ML11326A231

Request for Additional Information for Extended Power Uprate License Amendment Request

ML11326A231
Person / Time
Site:	Crystal River
Issue date:	12/07/2011
From:	Siva Lingam Plant Licensing Branch II
To:	Franke J Florida Power Corp
Lingam S
References
TAC ME6527
Download: ML11326A231 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 7,2011 Mr. Jon A. Franke, Vice President Crystal River Nuclear Plant (NA2C)

ATTN: Supervisor, licensing & Regulatory Programs 15760 W. Power Line Street Crystal River, Florida 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 (two letters), August 18 and 25, 2011, and October 11 and 25,2011, 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 (MWt) to 3014 MWt for Crystal River Unit 3 Nuclear Generating Plant.

The 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. Phil Rose of your staff on November 22,2011, and it was agreed that a response to the enclosed request for additional information would be provided within 45 days from the date of this letter.

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

Sincerely,

~<f'~

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 TO 3014 MEGAWATTS 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 (two letters), August 18 and 25, 2011, and October 11 and 25, 2011 (ADAMS Accession Nos. ML112010674, ML11228A032, ML11234A051, ML11234A427, ML11242A140, ML112860156, and ML113040176, respectively), Florida Power Corporation (the licensee),

doing business as Progress Energy Florida, Inc., submitted a license amendment request (LAR) for an extended power uprate (EPU) to increase thermal power level from 2609 megawatts (MWt) to 3014 MWt for Crystal River Unit 3 Nuclear Generating Plant (Crystal River 3 or CR-3).

In order to complete its review of the above documents, the Nuclear Regulatory Commission (NRC) staff needs the following additional information from the following technical branches:

Fire Protection Branch (AFPB)

1. Attachment 1 to Matrix 5 ("Supplemental Fire Protection Review Criteria, Plant Systems") of NRC RS-001, Revision 0, Review Standard for Extended Power Uprates, states that "power uprates typically result in increases in decay heat generation following plant trips. These increases in decay heat usually do not affect the elements of a fire protection program related to (1) administrative controls, (2) fire suppression and detection systems, (3) fire barriers, (4) fire protection responsibilities of plant personnel, and (5) procedures and resources necessary for the repair of systems required to achieve and maintain cold shutdown. In addition, an increase in decay heat will usually not result in an increase in the potential for a radiological release resulting from a fire. However, the licensee's LAR should confirm that these elements are not impacted by the extended power uprate."

The NRC staff notes that Attachment 5 of the LAR, "Crystal River Unit 3 Extended Power Uprate Technical Report," Section 2.5.1.4.2, page 2.5.1.4-2, addresses all items (1) through (5). The NRC staff requests that the licensee provide statements to address that the additional decay heat will not result in an increase in the potential for a radiological release resulting from a fire at EPU.

2. The NRC staff notes that Attachment 5 of the LAR, Section 2.5.1.4.2, states that CR-3 has reviewed the impact of EPU on various elements of the CR-3 fire protection program. The review concluded that the EPU does not adversely affect the elements of fire protection program including the procedures and resources necessary for the repair of systems required to achieve and maintain cold shutdown. The NRC staff requests the licensee to verify that additional heat in the plant environment from the EPU will not (1) interfere with required operator manual actions being performed at their designated time, or (2) require any new operator actions to maintain hot shutdown and then place the reactor in a cold Enclosure

-2 shutdown condition.

3. The NRC staff notes that Attachment 5 of the LAR, Section 2.5.1.4.2, page 2.5.1.4-2, states that, " ... EPU does not introduce any plant equipment failure modes which will adversely impact the ability to achieve any of the alternate shutdown functions ... " It is unclear to the NRC staff whether there are fire protection program plant modifications planned (e.g., adding new cable trays, or re-routing existing cables, or increases in combustible loading affecting fire barrier ratings, or changes to administrative controls) at EPU conditions. Clarify whether this request involves plant modifications, or changes to the fire protection program, including any proposed modifications to implement transition to Title 10 "Energy" of the Code of Federal Regulations (10 CFR), Section 50.48(c). If any, the NRC staff requests the licensee to identify proposed modifications and discuss the impact of these modifications on the plant's compliance with the fire protection program licenSing basis, 10 CFR 50.48, or applicable portions of 10 CFR Part 50, Appendix R.

4. The NRC staff notes that Attachment 5, "Crystal River Unit 3 Extended Power Uprate Technical Report," Section 2.5.1.4.2, on page 2.5.1.4-2, states that, " ... Safe Shutdown (SSD) Thermal-Hydraulic (T-H) results are impacted due to increased decay heat. .. " The NRC staff requests the licensee to verify that the plant can meet the 72-hour requirements in both 10 CFR Part 50, Appendix R, Sections III.G.1.b and /lI.L, with increased decay heat at EPU conditions.

