ML12333A089

Request for Additional Information for Extended Power Uprate License Amendment Request

ML12333A089
Person / Time
Site:	Crystal River
Issue date:	12/19/2012
From:	Siva Lingam Plant Licensing Branch II
To:	Franke J Progress Energy Carolinas
Lingam, Siva
References
TAC ME6527
Download: ML12333A089 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 19,2012 Mr. Jon A. Franke Vice President Crystal River Nuclear Plant (NA2C)

ArrN: Supervisor, Licensing and Regulatory Programs (NA 1B) 15760 W. Power Line Street Crystal River, FL 34428-6708 SU8.JECT: CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT - REQUEST FOR ADDITIONAL INFORMATION FOR EXTENDED POWER UPRATE LICENSE AMENDMENT REQUEST (TAC NO. ME6527)

Dear Mr. Franke:

By letter dated June 15, 2011, as supplemented by letters dated July 5, 2011, August 11, 2011 (two letters), August 18, 2011, August 25, 2011, October 11, 2011, October 25, 2011, December 15, 2011 (two letters), December 21,2011, January 5,2012 (two letters),

January 19, 2012 (two letters), January 31,2012, March 19, 2012, March 22,2012, April 4, 2012 (two letters), April 12, 2012, April 16, 2012, April 26, 2012, June 18, 2012, June 29,2012, July 17, 2012 (two letters), July 31,2012 (two letters), August 21,2012 (two letters), August 30,2012, September 6,2012, September 17,2012, September 27,2012, October 4, 2012, October 11, 2012, and November 7, 2012, Florida Power Corporation, doing business as Progress Energy Florida, Inc., submitted a license amendment request for an extended power uprate to increase thermal power level from 2609 megawatts thermal (MWt) to 3014 MWt for Crystal River Unit 3 Nuclear Generating Plant The U.S. Nuclear Regulatory Commission staff is reviewing the submittal and has determined that additional information is required to complete its evaluation. This request was discussed with Mr. Dan Westcott of your staff on December 18, 2012, and it was agreed that a response to the enclosed request for additional information (RAI) would be provided within 90 days for RAls 2.8.5.2.1.4, 2.8.5.2.3.4, 2.8.5.4.1.1, 2.8.5.4.2.1 and 2.8.5.4.5.4, and 45 days for rest of the RAls from the date of this letter.

If you have any questions regarding this matter, I can be reached at 301-415-1564.

Sincerely,

~.>.r..--tL ~. ~"'N~

Siva P. Lingam, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-302

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION REGARDING EXTENDED POWER UPRATE TO INCREASE THERMAL POWER LEVEL FROM 2609 MEGAWATTS THERMAL TO 3014 MEGAWATTS THERMAL CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302 By letter dated June 15, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML112070659), as supplemented by letters dated July 5, 2011, August 11, 2011 (two letters), August 18, 2011, August 25, 2011, October 11, 2011, October 25, 2011, December 15, 2011 (two letters), December 21, 2011, January 5, 2012 (two letters), January 19, 2012 (two letters), January 31, 2012, March 19, 2012, March 22, 2012, April 4, 2012 (two letters), April 12, 2012, April 16, 2012, April 26, 2012, June 18, 2012, June 29,2012, July 17, 2012 (two letters), July 31,2012 (two letters), August 21,2012 (two letters), August 30,2012, September 6,2012, September 17, 2012, September 27,2012, October 4, 2012, October 11, 2012, and November 7, 2012 (ADAMS Accession Nos.

