 205.

The approval for the analyses of record for these accidents is contained in Reference 2.3.4.

A separate analysis was performed to confirm that the current AMSAC arming setpoint remains valid. The description of this analysis is provided in Section 3.2.2.

2.2.19 ARTS Transients (FSAR Section 7.2.3.2.4)

The Anticipatory Reactor Trip System (ARTS) will trip the reactor if a turbine trip occurs with reactor power above 45%. ARTS was implemented after the Three Mile Island (TMI-2) accident to minimize the challenges to the pressurizer power operated relief valve (PORV) after a turbine trip. The ARTS trip function is not credited in the design bases accidents.

The ARTS power level setpoint is based upon the maximum core thermal power, wherein a reactor runback is capable of minimizing system pressures below the PORV setpoint. This power level is sensitive to the flow capacities of the turbine bypass (TBV) and the main steam safety valves (MSSV), and the reactor kinetics.

The current design basis analysis is a generic evaluation performed at a rated power condition of 2772 MWt and therefore bounds the MUR (Reference 2.3.3).

2.3 SECTION

2.0 REFERENCES

2.3.1 CR-3 Final Safety Analysis Report, Rev. 30.1.

2.3.2 51-9036887-000, "CR-3 MUR Summary Report."

2.3.3 BAW-1893A, "Basis for Raising Arming Threshold for Anticipatory Trip on Turbine Trip," August 1986.

2.3.4 Safety Evaluation Crystal River, Unit 3 Compliance With ATWS Rule 10 CFR 50.62, Docket No. 50-302, April 1989.

2.3.5 BAW-10179-P-A, Revision 5, "Safety Criteria and Methodology for Acceptable Cycle Specific Reload Analyses," December 2004

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 22 of 55 TABLE D. 2-1 Crystal River Unit 3 FSAR Accident Analyses Initial Core Bounded Supported FSAR Event Power Used in by Current / Bounded Discussion Section(s) FSAR Analysis FSAR by Other

(% of 2568) Analysis Analyses 14.1.2.1 Uncompensated 100.2% X This accident was originally analyzed to Operating Reactivity (2575 MWt) demonstrate the ability of control systems and Changes operators to compensate for slow variations in reactivity. The analysis was evaluated at 2575 MWt in Reference 2.3.1. A slight increase in power to the MUR conditions will not result in any appreciable change in the accident as previously analyzed. The reactivity changes for this event are also bounded by the reactivity changes in the startup accident. The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

14.1.2.2 Startup Accident 10' % X The analytical high flux reactor trip setpoint will (4.3.7) be reduced from 112% (of 2568 MWt) to 110.2%

(of 2609 MWt) to ensure that the reactor is tripped at the same net power level. The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

14.1.2.3 Rod Withdrawal at 100% X The analytical high flux reactor trip setpoint will Rated Power Operation 2 % uncertainty be reduced from 112% (of 2568 MWt) to 110.2%

Accident accounted for in (of 2609 MWt) to ensure that the reactor is the 112% power tripped at the same net power level. The analysis setpoint of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

14.1.2.4 Moderator Dilution 102% X The analytical high flux reactor trip setpoint will Accident be reduced from 112% (of 2568 MWt) to 110.2%

(of 2609 MWt) to ensure that the reactor is tripped at the same net power level. The shutdown margin calculation is evaluated as part of the standard reload process. The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 23 of 55 TABLE D. 2-1 Crystal River Unit 3 FSAR Accident Analyses Initial Core Bounded Supported FSAR Power Used in by Current / Bounded Section(s) Event FSAR Analysis FSAR by Other Discussion

(% of 2568) Analysis Analyses 14.1.2.5 Cold Water Accident 50% X Current FSAR analysis remains bounding as analyzed. The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

14.1.2.6 Loss-of-Coolant Flow 102% X The analysis of record is provided in Reference Accident 2.3.1 and remains acceptable for the MUR power uprate. In addition, the DNBR response is verified for each new fuel cycle. Current cycle analyses support the MUR power uprate.

14.1.2.7 Stuck-Out, Stuck-In, or 108% X The analysis of record is provided in Reference Dropped Control Rod 2.3.1 and is based on a core power level of 2772 Accident MWt. The analysis of record remains acceptable for the MUR power uprate.

14.1.2.8 Load Rejection 112% X Under the current configuration, the turbine trip Accident accident bounds the consequences of a load (Turbine Trip) rejection accident. The turbine trip analyses were evaluated at 112% of 2568 MWt. The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

14.1.2.9 Station Blackout 108% X SBO was evaluated at 2772 MWt. The analysis Accident of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 24 of 55 TABLE D. 2-1 Crystal River Unit 3 FSAR Accident Analyses Initial Core Bounded Supported FSAR Event Power Used in by Current / Bounded Discussion Section(s) FSAR Analysis FSAR by Other

(% of 2568) Analysis Analyses 14.2.2.1 Steam Line Failure 100%/102% X For the core response, as documented in the Accident FSAR, the core power was evaluated at 100% of 2568 MWt, to minimize the heat input to the reactor coolant system. The heat balance uncertainty of 2% was accounted for in the steam generator mass inventories. Therefore, the FSAR analysis bounds the MUR power uprate for the core and reactor coolant system response.

Steam Line Failure 102% X For the containment response, the steam line Mass & Energy break event was evaluated to 102 % of 2568 Releases MWt. Therefore, the conclusions reported in the FSAR regarding compliance to the reactor building pressure limit remains valid.

14.2.2.2 Steam Generator Tube 102% X The SGTR was evaluated at 102% of 2568 MWt.

Rupture Accident The dose consequences are evaluated each cycle as part of the standard reload process. These analyses bound the MUR power uprate. The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

14.2.2.3 Fuel Handling Accident 102% X The Fuel Handling Accident was evaluated at 102% of 2568. The dose consequences are evaluated each cycle as part of the standard reload process. These analyses bound the MUR power uprate The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 25 of 55 TABLE D. 2-1 Crystal River Unit 3 FSAR Accident Analyses Initial Core Bounded Supported FSAR Event Power Used in by Current / Bounded Discussion Section(s) FSAR Analysis FSAR by Other

(% of 2568) Analysis Analyses 14.2.2.4 Rod Ejection Accident 0.1% X The Hot Zero Power analyses are not impacted by a change in rated power level. The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

100% X The rod ejection at hot full power conditions were originally evaluated at 100% of 2568 MWt as documented in the FSAR. The neutron power response during a control rod ejection accident is not sensitive to the initial power conditions. Due to the rapid ejection time of 0.15 seconds, the transient is defined by the ejected rod worth and the kinetics parameters. The MUR will cause a slight increase in the fuel heat up of approximately 2 cal/g. There is approximately 80 cal/g margin to the fuel enthalpy limit for an ejected rod worth of 0.7%AK/K. The reload core design ensures that maximum ejected rod worth will not exceed 0.65%AK/K including a 15%

uncertainty. Therefore, the current fuel enthalpy margin more than compensates for the small power increase.

The radiological analyses for the rod ejection accident assumed that the fuel gap activity for 14% of the fuel rods is completely released. The dose release calculations are evaluated on a reload basis at a power level of 102% of 2568 MWt.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 26 of 55 TABLE D. 2-1 Crystal River Unit 3 FSAR Accident Analyses Initial Core Bounded Supported FSAR Event Power Used in by Current / Bounded Discussion Section(s) FSAR Analysis FSAR by Other

(% of 2568) Analysis Analyses 14.2.2.5 Loss-of-Coolant 102% X The spectrum of LOCAs was analyzed for CR-3 Accident at 102% of 2568 MWt. The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

LOCA Mass & Energy 102% X The mass and energy release was performed at Releases 102% of 2568 MWt. Therefore, the analysis of record remains valid for the MIUR 14.2.2.6 Makeup System 102% X The dose consequences were evaluated at 102%

Letdown Line Failure of 2568 MWt. These analyses bound the MUR Accident power uprate. The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

14.2.2.8 Waste Gas Decay Tank 102% X The analysis conservatively assumes that all three Rupture Accident available waste gas decay tanks rupture. Each tank is assumed to contain the maximum curie inventory allowed by the Offsite Dose Calculation Manual. The dose assessment for the WGDTRA is based on WGDT inventories of radioactive nuclides and are independent of power level.

14.2.2.9 Loss of Feedwater and 102% X The total loss of feedwater accident and the Main Feedwater Line feedwater line break was evaluated at 102% of Break Accident 2568 MWt. The analysis of record is provided in Reference 2.3.1 and remains acceptable for the MUR power uprate.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 27 of 55 TABLE D. 2-1 Crystal River Unit 3 FSAR Accident Analyses Initial Core Bounded Supported FSAR Event Power Used in by Current / Bounded Discussion Section(s) FSAR Analysis FSAR by Other

(% of 2568) Analysis Analyses 7.5 ATWS/DSS Setpoint 108% X The ATWS transients are considered beyond the (2772 MWt) original design basis of the B&W-designed plants.

The analyses were performed using nominal values and was evaluated at 2772 MWt. The approval of the analysis of record for CR-3 is provided in Reference 2.3.4 and remains acceptable for the MTUR power uprate.

ATWS/AMSAC 49.53% The ATWS LOFW was re-analyzed at a power Enabling Setpoint (50% of 2544 level of 52% of 2609 MWt to validate the arming MWt) setpoint of the AMSAC system and is discussed in Section 3.0.

7.2.3.2.4 ARTS 108% The current design basis analysis is a generic (2772 MWt) analysis performed at a rated power condition of 2772 MWt. The analysis of record is provided in Reference 2.3.3 and remains acceptable for the MUR power uprate.

N/A Flooding 102% X As discussed above, the various analyses applicable to flooding have been performed at 102% of 2568 MWt. The analyses of record are provided in Reference 2.3.1 and remain acceptable for the MUR power uprate.

N/A Natural Circulation N/A N/A The natural circulation cooldown time will Cooldown increase slightly based upon the power uprate from 2568 MWt to 2609 MWt. The time to cool the plant to 200'F will increase from 68.54 hrs to 70.38 hrs. This is still less than the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Appendix R requirement. No other safe shutdown parameters were impacted by the MUR uprate.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 28 of 55 TABLE D.2-2 COMPARISON OF OVER POWER REACTOR TRIP SETPOINTS, CURRENT VERSUS POST MUR UPRATE CURRENT POST MUR 112 OVERPOWER TRIP DESIGN LIMIT 111% OVERPOWER TRIP DESIGN LIMIT 112% 2876 110.2% 2876 110%/0 111% 4.7%/b ERROR 4.7% ERROR 109% (NOTE 1) 110% (NOTE 1) 108%

1090/0 0.4% MEASUREMENT UNCERTAINTY 107%/0 CALCULATED MAX.

108% ALLOWABLE VALUE 107.3% 2755 0 1 6 S10.5O 2752 107% 2742 2% MEASUREMENT 105%0/ 104.9%

1o5.1% 2737 CALCULATED MAX.

106% UNCERTAINTY rALLOWABLE VALUE TECH. SEPC. ALLOWABLE 'NOTE 2 105.3%/ 2704 104% VALUE 1050/° 1049% 2694 103% 103.3% (NOTE 3) 2694 TECH. SPEC. ALLOWABLE NOTE 2 104% VALUE 102%

103%

1010/0 102%

100% -100% LICENSED POWER 2609 101%

990/o 100% LICENSED POWER 2568 98.4% (NOTE 4) 2568 98%

Note 1: 4.7% error includes 4% bias errors, neutron measurement string errors, and bistable error.

Note 2: The margin between the ITS Allowable Value and the calculated maximum Allowable Value is available for future use if other non-conservatisms are introduced or discovered.

Note 3: If Caldon or associated instrumentation becomes inoperable, reactor trip setpoint is lowered below 103.3% RTP (2694 MWt), equivalent to the current 104.9% Allowable Value.

Note 4: If Caldon or associated instrumentation becomes inoperable, reactor power is lowered below 98.4% RTP (2568 MWt), equivalent to current 100%

RTP.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 29 of 55 3.0 Accidents and Transients for Which the Existing Analyses of Record do not Bound Plant Operation at the Proposed Uprated Power Level. (RIS 2002-03 Section III Questions)

1. This section covers the transient and accident analyses that are included in the plant's UFSAR (typically Chapter 14 or 15) and other analyses that are required to be performed by licensees to support licensing of their plants (i.e., radiological consequences, natural circulation cooldown, containment performance, anticipated transient without scrams, station blackout, analyses for determination of environmental qualification parameters, safe shutdown fire analysis, spent fuel pool cooling, flooding).

2. For analyses that are covered by the NRC approved reload methodology for the plant, the licensee should:

A. Identify the transient/accident that is the subject of the analysis B. Provide an explicit commitment to re-analyze the transient / accident, consistent with the reload methodology, prior to implementation of the power uprate C. Provide an explicit cormmitment to submit the analysis for NRC review, prior to operation at the uprated power level, if NRC review is deemed necessary by the criteria in 10 CFR 50.59 D. Provide a reference to the NRC's approval of the plant's reload methodology

3. For analyses that are not covered by the reload methodology for the plant, the licensee should provide a detailed discussion for each analysis. The discussion should:

A. Identify the transient or accident that is the subject of the analysis B. Identify the important analysis inputs and assumptions (including their values),

and explicitly identify those that changed as a result of the power uprate C. Confirm that the limiting event determination is still valid for the transient or accident being analyzed D. Identify the methodologies used to perform the analyses, and describe any changes in those methodologies E. Provide references to staff approvals of the methodologies in Item D. above F. Confirm that the analyses were performed in accordance with all limitations and restrictions included in the NRC's approval of the methodology G. Describe the sequence of events and explicitly identify those that would change as a result of the power uprate H. Describe and justify the chosen single-failure assumption

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 30 of 55 I. Provide plots of important parameters and explicitly identify those that would change as a result of the power uprate J. Discuss any change in equipment capacities (e.g., water supply volumes, valve relief capacities, pump pumping flow rates, developed head, required and available net positive suction head (NPSH), valve isolation capabilities) required to support the analysis K. Discuss the results and acceptance criteria for the analysis, including any changes from the previous analysis 3.1 Response to RIS 2002-03 Section III Questions All analyses of record bound the MUR power uprate except Anticipated Transient Without Scram (ATWS) Mitigation System Actuation Circuitry (AMSAC), which is discussed below and in Attachment F.

3.1.1 AMSAC Analysis The AMSAC analysis was last performed with an arming setpoint of 50% of 2544 MWt. To ensure that the AMSAC system arming setpoint remains valid for the MUR power level, a new analysis was performed based on a core power level of 52% of 2609 MWt. The limiting ATWS transient for the B&W-designed plant is a loss of feedwater initiated event. The nominal plant setpoint is 50% power. At a lower value, the AMSAC system would not be armed. The purpose of the transient is to demonstrate that without AMSAC, the peak RCS pressure will not exceed 3250 psia. The re-analysis indicated that the previous analysis was not bounding, however the existing setpoint (50%) is conservative and will not be changed. A more detailed discussion is provided in Attachment F of this report.

4.0 Mechanical/Structural/Material Component Integrity and Design (RIS 2002-03 Section IV Questions)

A discussion of the effect of the power uprate on the structural integrity of major plant components. For components that are bounded by existing analyses of record, the discussion should cover the type of confirmatory information identified in Section II, above. For components that are not bounded by existing analyses of record, a detailed discussion should be provided.

A. This discussion should address the following components:

i. reactor vessel, nozzles, and supports ii. reactor core support structures and vessel internals iii. control rod drive mechanisms iv. Nuclear Steam Supply System (NSSS) piping, pipe supports, branch nozzles

v. balance-of-plant (BOP) piping (NSSS interface systems, safety related cooling water systems, and containment systems) vi. steam generator tubes, secondary side internal support structures, shell, and nozzles

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 31 of 55 vii. reactor coolant pumps viii. pressurizer shell, nozzles, and surge line ix. safety-related valves B. The discussion should identify and evaluate any changes related to the power uprate in the following areas:

i. stresses ii. cumulative usage factors iii. flow induced vibration iv. changes in temperature (pre- and post-uprate)

v. changes in pressure (pre- and post-uprate) vi. changes in flow rates (pre- and post-uprate) vii. high-energy line break locations viii. jet impingement and thrust forces C. The discussion should also identify any effects of the power uprate on the integrity of the reactor vessel with respect to:

i. pressurized thermal shock calculations ii. fluence evaluation iii. heatup and cooldown pressure-temperature limit curves iv. low-temperature overpressure protection

v. upper shelf energy vi. surveillance capsule withdrawal schedule D. The discussion should identify the code of record being used in the associated analyses, and any changes to the code of record.

E. The discussion should identify any changes related to the power uprate with regard to component inspection and testing programs and erosion/corrosion programs, and discuss the significance of these changes. If the changes are insignificant, the licensee should explicitly state so.

F. The discussion should address whether the effect of the power uprate on steam generator tube high cycle fatigue is consistent with NRC Bulletin 88-02, "Rapidly Propagating Fatigue Cracks in Steam Generator Tubes," February 5, 1988.

4.1 Response to RIS 2002-03 Section IV Questions Table 4-1 (Reference 4.3.1) contains a summary of changes in operating conditions as a result of the MUR. As can be seen from Table 4-1, there are only minor changes in operating conditions resulting from the uprate at the current OTSG plugging limit.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 32 of 55 Table 4-1 Change in Operating Conditions for MUR Parameter Case A Case B Case C Core Thermal Power (MWt) 2568 2609 2609 Other RCS Power (MWt) (RCP-LD) 16 16 16 Total NSSS Power (MWt) 2584 2625 2625 Pressurizer Control Pressure (psig) 2155 2155 2155 SG A/B Tube Plugging % 2.4%/5.7% 2.4%/5.7% 20%/20%

0 Thot ( F) 601.9 602.2 602.9 Tcold (OF) 556.2 555.8 555.1 Tavg (F) 579 579 579 RCS Mass Flow (E6 lbm/hr) 144.05 144.08 139.81 RCS Vol. Flow (gpm) 386,873 386,729 374,896 Steam Temperature (F) 591.0 590.5 580.7 Steam Superheat (OF) 54.9 54.4 44.6 Feedwater/Steam Flow Rate (E6 lb/hr) 10.86 11.07 11.19 OTSG Steam Pressure (psia) (Input) 931.7 931.7 931.7 Feedwater Temperature (0F) (Input) 456.7 458.4 458.5 Case A Existing Tube Plugging at 2568 MWt Case B Existing Tube Plugging at 2609 MWt Case C 20 Percent Tube Plugging at 2609 MWt 4.2.1 Effect of Power Uprate on Major Components (RIS 2002-03 Section IV.1.A) 4.2.1.1 Reactor Vessel Structural Evaluation (RIS 2002-03 Section IV.1.A.i)

The revised design conditions were reviewed for impact on the existing design basis analyses for the reactor vessel. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Th0 t/TCId) are within design limits.

The design conditions in the existing analyses are based on the RCS functional specification.

The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the reactor vessel remain applicable for the uprated power conditions. (Reference 4.3.6).

4.2.1.2 Reactor Vessel Internals Structural Evaluation (RIS 2002-03 Section IV,1.A.ii)

The revised design conditions were reviewed for impact on the existing design basis analyses for the reactor vessel internals. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/TcoId) are within design

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 33 of 55 limits. The design conditions in the existing analyses are based on the RCS functional specification. The MIUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid.