5. Some plants credit aspects of their fire protection system for other than fire protection activities (e.g., utilizing the fire water pumps and water supply as backup cooling or inventory for nonprimary reactor systems). If CR-3 credits its fire protection system in this way, the LAR should identify the specific situations and discuss to what extent, if any, the EPU affect these "nonfire-protection" aspects of the plant fire protection system. If CR-3 does not take such credit, the NRC staff requests that the licensee verify this as well.

In your response discuss how any nonfire suppression use of fire protection water will impact the ability to meet the fire protection system design demands.

Health Physics and Human Performance Branch (AHPB)

Human Factors:

6. Table 2.12.1-8 of Attachment 5, "CR-3 Extended Power Ascension Test Plan," indicates that plant radiation surveys will be performed continuously between 0% and 100% full power to verify expected dose rates. Describe the scope of these surveys and provide a listing of plant areas where you will conduct radiation surveys following the proposed EPU implementation and describe your criteria for selecting these areas. Verify that they include surveys of all plant areas potentially affected by operations at the EPU full power level.

Indicate whether survey plans for these areas include collection of benchmark data needed to assess the impact of the EPU on radiation levels.

7. Page 2.10.1-3 provides the maximum dose to any individual at the plant from 2004 through 2008. Does "any individual at the plant" refer only to permanent employees or does it include contractors?

-3

8. Page 2.10.1-7 of Attachment 5 states that based on small break loss-of-coolant accident (LOCA) results, the time available for actions in the emergency diesel generator (EDG) rooms to maintain a dose less than 5 rem is reduced from 25 minutes to approximately 10 minutes, which still provides sufficient time to perform short compensatory actions in the EDG rooms. Provide a detailed analysis that demonstrates that 10 minutes is sufficient time for the operator to access and egress the area and perform the required actions in the EDG rooms under LOCA conditions (e.g., while wearing protective equipment).

9. Table 2.9.2-16 of Attachment 5 summarizes the CR-3 EPU LOCA Radiological Consequences. The EPU doses for the Exclusion Area Boundary, Low Population Zone and Main Control Room are greater than the pre-EPU doses by about a factor of 1.63. However, the EPU dose for the Technical Support Center (3.02 rem) is lower than the pre-EPU dose (4.71 rem). Provide additional information to explain the decrease in the dose for the Technical Support Center.

Instrumentation and Controls Branch (EICB)

10. Provide the diagrams or photographs of a" inadequate core cooling mitigation system (ICCMS), fast cooldown system, and atmospheric dump valves related controls and indications in the main control room, showing the information the operators have available to monitor and control.

Piping and NDE Branch (EPNB)

Applicable Application Section, 2.1.5, "Reactor Coolant Pressure Boundary (RCPB) Materials."

11. Under the heading, General CorrosionlWastage of Carbon Steel Components, Section 2.1.5.2 discusses the boric acid corrosion (BAC) program at Crystal River, Unit 3 (CR-3).

(1) Discuss how often the visual inspection will be performed under the BAC program.

(2) Clarify whether all RCPB components (piping and vessels) will be inspected and managed under the BAC program. Identify and justify any RCPB components that will not be inspected under the BAC program. (3) Explain why the BAC program will not be affected by the EPU. (4) Discuss the impact of EPU on general corrosioniwastage of RCPB components because the licensee discussed mainly the BAC program and Electric Power Research Institute (EPRI) water chemistry guidelines without addressing the impact of EPU on general corrosion/wastage of RCPB components.

12. Under the heading, SCC [Stress Corrosion Cracking] of Austenitic Stainless Steels, Section 2.1.5.2 focuses on the treatment of the reactor coolant inside the pipe. NRC Information Notice 2011-04, "Contaminants and Stagnant Conditions Affecting Stress Corrosion Cracking in Stainless Steel Piping in Pressurized-Water Reactors [PWRs],"

discusses stress corrosion cracking initiated from the outside diameter of austenitic stainless steel piping. The EPU conditions may cause higher stresses in the piping and can affect SCC. Therefore, discuss the impact of EPU on the potential of outside diameter SCC (ODSCC) on RCPB piping and the program to monitor the ODSCC on stainless steel piping.