ML112010674, ML11228A032, ML11234A051, ML11234A427, ML11242A140, ML112860156, ML113040176, ML11354A232, ML11354A233, ML11361A460, ML12011A035, ML12030A209, ML12024A300, ML12024A301, ML12032A280, ML12081A293, ML12086A107, ML120970114, ML12097A246, ML12107A216, ML12114A002, ML12118A498, ML12173A391, ML122060421, ML122050452, ML12205A268, ML12216A354, ML12216A355, ML12240A009, ML12240A010, ML12254A241, ML12251A250, ML12264A087, ML12272A344, ML12279A383, ML12286A329, and ML123140487, respectively), Florida Power Corporation (the licensee) submitted a license amendment request for an extended power uprate (EPU) to increase thermal power level from 2609 megawatts thermal (MWt) to 3014 MWt for Crystal River Unit 3 Nuclear Generating Plant (CR-3). Portions of the letters dated June 15, 2011, August 11, 2011 (ADAMS Accession No. ML11228A032), January 31,2012, June 18, 2012, July 17, 2012 (ADAMS Accession No. ML122050452), August 30, 2012, and September 17, 2012, contain sensitive unclassified non safeguards information and, accordingly, those portions have been withheld from public disclosure. In order to complete its review of the above documents, the U.S. Nuclear Regulatory Commission (NRC) staff requests additional information originating from its Reactor Systems Branch (SRXB). The following input begins at Section 2.8.4.2 of Attachment 5 or 7 of your original application dated June 15, 2011, and continues through the end of Section 2.8.

SRXB REQUEST FOR ADDITIONAL INFORMATION 2.8.4.3 Overpressure Protection During Low Temperature Operation 2.8.4.3.1 Please address the reduced exposure over which current low-temperature overpressure protection (LTOP) and pressure/temperature limits are valid by confirming that the Technical Specification (TS) is not limited by effective full-power year, or by revising the applicability period contained in TS 3.4.11.

Enclosure

- 2 2.8.4.3.2 Please verify that there were no changes to lower mode mass and energy input sources that would require revisiting the LTOP relief system capacity.

2.8.4.4 Residual Heat Removal System 2.8.4.4.1 Page 2.8.4.4-3 of the TR indicates that there are improved actions that could require that the plant be in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Please provide citations to the specific requirements for added clarity, and explain how these requirements "could require" cold shutdown in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, as stated in the TR.

2.8.4.4.2 In CR-3 Updated Final Safety Analysis Report Section 9.4 (Page 20), the following information is provided:

a. Decay Heat Removal Pumps Two 100% capacity pumps are arranged in parallel. Each is capable of continuous operation during the decay heat removal mode and during refueling operations. Both pumps are available for emergency operation.

The design flow is that required to cool the RC [reactor coolant] system from 280 of to 140 of in 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, assuming a Nuclear Service and Decay Heat Sea Water (RW) system temperature of 85 of. The steam generators are used to cool the RC system [RCS] from operating temperature to the 280 of temperature.

b. Decay Heat Removal Heat Exchangers During a routine shutdown one heat exchanger is capable of removing decay heat from the circulated reactor coolant. Both heat exchangers are operated to cool the circulated reactor coolant from 280 of to 140 of in 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, assuming a RW system temperature of 85 of. As RW system temperature rised (to a maximum of 95 OF) cooldown time will extend beyond 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />, but this increase in time is inconsequential. With a single heat exchanger in service, cooldown to 140 of will require 7 days.

c. Borated Water Storage Tank The Borated Water Storage Tank (BWST) is located outside the Reactor building and the Auxiliary Building. It contains a minimum of 2,270 ppm

[parts per million] boron in solution and is used both for emergency core injection and filling the fuel transfer canal during refueling. The BWST also supplies borated water for emergency cooling to the Reactor Building Spray (BS) system, the DH [decay heat] system, and the MU [make-up]

system.

In light of the above, please compare the Final Safety Analysis Report (FSAR) discussion to the evaluations discussed in this section of the TR. Explain whether these functional specifications remain applicable, whether this information is historical, or whether this information will be updated. If it will be updated, please provide the revised FSAR text.

- 3 2.8.5 Accident and Transient Analyses 2.8.5.1 Increase in Heat Removal by the Secondary System 2.8.5.1.1 Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow, and Inadvertent Opening of a Steam Generator Relief or Safety Valve 2.8.5.1.1.1 Provide an evaluation of events in this category relative to modifications to secondary heat removal capability. Include modifications to normal operational systems, such as main feedwater and main steam systems, as well as engineered safety systems, such as the proposed emergency feedwater initiation and control system.