(Reference 4.3.6) 4.2.1.3 Fuel Assembly The Crystal River 15x15 Mark-B fuel design was evaluated to determine the impact of the power uprate on the fuel assembly structural integrity. Since the core plate motions for the seismic and LOCA evaluations are not affected by the uprated conditions, there is no impact on the fuel assembly seismic/LOCA structural evaluation. The power uprate does not increase operating and transient loads such that they will adversely affect the fuel assembly functional requirements.

Therefore, the fuel assembly structural integrity is not affected, and the seismic and LOCA evaluations of the 15x15 Mark-B fuel design are still applicable for the power uprate.

4.2.1.4 Control Rod Drive Mechanism Structural Evaluation RIS Section 2002-03 Section IV.1.A.iii)

The revised design conditions were reviewed for impact on the existing design basis analyses for the control rod drive mechanisms. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/Tcold) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification. The MIUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the control rod drive mechanism remain applicable for the uprated power conditions. (Reference 4.3.6) 4.2.1.5 Reactor Coolant Piping and Supports Structural Evaluation (RIS 2002-03 Section IV.1.A.iv)

The revised design conditions were reviewed for impact on the existing design basis analyses for the reactor coolant piping and supports. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/TZcod) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification. The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the reactor coolant piping and supports remain applicable for the uprated power conditions. (Reference 4.3.6) 4.2.1.6 Balance-of-Plant (BOP) Piping (NSSS Interface Systems, Safety-Related Cooling Water Systems, and Containment Systems) (RIS 2002-03 Section IV.1.A.v)

The structural analyses of the piping attached to the RCS (decay heat line, makeup/high pressure injection lines) use anchor motions from the RCS structural analyses. As discussed in Section 4.2.1.4, these anchor motions do not change due to the uprated power conditions. The revised

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 34 of 55 design conditions were reviewed for impact on the existing design basis analyses for the reactor coolant system attached piping and supports. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/TWcod) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification. The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid.

The revised design conditions were reviewed for impact on the existing design basis analyses for the main steam and main feedwater piping and supports. No significant changes in OTSG design or operating pressure were made as part of the power uprate. The changes in the operating temperatures and flow rates due to the MUR power uprate have been evaluated. These changes were determined to have a negligible effect on the existing design basis analyses. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. (Reference 4.3.6) 4.2.1.7 Steam Generator Tubes, Secondary Side Internal Support Structures, Shell and Nozzles (RIS 2002-03 Section IV.1.A.vi)

The revised design conditions were reviewed for impact on the existing design basis analyses for the steam generator. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/Tcold) are within design limits.

The design conditions in the existing analyses are based on the RCS functional specification.

The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the steam generator remain applicable for the uprated power conditions.

Topical report BAW-10146 (Reference 4.3.2) established the minimum required steam generator tube wall thickness for the B&W 177-FA plants. Tube loads were calculated for normal operating and faulted conditions. Normal operating tube loads were determined using design operating transients and were combined with tube geometry to calculate minimum allowable tube wall thickness that satisfy the acceptance criteria of NRC Draft RG 1.121. Faulted condition tube loads are those arising from a safe shutdown earthquake, a loss of coolant accident, a main steam line break (MSLB) and a feedwater line break (FWLB). These loads were used to calculate minimum wall thickness based on the limits of NRC Draft RG 1.121 and ASME Code,Section III, Appendix F. The MUR Power Uprate Program operating conditions were compared with the existing design conditions. The comparison showed that the power uprate by itself will not result in operation outside the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the minimum required tube wall thickness for normal operating conditions will not be affected by the power uprate.

Tube loads for the faulted conditions were calculated for LOCA, MSLB, and FWLB accident conditions considering thermal and pressure loads on the steam generator. The MUR Power Uprate Program operating temperatures were compared with the existing design temperatures.

The comparison showed that the existing design temperatures bound the power uprate temperatures. This means that the existing tube loads due to LOCA, MSLB and FWLB will not

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 35 of 55 change as a result of the power uprate.

In addition, a review of calculations performed which assessed the integrity of tubes containing flaws of various types when subjected to operating and accident loads was conducted. This review ensured that existing structural margins are maintained for the MUR Power Uprate Program design conditions. (Reference 4.3.6) 4.2.1.8 Reactor Coolant Pump Structural Evaluation (RIS 2002-03 Section IV.1.A.vii)

The revised design conditions were reviewed for impact on the existing design basis analyses for the reactor coolant pumps. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (ThotITCold) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification. The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the reactor coolant pumps remain applicable for the uprated power conditions. (Reference 4.3.6) 4.2.1.9 Pressurizer Structural Evaluation (RIS 2002-03 Section IV.1.A.viii)

The revised design conditions were reviewed for impact on the existing design basis analyses for the pressurizer. No changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/TcoId) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification. The MUR power uprate conditions are bounded by the design conditions. Since the operating transients will not change as a result of the power uprate and no additional transients have been proposed, the existing loads, stresses and fatigue values remain valid. Thus, the existing stress reports for the pressurizer remain applicable for the uprated power conditions. (Reference 4.3.6) 4.2.1.10 Safety-Related Valves (RIS 2002-03 Section IV.1.A.ix)

The revised design conditions were reviewed for impact on the existing design basis analyses for the safety-related valves. The evaluation showed that the temperature changes due to the MUR uprate are bounded by those used in the existing analyses. Safety analysis confirmed the installed capacities and lift setpoints of the RCS and Main Steam relief valves to be valid for the MUR Conditions. Therefore, the existing loads remain valid and the stresses and fatigue values also remain valid. Safety-related valves were reviewed within the system and program evaluations. None of the safety-related valves required a change to their design or operation as a result of the MUR. (References 4.3.9 and 4.3.10) 4.2.2 Effect of Power Uprate on Stresses, Operating Conditions, and HELB (RIS 2002-03 Section IV.1.B) 4.2.2.1 Stresses (RIS 2002-03 Section IV.1.B.i)

The revised design conditions were reviewed for impact on the existing design basis. No

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 36 of 55 changes in RCS design or operating pressure were made as part of the power uprate. The effects of operating temperature changes (Thot/Tco1d) are within design limits. The design conditions in the existing analyses are based on the RCS functional specification. The stress reports including the tabulation of maximum stress intensities/stress ranges with a comparison to stress allowables, cumulative usage factors, and other special stress limits were reviewed. The MUR power uprate conditions are bounded by the design conditions. (Reference 4.3.6) 4.2.2.2 Cumulative Usage Factors (RIS 2002-03 Section IV.1.B.ii)

The revised design conditions for the NSSS components, piping and interface systems were reviewed for impact on the existing design basis analyses. For NSSS components, the evaluation showed that the operating conditions due to the MUR uprate are bounded by those used in the existing analyses. Further, since the evaluated transients listed in FSAR Table 4-8 will not change as a result of the power uprate, the existing loads remain valid and the stresses and fatigue values (cumulative usage factors) also remain valid. (Reference 4.3.6) 4.2.2.3 Flow Induced Vibration (FIV) (RIS 2002-03 Section IV.1.B.iii)

As shown in Table 4-1, the RCS flow rate changes insignificantly compared to the RCS flow rate prior to the uprate for the same steam generator tube plugging conditions. These flow rates were evaluated against flow rates used in Topical Report BAW-10051 (Reference 4.3.3), which presents the design analysis of the RV internals and incore instrument nozzles subjected to operational flow-induced vibration loading for the B&W 177-FA plants. A comparative analysis was performed to evaluate the effects of the operating conditions. This evaluation concluded that those components remain structurally adequate for the observed flow conditions. (Reference 4.3.6)

An evaluation was performed (Reference 4.3.13) that concluded that there currently exists a minimum of 13.6% margin against detrimental effects inside the OTSG due to flow induced vibrations for the 2609 MWt uprate considering 20% tube plugging. The limiting FIV mechanism is turbulence and the resulting mid-span tube impacts.

4.2.2.4 Changes in Temperature (Pre- and Post-Uprate) (RIS 2002-03 Section IV.1.B.iv) 4.2.2.4.1 Temperature Changes The changes in operating temperatures are provided in Table 4-1. The average temperature is unchanged and the cold leg decreases 0.4°F while the hot leg temperature increases 0.3°F. These changes as discussed elsewhere have minimal impact on the MUR.

4.2.2.4.2 Evaluation of Potential for Thermal Stratification Thermal stratification in the lines attached to the primary side of the RCS occurs mainly during heatup and cooldown. The 100% power hot and cold leg temperatures that the plant has been designed to are essentially the same as those for the MUR Power Uprate Program. This means that the effects of thermal stratification will not change as a result of the power uprate.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 37 of 55 NRC Bulletin 88-08, "Thermal Stresses in Piping Connected to Reactor Coolant Systems,"

addresses the issue of thermal stresses in piping attached to the primary loop that cannot be isolated. The temperature changes as a result of the MUR Power Uprate Program compared to current operation are negligible and will not have an effect on existing or potential thermal stratification conditions. In addition, the design RCS flow rates are essentially the same as those for the MUR Power Uprate Program and thus the effects of turbulent penetration will not change as a result of power uprate.

NRC Bulletin 88-11, "Pressurizer Surge Line Thermal Stratification," addresses the issue of surge line thermal stratification. Thermal stratification in the surge line occurs mainly during plant heatup and cooldown and is driven by the temperature difference between the hot leg and the pressurizer. The current operating temperature of the hot leg will increase very slightly due to the MIUR Power Uprate Program. A higher hot leg temperature gives a lower temperature differential between the hot leg and the pressurizer which in turn lessens the stratification effects.

This means that stress and fatigue in the surge line which is attributed to thermal stratification is bounded by the existing analyses. (Reference 4.3.6) 4.2.2.5 Changes in Pressure (Pre- and Post-Uprate) (RIS 2002-03 Section IV.1.B.v)

The changes in operating pressures are provided in Table 4-1. As discussed in Section 2.2 the accident analyses is unchanged. The RCS pressure and pressurizer pressure control setpoint remains the same.

4.2.2.6 Changes in Flow Rates (Pre- and Post-Uprate) (RIS 2002-03 Section IV.1.B.vi)

The changes in RCS flow are provided in Table 4-1. The MUR power uprate does not have an appreciable effect on RCS mass flow (<0.1%). Therefore, the changes in mass flow rates (pre-and post-uprate) will have a negligible impact on core design and safety analyses.

4.2.2.7 High Energy Line Break Locations (HELB) (RIS 2002-03 Section IV.1.B.vii)

An engineering evaluation was performed (Reference 4.3.12) which evaluated the impacts of HELB systems inside and outside containment at CR-3. High energy piping is defined as piping carrying fluid above 275 psig and 200°F inside containment and above 275 psig and/or 200°F outside containment. The HELB evaluations were performed at 2619 MWt to bound the expected range of operation resulting from the MUR uprate.

There are no HELB impacts on the systems reviewed inside containment, nor for flooding inside containment. For high energy systems reviewed outside containment, there are no outliers as a result of the proposed MUR uprate. Flooding events outside containment in the Intermediate Building and the Auxiliary Building are not affected by the uprate.

There were no new line breaks postulated for current HELB systems inside or outside containment as pressures and temperatures did not increase. There are no new systems inside or outside containment that qualify as HELB systems as a result of the uprate.

4.2.2.8 Leak-Before-Break Evaluation The Leak-Before-Break (LBB) concept applies known mechanisms for flaw growth to piping designs with assumed through-wall flaws and is based on the plant's ability to detect an RCS

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 38 of 55 leak. Topical Report BAW-1847, Rev. 1, (Reference 4.3.4) presents the LBB evaluation of the RCS primary piping. It showed that a double-ended guillotine break will not occur and that postulated flaws producing detectable leakage exhibit stable growth, and thus, allow a controlled plant shutdown before any potential exists for catastrophic piping failure. The major areas that contributed to this evaluation were: RCS piping structural loads; leakage flaw size determination; flaw stability analysis; and, RCS piping material properties. An evaluation was performed which determined the impact of the MUR uprate design conditions on the inputs to the LBB analyses is negligible and the LBB conclusions remain unchanged. (Reference 4.3.6) 4.2.2.9 Reactor Coolant System Loss of Coolant Accident Forces Evaluation Topical report BAW-1621 (Reference 4.3.5) addresses the RCS components for primary break Loss-of-Coolant Accident (LOCA) loadings. The breaks considered were limited break ruptures of the primary piping. Due to LBB qualification of the hot and cold legs, the RCS was requalified for snubber removal. The MUR Power Uprate Program design conditions were reviewed for impact on the existing hydraulic forcing functions and the HELB locations in the primary RCS piping and the piping attached to the primary RCS to the first anchor. The evaluation showed that the asymmetric cavity pressure forces, thrust loads, and jet impingement loads remain bounded by the values in the existing analyses. The evaluation also showed that there are no additions or changes to the HELB locations or loads. (Reference 4.3.6) 4.2.3 Effect of Power Uprate on Reactor Vessel Integrity (RIS 2002-03 Section IV.1.C) 4.2.3.1 Pressurized Thermal Shock (PTS) (RIS 2002-03 Section IV.1.C.i)

The reference temperature for pressurized thermal shock (RTprs ) values in support of a power uprate applicable to the projected end-of-life period (32 EFPY) for the reactor vessel beltline materials were re-evaluated. These values were calculated in accordance with the requirements in 10 CFR 50.61. A 7% increase in 32 EFPY neutron fluence was used to bound the effects of the MUR power uprate on RTprs. The limiting reactor vessel beltline material has a RTpTs value of 206'F at 32 EFPY. The screening criterion for this weld metal is 270'F. Therefore, the reactor vessel will remain within its limits for PTS after the MUR power uprate. (Reference 4.3.7) 4.2.3.2 Fluence Evaluation (RIS 2002-03 Section IV.1.C.ii)

The impact of a MUR power uprate on the high energy neutron leakage (neutrons with energies greater than 1.0 million electron volts (MeV) or E > 1.0 MeV) from the core to the internals and reactor pressure vessel will be minimal. The neutron leakage directly impacts the pressurized thermal shock criteria, the pressure-temperature (P-T) limits (including those for low temperature over pressurization), and the baffle bolts or other internals. A 7% increase in 32 EFPY end of life fluences was used to bound the effects of the MUR power uprate, for both the reactor vessel and the reactor internals. Clearly, the assumed 7% increase in neutron fluence conservatively bounds the actual anticipated increase of the reactor thermal power of 2% or less based on the MUR power uprate. (Reference 4.3.7)

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 39 of 55 4.2.3.3 Heatup and Cooldown Pressure / Temperature Limit Curves (RIS 2002-03 Section IV.1.C.iii)

The current P-T limit curves are licensed through 32 effective full power years (EFPY) and are based on adjusted reference temperatures at the 1/4/-thickness (I/T) and 3/4-thickness ( 3/T) wall locations for the limiting reactor vessel beltline material. Adjusted reference temperature (ART) values were calculated in accordance with Regulatory Guide 1.99, Revision 2. Inputs affecting the adjusted reference temperatures and P-T curves remain unchanged under the MUR power uprate, with the exception of neutron fluence. Changes to the core power level will affect neutron flux, which will affect neutron fluence, and could have ultimately affected the validity period of the current P-T curves.

The impact of the MUR power uprate on the P-T curves was assessed by performing a revised 32 EFPY ART calculation in accordance with Regulatory Guide 1.99, Revision 2, which considered recent reactor vessel surveillance data and an assumed 7% increase in 32 EFPY fluence due to a power uprate. The assumed 7% increase in neutron fluence conservatively bounds the actual anticipated increase of neutron fluence at 2% or less based on the MUR power uprate. Based on the additional credible reactor vessel surveillance data, the chemistry factors utilized in the ART calculations were reduced leading to an overall reduction in ART at 32 EFPY. The limiting ART values at 1/4/4T and 3/T were reduced from 213'F and 144.5'F to 195.7°F and 144.1'F, respectively. Therefore, the existing 32 EFPY P-T curves and Low Temperature Overpressure Protection (LTOP) limits remain valid for the MUR power uprate. (Reference 4.3.7) 4.2.3.4 Low-Temperature Overpressure Protection (RIS 2002-03 Section IV.l.C.iv)

As described above, the current LTOP limits in the 32 EFPY P-T curves do not need to be modified for the MUR. (Reference 4.3.7) 4.2.3.5 Effect on Low Upper Shelf Energy (RIS 2002-03 Section IV.1.C.v)

Due to the increase in fluence from a power uprate, low upper-shelf toughness was evaluated to ensure compliance with Appendix G to 10 CFR Part 50. If the limiting reactor vessel beltline material's Charpy upper-shelf energy (USE) is projected to fall below 50 ft-lb, an equivalent margins assessment must be performed. The limiting reactor vessel beltline materials for CR-3 are welds WF-70 (upper shell to lower shell circumferential weld) and WF-8/WF-18 (upper shell longitudinal weld).

An equivalent margin assessment was performed for these welds in a 1994 B&W Owners Group generic analysis. These welds were evaluated for ASME Boiler & Pressure Vessel Code Levels A, B, C, and D Service Loadings based on the evaluation acceptance criteria of ASME Section XI, Code Case N-512, which later became ASME Section XI, Appendix K.

The analysis demonstrated that the limiting reactor vessel beltline welds at CR-3 satisfy the ASME Code requirements of ASME Code Case N-512 (ASME Section XI, Appendix K) for ductile flaw extensions and tensile stability using projected low upper-shelf Charpy impact energy levels for the weld material at 32 EFPY considering a fluence which bounds the MUR power uprate. (Reference 4.3.7)

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 40 of 55 4.2.3.6 Surveillance Capsule Withdrawal Schedule (RIS 2002-03 Section IV.1.C.vi)

A withdrawal schedule is developed to periodically remove surveillance capsules from the reactor vessel to effectively monitor the condition of the reactor vessel materials under actual operating conditions. FPC has completed withdrawal of capsules for CR-3. As discussed above, projections based upon these withdrawals has been factored into fluence calculations and have demonstrated acceptable operation through 32 EFPY.

4.2.4 Code of Record Used in Associated Analyses (RIS 2002-03 Section IV.1.D)

No new structural or fluence analyses were performed. Analyses and codes of record remain unchanged except as discussed in Section 3.0.

4.2.5 Impact of Uprate on Inspection and Testing Programs Including Erosion/Corrosion Programs (RIS 2002-03 Section IV.1.E) 4.2.5.1 Alloy 600 Primary Water Stress Corrosion Cracking (PWSCC)

The effects of an RCS temperature increase resulting from the power uprate on Alloy 600 PWSCC have been evaluated. For the limiting case of 20% OTSG tube plugging, it is estimated that the increase of Thor from 601.7 0 F to 603.3TF decreases the time to PWSCC initiation by 6%

and increases the crack growth rate by 4%. Because the power uprate does not increase the T 0old and Tavg, or the RCS pressure and Tsat, the impact is limited to Alloy 600 components and welds operating near Thot. Examination of the AREVA NP Alloy 600 ranking model shows that the current relative PWSCC ranking of Alloy 600 components will not change after the power uprate. CR-3 is mitigating the pressurizer butt welds during the Fall 2007 refueling outage.

These welds are the most susceptible components to PWSCC. Hence, the impact of the power uprate on Alloy 600 PWSCC is considered very limited and addressed by current CR-3 aging management programs for Alloy 600.