13. Under the heading "Alloy 600/82/182 Components at CR-3," Section 2.1.5.2 discusses impact of EPU on primary water stress corrosion cracking (PWSCC) of Alloy 82/182

-4 dissimilar welds. (1) Identify any RCPB piping that have unmitigated Alloy 821182 dissimilar butt welds. (2) Discuss the impact of EPU on the integrity of those pipes with unmitigated Alloy 82/182 welds and how the unmitigated pipes will be monitored to minimize potential PWSCC. (3) Section 2.1.5.2 states that the pressurizer surge line was mitigated in 2009.

The NRC staff understands that the licensee installed a weld overlay on the surge line.

Based on the NRC-approved weld overlay relief request, a flaw growth calculation must be performed for the surge line. Discuss whether the flaw growth calculation has been updated to include the EPU conditions. If not, provide justification. (4) Discuss whether the safety, relief, and spray nozzles of the pressurizer have been mitigated. If yes, subquestion (3) above applies. (5) Discuss whether an Alloy 600 Program has been implemented to monitor and manage PWSCC of Alloy 600/821182 material. If not, provide justification.

(6) Discuss whether EPU will increase the potential of PWSCC at the bottom-mounted instrument nozzles of reactor pressure vessel.

14. Under the heading "Alloy 600/82/182 Components at CR-3," Section 2.1.5.2 states that

"... [a] review of operating experience regarding effects of lithium and pH on PWSCC by EPRI PWR Water Chemistry Guidelines suggests that there are no adverse effects on PWSCC from the lithiumlboron concentration range. Therefore, there is no PWSCC impact from the EPU water chemistry .... "

The licensee's statements are not clear regarding the impact of EPU on the potential for PWSCC on RCPB piping. (1) Clarify whether EPU will change the lithium and boron concentration. If yes, discuss whether the change will increase the potential for PWSCC in the RCPB piping. (2) Discuss whether EPU will cause unfavorable changes in reactor coolant system (RCS) water chemistry that would lead to the RCPB piping being more susceptible to PWSCC. For example, EPU may cause chemicals to be added or oxygen and hydrogen concentration to be changed. (3) If EPU causes an unfavorable change in RCS water chemistry, discuss how CR-3 will manage the unfavorable changes to minimize degradation.

15. (a) Address the impact of EPU on other degradation mechanisms in the RCPB piping that were not discussed in Section 2.1.5, such as water hammer, flow accelerated corrosion, high cycle fatigue, and environmentally assisted fatigue of the RCPB piping (e.g., thermal stratification in the pressurizer surge line). (b) Discuss the impact of EPU on the fatigue cumulative usage factors of the RCPB piping. Discuss whether the cumulative usage factors have been updated due to EPU. If not, provide justification.

16. In Section 2.1.5.2, under the heading "Thermal Aging," the licensee stated that "A review of the RCS pressure boundary components shows no CASS [cast austenitic stainless steel]

material in the RCS pressure boundary exposed to the T HOT. Therefore, there will be no impact on thermal aging by the EPU." However, reactor coolant pump (RCP) caSing is made of CASS material as discussed in Section 2.1.6. RCP experiences TCOLD temperature, not T HOT. Discuss whether EPU affects the thermal aging of the cold leg piping. If yes, discuss how thermal aging of the RCP casing will be monitored.

Applicable Application Section 2.1.6, "Leak-Before-Break."

17. (a) Confirm that the only piping that CR-3 has been approved for leak-before-break (LBB) is

- 5 the primary loop, and no branch lines. (b) Discuss whether the primary loop piping contains nickel-based Alloy 82/182 welds and discuss whether the primary loop piping has been mitigated to minimize PWSCC. Discuss the mitigation method. (c) If the primary loop piping has not been mitigated, discuss the measures to minimize its potential of PWSCC under EPU. (d) If the primary loop piping has been mitigated with a weld overlay discuss whether the original LBB evaluation has been updated per NRC Regulatory Issue Summary 2010-07.

18. Section 2.1.6.2 specifies an acceptance criterion that requires the "Minimum Moment" loads calculated for the EPU be greater than the "Minimum Moment" loads in the original LBB analysis (BAW-1847). Note (2) to Table 2.1.6-2 (minimum moment loads) states that the minimum moment load calculated using EPU conditions for the 36-inch inside diameter hot leg straight pipe is less than the minimum moment load in the original LBB analysis. This implies that the 36-inch hot leg pipe does not satisfy the above acceptance criterion.

However, Note (2) explains that Table 2.1.6-1 (maximum moment loads) shows that the EPU maximum bending moment at this location is much smaller than the maximum moment load in the original LBB analysis, meaning that the flaw size required for a flow rate of 10 gallons per minute is stable enough to not experience growth under the EPU maximum moment loading. Discuss whether the 36-inch hot leg pipe under the EPU conditions will satisfy the margin of 2 for crack size and margin of 10 for the leakage as specified in Standard Review Plan 3.6.3. If not, provide justification for the proposed EPU.