2.8.5.1.1.2 Provide information to demonstrate that, at planned EPU conditions, events of this category remain non-limiting, such that planned EPU modifications would not, for example, create a possibility for an accident of a different type than any previously evaluated in the final safety analysis report or result in more than a minimal increase in the consequences of a malfunction of a system, structure, or component important to safety, two criteria that, if satisfied, would cause otherwise cause the proposed modifications to require a license amendment.

2.8.5.1.1.3 Relative to the information contained in the TR, provide additional and more specific information concerning axial offset limit determination. Describe the limiting overcooling events that are specifically analyzed, and provide information to demonstrate that the increase in secondary heat removal events remain bounded by other anticipated operational occurrences (AOOs) that are within the CR-3 current licensing basis.

2.8.5.1.1.4 Provide an evaluation of the effects of an untripped overpower transient relative to fuel cladding strain acceptance criteria.

2.8.5.1.2 Steam System Piping Failures Inside and Outside Containment NRC Staff review of this section continues.

2.8.5.2 Decrease in Heat Removal by the Secondary System 2.8.5.2.1 Loss of External Load, Turbine Trip, Loss of Condenser Vacuum, and Steam Pressure Regulator Failure 2.8.5.2.1.1 The TR, in this and other sections, describes the reactor trip on high pressure as follows: "Reactor trip was modeled to occur on a nominal high RCS pressure setpoint plus uncertainty (2400 pounds per square inch absolute (psia))." Please clarify whether 2400 psia is the nominal value or the model value. Also, please discuss the relationship between the nominal value, the modeled value, and the setpoint values contained in the TSs.

-4 2.8.5.2.1.2 Provide information concerning the basis for the assumed trip delay times for both high neutron flux and high pressure.

2.8.5.2.1.3 The TR states that 0 percent steam generator tube plugging was modeled. It is apparent to the reviewer that this assumption would maximize the primary-to secondary heat transfer, and therefore result in the most prompt delivery of core energy to the main steam system. It is not, however, apparent that such an assumption is conservative and appropriate for modeling to determine the peak ReS pressure. Please justify this assumption, or quantify the effect of changing the input to a more limiting value.

2.8.5.2.1.4 The supplemental analysis for the main feedwater line break identifies a number of current licensing basis modeling assumptions that are not necessarily reflective of the most limiting permissible operating conditions relative to the proposed EPU operating condition. In light of this information, please evaluate the transients in this section with respect to initial reactor coolant flow, reactor coolant system average temperature, initial turbine header pressure, and steam generator operate range level, to confirm that the initial conditions are chosen to maximize the reactor coolant and/or main steam system pressure response, as appropriate.

2.8.5.2.2 Loss of Nonemergency AC [Alternating Current] Power to the Station Auxiliaries No RAls for this section at this time.

2.8.5.2.3 Loss of Normal Feedwater Flow 2.8.5.2.3.1 The TR states that the loss of feedwater AOO is the limiting transient in terms of establishing the minimum emergency feedwater (EFW) flow requirements. Please discuss this statement in the context of Title 10 of the Code of Federal Regulations, Part 50, Section 50.36 requirements. For example, cite the applicable TS requirements that pertain to EFW, discuss what changes are necessary. If this information is contained in material that has already been submitted, a reference to this material is acceptable.

2.8.5.2.3.2 Pressurizer safety valves, for this and other transients, are modeled with 3 percent lift tolerance, with 0 percent accumulation and 4 percent blowdown. Please justify this modeling assumption relative to valve design and observed performance capabilities.

2.8.5.2.3.3 Provide similar justification as in 2.8.5.2.3.2 relative to the main steam safety valves modeling assumptions.

2.8.5.2.3.4 The supplemental analysis for the main feedwater line break identifies a number of current licensing basis modeling assumptions that are not necessarily reflective of the most limiting permissible operating conditions relative to the proposed EPU operating condition. In light of this information, please evaluate the transients in this section with respect to initial reactor coolant flow, reactor coolant system average

-5 temperature, initial turbine header pressure, and steam generator operate range level, to confirm that the initial conditions are chosen to maximize the reactor coolant and/or main steam system pressure response, as appropriate.