4.2.5.2 Inservice Testing (IST) Program 10 CFR 50.55a(f), "Inservice Testing Requirements," requires the development and implementation of an Inservice Testing (IST) Program. CR-3 has developed, and is implementing an Inservice Testing (IST) Program for Pumps and Valves per the applicable requirements. This evaluation reviewed the impact to the Inservice Testing Program as part of the MUR uprate conditions up to the original licensed reactor thermal power of 2609 MWt and concluded that the MUR uprate is bounded by current analysis and any changes are insignificant.

4.2.5.3 Inservice Inspection (ISI) Program 10 CFR 50.55a(g), "Inservice Inspection Requirements," requires the development and implementation of an Inservice Inspection (ISI) Program. The applicable program requirements are specified in ASMIE Section XI. CR-3 has developed and is implementing an Inservice Inspection (ISI) Program per these requirements. This evaluation evaluated the impact to the Inservice Inspection Program as part of the MUR uprate conditions up to of the original licensed reactor thermal power 2609 MWt and concluded that the MUR uprate is bounded by current analysis and no changes are required.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 41 of 55 4.2.5.4 Erosion / Corrosion (FAC) Program The CR-3 FAC model has been revised to reflect the 1.6% MUR conditions. Therefore, the predicted increases in maximum component wear rates and reductions in service lives will be managed by the CR-3 FAC program. The most limiting piping segment is in the feedwater system. It was explicitly re-evaluated based on the revised model. The results support continued operation until its scheduled replacement concurrent with steam generator replacement.

4.2.6 Impact of NRC Bulletin 88-02 "Rapidly Propagating Fatigue Cracks in Steam Generator Tubes" and NRC Information Notice 2002-02 (including Supplement 1)

"Recent Experience with Plugged Steam Generator Tubes" Upon the CR-3 MUR Power Uprate (RIS 2002-03 Section IV.1.F)

NRC Bulletin 88-02 implements actions to be taken by the holders of operating licenses of Westinghouse Replacement Steam Generator designs (Specifically models 13, 27, 44, 51, D1, D2, D3, D4 and E) to minimize the potential for a steam generator tube rupture event caused by rapidly propagating fatigue cracks such as occurred at North Anna Unit 1 on July 15, 1987. The tube rupture occurred in the ubend region of a row 9 tube at the top Tube Support Plate (TSP).

The cause of the rupture was high cycle fatigue. The source of the loads was a combination of a high mean stress level in the tube and a superimposed alternating stress. The mean stress was produced by denting of the tube at the upper most TSP and the alternating stress was the result of out-of-plane defection of the ubend portion of the tube above the uppermost support plate caused by flow-induced vibration.

The most significant contributors to this occurrence was a high fluid-elastic stability ratio (not margin as addressed in this document) resulting from a reduction in damping at the tube-to-tube support plate intersection caused by denting and a locally high flow velocity caused by non-uniform anti-vibration bars penetrations into the u bend tube bundle region.

Since NRC Bulletin 88-02 is not applicable to the OTSG designs, there is no impact upon the Appendix K power uprate and no action is required. A more relevant NRC Generic Communication for OTSG designs to consider would be Information Notice 2002-02, "Recent Experience with Plugged Steam Generator Tubes," dated January 2002 and July 2002 for Supplement 1. EPRI Topical Report 1008438, "Three Mile Island Plugged Tube Severance (A Study of Damage Mechanism)," addresses the concerns identified with Information Notice 2002-02.

The results and findings of the EPRI Report 1008438 concluded that certain types of tube degradation can continue to occur in any steam generator after the tube has been taken out of service. For the B&W OTSGs, it was concluded that the only real vulnerability for tube severance is the growth of circumferential cracks due to high cycle fatigue. However, for a swollen, plugged tube, any degradation mechanism has the potential to provide an initiating site for failure.

In response to the findings, AREVA NP has implemented steam generator plugging and de-plugging maintenance procedures that will prevent such incidences from occurring in the future, and CR-3 has complied with these and all other recommendations to mitigate the consequences

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 42 of 55 of over-pressurized tubes in the OTSGs. To address tubes that were plugged prior to NRC Information Notice 2002-02 that may be susceptible to tube swelling, CR-3 has plugged and stabilized all of the adjacent/neighboring tubes.

To address the possibility of circumferential tube cracks eventually severing due to high cycle fatigue, the OTSG stabilization criteria have historically required stabilization of all circumferential crack-like indications regardless of the radial location or elevation. In addition, the OTSG stabilization criteria have historically required stabilization of circumferentially-oriented volumetric indications in regions of high cross flows. Therefore, the finding of the EPRI Report 1008438 have always been employed for these degradation types.

Therefore, there are no Flow Induced Vibration concerns related to the tube bundle associated with the Appendix K power uprate relevant to findings provided by NRC Information Notice 2002-02 or the EPRI Report 1008438 that have not already been evaluated in this and earlier revisions of this document. (Reference 4.3.8) 4.3 Section 4.0 References References 4.3.2 through 4.3.5 are part of the CR-3 design and licensing basis and are also referenced in various locations in the CR-3 FSAR.

4.3.1 .32-5012972-001, "CR-3 Power Uprate Operating Conditions."

4.3.2 Topical Report BAW-10146, "Determination of Minimum Required Tube Wall Thickness for 177-FA Once-Through Steam Generators," November 1980.

4.3.3 Topical Report BAW-10051, "Design of Reactor Internals and Incore Instrument Nozzles for Flow-Induced Vibration," September 1972.

4.3.4 Topical Report BAW-1847, Rev. 1, "The B&W Owners Group Leak-Before-Break Evaluation of Margins Against Full Break for RCS Primary Piping of B&W Designed NSS," September 1985.

4.3.5 Topical Report BAW-1621, "Effects of Asymmetric LOCA Loads - Phase II," July 1980.

4.3.6 51-9036246-000, "CR-3 MUR Power Uprate RCS Structural Assessment."

4.3.7 51-9040378-000, "CR-3 MUR RV Integrity Summary."

4.3.8 51-5000475-002, "CR-3 OTSG FIV Margins."

4.3.9 51-9041016-000, "CR-3 MUR Power Uprate (2609 MWt) Mechanical Systems Review."

4.3.10 51-9036887-000, "CR-3 MUR Summary Report."

4.3.11 Amendment No. 205 to the Crystal River Unit 3 Facility Operating License (TAC NO.

MB 5289), dated December 4, 2002.

4.3.12 51-9046833-000, "CR-3 MUR 101.6% HELB Evaluation."

4.3.13 51-5000475-002, "CR-3 OTSG FIV Margins."

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 43 of 55 5.0 Electrical Equipment Design (RIS 2002-03 Section V Questions)

1. A discussion of the effect of the power uprate on electrical equipment. For equipment that is bounded by the existing analyses of record, the discussion should cover the type of confirmatory information identified under Section II, above. For equipment that is not bounded by existing analyses of record, a detailed discussion should be included to identify and evaluate the changes related to the power uprate. Specifically, this discussion should address the following items:

A. emergency diesel generators B. station blackout equipment C. environmental qualification of electrical equipment D. grid stability E. transformers 5.1 Response to RIS 2002-03 Section V Questions 5.2.1 Emergency Diesel Generators (RIS 2002-03 Section V.1.A)

The emergency diesel generator system (System Code - EG) provides emergency electrical power for the plant Engineered Safeguards (ES) plus selected balance of plant emergency loads.

Margin currently exists on each emergency diesel generator (EGDG-IA and EGDG-1B) and the alternate AC diesel. The uprate will not change the loading of the emergency diesel generators or the alternate AC diesel. Therefore, EG System equipment capacity and capability for plant operation under MUR power uprate conditions are bound by the generator loading tables which are supported by the existing analysis of record. As a result, the EG System will continue to have adequate capacity and capability to operate the plant equipment. Relative to the EG System, there are no changes to plant technical specifications, protection system settings, and/or emergency system settings needed to support the MUR power uprate. (Reference 5.3.1)

The alternate AC diesel, with its separate fuel supply, has the capability of being aligned to either safety-related AC distribution bus. This provides additional assurance that AC power remains available. The alternate AC diesel provides defense in depth however, this diesel was not credited in the Station Blackout analysis.

5.2.2 Station Blackout Equipment (RIS 2002-03 Section V.1.B)

The DC power system (System Code - DP) supplies required and expected loads (during the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months load profile) in the event of a Station Blackout. The MUR uprate will have no impact on the design of or the loads supplied from the DP System. Therefore, DP System equipment capacity and capability for plant operation under MiUR power uprate conditions are bound by the load profiles which are supported by the existing analysis of record. As a result, the DP System will continue to have adequate capacity and capability to operate the plant equipment. Relative to the DP System, there are no changes to plant technical specifications, protection system settings, and/or emergency system settings needed to support the MUR power uprate.

(Reference 5.3.1)

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 44 of 55 5.2.3 Environmental Qualification of Electrical Equipment (RIS 2002-03 Section V.1.C)

The limiting post-accident reactor building conditions were demonstrated to not be increased as part of the LOCA Mass and Energy and Reactor Building analyses described earlier. Thus, the accident profile to which equipment is qualified remains bounding.

5.2.4 Grid Stability (RIS 2002-03 Section V.1.D)

CR-3 output is provided to the 500 kV switchyard. The power to both the preferred and alternate sources of offsite power, offsite power transformer (OPT) and Back-up Engineered Safeguards ES transformer (BEST), is provided from the 230 kV switchyard. The two switchyards are electrically separate and the loss of CR-3 will only affect the voltage for the 500 kV switchyard, not the 230 kV switchyard. Hence, the 230 kV switchyard and the two offsite power sources (OPT and BEST) are unaffected by the loss of CR-3 generator output.

These switchyards are electrically separated by approximately thirty-five miles of 230 kV transmission line, approximately thirty-five miles of 500 kV transmission line, and a 230:500 kV transformer located at the Brookridge substation about thirty-five miles away from the plant.

The separation between these switchyards and the generating capacity of the three local 230 kV fossil generators at the Crystal River Energy Complex protect the 230 kV switchyard from large voltage perturbations induced by CR-3 tripping off-line (disconnecting from the 500 kV switchyard).

Crystal River Unit 4, a 720 MW coal station, is the largest unit that supplies the 230 kV switchyard. Predicted results for a trip of Unit 4 and other potential contingencies have compared favorably with actual event data.

Transmission Planning has reviewed the 180 MW capacity increase for CR-3. Specifically, requests were made through the FERC Large Generator Interconnection procedures to uprate CR-3 in three phases: a) 14 MW in December 2007, b) 26 MW in December 2009, and c) 140 MW in December 2011. Various dynamic and steady state scenarios were studied. The uprates were analyzed as a single increase of 180 MW.

The transient stability simulation for NERC category D5 type faults (breaker failure for three phase fault) does not indicate any adverse effect of the capacity increase on the bulk grid system.

The critical clearing time simulation indicated a reduction of approximately of 1 cycle that is from 9.5 cycles to 8.5 cycles.

The steady state analysis was conducted in two parts: a) single contingency analysis was performed for all facilities on the Florida transmission system to determine if the CR-3 uprates would require transmission upgrades, and b) due to CR-3's proximity to the Florida-Southern Company Interface, a study of 56 line outages both for existing capacity of 812 MW and projected capacity of 992 MW were studied at a 3600 MW import scenario to determine if the CR-3 uprates would have any adverse impact on the Total Transfer Capabilities (TTC) of the Interface. Again, the capacity increase did not adversely influence any of the contingencies.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 45 of 55 The analysis was performed using the following methodologies and assumptions:

Transient Stability Study For the transient stability analysis, a 2012 summer dynamic case developed by the Stability Working Group (SWG) of the Florida Reliability Coordinating Council (FRCC) was used as a starting point. The case (y06_12s-dynd.sav) was modified to reflect the necessary changes. The following scenarios were studied:

1. Three phase fault on Crystal River 500 kV bus, cleared normally. NERC Category B 1.

2. Three phase fault on Crystal River 500 kV bus, with breaker failure condition, NERC Category D 1.

3. Three phase fault on Crystal River 500 kV bus, critical clearing time.

A comparison of the results indicated that an increase in capacity will not have an adverse effect on the power grid. A NERC category C fault was not studied since a more severe category D event was simulated and results are satisfactory. Note that Progress Energy Florida's Transmission Planning unit is required to demonstrate compliance with the NERC Transmission Planning Standards TPL-001-0, TPL-002-0, TPL-003-0 and TPL-004-0.

Steady State Load Flow Study For Part (a), analysis for typical peak load scenarios was performed earlier in 2007 in a separate study and the total 180 MW uprate was determined to have no adverse impact on normal condition or contingency flows anywhere on the Florida transmission system. For Part (b), a maximum 3600 MW import from Southern Company was analyzed using a 2013 summer peak load case created by FRCC. Two types of outages were examined: 1) an outage of the largest unit in Florida, which by 2011 would be CR-3, and 2) single contingency loss of transmission facilities that are critical to the Florida-Southern Interface. The line outage list was the same that was used in the biannual Florida-Southern Interface studies.

1. For the unit outage, the loss of generation was made up of 87.5% coming from the Southern Interface and the rest, 12.5%, being contributed from the FRCC region. This was repeated for 812 MW as well.

2. There were 56 contingencies studied. Each one consisted of one element out of service at a time. This was performed for the 992 MW case as well as for 812 MW. This provided a comparison to assess the impact of the capacity increase.

Conclusion Based on the analysis performed, it is concluded that the scenarios studied do not indicate any adverse effect on grid stability or various contingencies as listed in the report.

All the simulations and studies are based upon the full 180 MW increase, which is considered a worst case situation for transient stability. Since the analysis demonstrated that a full uprate of 180 MW may be implemented with no violation of NERC Reliability Standards, it is clear that the 14 MW uprate would not adversely impact grid stability.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 46 of 55 5.2.5 Station Auxiliary Electric Power Distribution System The AC power system (System Code - AC) will experience minor load changes (additions) as a result of the MiUR uprate. The installation of the Caldon equipment by Engineering Change (EC) 65626, and additional Main Steam pressure and temperature instrumentation by EC 65629, will add negligible loads which will be addressed in the modification documents. Condensate pump motor load will increase slightly but remain within the design rating of the motor, associated electrical components and protective relay settings. Feedwater booster pump motor load will increase slightly; however, the motor power required at the uprate condition will remain well within design. Therefore, the AC System will continue to have adequate capacity and capability to operate the plant equipment. Relative to the AC System, there are no changes to plant technical specifications, protection system settings, and/or emergency system settings needed to support the MUR power uprate. (Reference 5.3.1) 5.2.6 Step-Up and Auxiliary Transformers The Main Power Transformer (Step-Up) is being replaced during the upcoming refueling outage (RFO-15). At that time the Step-Up Transformer will have the nominal rating of 1200 MVA.

The transformer will have more than enough capability to accept the approximately 14 MVA from the MUR uprate. Current Step-Up transformer rating is approximately 950 MVA, while current load is approximately 900MWe. The Unit Auxiliary Transformer is capable of handling full in-house loads before and after uprate.

5.3 Section 5.0 References 5.3.1 51-9037444-000, "CR-3 MUR Power Uprate (2609 MWt) BOP Electrical Systems Review."

6.0 System Design (RIS 2002-03 Section VI Questions)

1. A discussion of the effect of the power uprate on major plant systems. For systems that are bounded by existing analyses of record, the discussion should cover the type of confirmatory information identified under Section II, above. For systems that are not bounded by existing analyses of record, a detailed discussion should be included to identify and evaluate the changes related to the power uprate. Specifically, this discussion should address the following systems:

A. NSSS interface systems for pressurized-water reactors (PWRs) (e.g., main steam, steam dump, condensate, feedwater, auxiliary/emergency feedwater) or boiling-water reactors (BWRs) (e.g., suppression pool cooling), as applicable B. containment systems C. safety-related cooling water systems D. spent fuel pool storage and cooling systems

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 47 of 55 E. radioactive waste systems F. Engineered safety features (ESF) heating, ventilation, and air conditioning systems 6.1 Response to RIS 2002-03 Section VI Questions A comparison between operating requirements for the 2609 MWt MUR conditions generated by the PEPSE heat balance and the 2568 MWt heat balance conditions demonstrates that the major plant systems that meet the requirements identified in Section 4.1 above and discussed below have sufficient design and operational margin to accommodate the MUR uprate.

A review of the FSAR (Reference 6.3.1) Chapter 14 accidents was performed to determine if the analyses of record for CR-3 remained applicable and bounding for the power uprate. The results of this review are described herein. It was concluded that the existing analyses as described in the FSAR were performed with a bounding core power level or the consequences for a given event were bounded by other analyses presented in the FSAR. Based on the results of this review, and in order to ensure protection for the high flux analytical limit of 2876 MWt, the reactor over power limit value will be adjusted to reference the new Reactor Thermal Power value of 2609 MWt. This will require a change from the current value of 112% RTP (2568 MWt) to 110.2% RTP (2609 MWt). This change maintains the reactor trip at the same net power level modeled in the safety analysis. After accounting for Measurement Uncertainty, instrumentation, and process errors, with the reduced heat balance uncertainty using the LEFM CheckPlus TM system, the Technical Specification RPS High Flux trip allowable value can be maintained at 104.9% of the new RTP with four reactor coolant pumps operating. These limits are applicable when power range nuclear instrumentation is verified consistent with the heat balance results calculated using the LEFM system for feedwater flow measurement. Specific requirements for operation with inoperable LEFMs will be placed in CR-3 procedure CP-500.

The individual systems are discussed in more detail below.

6.2.1 NSSS Interface Systems (RIS 2002-03 Section VI.1.A) 6.2.1.1 Main Steam (MS) System The MS System performs the following safety functions; provides automatic isolation of the OTSGs after a steam line failure, provides overpressure relief capacity in the event of accidents, provides pressure control for decay heat removal in case of accidents, provides steam to the Emergency Feedwater (EF) System as required for accidents and provides capability for RCS cooldown following a steam generator tube rupture event. As discussed in Sections 4 and 5 above, the MS system will support the MUR and the safety functions of this system are not impacted by the uprate. The MS System also functions during normal operation. While steam flow increases with increasing power, no changes in design are required and all parameters remain within design. (Reference 6.3.2)

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 48 of 55 6.2.1.2 Steam Dump The CR-3 equivalent of a steam dump system includes the Atmospheric Dump Valves (ADVs) and Turbine Bypass Valves (TBVs).

6.2.1.2.1 Atmospheric Dump Valves An ADV is located in each of the Main Steam Lines, upstream of the MSIVs (MSV-25 in Steam Line A-1 and MSV-26 in Steam Line B-2). The valve function is to provide a controlled path for venting of main steam to the atmosphere. These valves were evaluated for power uprate impact on three functions: (1) close to isolate containment; (2) open and modulate to relieve steam to the atmosphere; and, (3) maintain pressure boundary to transport steam to safety and non-safety related loads. There are no changes in function. Power uprate conditions are bounded by existing design. The evaluation concludes the functional performance requirements of the Main Steam ADVs will be unaffected by the power uprate. (Reference 6.3.2) 6.2.1.2.2 Turbine Bypass Valves Four TBVs are located in the Main Steam lines downstream of the MSIVs. The Turbine Bypass Valves are piston-operated globe valves which actuate in response to a hand generated signal or an Integrated Control System (ICS) generated signal. The valves primary function is to maintain stable turbine header pressure during load swing events. The flow rate is not being changed significantly and the function of the TBVs is not being changed. For the power uprate, the ICS control will use the existing TBVs. Power uprate parameters are bounded by existing design conditions at 102% (2619 MWt). There is no impact on the TBVs for the MUR power uprate.