2.8.5.2.3.5 Please provide additional information to justify the chosen steam generator tube plugging modeling selection.

2.8.5.2.3.6 The TR discusses a separate analysis, performed in a nominal condition, which is used to confirm the TS value for the pressurizer water level upper limit. Please provide additional information explaining how this analysis accomplishes that purpose.

2.8.5.2.4 Feedwater System Pipe Breaks Inside and Outside Containment By letter dated July 17, 2012, the licensee provided ANP-3114(P}, which discussed sensitivity studies performed on the initial conditions. The sensitivity studies identified a new set of limiting initial conditions. It is this analysis, and the associated initial conditions, that the NRC staff evaluated in support of the proposed EPU.

2.8.5.2.4.1 The limiting results from ANP-3114(P) are slightly less severe than those presented in the TR. The base case evaluated in ANP-3114(P) is significantly less severe than the analysis reported in the TR. Although ANP-3114 indicates that the TS minimum value for EFW was used in the analysis, the peak pressure occurs before EFW flow initiates, both in the TR analysis and in ANP-3114(P). Please identify the modeling assumptions that differed between the TR analysis and ANP-3114(P), which caused the ANP-3114(P) base case results to be significantly less severe.

2.8.5.2.4.2 Explain why the feedwater line break (FWLB) results are significantly more severe than the loss of normal feedwater (LONF) results.

2.8.5.2.4.3 Discuss any additional available analytic operating experience with the FWLB for other Babcock & Wilcox plants. How do the CR-3 results differ from results for other plants, and why are they different?

2.8.5.2.4.4 The TR states that the LONF event is used to establish TS requirements for the EFW system. Explain what role the FWLB event analysis plays in establishing TS requirements.

2.8.5.2.4.5ANP-3114(P) indicated that steam generator initial inventory had a significant effect on the results of the FWLB analyses. Please provide additional information concerning the steam generator operating characteristics, both at pre-EPU and post EPU power levels. Additional information concerning the original once through steam generators would also facilitate further comparison between the current licensing basis results and the EPU results.

- 6 2.8.5.3 Decrease in Reactor Coolant System Flow 2.8.5.3.1 Loss of Forced Reactor Coolant Flow No RAts for this section at this time.

2.8.5.3.2 Reactor Coolant Pump Rotor Seizure and Reactor Coolant Pump Shaft Break 2.8.5.3.2.1 The Locked Rotor Analysis in the CR-3 EPU TR does not clearly show the acceptance criteria and the analytical basis to demonstrate that the criteria are met.

TR section 2.8.5.3.2 states that the thermal design limit is not met, therefore indicating that fuel failure may occur. However, the locked rotor maximum allowable peaking analyses demonstrate that fuel failure does not occur. The alternate source term dose acceptance criteria are used to back-calculate the amount of pins allowed to fail without exceeding the 90 percent limit in TR section 2.9.2. It is not clear what acceptance criteria are being proposed, nor is it clear how the acceptance criteria are being met for the locked rotor event. Please clarify.

2.8.5.4 Reactivity and Power Distribution Anomalies 2.8.5.4.1 Uncontrolled Control Rod Assembly Withdrawal from a Subcritical or Low Power Startup Condition 2.8.5.4.1.1 What is the sensitivity of initial pressure (nominal vs. higher or lower) on the transient that is terminated with an overpressure trip?

2.8.5.4.1.2 If the transient were terminated instead by a neutron flux trip, please explain what sensitivity the trip timing, and corresponding results, would have to the selected initial conditions.

2.8.5.4.2 Uncontrolled Control Rod Assembly Withdrawal at Power 2.8.5.4.2.1 Please evaluate the results of this AOO at a spectrum of power levels, to provide assuance that 3026.1 MWt is the limiting case.

2.8.5.4.2.2 Justify the selection of 4.62 percent millinile [pcm]/sec as a conservative reactivity insertion limit.

2.8.5.4.3 Control Rod Misoperation 2.8.5.4.3. 'I Please explain how the stuck-out control rod assembly and stuck-in control rod assembly events are dispositioned through analyses. How is it confirmed that the dropped rod is the most limiting control rod mis-operation?