(Reference 6.3.2)

The MUR power uprate conditions remain bounded by the design basis of the CR-3 FSAR.

6.2.1.3 Condensate (CD) System The primary function of the CD system is to supply preheated condensate to the FW System.

The Condensate system was evaluated for a power uprate from 2568 MWt to 2609 MWt. The uprate will have no impact on the design functions of the CD system. The condenser load limiting back pressure is 9 inches of mercury (" Hg) absolute and the current maximum operating pressure has been 3-4" Hg absolute. No design changes will be required. (Reference 6.3.2)

The feedwater heaters were evaluated and determined to be adequate for the 2609 MWt operating conditions. No operational changes are required. The Condensate system analysis of record is not impacted by the MUR power uprate. (Reference 6.3.2) 6.2.1.4 Main Feedwater (FW) System The FW System provides isolation capability of the feedwater during accidents. It also provides feedwater to the OTSGs during normal operation. The CR-3 accident analyses are discussed in Sections 2 and 3 above. The safety functions of this system are not impacted by the uprate. In addition, the main feedwater pumps and the booster pumps have been determined to have

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 49 of 55 adequate margin for a 1.6% power increase. The feedwater heaters were evaluated and determined to be adequate for the 2609 MWt operating conditions. No changes in design are required and all parameters remain within design. The FW system will therefore support a power uprate to 2609 MWt. (Reference 6.3.2) 6.2.2 Containment Systems (RIS 2002-03 Section VI.1.B)

The containment systems include the building spray system, penetrations and hatches. The building spray system's functions are to remove fission products from the post-accident containment atmosphere, and to assist in post-accident pressure and temperature control. The safety function of the penetrations and hatches is to maintain containment integrity under accident conditions. As indicated in Sections 2 and 3 above, the transients continue to be maintained within design limits. As such, these systems are not impacted by the MUR.

(Reference 6.3.3) 6.2.3 Safety-Related Cooling Water Systems (RIS 2002-03 Section VI.1.C) 6.2.3.1 Decay Heat Closed Cycle Cooling (DC)

The DC System removes heat from the reactor core via the Low Pressure Injection/Decay Heat system as well as various pumps and motors following a LOCA and transfers it to the Raw Water (RW) system. The applicable CR-3 accident analyses were evaluated at 102% reactor thermal power and bound the 1.6% power increase (Sections 2 & 3). Therefore, the safety functions of this system are not impacted by the uprate. There are no design changes required.

As such, this system is not impacted by the MIUR. (Reference 6.3.2) 6.2.3.2 Nuclear Services Closed Cycle Cooling (SW)

The SW System removes heat from various safety-related equipment following ES actuation and transfers this heat to the RW system. The applicable CR-3 accident analyses were evaluated at 102% reactor thermal power and bound the 1.6% power increase (Sections 2 and 3). Therefore, the safety functions of this system are not impacted by the uprate. There are no design changes required. As such, this system is not impacted by the MUR. (Reference 6.3.2) 6.2.3.3 Nuclear Services & Decay Heat Seawater (RW)

The RW System provides cooling water to the SW and DC Systems for heat removal during accidents and normal operation. The CR-3 accident analyses were evaluated at 102% reactor thermal power and bound the 1.6% power increase (Sections 2 and 3). Therefore, the safety functions of this system are not impacted by the uprate. There are no design changes required.

As such, this system is not impacted by the MUR. (Reference 6.3.2) 6.2.3.4 Emergency Feedwater System (EF)

The EF System provides emergency feedwater in the event of loss of main feedwater. The CR-3 accident analyses were evaluated at 102% reactor thermal power and bound the 1.6% MIUR power uprate (Sections 2 & 3). There are no design changes required for the EF system to

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 50 of 55 operate at 2609 MWt. There are no design changes required. As such, this system is not impacted by the MIUR. (Reference 6.3.3) 6.2.4 Spent Fuel Pool Storage and Cooling Systems (RIS 2002-03 Section VI.1.D)

The principal function of the Spent Fuel (SF) system is to provide for the cooling and storage of irradiated fuel. The system is described Section 9.3 of Reference 6.3.1. The functions of the system were reviewed and were found to be unaffected by the MUR uprate. There are no design changes required. As such, this system is not impacted by the MUR. (Reference 6.3.3) 6.2.5 Radioactive Waste Systems (RIS Section 2002-03 Section VI.1.E)

The Waste Decay (WD) system provides the means to sample, collect, process, store/hold, re-use or release gaseous and liquid low-level effluents generated during normal operation. The WD system consists of the gaseous waste disposal (WD-GW) and the liquid waste disposal (WD-LW) sub-systems. These systems are discussed below.

6.2.5.1 Gaseous Waste Disposal (WD-GW)

The WD-GW system is used to control low-level gas releases to the environment, and to permit the venting of excess gas to the Reactor Building in a post-accident situation. Portions of the system are required to be operational and intact to provide containment isolation upon an Engineered Safeguard (ES) actuation signal. This system is unaffected by the MUR uprate.

There are no design changes required. As such, this system will support the MUR. (Reference 6.3.3) 6.2.5.2 Liquid Waste Disposal (WD-LW)

The WD-LW system is required to collect, store and process, for disposal or reuse, radioactive liquid waste. The WD-LW system provides a means to process radioactive liquid waste prior to release. The WD-LW system tank volumes and processing components (i.e., demineralizers and filters) capacity are adequate to process radioactive liquid waste prior to release. The volume of liquid waste is primarily dependent on RC bleed and SG draindown as well as leakage from various components; the volume generated during normal operation is not expected to change due to the uprate. However, the activity of fission products in the liquid waste is dependent on the power level and will increase slightly due to the uprate. However, the impact of the higher activity on the operation of the WD-LW system will not be significant. The resins in the demineralizers may require replacement or regeneration at slightly higher frequencies, which would affect the volume of generated solid waste slightly, but this would not be a constraint to implementing the uprate. There are no design changes required. As such, this system will support the MIUR. (Reference 6.3.3) 6.2.6 Engineered Safety Features (ESF) Heating, Ventilation and Air Conditioning Systems (RIS 2002-03 Section VI.1.F)

The ESF Heating, Ventilation, and Air Conditioning Systems remain bounded by the design basis (102% of 2568 MWt) of the CR-3 FSAR (Reference 6.3.1) for MUR power uprate

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 51 of 55 conditions. The CR-3 accident analyses were evaluated at 102% reactor thermal power and bound the 1.6% power increase (Sections 2 and 3). Therefore, the safety functions of these systems are not impacted by the uprate. There are no expected changes in containment cooling operation at the MUR uprate power level. The containment accident analysis has been performed at a bounding power level with the containment air coolers and fan flow rates and found to be acceptable. The containment cooling system has adequate margin to cool the containment at MUR conditions. There are no design changes required for any of the ESF Heating, Ventilation, and Air Conditioning Systems. As such, these systems will support the MUR. (Reference 6.3.4) 6.3 Section 6.0 References 6.3.1 CR-3 Final Safety Analysis Report, Rev. 30.1.

6.3.2 51-9036486-000, "CR-3 MUR Power Uprate (2609 MWt) BOP Mechanical Systems Review."

6.3.3 51-9041016-000, "CR-3 MUR Power Uprate (2609 MWt) NSSS Mechanical Systems Review."

6.3.4 51-9036250-000, "Crystal River 3 MUR HVAC System Evaluation Report."

7.0 Other (RIS 2002-03 Section VII Questions)

1. A statement confirming that the licensee has identified and evaluated operator actions that are sensitive to the power uprate, including any effects of the power uprate on the time available for operator actions.

2. A statement confirming that the licensee has identified all modifications associated with the proposed power uprate, with respect to the following aspects of plant operations that are necessary to ensure that changes in operator actions do not adversely affect defense in depth or safety margins:

A. emergency and abnormal operating procedures B. control room controls, displays (including the safety parameter display system) and alarms C. the control room plant reference simulator D. the operator training program

3. A statement confirming licensee intent to complete the modifications identified in Item 2 above (including the training of operators), prior to implementation of the power uprate.

4. A statement confirming licensee intent to revise existing plant operating procedures related to temporary operation above "full steady-state licensed power levels" to reduce the magnitude of the allowed deviation from the licensed power level. The magnitude should be reduced from the pre-power uprate value of 2 percent to a lower value corresponding to the uncertainty in power level credited by the proposed power uprate application.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 52 of 55

5. A discussion of the 10 CFR 51.22 criteria for categorical exclusion for environmental review including:

A. A discussion of the effect of the power uprate on the types or amounts of any effluents that may be released offsite and whether or not this effect is bounded by the final environmental statement and previous Environmental Assessments for the plant.

B. A discussion of the effect of the power uprate on individual or cumulative occupational radiation exposure.

7.1 Response to RIS 2002-03 Section VII Questions 7.1.1 Operator Actions (RIS 2002-03 Section VII.1)

Operator actions that are sensitive to the power uprate, including any effects of the time available for operator actions were reviewed. It was determined that the additional power increase afforded by the MUR will have no impact on operator actions 7.1.2 Modifications Associated With the Power Uprate (RIS 2002-03 Section VII.2) 7.1.2.1 Emergency and Abnormal Operating Procedures (RIS 2002-03 Section VII.2.A)

Emergency and Abnormal Operating Procedures (EOPs and AOPs) have been reviewed for potential impact from the proposed power uprate. No adverse impact on these procedures with this power uprate should occur. Any EOP changes will be documented and implemented as part of the Engineering Change process.

7.1.2.2 Control Room Controls, Displays (including safety parameter display system) and Alarms (RIS 2002-03 Section VII.2.B)

The following changes/modifications are associated with implementation of the power uprate that affect control room controls:

" Power Production Heat Transfer will be increased slightly but within the ability of the operator to maintain prescribed parameters less than the required limits, thus having no adverse effect on defense in depth or safety margins.

" Cooldown time on Decay Heat is increased slightly, which increases the amount of time the operators must control the cooldown of the RCS. The increase in the amount of time Decay Heat is controlled is within the ability of the operators thus, having no adverse effect on defense in depth or safety margins.

" Changes to be made to the calibration of the nuclear instrumentation due to power uprate will be accommodated by corresponding changes being made to the Integrated Control System (ICS) due the range of the instruments remaining the same whether maintaining 2568 MWt or 2609 MWt. The change in nuclear instrumentation calibration will have no effects on any control room controls or the operator's ability to monitor core power production, thus having no adverse effect on defense in depth or safety margins.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 53 of 55

Required changes to the settings of the pre-identified ICS modules are associated with maintaining the plant within normal operating parameters, thus having no adverse affect on defense in depth or safety margins.

The following modifications are associated with implementation of the power uprate that affect operator displays (including Safety Parameter Display System):

" The Fixed Incore Monitoring System functions may require changes to the plant computer software. The software changes will be transparent to the operators; their response to abnormal indications by the software will remain unchanged, thus having no adverse affect on defense in depth or safety margins.

" The AULD Panel PC displays will be modified to provide either the improved calorimetric or the existing calorimetric values on the main display. The operator will use the modified AULD display to select which calorimetric is to be used in the AULD as the controlled parameter used to establish thermal demand. The AULD display, in conjunction with the plant annunciator, will also alert the operator when the AULD has automatically transferred out of Automatic upon the detection of a sufficient differential in the available secondary heat balance calculations. When this occurs, the operator will be required to override this transfer by adjusting the differential value to return the AULD to Automatic using the improved calorimetric. In any event, the AULD will be allowed to be placed in Automatic using the existing calorimetric with a maximum core thermal power setpoint of 2568 MWt.

The following modifications are associated with implementation of the power uprate that affect alarms:

A control room alarm has been added to alert the operators when the LEFM system has self diagnosed a condition that has resulted in an internal alert or failure.

The following modifications are required to support the MUR, but are not expected to otherwise have significant operational impact:

" Installation of the Caldon system.

" Addition of new Feedwater and Main Steam pressure and Main Steam Temperature instrumentation.

" Modification of AULD software.

7.1.2.3 Control Room Plant Reference Simulator (RIS 2002-03 Section VII.2.C)

The Control Room plant reference simulator will be modified due to this uprate. While there is minimal impact on plant response due to this uprate, changes are needed to be properly modeled on the simulator. These include modifying Integrated Control System function curves to match the increased power output of the plant as well as the equipment modifications discussed above.

The simulator modifications will be completed in time for operator training support.

7.1.2.4 Operator Training Program (RIS 2002-03 Section VII.2.D)

The Operations Department has been integrated into the uprate process. An Operations Department representative has been assigned to the uprate team. The design change process requires the Operations Department review and sign-off on design change packages.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 54 of 55 The Operator Training program will need to be modified due to this uprate. While there will be minimal impact due to this uprate, changes are being made that the operator will need to be properly trained on. Training on operation and maintenance of the Caldon LEFM CheckPlus Tm System, will be developed and completed prior to implementation of the MIUR.

7.1.3 Confirmation of Intent to Complete Modifications and Training (RIS 2002-03 Section VII.3)

FPC will complete all modifications identified above (including the training of operators), prior to implementation of the power uprate.

7.1.4 Confirmation of Intent to Revise Operating Procedures Related to Power Levels (RIS 2002-03 Section VII.4)

FPC will revise existing plant operating, maintenance, alarm response, and other procedures to reflect the modifications noted above, as well as appropriate administrative controls necessary to assure timely response to the loss of equipment availability.

7.1.5 10 CFR 51.22 Discussion (RIS 2002-03 Section VII.5) 10 CFR 51.22(c)(9) provides criteria for, and identification of, licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (A. 1) involve a significant hazards consideration; (A.2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite; or, (B) result in a significant increase in individual or cumulative occupational radiation exposure.

7.1.5.1 Environmental Assessment (RIS 2002-03 Section VII.5.A)

It has been determined that this License Amendment Request (LAR) meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the proposed license amendment. The basis for this determination is as follows:

1. The proposed license amendment does not involve a significant hazards consideration as described previously in the no significant hazards evaluation (Attachment 1) for this LAR

2. The proposed changes will allow CR-3 to operate at an uprated power level of 2609 Megawatts Thermal (MWt). This represents an increase of approximately 1.6 percent over the current 100 percent power level of 2568 MWt.

The proposed changes do not significantly impact installed equipment performance or require significant changes in system operation. Changes in maintenance and operational practices will not impact the release of solid, liquid or gaseous effluents. The specific activity of the primary and secondary coolant is expected to increase by no more than the percentage increase in power level. Therefore, the amount and specific activity of solid waste is not expected to change significantly.

U.S. Nuclear Regulatory Commission Attachment D 3F0607-05 Page 55 of 55 Gaseous and liquid effluent releases are expected to increase from current values by no more than the percentage increase in power level. Offsite release concentrations and doses will continue to be maintained within the limits of 10 CFR 20 and 10 CFR 50, Appendix I, in accordance with the requirements of the CR-3 Offsite Dose Calculation Manual (ODCM). The ODCM contains offsite dose calculation methodologies, the radioactive effluent controls program, and radiological environmental monitoring activities. The ODCM contains the methodologies and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, the methodologies and parameters used in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and the controls for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable in accordance with 10 CFR 50.36a. The proposed changes will not result in changes in the operation or design of the gaseous, liquid or solid waste systems, and will not create any new or different radiological release pathways.

Therefore, the proposed license amendment will not result in a significant change in the types or increase in the amounts of any effluents that may be released off-site.

7.1.5.2 Environmental Assessment (RIS Section 2002-03 Section VII.1.5.B)

The proposed changes will not cause radiological exposure in excess of the dose criteria for restricted and unrestricted access specified in 10 CFR 20. Radiation levels in the plant are expected to increase by no more than the percentage increase in power level. Individual worker exposures will be maintained within acceptable limits by the CR-3 as-low-as-reasonably-achievable (ALARA) program. Therefore, the proposed license amendment will not result in a significant increase to the individual or cumulative occupational radiation exposure.

8.0 Changes to Technical Specifications, Protection System Settings, and Emergency System Settings (RIS 2002-03 Section VIII Questions)

I. A detailed discussion of each change to the plant's technical specifications, protection system settings, and/or emergency system settings needed to support the power uprate:

A. a description of the change B. identification of analyses affected by and/or supporting the change C. justification for the change, including the type of information discussed in Section 3, above, for any analyses that support and/or are affected by change 8.1 Response to RIS 2002-03 Section VIII Questions 8.1.1 There will be no technical specification changes resulting from this LAR, other than the changes to Section 1, DEFINITIONS.

8.1.2 There will not be any protection system or emergency system setpoint changes resulting from this LAR, although several instruments will require rescaling in order to support implementation of this LAR.

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #296, REVISION 1 ATTACHMENT E UNCERTAINTY CALCULATION

20697-10 (3/30/06)

CALCULATION

SUMMARY

SHEET (CSS)

AREVA Document Identifier 32- 9042687 -001 Title Heat Balance Uncertainty for CR-3 MUR PREPARED BY: REVIEWED BY:

METHOD: X DETAILED CHECK F1 INDEPENDENT CALCULATION NAME J A Weimer NAME Suzanne Palmer SIGNATURE a6/ tf/* .___ SIGNATURE J&A jL Qi -

TITLE Advi ry Engineer DATE / TITLE Principal Engineer DATE 41111abb1 COST REF. TM STATEMENT:

CENTER 41306 PAGE(S) 24 REVIEWER INDEPENDENCE NAME . i 4.'fE C ;.

PURPOSE AND

SUMMARY

OF RESULTS:

PURPOSE The objective of this calculation is to determine the full-power reactor core power uncertainty value, also referred to as the "heat balance uncertainty," for Crystal River Unit 3 (CR-3) based on the installation of CALDON ultrasonic feedwater flow metering equipment and other new secondary side instruments and equipment. The "Square Root Sum of the Squares" (SRSS) approach will be used in this calculation along with request from CR-3 to apply some of the uncertainties as "biases" (algebraically added to the SRSS total). This calculation will also (1) predict the nominalRC pump heat into the RCS and the uncertainty associated with pump heat, and (2) the energy added to the MU tank flow from the makeup pump.

RESULTS The ASME Performance Test Code Methodology was used to calculate the expected core thermal power uncertainty to be achieved using the Caldon CheckPlusTM System ultrasonic flow meter. The analysis concluded that using the following instrument uncertainty values, the core thermal power uncertainty would be 0,394% of 2609 Mwt, thus allowing a power uprate of 1 6% to be pursued. The feedwater flow and temperature measurement is the bulk of this uncertainty (0 34% absolute and -84% of the total uncertainty) The new steam temperature/pressure instrumentation results in -4% of the total uncertainty and the steam pressure measurement uncertainty is -1% of the total. The RC pumps energy uncertainty, ambient loss uncertainty and an atmospheric pressure correction uncertainty were chosen to be treated as a bias (algebraically added and not SRSS) and they are -11% of the total uncertainty (see Attachment 1). After the final feedwater flow/temperature uncertainty is determined for the CR.-3 specific equipment (post fabrication testing), the total uncertainty may be reduced.

Other pertinent output of this calculation include (1) the RC pump power (to be used in the HB equation) is 20.96 MWt, (2) the temperature increase due to the MU pump to be added to the makeup measured temperature is 5.6F, (3) the pressure to be used for MU enthalpy is a constant (2400 psia), and the letdown pressure will be the measured RCS pressure minus 20 psi (noting that this input has essentially no impact on the final HB uncertainty),

Revision 1 replaces revision 0 (editorial changes).

THE DOCUMENT CONTAINS ASSUMPTIONS THAT MUST BE VERIFIED PRIOR TO USE ON

'THE FOLLOWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT: SAFETY-RELATED WORK CODENERSION1REV

CODENERSION/REV x YES NO AREVA NP Inc., an AREVA and Siemens company Page I of 36

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR RECORD OF REVISION Revision Description Date 00 Original release 04/2007 01 Minor editorial changes 04/2007 2

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR TABLE OF CONTENTS 1.0 PURPOSE 4 2.0 ASSUMPTIONS 4 2.1 Key Assumptions 4 2.2 Other Assumptions 4 3.0 METHODOLOGY 6 3.1 Industry Standard 6 3.2 AREVA NP Experience 6 3.3 Crystal River 3 Heat Balance Equations 7 3.4 Crystal River 3 Heat Balance Instruments 7 4.0 INPUTS TO THE UNCERTAINTY EVALUATION 7 4.1 Referenced Instrumentation Inputs 8 4.1.1 CALDON Equipment 8 4.1.2 New CR-3 Installed Instrumentation 8 4.1.3 Letdown/Makeup Energy 9 4.1.4 Ambient Heat Losses 11 4.2 Calculated Uncertainty Inputs 11 4.2.1 RC Pump Heat Input and Uncertainty 11 4.2.2 Partial Derivatives at Nominal Full Power Heat Balance Parameters 18 5.0 CALCULATION OFHEATBALANCE UNCERTAINTY 20 6.0

SUMMARY

OFRESULTS 23

7.0 REFERENCES

24 A TTACHMENT 1 - HEAT BALANCE SPREADSHEET 26 A TTA CHMENT 2 - CALDON UNCERTAINTY INPUTS -PRELIMINAR Y VAL UES 227 ATTACHMENT 3 TRANSCRIBED RC PUMPDATA 28 A TTA CHMENT 4 ATMOSPHERIC PRESSURE DATA FROM CR-3 36 3

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR 1.0 PURPOSE The objective of this calculation is to determine the full-power reactor core power uncertainty value, also referred to as the "heat balance uncertainty," for Crystal River Unit 3 (CR-3) based on the installation of Caldon ultrasonic feedwater flow metering equipment and other new secondary side instruments and equipment. The "Square Root Sum of the Squares" (SRSS) approach will be used in this calculation along with request from CR-3 to apply some of the uncertainties as "biases" (algebraically added to the SRSS total). This calculation will also (1) predict the nominal RC pump heat into the RCS and the uncertainty associated with pump heat, and (2) the energy added to the MU tank flow from the makeup pump.

2.0 ASSUMPTIONS 2.1 Key Assumptions There are no key assumptions that need to be verified to use the results of this calculation. The assumptions used in these calculations are presented in this section.

2.2 Other Assumptions (1) Since CR-3 will modify the Automatic unit load demand (AULD) to simulate the online computer program "Fixed Incore Detector Monitoring System" (FIDMS - to be installed),

this heat balance uncertainty will be applicable to the heat balance performed in both calculations.

(2) All the uncertainties discussed herein are in absolute values (i.e., psi absolute for steam pressure at the maximum possible value, not percent of full power pressure). Therefore, the full power values are not pertinent to the uncertainty values. However, since the partial derivatives used in the uncertainty calculation require approximate full power values, they will be presented or calculated herein. Since these derivatives (i.e., aH/aTFW) will not change significantly within 10% of the anticipated values, these approximate plant values at full power are acceptable.

(3) Letdown mass flow rate will be assumed equal to makeup (MU) plus seal injection (SI).

This is a reasonable steady state assumption since pressurizer level is essentially constant. Also, since the contribution of the MU/LD energy to the total heat balance uncertainty is -0.5% after SRSS (-0.5% of 0.39% or 0.002% of the total HB uncertainty),

a small variation in the MU vs. LD flow will be negligible on the final uncertainty.

(4) Since the ambient RCS loss is not a "measured" value and therefore has no "instrument" uncertainty, the uncertainties associate with this value will be estimated in this document.

CR-3 has requested that the ambient heat loss uncertainty be applied as a bias and not be SRSS with the other random uncertainties.

(5) The RC pump heat into the RCS is based on the brake horsepower test values of the original pump impellers. Based on BHP test data from different Byron Jackson (BJ) pumps at CR-3 and other B&W plants, it is assumed that the BHP of replacement impellers at CR-3 (also made by BJ) are very similar to be the original BJ impellers (tested). The similarity of BJ pump impellers is shown below.

(6) The pump power input is based on the BHP test data which in turn is based on cold leg temperature, pump flow rate, and any test instrument uncertainty. The uncertainty in the BHP (impeller characteristics) will be based on the uncertainty in the testing instrumentation (of which DB-1 and ANO-1 testing accuracies were found and used for 4

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR CR-3). The RC flow uncertainty on BHP at CR-3 is shown below to be negligible. Since lower cold leg temperatures are conservative for the BHP power in the HB equation, the actual BHP at full power will be based on a maximum expected cold leg temperature (in all four cold legs) at full power. This is 559F (2F greater than the nominal expected 557F).

(7) Water/steam properties in AULD will be made to be essentially identical to the values generated in FIDMS and therefore no uncertainties or adjustments will be needed to account for water/steam properties. Water properties used in the file are based on Reference 12 which is also the basis for the 1997 ASME published values (noting that the IFC-67 standard used in the ASME 1967 version has been declared obsolete by the ASME).

(8) The 14.7 psi (22.93 in Hg) constant added to the gage pressure to convert to absolute pressure (psia) will vary slightly as atmospheric pressure varies. CR-3 has chosen to address this error and apply it as a bias (not a random, independent uncertainty).

Attachment 4 shows a summary of 7 years worth of average barometric pressure indications in central .Florida (as provided by CR-3). This data shows the maximum pressure in this time period was 0.67 in Hg greater than 29.93 and CR-3 has chosen to bias this impact in the HB uncertainty. It is applied only for steam properties only since this pressure difference will impact FW enthalpy or RCS cold leg enthalpies insignificantly.

Should all the steam pressure transmitters be calibrated during a high atmospheric pressure, the steam pressure could be in error by .67 in Hg (.33 psi). Using the 1.29E6 BTU/hr/psi (1.284E6 calculated in section 5.0 below), this is an error of 425700 BTU/hr.

This results in a HB uncertainty of 4.257E5/8.905E9 BTU/hr or .000048 (.0048%).

(9) The FW pressure has previously never been measured near the SG inlet at CR-3 and therefore no plant data is available for the nominal full power value (at the new FW pressure instrument is located at the FE-34A and FE-34B flow measurement locations).

The indicated full power FW pressure is estimated to be -980 psia at the new measurement location (engineering judgment). This does not impact the uncertainty on this pressure since it is based on 1200 psig (maximum range pressure). Also, since the partial derivative used for the impact on heat balance is based on a range from 925 to 1025 psia, this assumption will have no impact on the partial derivative and consequently no impact on the final HB uncertainty.

(10) Reference 8 shows the expected steam pressure at the bundle outlet is 930 psia.

Reference 9 shows that the pressure drop from the tube bundle outlet through the SG nozzle is -5 psid. The velocity head in the steam pipe is -5 psi. This results in a steam pressure of -920 psia in the piping just outside the SG. Note that the uncertainty for this value is conservatively based on 1000 psig (and not percent of full power) and therefore this parameter is presented only to show expected range at full power. This is used only to determine the partial derivatives. These partial derivatives will not change substantially in the 800-1000 psi range and is therefore an acceptable assumption for this calculation.

(11) No "pressure" location correction is needed for the steam temperature indication since they are assumed to be within -2 to 3 feet of the steam pressure indication. This assumption has been verified verbally and the appropriate drawing will be referenced when it is available.

(12) These uncertainty calculation are for the Caldon LEFM equipment fully operational (not in the maintenance mode).

5

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR 3.0 METHODOLOGY A discussion of heat balance uncertainty methodology used for the CR-3 heat balance uncertainty is presented herein.. This information is presented to show the acceptability of the "Square Root Sum of the Squares" (SRSS) approach for this type of calculation. Some of the uncertainties were chosen to be biased uncertainties (a non-random constant error in the HB equation).

3.1 Industry Standard The ASME provides a standard methodology for estimating instrument-related uncertainties, Reference 1. Both individual instruments as well as resultants from multiple instruments are treated in this reference.

3.2 AREVA NP Experience AREVA NP has performed secondary heat balance calculations including uncertainty calculations for secondary thermal power, core thermal power, and RCS flow for a number of B&W plants.

The methodology used in these calculations is consistent with those of the ASME, Reference 1.

The governing equation is presented and then differentiated with respect to the contributing measurements. The products of the partial derivatives and individual measurement uncertainties are squared, summed, and then square-rooted to solve for the core thermal power uncertainty.

For example, from Reference 6, the uncertainty in steam generator secondary power is:

E(Q) = [(aQ/8Wtw X 6wfw) 2 + (aQIaTs x ETs) 2 + (aQ/aTfwx STfw)2 + (pQ/aPsx 6Ps) 2 + (aQ/Pafw x Spfw)2]°5 Where E(Q) = steam generator thermal power uncertainty Q = steam generator thermal power Wfw = feedwater flow Ts = steam temperature Tfw = feedwater temperature Ps = steam pressure Pfw = feedwater pressure 6= measurement uncertainty for feedwater flow, feedwater pressure, feedwater temperature, steam pressure, and steam temperature Since the FW flow and temperature uncertainties were combined by CALDON (Attachment 2) into one power uncertainty (vwfwrr) in BTU/hr, the equation becomes 2

E(Q) =[ cwrr)2 + (aQ/aTs x ETs)2 + (pQ/aPsx EPs) + (aQ/aPr, x Epfw)2]05 The nominal heat balance uncertainty would be:

E2 (core) =[EQ2 + ERCpump energy 2 + Eambient heat loss 2 + EMakeup/letdown 2 ] if all components were assumed independent and random (with the secondary heat balance "EQ " comprising the large majority of the uncertainty).

Since the measured pressures (in particular the steam pressure) require a conversion to psia by adding the atmospheric pressure, CR-3 has requested an atmospheric pressure uncertainty be added to this calculation. They also requested that the ambient heat loss uncertainty and the RC pump energy uncertainty be treated as bias error (added algebraically to the SRSS error).

This final error in the heat balance uncertainty equation will be E (core) =[EQ2 + EMakeup/letdown 2 ]0 5 + Eambient heat loss + EAtmos press + ERC Pump 6

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR 3.3 Crystal River 3 Heat Balance Equations Crystal River 3 uses software in the AULD module to calculate core thermal power. This calculation is based on the following equation:

Q core = WfwA (HsA - HFwA)+ WFWB (HsB - HFWB )+ QLD-MU - QRCP + QLOSS Where; WFWA, WFWB Feedwater flows in Loop A & B HSA, HFWA, HSB, HFWB Steam & feedwater enthalpies for Loops A & B QLD-MU Heat loss due to primary side letdown flow QRCP Heat added due to RC pumps QLOSS Ambient heat losses from the RCS 3.4 Crystal River 3 Heat Balance Instruments A listing of Crystal River 3 AULD computer points that are input to the current core thermal power calculation is provided for information In Table 1.

Table I - Crystal River 3 Heat Balance Input Listing Instrument Description Units Range FW 1 COMP FLOW, Caldon KPPH FW 2 COMP FLOW, Caldon KPPH RCS LETDOWN FLOW GPM 0-160 (Ref. 25)

RCS PRESSURE PSIG 1700-2500 (Ref 15,19)

TEMP @ LETDOWN FLOW MEASUREMENT Deg F 0-200(Ref. 23)

MAKEUP TEMPERATURE (AT MU TANK) Deg F 40-200(Ref. 23)

LETDOWN TEMPERATURE (AT CL LD NOZZLE) Deg F 520-620(Ref. 24)

SG 1 FW PRESS FW-297-PT PSIG 600-1200(Ref. 7)

SG 1 OUT STM PRESS, MS-114-PT PSIG 800-1000(Ref. 4)

SG 1 OUT STM PRESS, MS-1 15-PT PSIG 800-1 000(Ref. 4)

SG 2 FW PRESS FW-298-PT PSIG 600-1200(Ref. 7)

SG 2 OUTLET STM PRESS, MS-1 16-PT PSIG 800-1000(Ref. 4)

SG 2 OUTLET STM PRESS, MS-1 17-PT PSIG 800-1000(Ref. 4)

SG 1 OUTLET STM TEMP MS-001-TE1,MS-001-TE2 Deg F To 610F(Ref. 5)

SG 1 OUTLET STM TEMP MS-002-TE1,MS-002-TE2 Deg F To 610F(Ref. 5)

SG 2 OUTLET STM TEMP MS-003-TE1,MS-003-TE2 Deg F To 610F(Ref. 5)

SG 2 OUTLET STM TEMP MS-004-TE1,MS-004-TE2 Deg F To 610F(Ref. 5)

SG 1 FW TEMP Caldon Deg F >250F SG 2 FW TEMP Caldon Deg F >250F

The range of this instrument (at the specified accuracy) is limited only to the feedwater being above 250TF (Reference 3) 4.0 INPUTS TO THE UNCERTAINTY EVALUATION The inputs to the final heat balance uncertainty are (1) instrument uncertainties, (2) non instrument uncertainties (pump heat, ambient losses, atmospheric pressure), and (3) the full 7

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR power partial derivative. These three groups of inputs are taken from references or calculated below.

4.1 Referenced Instrumentation Inputs 4.1.1 CALDON Equipment The Caldon LEFM CheckPlusTM System ultrasonic feedwater flow meter provides a measurement of the feedwater flow and feedwater temperature (Reference 17) with appropriate reference page in Attachment 2. The preliminary values used are for the combine temperature and flow uncertainty converted to a power uncertainty:

Feedwater Flow Rate Combined with Feedwater Temperature Uncertainty = +/-0.34% of nominal Power This uncertainty is based on the total power (both SGs combined).

Note that the instrument specific value is expected to decrease when final testing of these instruments is performed. The value used herein is a nominal (maximum) value expected.

4.1.2 New CR-3 Installed Instrumentation The following uncertainties for steam temperature, steam pressure, and feedwater pressure were provided by Crystal River 3, References 4 (pg 6), 5 (pg 6), and 12. These are new instrumentation installed at CR-3.

The feedwater pressure instrument is in the FW piping approximately 150 feet (with pressure losses due to friction and 20 elbows) before the feedwater ring header and at the same approximate elevation (see Assumption 9). The FW pressure at the point where primary energy is being added (at the spray nozzle) is the most correct pressure to use for the FW enthalpy, however, the FW enthalpy dependency on FW pressure is very small and the location of the pressure input will not change the heat balance significantly. This DP will have no impact on the partial derivatives calculated herein for the final heat balance uncertainty.

The steam pressures and temperatures are measured approximately just before the piping to the MSSVs (4 places). The pressure at point where primary energy stops being added to the FW is just inside the SG steam nozzles will be 12 psi higher than the measured point. This change will not impact the heat balance as long as the pressure and temperature are measured at approximately the same location (which it is based on assumption 11) and therefore will have no impact on the final heat balance uncertainty.

Steam Temperature Uncertainty (per instrument) = +/-2.02°F (up to 61 0°F) (see Reference 5, eight Indications per Table 1).

Steam Pressure Uncertainty = +/-4.57 psi (uncertainty of maximum range [200 psid] at 1000 psig with four indications per Table 1, and Reference 4).

Feedwater Pressure Uncertainty = 7.11 psi (per FW line at maximum range [600 psid] at 1200 psig and one indication per Table 1, and Reference 7). As discussed above in the assumptions section, the atmospheric pressure change will be insignificant on this portion of the calculation.

The steam temperature is monitored by eight RTDs. There are two RTDs at each measurement location and the combined uncertainty using the square root sum of the squares is (2*2.022)05/2 or 1 .430F in each location. Since both SG outlet pipe temperatures are the same, combining the

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR two pipes results in (2*1.432)05/2 or 1.01OF uncertainty for each SG. Combining the two pressure sensors for each SG, the uncertainty is (2*4.572)°5/2 or 323 psi uncertainty for the total heat balance per SG.

Summarizing the Total Uncertainty of pressure and temperature per SG (statistically combining the two steam line values)

Steam Temperature (per SG) = +/-1.01°F (up to 610°F)

Steam Pressure (per SG) = +/-3.23 psi (up to 1200 psig)

Feedwater Pressure has only one indication and will not change from above (7.11 psi)

Since the FW flow and temperature uncertainties are for both SGs (total RCS uncertainties), as are the letdown flow, RC pump uncertainties, the other secondary heat balance parameters will also be calculated for both SGs. Therefore, when the heat balance is performed on the entire RCS, all the secondary side instruments can be used (and statistically combined again). The resulting uncertainties are; Total Uncertainty (statistically combining the two steam line values)

Steam Temperature (total both SGs) = 0.71°F (up to 610 0 F)

Steam Pressure (total both SGs) = 2.28 psi (up to 1000 psig)

Feedwater Pressure (total both SGs) 5.03 psi (up to 1200 psig) 4.1.3 Letdown/Makeup Energy The letdown energy is based on the letdown flow times the LD-MU energy (enthalpy). The equation for this uncertainty is; E(QLD) = [(OQ/oWLD X 6WLD)2 + (aQ/DTLD X GTLD)2 + (aQ/aTMu X rTMU) 2 + (aQ/CPLD X SPLD)2 ]05 This assumes that the MU plus SI flow (lb/hr) is equal to the letdown flow (see assumption 3).

Where E(QLD) = Letdown energy loss uncertainty Q LD = Letdown thermal power Loss WLD = Letdown Flow (mass flow)

TLD = Letdown temperature (Cold leg temperature)

TMU = Makeup temperature PLD = Letdown Nozzle Pressure in RCS S= measurement uncertainty for letdown flow, letdown temperature, and makeup temperature.

Temperature Inputs The makeup temperature is a measured parameter, however HPI/MU pump heat added at the typical flow rates is not insignificant. Reference 20 shows the MU/HPI pump curves at CR-3. The pertinent data from these curves is the efficiency and total head. Reference 21 notes that the minimum allowed pump flow (recirculation back to the MU tank) is 100 gpm. With the nominal 35000 lb/hr MU/SI flow rate (see Table 2 below), this is also -70 gpm SI/MU flow resulting in 170 9

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR gpm total pump flow. At 170 gpm Reference 20 shows the pump efficiency is between 49% and 52% and a pump head of 6400 ft.