2.8.5.4.4 Startup of an Inactive Loop at an Incorrect Temperature 2.8.5.4.4.1 Justify the selected initial conditions in light of the predicted response. How could other permissible initial conditions make the results of this event more severe?

-7 2.8.5.4.5 Chemical and Volume Control System Malfunction that Results in a Decrease in Boron Concentration in the Reactor Coolant 2.8.5.4.5.1 Describe the effects of steam generator tube plugging on this transient. Why is o percent steam generator tube plugging conservative? Does this conservatism exceed the reduction in RCS volume that would cause an increase in the dilution rate?

2.8.5.4.5.2 Describe the extent of operator action required to terminate the event.

2.8.5.4.5.3 Please address the effects of moderator dilution in lower modes of operation.

2.8.5.4.5.4 Please explain the RCS pressure response as a function of moderator dilution rate.

2.8.5.4.6 Spectrum of Rod Ejection Accidents 2.8.5.4.6.1 Please provide a plot of pressure vs. time for the pressure-limiting rod ejection event, and explain why there is such a significant pressure excursion.

2.8.5.6 Decrease in Reactor Coolant Inventory 2.8.5.6.1 Inadvertent Opening of Pressurizer Pressure Relief Valve 2.8.5.6.1.1 Recent NRC staff review experience has indicated that a spurious PORV opening can cause an engineered safety features actuation, associated with the depressurization. This actuation can challenge the RCS by overfilling the pressurizer. Please provide information to address this concern.

2.8.5.6.2 Steam Generator Tube Rupture No RAls related to this section at this time.

2.8.5.6.3 Emergency Core Cooling System and Loss-of-Coolant Accidents NRC Staff review of this section continues 2.8.5.7 AntiCipated Transients without Scram No RAls related to this section at this time.

December 19, 2012 Mr. Jon A. Franke Vice President Crystal River Nuclear Plant (NA2C)

ATTN: Supervisor, Licensing and Regulatory Programs (NA 1B) 15760 W. Power Line Street Crystal River, FL 34428-6708

SUBJECT:

CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT - REQUEST FOR ADDITIONAL INFORMATION FOR EXTENDED POWER UPRATE LICENSE AMENDMENT REQUEST (TAC NO. ME6527)

Dear Mr. Franke:

By letter dated June 15, 2011, as supplemented by letters dated July 5, 2011, August 11, 2011 (two letters), August 18, 2011, August 25, 2011, October 11, 2011, October 25, 2011, December 15, 2011 (two letters), December 21, 2011, January 5, 2012 (two letters),

January 19, 2012 (two letters), January 31, 2012, March 19, 2012, March 22, 2012, April 4, 2012 (two letters), April 12, 2012, April 16, 2012, April 26, 2012, June 18, 2012, June 29,2012, July 17, 2012 (two letters), July 31,2012 (two letters), August 21,2012 (two letters), August 30,2012, September 6,2012, September 17,2012, September 27,2012, October 4, 2012, October 11, 2012, and November 7, 2012, Florida Power Corporation, doing blJsiness as Progress Energy Florida, Inc., submitted a license amendment request for an extended power uprate to increase thermal power level from 2609 megawatts thermal (MWt) to 3014 MWt for Crystal River Unit 3 Nuclear Generating Plant.

The U.S. Nuclear Regulatory Commission staff is reviewing the submittal and has determined that additional information is required to complete its evaluation. This request was discussed with Mr. Dan Westcott of your staff on December 18, 2012, and it was agreed that a response to the enclosed request for additional information (RAI) would be provided within 90 days for RAls 2.8.5.2.1.4, 2.8.5.2.3.4, 2.8.5.4.1.1, 2.8.5.4.2.1 and 2.8.5.4.5.4, and 45 days for rest of the RAls from the date of this letter.

If you have any questions regarding this matter, I can be reached at 301-415-1564.

Sincerely, IRA!

Siva P. Lingam, Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-302

Enclosure:

Request for Additional Information cc w/encl: Distribution via Listserv DISTRIBUTION:

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