Per Reference 22, the pump heat input is calculated as T(degree rise) = [total head (ft)]/778 (1/efficiency -1)

TI = 6400/778(1/.5 - 1) = 8.22 x 1 = 8.2F At 240 gpm pump flow (-140 gpm LD flow),

T2 - 6000/778(1/.58 - 1) = 7.7 x 0.72 = 5.6F Therefore, since high LD/MU energy is conservative for the core power calculation, and low MU temperature results in high MU/LD energy, 5.6F will be a constant added to the measured MU temperature to account for pump heat.

Reference 23 calculates the maximum uncertainty on the measured MU temperature at +/-4.82F (increasing to 5.OF per Table 4 below).

The letdown temperature is the cold leg temperature and the uncertainty is based on the full span of 10OF (520F-620F) and is 3.42F per Reference 24. As discussed below, this will be rounded up to 4F in the uncertainty calculation.

Pressure Inputs The MU pressure is essentially at the MU pump discharge where the -5.6F temperature energy is added to the MU water. This is typically atmospheric pressure plus the pump head or between 2350 and 2400 psia. At 11OF MU temperature 1 , 2350 and 2400 psia, the enthalpies are 84.106 and 84.235 BTU/Ib. At 557F LD temperature and, 2230 psia LD pressure, the enthalpy is 555.993 BTU/lb. The LD-MU enthalpy difference is 471.887 BTU/Ib at a 2350 psia MU pressure and 471.758 BTU/Ib at a 2400 psia MU pressure. This is 471.758/471.887 = .99973 difference in the total MU/LD energy (typically -3 Mwt) or -. 0008 Mwt which is insignificant relative to the 2609 MWt core power. Therefore, a constant pressure of 2400 psia along with the measured MU temperature (plus a constant 5.6F) will be used to determine the MU enthalpy.

In order to be consistent with FIDMS, the letdown pressure is the measured pressure (measured at the "A" hot leg tap) minus -20 psi (the reduction in pressure after passing through the SG at full power plus an elevation pressure increase) to the let down nozzle. The uncertainty in the LD pressure per Reference 19 is 22.48 psi, (use 23 psi). This uncertainty and location correction will essentially have no impact the final HB uncertainty and since even +/- 100 psi will not impact the HB uncertainty (see discussion on MU pressure), no further discussion will is required on this parameter.

Flow Inputs The HB equation requires mass flow (lb/hr) and the measured flow at CR-3 is volumetric (gpm).

This will introduce two uncertainties associated with LD flow, one in the measurement of GPM and one in the measurement of the temperature (needed to convert to lb/hr). The maximum flow uncertainty from Reference 25 is 6.77 gpm (based on the maximum flow of 160 gpm). The maximum temperature uncertainty on the location where the flow is measured is 5.64F This is an assumed nominal temperature (see assumptions 2) at the MU pump discharge and is used only for relative difference in heat inputs.

10

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR (Reference 23) based on the maximum range temperature of 200F. At a typical maximum LD3 3

temperature of -135F, 5.64F will impact the density by 61.5 lb/ft [135F, 150 psia]161.6 lb/ft

[129F, 150 psia] =0.999. At a maximum of 140 gpm flow (see Reference 25), this is 0.001 x 140

- 0.2 gpm. Therefore, the flow uncertainty will be 6.77 + 0.2 = 7 gpm. This temperature uncertainty is essentially a bias added to the flow measurement uncertainty (and not SRSS with the flow measurement).

4.1.4 Ambient Heat Losses Reference 6 estimated an ambient loss uncertainty at 2.5E6 BTU/hr (out of 5.12E6 BTU/hr [1.5 MWt] estimated total heat loss). This uncertainty is essentially 50% of the expected maximum heat losses predicted in Reference 16 (which already had -+/-20% uncertainty included in the 1.5 MWt) and was based on engineering judgment and has no analytical basis. The 0.75 MWt uncertainty 2 was a conservatively high estimate added again to this ambient loss since it did not impact the final HB uncertainty when it was SRSS with the other plant parameters. CR-3 has chosen to treat this uncertainty as a bias of 0.75/2609 = .00029 or 0.029%.

Note that this ambient loss does not include pressurizer ambient losses because the pressurizer heaters cycle on to compensate for these losses.

4.2 Calculated Uncertainty Inputs 4.2.1 RC Pump Heat Input and Uncertainty The brake horsepower is a function of flow (gpm) and temperature and is the total energy the impeller puts into the water and is a published (test) value of each original impeller at CR-3. The temperature function is the density difference between the test temperature and the temperature expected during operation (at 100% power). The flow function is shown in the following curves based on the test data. Note that the impellers were designed to have the approximate peak BHP at the expected flow and at this flow, the BHP is relatively steady (not changing significantly in a

+/- 2% flow range). The attached BHP Curves (test data) originally installed at CR-3 show the following (Attachment 3 shows original data transcribed for clarity). Some conclusions drawn from this data are;

1. The BHP was apparently designed to peak at approximately the initial design flow (-90000 gpm)

2. Each Impeller provides approximately the same total energy at 96720 gpm (CR-3 best estimate flow at power) or -5.24 MWt (or 1.8E7 Btu/hr, -7030 brake horsepower) per pump. This is the average of 12 BJ pump impellers shown below. This results in a best estimate total energy is 20.96 MWt (or 7.16E7 Btu/hr) for the 4 pumps.

3. A 20 F difference in Tcold is equivalent to -0.008 MWt per pump or -0.03 MWt for 4 pumps and a +/- 2% difference in RCS flow will not change BHP significantly (4-pump only).

Density correcting the test data to 556 0 F and 558°F (the range of cold leg temperatures expected at CR-3 at full power and 2250 psia) are; 2 This uncertainty was based on the judgment that any instrumentation used to actually measure/calculate the primary system ambient losses would likely have less than the 0.75MWt random uncertainty (since it is approximately +/- 50% of an anticipated value).

I1

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR Impeller ID 0532 Correctedtotemp Test Data Flow(gpm) BHP(556) BHP(558F) BHP Temp (F) 80350 6990.25 6970.22 6950 560 97800 7141.99 7121.53 7080 562 100700 7061.29 7041.07 7000 562 111800 6864.59 6844.92 6805 562 120150 6658.04 6638.97 6610 561 Impeller ID 0531 Corrected totemp Test Data Flow(gpm) BHP(556) BHP(558F) BHP Temp (F) 79550 7010.85 6990.77 6950 562 95000 7121.82 7101.42 7060 562 100400 7061.29 7041.07 7000 562 109050 6950.33 6930.42 6890 562 119500 6647.97 6628.92 6600 561 Impeller ID 0530 Corrected to temp Test Data Flow(gpm) BHP(556) BHP(558F) BHP Temp (F) 74500 6984.49 6964.48 7110 543 89350 7043.43 7023.26 7170 543 98750 7043.43 7023.26 7170 543 108500 6984.49 6964.48 7110 543 118800 6688.80 6669.64 6800 544 Impeller ID 0533 Corrected totemp Test Data Flow(gpm) BHP(556) BHP(558F) BHP Temp (F) 70800 6592.78 6573.89 6630 552 80600 6851.32 6831.69 6890 552 90750 6970.64 6950.68 7010 552 99950 6951.41 6931.49 7010 550 108850 6574.58 6555.75 6630 550 12

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR BHP Pump 0530 (543F test) 7200 I FI 1

556 7100

-7035 HP 0 557 1 I I I -*-- 558 7000 6900 0

6800 6700 6600 I 70000 75000 80000 85000 90000 95000 100000 105000 110000 115000 120000 125000 Pump Flow gpm BHP Pump 0531 (562F test) 7200 -7105 HP or -U7060 curve fit 7100o 7000 -... ..--

68900- -

556 0 6800 - 558 6700 -- - x--- smooth curve fit __ _

6800 _II_

70000 75000 80000 85000 90000 95000 100000 105000 110000 115000 120000 125000 Pump Flow gpm BHP Pump 0532 (562F test) 0)

70000 75000 80000 85000 90000 95000 100000 105000 110000 115000 120000 125000 Pump Flow gpm 13

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR BHP Pump 0533 (552F test) 7200

- 556 7100 -6990 HP 55 557 5 I1_5

7 0 00 aa6900 __

o 6800 6700_l 6600 70000 75000 80000 85000 90000 95000 100000 105000 110000 115000 120000 125000 Pump Flow gpm Finally, note that the Byron Jackson (Flow Server) design was relatively consistent in the other impellers tested for ANO-1 and DB-1. Some of these impeller BHP curves are shown below (based on References 13 and Attachment 3).

ANO BHP Pump 0532A 7200

-76905 HP @ 557 -*-- 556 7100 -o--- 558 7000 ____

6900 -. ---

6800 6700 _____

6600 75000 80000 85000 90000 95000 100000 105000 110000 115000 Pump Flow gpm ANO BHP Pump 0532B 7200 1 _ 1 _ 1 _ 1 71001 -7010 HP @557 _ 5516 7000 '_-58_

6800- -

6700 -

6600 6500 70000 75000 80000 85000 90000 95000 100000 105000 110000 115000 Pump Flow gpm 14

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR ANO BHP Pump 0532C a

75000 80000 85000 90000 95000 100000 105000 110000 115000 Pump Flow gpm DB BHP Pump 0240 7100 -

5561 7000 - 5581 C. 6800.

0 -7050 HP @ 557 (Curve Fit) -

, 6700 6600 . --

6500 6400 70000 75000 80000 85000 90000 95000 100000 105000 110000 115000 Pump Flow gpm DB BHP Pump 0241 1!

0 ak.

0 0

70000 75000 80000 85000 90000 95000 100000 105000 110000 115000 Pump Flow gpm 15

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR DB BHP Pump 0242 7200 7080 HP @557 7150

,- 7100 7050 0

S7000___

6950 6900 75000 80000 85000 90000 95000 100Q000 105000 110000 115000 Pump Flow gpm DB BHP Pump 0243 7200 _____

7100 -7020 HP @ 5.57 556 7000 - _' _ .

6900 0

x 6800 6700 ___

6600-75000 80000 85000 90000 95000 100000 105000 110000 115000 Pump Flow gpm The following table shows the average of these 12 impellers.

HP MWt DB 7050 5.26 7055 5.26 7020 5.23 7080 5.28 ANO 7140 5.32 6905 5.15 6940 5.18 7010 5.23 CR-3 7060 5.26 7035 5.25 6990 5.21 7050 5.26 Avg 7027.917 5.24 Avg (CR-3 only) 7033.75 5 25 16

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR Given that all the BJ test data on impellers averaged 5.24 MWt range at the CR-3 flow rate, it will be assumed that the replacement impellers (manufactured by BJ) will all be about the same and the original impeller data will be valid (see assumption 5). Therefore, the total heat input of the CR-3 pumps will be 4 x 5.24 = 20..96 Mwt Uncertainty On Pump Power The uncertainty in the BHP used for CR-3 is based on two components (1) the validity of the test data and the applicability of the tested impellers relative to the actual replacement impellers at CR-3, and (2) the actual cold leg temperature during full power operation.

Test Data:

At the CR-3 expected RCS flow rate (see curves above), the range of BHP spans -5.15 to 5.32 MWt per pump. Data from the ANO-1 and Davis Besse pump test (see Attachment 3) revealed the flowing instrument uncertainties in the test instrumentation used by Byron Jackson.

Kilowatt Meter +/- 0.5%

Potential Transformers +/- 0.5%

Current Transformers +1- 0.5%

Flow rate +/- 1%

Temperature +/- 0.5%

Pressure +/- 0.25%

The uncertainty on the power measuring Instrumentation requires two potential transformers 3, the current transformer, and the watt meter resulting in (0.52 +0.52 +0.52 + 0.52)05 = 1.0% of the pump energy or 0.0524 Mwt per pump The flow uncertainty of 1% of the nominal flow (1000 gpm) has essentially no impact on the final pump energy since as shown above, the BHP at CR-3 is in the "flat" portion of the curve of BHP vs flow. The temperature impact is -0.005 x 560F = 2.8F (use 3F). As shown on the table for pump Impeller ID 0532 at 97800 gpm, the impact of 3F is (7142-7121)3/2 HP = 31.5 HP = .023 Mwt per pump.

The impact of the RCS pressure uncertainty (.0025 x 2250 psia= 6 psia) will change the pump power by less than 1 horse power and will be ignored. This is based on the cold leg density impact of 2250 vs. 2256 psia. This is -46.437/46.433 x -7000 HP < 7001 HP.

The potential variability in the pump impellers actually installed at CR-3 will use a conservative application of the nominal pump power minus the minimum measured power (the conservative direction for the HB). This is 5.24-5.15= 0.09 Mwt Since each of these uncertainties are independent and random, the total pump energy uncertainty is (0.05242 +.0232 + .092)0° = 0.107 MWt. Since the four pumps are independent, the total uncertainty on four pumps is (4 x .1072)0.5= 0.21 Mwt.

Plant Conditions Similar to the discussion above, the uncertainty in the actual RCS flow (within - +/- 2%) and RCS pressure will both have an insignificant impact on the final total pump heat. The pump heat calculated above (20.96 Mwt) is based on 557F Tcold leg and actual Tcold is the dominant uncertainty parameter in the pump heat calculation during operation. At full power, Tcold can 3 Reference 18 pg 16-54 describes a normal measurement setup for 3-phase current with a balanced load (which includes a motor). This normal setup requires two potential transformers and one current transformer.

17

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR vary due to the ICS control over reactor vessel Tavg and loop A to loop B delta Tc. If delta Tc (cold leg delta temperature difference) varies at a constant Tavg, the two pumps with an increased temperature and the two pumps with a decreased temperature will essentially offset each other, maintaining a constant total pump energy. During an end of cycle coastdown (reduced Tcold), the actual pump power will increase. If this increase was used in the HB equation, the allowable Qsec would actually increase, resulting in the constant Qpump at 557F conservative for this situation. An increase Tcold will reduce pump energy which in turn will decrease the allowable SG energy to maintain a 2609 Mwt core power making the 20.96 Mwt constant pump power (at 557F) non-conservative ifTavg (Tcold) were to increase.

During normal full power operation, the ICS uses all the cold leg RTDs and the hot leg RTDs to establish Tavg. The normal operation permits Tavg +/- 2F. At +2F Tavg the cold leg temperature would be 559F. Per pump impeller data above, 2F equates to a little less than 21 HP or 0.015 MWt per pump or 0.06 Mwt for 4 pumps. The total pump energy uncertainty is 0.21+0.06 = 0.27 Mwt. Per CR-3 request, this will be conservatively applied as a bias or it will add 0.27/2609

=0.01% to the uncertainty.

4.2.2 Partial Derivatives at Nominal Full Power Heat Balance Parameters The partial derivatives in the uncertainty equation are based on water/steam properties at or near the values they are calculated for. Table 2 presents the nominal expected full power values at 2609 MWt.

TABLE 2 - Nominal Heat Balance Parameter Values Symbol Description Units Nom. Value Basis (at -location of measurement)

WFW Feedwater Flow Rate Lbm/hr 1.107E+07 Ref. 8 TS Steam Temperature F 590.5 Ref. 8 PS Steam Pressure psia -920 Assumption 10 TFW Feedwater Temperature F 458.4 Ref. 8 PFW Feedwater Pressure psia -980 Assumption 8 WLD Letdown Flow Rate Lbm/hr 35000 70 gpm @-100F, 150 psia*

gpm 70 Ref 6 TLD Letdown Temperature F 557 Ref. 8 (CL temp)

TMU Makeup (tank+pump heat) temperature F 110 See Sec 4.1.3 PMU Makeup Press psia 2400 See Sec 4.1.3 PLD Letdown Pressure psia 2230 See Sec 4.1.3 QRCP RCP Power Btu/hr 7.158E+07 See Sec 4.2.1 QLOSS Ambient Heat Loss Btu/hr 5.12E+06 See Sec 4.1.4

This neglects density increases due to boron which is acceptable since these are approximate full power nominal operating conditions.

Note that if the actual values of these parameter vary a few percent when the AULD monitors them, the partial derivatives calculated below will remain applicable.

The water/steam properties at these nominal values are; Steam Enthalpy = 1250.386 Btu/Ibm at 590.5 0 F and 920 psia Feedwater Enthalpy = 439.904 Btu/Ibm at 458.40 F and 980 psia Letdown Enthalpy = 555.993 Btu/Ibm at 5570 F and 2230 psia 18

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR Makeup Enthalpy = 84.235 Btu/Ibm at 110°F and 2400 psia The partial derivatives based on these values are:

For steam at 920 psia:

At T = 5850 F, H = 1245.687 Btu/Ibm At T = 595 0 F, H = 1254.157 Btu/lbm aH/aTs = (1254.157-1245.687)/(595 - 585) = 0.847 Btu/lbm/°F For steam at 590.50 F:

At P = 915 psia, H = 1250.965 Btu/lbm At P = 925 psia, H = 1249.805 Btu/lbm aH/aPs -(1250.965 - 1249.805)/(915-925) = -0.11601 Btu/Ibm/psia For feedwater at 458.40 F:

At P = 1025 psia, H = 439.9242 Btullbm At P = 925 psia, H = 439.8797 Btu/lbm aHaPf,, -=(439.9242 - 439.8797)/(1025 - 925) = 4.45E-4 Btu/Ibm/psia For MU at 2400 psia:

At T = 1051F, H = 79.289 Btu/lbm At T = 1 10°F, H = 84.235 Btu/lbm aH/aTMu = (79.289 - 84.235)/(105 - 110) = 0.989 Btu/lbm/°F For LD at 2250 psia:

At T = 555 0F, H = 553.442 Btu/Ibm At T = 560 0F, H = 559.749 Btu/lbm aH/laTLD -=(559.749 - 553.442)/(560 - 555) = 1.261 Btu/lbm/°F For LD at 557 F:

At P = 2200 psia, H = 556.048 Btu/Ibm At P = 2250 psia, H = 555.957 Btu/Ibm aH/aPLD = (556.048 - 555.957)/(2250 - 2200) = -1.82E-3 Btu/Ibm/psi The following terms are include for potential future use but are not used in this calculation since the feedwater flow and temperature were combined by CALDON into one power uncertainty For feedwater at 975 psia:

At T = 455 0F, H = 436.085Btu/Ibm At T = 465 0F, H = 447.348 Btu/lbm aH/aTf _=_(447.348- 436.085)/(465 - 455) = 1.126 Btu/lbm/°F The water property derivatives are summarized in Table 3 below.

19

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR TABLE 3 - WATER PROPERTY DERIVATIVES alH/DT, Btu/(lbm°F) 8aHISP, Btul(Ibm psi)

Steam (590.51F, 920 psia) 0.847 -0.116 Feedwater (458.40 F, 975 psia) 1.126 4.45E-4 Makeup(1 101F, 2400 psia) 0.989 NA (Const Press per Sec 4.1.3)

Letdown(5570 F, 2230 psia) 1.261 -1.82E-3 5.0 CALCULATION OF HEAT BALANCE UNCERTAINTY Inputs were calculated for operating conditions for the Appendix K power uprate. Reference 1 provides step-by-step instructions for calculating the uncertainty of a result. The independent measurement parameters and their nominal values are comprised of the values in Table 2 above.

The expression for core power in terms of a secondary side heat balance is re-stated below.

QC= WFWA )

(HsA - HFwA )+ W wB(HsB - HfwB + QLD-MU -QRcP + QLOSS Where:

WFWA, WFWB Feedwater flows in Loop A & B HSA, HFWA, HSB, HFWB Steam & feedwater enthalpies for Loops A & B QLD-MU Heat loss due to primary side letdown flow QRCP Heat added due to RC pumps QLOSS Ambient heat losses from the RCS WLD, WMU Letdown and Makeup Flow Rates HLD, HMU Letdown and Makeup Enthalpies Since the SG uncertainties were combined to one uncertainty QC = Wfw(Hs - Hfiv)+ QLD-MU - QRCP + QLOSS The uncertainties for each parameter are also summarized below:

TABLE 4- HEAT BALANCE PARAMETER UNCERTAINTY VALUES Symbol Description Units Uncertainty Value Basis WFW/TFW Combined BTU/Ib 0.34% of nominal flow (both SGs) Section 4.1.1 Uncertainty TS Steam Temp F 1.43(single) [2.02F per RTD, 4 RDTs per SG] Section 4.1.2 1.01 (one SG) 0.71 (both SGs)

PS Steam Pressure psi 4.57 (single) Section 4.1.2 3.23 (one SG) 2.28 (both SGs)

PFW Feedwater Press psi 7.11 (one SG) Section 4.1.2

_______ Flow5.03 (both SGs)

WLD Letdown Flow Ibm/hr 7 gpm @ 135F, 150 psia or 3.5E3 lb/hr Section 4.1.3 TLD Letdown Temp(CL) F 3.42 (use 4.0 to assure worst case) Ref. 24 TMU Makeup Temp F 4.82 (use 5.0 to assure worst case) Section 4.1.3 20

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR PLD Letdown Press psia 22.48(use 23) Section 4.1.3 QRCP RCP Power Btu/hr 9.22E5 (0.27 Mwt) Section 4.2.1 QLOSS Ambient Heat Loss Btu/hr 2.5E6 Section 4.1.4 QATMOS Psig to Psia error psi 0.33 (converts to 4.257E5 BTU/hr -Steam Section 2.2 1 _Press)

The uncertainty calculation from section 3.2 for the secondary side heat balance is 2

E(Q) = rr)2 + (aQ/Ts x CT5)2 + (aQ/aPs x sps) + (aQ/aPW X 8p )2]o0.

The MU/LD uncertainty is per section 4.1.3 Eletdown = [(aQ/aWLD X EWLD)2 + (DQ/aTLD X ETLD)2 + (aQ/aTMU X :TMU)2 ]05 Eambient = Constant ERCpump= Constant EATMOS = Constant The final heat balance uncertainty will be:

E (core) =[EQ2 + Eletdown 2 ]0 5 + Eambient heat loss + EAtmos press + ERC Pump with EQA comprising the large majority of the uncertainty The core thermal power equation was differentiated with respect to the individual measured parameters to yield the following sensitivity coefficients (values from Section 4.2.2):

Opw = aQc/aPFw = WFW aH/aPW = (1.1 07E+07 lb/hr x 4.45E-4 Btu/lb/psi) = 4.926E3 Btu/hr/psi ps= aQc/aPs = rw a8H/aPs = (1.107E+07 lb/hr x -0.11601 Btu/lb/psi) = -1.284E6 Btu/hr/psi OTs = aQc/aTs = WFw aH/lTs = (1.1 07E+071b/hr x 0.847 BtuIlb/°F) = 9.376E6 BTU/hr F 0

W\d = aQc/aVLD = HLD - Hmu = 555.993-84.235 = 471.76 BTU/Ibm OTmu = aQc/OTMu= WMU aH/aTMu = (35000 lb/hr x 0.989 Btu/lb/°F) = 3.462E4 BTU/hr F OPLD = iQc/aPLD= WMU aH/OlPLD = (35000 lb/hr x -0.00182 Btu/lb/psi) = -6.37E1 BTU/hr/psi OTLD = aQc/aTLD= WLD 8H/aTILD = (35000 lb/hr x 1.261 Btu/lb/°F) = 4.414E4 BTU/hr F OQrcp = aQC/aQRCPs = 1 QIoss = aQC/aQLoss 1 The following terms are include for potential future use but are not used in this calculation since the feedwater flow and temperature were combined by CALDON into one power uncertainty.

ewf = aQc/aWFw = (Hs - HEW) = (1250.384 Btu/lb - 439.904 Btu/lb) = 810.080 Btu/lbm 0

Tfw = aQc/aTFW=WFW aH/aTFw= (1.107E+07 lb/hr xl.126 Btu/Ib/°F) =1.246E7 BTU/hr F Sensitivity Coefficients and Uncertainty Contributions 21

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR The sensitivity coefficients and the uncertainty contributions were calculated using the values in Tables 2 and 3 as follows:

Feedwater Flow Rate and Feedwater Temperature Using the uncertainty of 6wfwn- = (0.34/100)

8.905E9 = 3.028E7 BTU/hr, the systematic uncertainty contribution is:

2

[swffrr] 2 = 9.166E14 (Btu/hr)

Feedwater Pressure Using the uncertainty of Fpf, = 5.03 psi, the systematic uncertainty contribution is:

2

[0pf

Epf]2 = [4.926E3

5.03]2 = 6.140E8 (Btu/hr)

Steam Pressure Using the uncertainty of EPs = 2.28 psi, the systematic uncertainty contribution is:

2 2

[0ps

Sps] = [-1.284E6

2.28]2 = 8.573E12 (Btu/hr)

Steam Temperature Using the uncertainty of CTs = 0.71OF, the systematic uncertainty contribution is:

2

[OTsa

ETs ]2 = [9.376E6

0.71]2 = 4.432E13 (Btu/hr)

Letdown Flow Using the uncertainty of EWLD = 3.5E3 Ibm/hr 2

[OWLD

SWLD]

2

= [471.76* 3.5E3]2 = 2.726E12 (Btu/hr)

Letdown Pressure Using the uncertainty of EPLD = 23 psi 2

[OPLD

8PLD]

2

= [-63.7* 23]2 = 2.147E+06 (BtU/hr)

Letdown (CL) Temperature Using the uncertainty of ETLD = 4.0 OF, the systematic uncertainty contribution is:

[OTLD

STLD]2 = [4.414E4

4.0]2 = 3.117E10 (Btu/hr) 2 Makeup Temperature Using the uncertainty of STLD = 5.0 OF, the systematic uncertainty contribution is:

2

[ETLD

STLD] =

2

[3.462E4

5.0]2 = 2.995E10 (Btu/hr)

RCP Power This is a biased uncertainty calculated in section 4.2.1 of 0.27 Mwt.

Ambient Heat Loss This is a biased uncertainty calculated in section 4.1.4 of 0.75 Mwt.

22

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR Pcqage to Pabsolute Conversion This is a biased uncertainty shown in section 2.2 of 425700 BTU/hr (0.125 Mwt)

The uncertainty contributions are summarized below in Table 5.

TABLE 5 - HEAT BALANCE PARAMETER UNCERTAINTY CONTRIBUTIONS Uncertainty Contribution Symbol Description Absolute Relative Relative Percent (Btu/hr)2 (SRSS) (Total) of Total Power WFW/TFW Feedwater Flow/Temp 9.166E+14 94.2733% 83.7722% 0.3301%

TS Steam Temperature 4.432E+13 4.5582% 4.0504% 0.0160%

PS Steam Pressure 8.573E+12 0.8818% 0.7836% 0.0031%

PFW Feedwater Pressure 6.140E+08 0.0001% 0.0001% 0.0000%

TLD Letdown Temperature 2.147E+06 0.0000% 0.0000% 0.0000%

PLD Letdown Pressure 3.117E+10 0.0032% 0.0028% 0.0000%

WLD Letdown Flow Rate 2.726E+12 0.2804% 0.2492% 0.0010%

TMU Makeup Temperature 2.995E+10 0.0031% 0.0027% 0.0000%

TOTAL 9.723E+14 100%

Bias Corrections (Btu/hr) __________

QRCP RCP Power 9.215E+04 NA 2.6261% 0.0103%

QLOSS Ambient Heat Loss 2.560E+06 NA 7.2948% 0.0287%

PATMOS Pqaqe to Pabsolute 4.274E+05 NA 1.2181% 0.0048%

Totals 100% 0.394%

6.0

SUMMARY

OF RESULTS The ASME Performance Test Code Methodology was used to calculate the expected core thermal power uncertainty to be achieved using the Caldon CheckPlusTM System ultrasonic flow meter. The analysis concluded that using the following instrument uncertainty values, the core thermal power uncertainty would be 0.394% of 2609 Mwt, thus allowing a power uprate of 1.6% to be pursued. The feedwater flow and temperature measurement is the bulk of this uncertainty (0.34% absolute and -84% of the total uncertainty). The new steam temperature/pressure instrumentation results in -4% of the total uncertainty and the steam pressure measurement uncertainty is -1% of the total. The RC pumps energy uncertainty, ambient loss uncertainty and an atmospheric pressure correction uncertainty were chosen to be treated as a bias (algebraically added and not SRSS) and they are -11% of the total uncertainty (see Attachment 1). After the final feedwater flow/temperature uncertainty is determined for the CR-3 specific equipment (post fabrication testing), the total uncertainty may be reduced.

Other pertinent output of this calculation include (1) the RC pump power (to be used in the HB equation) is 20.96 MWt, (2) the temperature increase due to the MU pump to be added to the makeup measured temperature is 5.6F, (3) the pressure to be used for MU enthalpy is a constant (2400 psia), and the letdown pressure will be the measured RCS pressure minus 20 psi (noting that this input has essentially no impact on the final HB uncertainty).

23

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR

7.0 REFERENCES

(1) ASME PTC 19.1-1998, Test Uncertainty, Instruments and Apparatus, American Society of Mechanical Engineers, NY, NY, 1998.

(2) Caldon, Inc. Engineering Report-80P Revision 0 (Proprietary Version), Topical Report -

"Improving Thermal Power Accuracy and Plant Safety While Increasing Operating Power Level Using the LEFMVI TM System," March 1997. (For Information Only)

(3) Caldon Topical Report Caldon, Inc. Engineering Report-157P Revision 5, Topical Report -

"Supplement to Topical Report ER-80P: Basis for a Power Uprate With the LEFM,/TM or LEFM CheckPlusTM System," February 2001.

(4) AREVA Doc. 32-9040357-001 "Loop Accuracy Calculation for CR3 Steam Press Computer Ind for MUR Heat Balance Computation" (5) AREVA Doc 32-9040364-000 "Loop Accuracy Calculation for CR3 Steam Temperature Computer Ind for MUR Heat Balance Computation,"

(6) AREVA Document 32-5001078-01, "CR-3 Heat Balance Uncertainty Calc," March 1998.

(7) AREVA Document 32-9041510-001 "CR3 MUR Feedwater Pressure Uncertainty Calculation" (8) AREVA Document 32-5012972-01, "CR-3 Power Uprate Operating Conditions" (9) AREVA Document 32-5018220-02 "ANO-1 ETOSG Secondary Pressure Distribution" (10) AREVA Document 32-9034910-00, "CR-3 Main Steam Line DP Estimate" (11) Deleted (12) WinSteam program for water properties based on IAPWS-IF95 by ChemicaLogic Corporation.

(13) AREVA Document 32-1174376-00, "ANO-1 RCS Delta Pressure," 1989.

(14) Deleted (15) CR-3 Document 191-0023 Revision 2 "Narrow Range RC Pressure (PIR) Instrument Loop Accuracy" (16) AREVA Document 32-1240860-00, "Ambient Heat Loss estimate" 12/1995 (17) Caldon, Inc. Engineering Report: ER-579 Revisiop 1, "Bounding Uncertainty Analysis for Thermal Power Determination at crystal River Unit 3 Using the LEFM,+ System,"

February 2007. (Pertinent pages attached as Attachment 4).

(18) Standard Handbook for Electrical Engineering Eleventh Edition 1978, McGraw Hill Publishing (19) CR-3 Document 101-0002 Revision 0 "TRICON Instrument Accuracy", 1012001 24

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR (20) AREVA Document 32-1257388-00 "CR-3 HPI Pump HQ Response" 1996 (21) AREVA Document 32-5002731-03 "CR-3 HPI Hydraulics Analysis" 2000 (22) Karrassik, Igor J. Centrifugal Pump Clinic, Marcel Dekker Inc. (1981)

(23) AREVA Document 32-9043792-000 "Loop Accuracy Calculation for CR-3 Makeup Tank (MUT) Temperature and Letdown Flow temperature Computer indications for MUR Heat balance Computation," 3/2007 (24) AREVA Document 32-9043797-000 "Loop Accuracy Calculation for CR-3 T-Cold Temperature indications for MUR Heat balance Computation," 3/2007 (25) AREVA Document 32-9049930-000 "Loop Uncertainty Calculation for Let down Flow to AULD Plant Computer," 3.2007 References 15 and 19 are retrievable in the CR-3 documentation system :t* Date!_/0/*'7 25

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR ATTACHMENT I - HEAT BALANCE SPREADSHEET The methodology developed in Section 4 was programmed in Excel for ease of evaluating various inputs. The Excel spreadsheet was verified by comparing the results of these results with data listed in Section 5.

BASE CASE Absolute Relative Relative DescriptionNominal Absolute Absolute Uncertainty Uncertainty Uncertainty C ion Value Uncertainty Sensitivity Contribution( Contribution Contribution squared)

(SRSS Only) Tot Uncrt Feedwater Flow Rate& Temp BTU/hr 8.905E+09 3.028E+07 BTU/hr 3.028E+07 BTU/hr 9.166E+14 BTU/hr112 94.2733% 83.7722% 8.613 Steam temperature F 590.5 0.71 F 9.376E+06 BTU/hr /F 4.432E+13 BTU/hrA2 4.5582% 4.0504% 0.416 Steam Pressure osia 920 2.28 psi -1.284E+06 BTU/hr/psi 8.573E+12 BTU/hrA2 0.8818% 0.7836% 0.081 Feedwater Pressure osia 980 5.03 psi 4.926E+03 BTU/hr/psi 6.140E+08 BTU/hrA2 0.0001% 0.0001% 0.000 Letdown Pressure psia 2230 23 psi -6.370E+01 BTU/hr/psia 2.147E+06 BTU/hrA2 0.0000% 0.0000% 0.000 Letdown Temperature F 557 4 F 4.414E+04 BTU/hr/ F 3.117E+10 BTU/hrA2 0.0032% 0.0028% 0.000 Letdown Flow Rate Ibm/hr 35000 3.50E+03 Ibm/hr 471.76 BTU/Ib 2.726E+12 BTU/hrA2 0.2804% 0.2492% 0.026 Makeup Temperature F 110 5 F 3.462E+04 BTU/hr/ F 2.995E+10 BTU/hrA2 0.0031% 0.0027% 0.000 SUM= 9.723E+14 100.0%

Bias for pump power 9.215E+05 BTU/hr 2.6261% 0.270 Bias for Ambient Loss I 2.560E+06 BTU/hr 7.2948% 0.750 Bias for Atmospheric Press 4.274E+05 BTU/hr 1.2181% 0.125 100.0% 10.28 Nominal Nominal Absolute Relative Bias for 0.75 Description Value (MWt) Value Uncertainty Uncertainty Bias for Pump Mwt t Atmospheric Total (Btu/hr) (Btu/hr) (SRSS (%) SRSS Power loss pressure bias Components) Uncertainty components Core Thermal Power 2609 8.905E+09 3.118E+07 0.35017% 0.0103% 0.0287% 0.0048% 0.3941%

3.509E+07 BTU/hr Steam Enthalpy Btu/Ibm 1250.386 Steam art/saT 0.847 aHlaP -0.11601 10.28 MWt Feedwater Enthalpy Btu/lbm 439.904 Feedwater aHlw/aT 1.126* aHFw/aP 4.450E-04 Makeup Enthalpy(110,2400) Btu/lbm 84.235 Makeup aHMulaT 0.989 Nom FW Flow 11070000 lb/hr (both SGs)

Letdown Enthalpy(557,2230) Btu/Ibm 555.993 Letdown aHLD/aT 1.261 aHlLO/aP -0.00182

Not used in this spreadsheet 26

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR ATTACHMENT 2 - CALDON UNCERTAINTY INPUTS -PRELIMINARY VALUES 2.0

SUMMARY

For CrystalRiver Unit 3, Revision 0 results are as follows:

1. The mass flow uncertainty approach is documented in Reference 3. The uncertainty in the LEFM,f+'s mass flow of feedwater is as follows:

o Fully Functional LEFM,/ + system mass flow uncertainty is +/- 0.32%

o Maintenance Mode LEFM.I+ system mass flow uncertainty is +/- 0.38%.

Note: The LEFM, + system is in maintenance mode when only one of the two LEFMI" +

subsystems is fully functional, i.e., LEFM/"+ System is operating as an LEFM,[ System. The uncertainty of the LEFMI/+ when in maintenance mode may be re-evaluated and will likely be reduced after site specific hydraulic experience has been taken into account.

2. The uncertainty in the LEFM" + feedwater temperature is as follows:

o Fully Functional LEFM,/ + system temperature uncertainty is _+0.57 0 F (+/-0.320 C) o Maintenance Mode LEFMI/+ system the uncertainty is +/-0.57°F (10.32°C)

3. The thermal power uncertainty approach is documented in Reference 3. The uncertainty in the calculation of thermal power due to the LEFMI + is as follows:

o Fully Functional LEFMI + system thermal power uncertainty is +/- 0.34%

o Maintenance Mode LEFMf + system thermal power uncertainty is +/- 0.40%.

Note: Because some elements of the temperature uncertainty are systematic, the total power uncertainty due to the LEFMf + is not the root sum squares of the uncertainties due to items I and 2 above.

4. For an overall thermal power uncertainty analysis in which mass flow and temperature errors are to be treated separately (i.e., the thermal power uncertainty above is not used), the bounding mass flow and temperature errors must be divided into components that are systematically and randomly related to the mass flow error, as follows:

o The fully operational LEFM/ + systematic temperature error related to the mass flow error is +/-0.070F (+/-0.040 C) and the random temperature error related to the mass flow error is +0.56TF (4O.31QC). The mass flow uncertainty remains at 10.32%. The thermal power error due to this term is combined as the root sum square with other elements of the thermal power uncertainty.

o The maintenance mode LEFM./ + systematic temperature error related to the mass flow error is +/-0.051F (+/-0.031C) and the random temperature error related to the mass flow error is +0.570 F (=0.32°C). The mass flow uncertainty remains at +/-0.38%. The thernal power error due to this term is combined as the root sum square with other elements of the thermal power uncertainty.

ER-579 Rev. 0 Prepared by: RH Reviewed by: *A-27

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR ATTACHMENT 3 TRANSCRIBED RC PUMP DATA

.1 Pump Discharge Eq pressure flow Watt HP Volts Amps Pump Temp Indicated Corr Pump DP Press psig FPress psid Ppm WHP BHP Reading Input Eff.

2242 2242.9 122.3 89350 6370 7170 5650 7560 6400 532 88.9 543 2236 2236.9 110.2 98750 6350 7170 5650 7560 6400 532 88.5 543 2223 2223.8 57.1 108500 6150 7110 5600 7500 6400 528 86.5 543 2193 2143.8 81.2 118800 5630 6800 5350 7160 6400 510 82.8 543 2153 2153.7 60.1 129700 4520 6350 5000 6700 6400 480 71.2 544 2254 2254.9 136.3 79500 6325 7110 5600 7500 6400 525 88.3 543 2256 2256.9 143.3 69650 5830 6900 5430 7280 6400 512 89.5 542 2246 2246.9 149.3 57950 5050 6540 5150 6900 6400 489 77.2 541 2245 2245.9 153.3 49030 4390 6410 5050 6760 6400 485 68.5 540 2264 2264.9 119.2 91300 6350 7150 5030 7540 6400 530 88.8 540 ipeller I I / /rl I F*I I I . .......... . J Pump Discharge Eq pressure flow Watt HP Volts Amps Pump Temp Indicated Corr Pump DP Press Dsia Press Dsid gpm WHP BHP Reading Input Eff.

2312 2303.0 118.2 90750 6250 7010 5500 7370 6500 519 89.0 552 2320 2311.1 128.3 80600 6040 6890 5400 7240 6500 515 87.6 552 2321 2312.1 135.3 70800 5590 6630 5200 6970 6500 500 84.2 552 2307 2298.0 140.3 59790 4885 6250 4900 6560 6500 470 78.1 552 2306 2299.0 146.3 50700 4330 6120 4800 6430 6500 455 70.7 552 2317 2308.0 105.2 99950 6140 7010 5500 7370 6500 510 87.5 550 2275 2266.0 99.2 108850 5860 6630 5200 6970 6500 505 88.4 550 2248 1238.9 75.2 119550 5240 6630 5200 6970 6500 490 79.0 550 2223 2213.8 60.1 128150 4500 6250 4900 6560 6500 470 72.0 550 2306 2297.0 113.2 94700 6255 7010 5500 7370 6500 520 89.3 549 28

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR Pump Discharge Eq pressure flow Watt HP Volts Amps Pump Temp Indicated Corr Pump DP Press psig Press psid qpm WHP BHP Readina Input Eff.

2273 2270.5 109.8 95000 6090 7060 5540 7420 6780 517 86.2 562 2265 2262.5 104.8 100900 6190 7000 5500 7360 6780 518 87.6 562 2248 2295.5 94.8 109050 6030 6890 5400 7240 6780 510 87.5 562 2212 2209.6 75.9 119500 5290 6600 5175 6940 6770 485 80.0 562 2173 2170.6 59.8 128050 4965 6120 4800 6440 6800 448 73.0 561 2293 2290.4 129.7 79550 6015 6950 5450 7300 6800 510 86.5 560 2295 2292.4 136.7 69650 5560 6850 5370 7200 6800 500 81.3 560 2298 2295.4 142.7 59400 4950 6530 5130 6880 6800 480 75.8 560 2297 2294.4 146.7 49600 4250 6280 4940 6610 6800 460 67.7 560 2282 2279.4 115.7 42500 6220 7120 5575 7480 6790 519 87A4 559

-InTelrc Pump Discharge Eq pressure flow Watt HP Volts Amps Pump Temp

.Indicated Corr Pump DP Press psia Press psid pDm WHP BHP Readina Input Eff.

2242 2239.5 108.8 97800 6210 7080 5550 7440 7000 525 87.8 562 2236 2233.5 105.8 100700 6220 7000 5490 7350 6950 510 88.7 562 2215 2212.6 89.8 111800 5855 6805 5340 7150 6950 500 86 562 2196 2193.6 75.8 120150 5310 6610 5180 6950 6950 485 80.3 561 2181 2178.6 64.8 127950 4820 6400 5040 6710 6975 470 75.4 560 2265 2262.5 127.8 80350 5960 6950 5450 7300 6975 510 86 560 2278 2275.4 136.7 70100 5600 6720 5270 7060 7000 490 83.2 560 2291 2288.4 143.7 60150 5040 6360 4990 6690 7000 465 79.1 560 2300 2297.4 147.7 49350 7240 6000 4700 6300 7000 440 70.7 561 2255 2252.5 114.8 92850 6210 7010 5500 7360 7000 515 88.6 562 29

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR BJ Test Data Uncertainties for ANO-1 FIGURE II PUMP LOOP TEST INSTRUMENTATION LIST PARAMETER INSTRUMENT RATED ACCURACY Motor Kilowatt Input Weston Polyphase Wattmeter Voltage Weston Voltmeter Model 341 + 0.25%

Amperage Weston Ammeter Model 370 +0.25%

Current Transformers Westinghouse Type CLA- 10 Potential Transformers Westinghouse Type PTM- 15 Winding Temp. Leeds & Northrop Ohmmeter + 0.5%

& 10 RTD Bearing Temp. Honeywell Recorder and T.C. + 1.5°F.

Cooling Water Flow System Calibration AP vs Q +2%

Cooling Water Temp. Honeywell Recorder and T.C. + 1.50 F Pump Flow Herschel Type 24" x 17-1/2" + 1%

Venturi (four)

Validyne X-Ducer Mod. DP 15 + 0.5%

Barton AP Unit Model 200 +0.5%

Static Pressures Pump Suction and Discharge Crosby Bourdon Tube Type +0.5%

Roylyn Coil Tube Type + 0.25%

System Loop Water Temp. Bailey Recorder and 100 "RTD Pump Speed Hewlett Packard Elec. Counter Pump Seals Controlled Bleed-Off Flow Fischer Porter Float Type +2%

Controlled Bleed-Off Temp. Honeywell Recorder and T..C. + 1..5 0F Seal Cavity Pressure (3) Bailey Recorder Model El 01 +0.5%

Seal Cavity Temp. Honeywell Recorder and T.C. + 1.5°F Cooling Water Flow Fischer Porter Float Type +2%

Cooling Water Temp. Honeywell Recorder and T.C. + 1.50F Inlet and Outlet Differential Pressure Across Pump Valydyne X-Ducer Mod. DP. 15 Bailey AP Unit Model 200 150PSI +0.5%

Additional Static Pressure measurements Crosby Bourdon Tube Type + 1.0%

Additional Temperature Measurements Honeywell Recorder and T.C..

Up to + 200'F + 1.5°F 200°F to 700°F + 0.75%

Pump and Motor Vibration B & K Accelerameter +2%

IR.D. Model 306 Meter +2%

30

AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR BJ Test Data Uncertainties for Davis Besse Exerpt From Byron Jackson Report On Davis Besse Pump Tests June 1, 1973 PARAMTER MFG. ACCURACY Killowatt Input Weston 0.5%

Pump Capacity Barton (DP) 0.75%

Byron Jackson (Ventuwi) 0.5%

Loop Water Temperature Baily Meter 0.5%

Static Pressure Pump Suction Roylyn Linsay 0.25%

Pump Discharge Roylyn Linsay 025%

31

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THE ONLY DATA USED FROM THE PAGE WAS . I INTERPRETED CORRECTLY FROM THE ORIGINAL COPY OF THIS PAflF AND IS SHOWN BELOW Impeller 32-9042687-,001 Pg II ID 0240 Corrected to temp Test Data GPM 556 558 BRAKE HP Temp (F) 78500 7094.77 7074.45 7260 539 90500 7114.94 7094.56 7290 538 99200 6988.06 6968.05 7160 538 109100 6953.91 6933.99 7125 538 119§00 6744.07 6724.75 6910 538 Date V//,P7

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110300 6939.893 6920.014 6910 559 DATE-.0~ý,:-j 11860Q,6688.81266Q9.551 6660 559 R Q"7 P-366 ..............

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COPY OF THIS PAGE AND IS SHOWN BELOW .Z.., 9 Impeller

'Impele ID 32-9042687-0.01 Pg._ X"A.... * '- :"" "::"

0243 Corrected to temp Test Data : . ...-. -- -

GPM 556 558 BRAKEHP Temp (F) 81250 - "- .,

6922.59 6902.76 7120 535

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91900 7011.214 6991.131 7080 549 -2 101300 7050.826 7030.629 7120 549 109450 7001.311 6981.256 . ,-

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AREVA NP 32-9042687-001 Heat Balance Uncertainty for CR-3 MUR ATTACHMENT 4 ATMOSPHERIC PRESSURE DATA FROM CR-3 The attached data gives a high over a 7 year period from Ocala and Gainesville Florida and 4 years from Holder Florida. The high for that period was 0.67 in Hg. Use this number. It is the only number that can be defended based on engineering data.

14.7 2.036022 29.930 range OCF low "Hg high "Hg avg "hg "Hg neg "Hg pos "hg neg psi pos psi 2000 29.66 30.49 30.04 0.83 -0.27 0.56 -0.13 0.28 2001 29.54 30.50 30.16 0.96 -0.39 0.57 -0.19 0.28 2002 29.67 30.55 30.13 0.88 -0.26 0.62 -0.13 0.30 2003 29.57 30.54 30.10 0,97 -0.36 0.61 -0.18 0.30 2004 28.94 30.55 30.15 1.61 -0.99 0.62 .-0.49 0.30 2005 29.50 30.59 30.16 1.09 -0.43 0.66 -0.21 0.32 2006 29.66 30.53 30.13 0.87 -0.27 0.60 -0.13 0.29 GNV 2000 29.60 30.49 30.02 0.89 -0.33 0.56 -0.16 0.28 2001 29.60 30.50 30.15 0.90 -0.33 0.57 -0.16 0.28 2002 29.63 30.56 30.15 0.93 -0.30 0.63 -0.15 0.31 2003 29.54 30.53 30.10 0.99 -0.39 0.60 -0.19 0.29 2004 29.11 30.56 30.12 1.45 -0.82 0.63 -0.40 0.31 2005 29.45 30.60 30.15 1.15 -0.48 0.67 -0.24 0.33 2006 29.63 30.55 30.12 0.92 -0.30 0.62 -0.15 0.30 Holder 2003 29.68 30.55 0.87 -0.25 0.62 -0.12 0.30 2004 28.79 30.55 1.76 -1.14 0.62 -0.56 0.30 2005 29.51 30.59 1.08 -0.42 0.66 -0.21 0.32 2006 29.66 30.52 0.86 -0.27 0.59 -0.13 0.29 Max -1.14 0.67 -0.56 0.33 36

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #296, REVISION 1 ATTACHMENT F ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY (AMSAC) ARMING SETPOINT EVALUATION

U.S. Nuclear Regulatory Commission Attachment F 3F0607-05 Page 2 of 7 Table F-i: Key Inputs and Boundary Conditions LOFW ATWS for AMSAC Parameter Arming Value Core Power (MWt) (52% of 2609 MWt) 1356.68 Core Decay Heat 1.0*ANS71 +

B&W actinides RCP Heat Addition, net (MWt per pump) 4.0 Average RCS Temperature ('F) 579 RCS Pressure (psig) - hot leg 2155 Initial Pressurizer Level (in) 220 Spray Setpoint - on/off (psig) 2205/2155 Spray Capacity - design (gpm) 190 PORV Setpoint - open/close (psig) 2450 / 2380 PORV Capacity - nominal (lbm/hr) 148,306

@ 2450 psig + 3% accumulation @2538 psia Number of Pressurizer Code Safety Valves (PSV) 2 PSV Setpoint - nominal (psig) 2500 PSV Setpoint Tolerance (%) +2.0 / -4.0 PSV Capacity - nominal (lbm/hr/valve) 317,973

@2750 psig MSSV Setpoint Lift Tolerance (%) +1.0 MSSV Accumulation (%) +3.0 MSSV Blowdown (%) -0.0 Tube Plugging - average (%) 20 MFW coastdown time (sec) 7 Min. EFW Flow Rate (gpm) 550 EFW Delay Time (sec) 60 Low SG level setpoint for EFW actuation (in. above UFLTS) 6 Moderator Temperature Coefficient (AK/K/°F) +0.315x 10-4 Doppler Coefficient (AK/K/IF) -1.17x10.5

U.S. Nuclear Regulatory Commission Attachment F 3F0607-05 Page 3 of 7 LOFW ATWS for AMSAC Parameter Arming Value Group 5-7 Insertable Worth for DSS trip (%AK/K) 1.7 RPS High RCS Pressure Setpoint (psia) N/A DSS High RCS Pressure Setpoint (psia) 2464.7 DSS High RCS Pressure Response Time (s) 2.0 Offsite Power Available Single Failure RPS Operator Actions None

U.S. Nuclear Regulatory Commission Attachment F 3F0607-05 Page 4 of 7 A 120-second LOFW ATWS event transient with the AMSAC disabled was simulated with RELAP/MOD2-B&W to confirm that the peak RCS pressure does not exceed the conservative estimate of the ASME Service Level C limit, 3250 psia. A loss of main feedwater was simulated at time zero with a 7-second coastdown. The RCS temperature and pressure begin to increase due to the reduction in heat transfer. The RCS reaches the DSS trip setpoint and the PORV open setpoint within the first 20 seconds. The peak RCS pressure reached a maximum of 2616.0 psia at approximately 22.9 seconds, which is below the pressure limit of 3250 psia. The EFIC low SG level EFW actuation setpoint is reached at about 33 seconds with EFW available 60 seconds later. These results confirm that the AMSAC is not required at or below 52% of 2609 MWt to prevent the RCS pressure from exceeding ASME acceptance criteria. Therefore, the 50% power arming setpoint remains valid for CR-3. The sequence of events for this transient is provided in Table F-2. A plot of the RCS pressure and temperature responses are provided in Figures F-1 and F-2. A second case was also run assuming that the DSS was not available. In this case, the peak RCS pressure was 3164.9 psia, which is also less than the 3250 psia limit.

The pressure response for the second case is included as Figure F-3.

Table F-2: Sequence of Events - LOFW ATWS w/ AMSAC Disabled Transient Parameter Time, seconds Loss of Main Feedwater 0.0 MFW terminated 7.0 Pressurizer Spray flow begins -13.0 DSS High RCS Pressure signal 18.6 PORV initial lift -19.0 Control Rods begin to fall 20.6 Peak RCS Pressure (2615.5 psia) 22.9 SG Level Reaches 6 inches, EFIC Setpoint -33.2 EFW initiated SG-A/SG-B -93.2 Peak RCS Average Temp. -97.0 Analysis Terminated 120.0 F.

2.0 REFERENCES

F.2.1 BAW-10164P-A, Revision 4, "RELP5/MOD 2-B&W - An Advanced Computer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis."

F.2.2 BAW-10193P-A, Revision 0, "RELAP5/MOD2-B&W for Safety Analysis of B&W-Designed Pressurized Water Reactors."

U.S. Nuclear Regulatory Commission Attachment F 3F0607-05 Page 5 of 7 Figure F-i: RCS Pressure for LOFW ATWS for AMSAC with DSS at 52% Power 3300 ___________________

' 3250 osia (LOFW ATWS Accebtance Lin-fit) -_:D Leg 1

-+Hot

-HotiLeg 2 Core o- Exit i Top of Pressurizer

ý-I Lo er RV Plenum, Maximnum RCS Pressure 3100 --- -- -----Wý------ ---- --

ý500 ... ... ... ...A

... [,. . . . . . .......... . ......

2700 ...... ... .... ., .. . . . . . .. ...... .re s e 2500 :,:i

-2300 .. ...- - ------- -------- -- -----

kVý 2100 0 20 40 60 80 100 120 140 160 Tin-*- )

U.S. Nuclear Regulatory Commission Attachment F 3F0607-05 Page 6 of 7 Figure F-2: RCS Temperature for LOFW ATWS for AMSAC with DSS at 52% Power nn

-:3,RCS Ave'Temp

-E" Loop -lot I* Leg

-CLoop 2 Hot Leg

-- ts Loop lA ,ColdLeg Loop 18 Cold Leg Loop. 2A Cold Leg q701 Loop 28 CoidLe 640

. ...... . -.' .......... * ............ "....................... Ir.--. --

610 7i 580 550 520, ' .....

-0 -o 40 60 80 i~o 120 l4b '160 T"ie,(s)

U.S. Nuclear Regulatory Commission Attachment F 3F0607-05 Page 6 of 7 Figure F-2: RCS Temperature for LOFW ATWS for AMSAC with DSS at 52% Power Inn,. -

RCS,Ave'Temp

- Lop1Hot, Leg

-Loop 2 Hot Leg

-- Loop 1A Cold Leg

-I Loop IB Cold Leg Loop 2A Cold Leg 670 -----------------------

-Loop 2B C61oLeg 6401 ------------------------

r -----------------------

Li..

610 ------------------------------------------

-v H --- --- -------------- ----- -- ------ ---- -

580 ---------------------------

550 .............-----------------------

200

... .0 "0 "010 1o 40 60 so too 120 14b 1'60 Tine(s)

U.S. Nuclear Regulatory Commission Attachment F 3F0607-05 Page 7 of 7 Figure F-3: RCS Pressure for LOFW ATWS for AMSAC without the DSS at 52% Power 3300 -____________________

3250. sia (LO*W-ATWS Accebtnce Lirnit) Hot Leg 1

-C)__

---aHOtLeg2 310 --------

SCore Exit.

31.'

Top of Pressurizer, Lower RV Plenum Maximum ROS Pressure

,2900 --------;---*- . - - -----

,.,.-.,-,~~ ~ r---7 7--------.............................

2700------------------------------------------------------

.2500. . . ..

2100 20 40 60 i0 100 120 .140 6 Tilli (5)

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #296, REVISION 1 ATTACHMENT G LIST OF REGULATORY COMMITMENTS

U.S. Nuclear Regulatory Commission Attachment G 3F0607-05 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Florida Power Corporation (FPC) in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Mr. Paul Infanger, Supervisor, Licensing & Regulatory Programs at (352) 563-4796.

Regulatory Commitments Due Date Administrative controls will be added to CP-500 for situations where the prior to LEFM CheckPlusTM system or other specific heat balance uncertainty inputs implementation is unavailable. These controls will address maximum power and inputs into the heat balance calculation as well as the allowed outage time for the LEFM CheckPlusTM system to be inoperable.

This requirement will state that if either LEFM or any low-uncertainty heat balance input parameters are inoperable, then reduce power to < 2568 MWt within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reduce the nuclear overpower - high setpoint to < 103.3%

RTP within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Should the LEFM system become unavailable, the current flow nozzle-based feedwater flow and RTD feedwater temperature instrumentation will be used as input to the core power calorimetric, and the core power will be limited to the current licensed power level of 2568 MWt.

A preventative maintenance program will be developed for the LEFM using prior to the vendor's maintenance and troubleshooting manual. implementation CR-3 will complete all LEFM and associated modifications. prior to

" Installation of the Caldon system. implementation

" Addition of new Feedwater and Main Steam pressure and Main Steam Temperature instrumentation.

" Modification of AULD software.

" The Control Room plant reference simulator.

CR-3 will complete the training of operators on the LEFM modification and prior to actions to be taken if the system is inoperable. implementation Alden Labs calibration and test data will be provided to the NRC once 09/01/2007 completed.

3F0607-05, License Amendment Request 296, Revision 1, Measurement Uncertainty Recapture Uprate

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1.0 DESCRIPTION

2.0 BACKGROUND

3.0 TECHNICAL ANALYSIS

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3F0607-05, License Amendment Request 296, Revision 1, Measurement Uncertainty Recapture Uprate

Text

Subject:

References:

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1.0 DESCRIPTION

2.0 BACKGROUND

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1.0 REFERENCES

Subject:

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SUMMARY

SUMMARY

SUMMARY

7.0 REFERENCES

SUMMARY

7.0 REFERENCES

SUMMARY

2.0 REFERENCES

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