Component Performance and Testing Branch (EPTB)

19. On page 2.2.4-2, it is stated:

"If not discussed below, no adverse impact was found to the pumps or valves."

This statement is very vague and does not describe the adequacy of the evaluation for EPU conditions for pumps and valves. Please list and describe each plant system and component that did not experience adverse impacts to safety-related pumps and valves when reviewed for EPU conditions.

20. Please identify any modifications to your Appendix J program as a result of the EPU.

21. Please describe your air-operated valve (AOV) program and any modifications as a result of the EPU.

22. On page 2.2.4-4, it is stated:

"Calculations reviewing the stroke times for MOVs [motor operated valves] predict that the stroke times will increase minimally, therefore the impact to stroke time is negligible."

This statement is unclear. Please better define "minimally" and "negligible" in this statement.

23. Please describe the lessons learned programs for MOVs and AOVs.

- 6

24. Throughout section 2.2.4, often it is stated at the end of the section that "Changes are reflected in the Inservice Testing Program." This statement is very vague. For each instance of a change to the Inservice Testing (1ST) Program in this section, please provide a more detailed description of the change that will be instituted and how it impacts the 1ST program.

25. The weak link component capability and maximum actuator capacities of the MOVs, post-EPU, are provided on Table 2.2.4-2 on page 2.2.4-6. For valves FWV-33 and FWV-36 , please identify if analysis was performed to determine their susceptibility to a "hot short" since the weak link component capabilities are less than the maximum actuator capabilities.

Steam Generator Tube Integritv and Chemical Engineering Branch (ESGB)

26. On page 2.1.7-2 of attachment 5 of its letter dated June 15, 2011, the licensee stated that a new design-basis accident (DBA) test was performed to qualify the use of the Carboline Carboguard 2011 SN surface topcoated with Carboline Carboguard 890N for concrete substrates. It was demonstrated that the DBA qualification test report provides the basis for qualification for these coating systems and bounds EPU conditions. Please clarify whether all Service Level 1 coatings have been qualified to meet design basis LOCA containment EPU conditions for temperature, pressure and radiation.

27. On page 2.8.6.2-1 of Attachment 5 of its letter dated June 15, 2011, the licensee stated that Spent Fuel Pools A and B utilize boron carbide and Boral, respectively, as the neutron absorbing materials at CR-3. It is not clear to the staff what surveillance approach will be implemented and how it will demonstrate that the neutron absorbing materials will continue to perform their intended function. As such, please discuss in detail the surveillance approach that will be used for monitoring the neutron absorber materials, specifically the methods of neutron attenuation testing, frequency of inspection, sample size, data collection, and acceptance criteria.

28. In its letter dated, June 15, 2011, the licensee stated the following about the flow accelerated corrosion (FAC) program for CR-3:

If a component is considered susceptible to FAG but cannot be inspected, it is analytically evaluated using the GHEGKWORKS Pass 2 results. The analytical predictions are then compared to actual wear rate results for actually inspected, usually adjacent, components which have the same fluid conditions. These results are used to trend the un-inspected component and if possible, a visual inspection to confirm them.

The CHECWORKS Pass 2 analysis uses plant inspection data to refine the Pass 1 wear rate predictions. Please explain how a component can be analytically evaluated using the CHECWORKS Pass 2 results from a different component.

29. The FAC monitoring program includes the use of a predictive method to calculate the wall thinning of components susceptible to FAC. In order for the staff to evaluate the accuracy

-7 of these predictions, the staff requests a sample list of components for which wall thinning is predicted and measured by ultrasonic testing or other method. Include the initial wall thickness (nominal), current (measured) wall thickness, and a comparison of the measured wall thickness to the thickness predicted by the CHECWORKS FAC model.

Vessels and Internals Integrity Branch (EVIB)

30. The EPU Technical Report, Table 2.1.1-3 indicates that Babcock & Wilcox Owners Group (B&WOG) Capsule A5 fulfills the requirement of American Society for Testing and Materials

[ASTM] E185-82 for the last capsule to receive not less than once nor greater than twice the peak EOl vessel fluence. Table 2.1.1-1 of the EPU Technical Report shows the peak vessel inner surface fluence at EOl (50.3 EFPY [effective full year power]) as 1.57 x 1019 n/cm 2. However, BAW-1543, Revision 4, Supplement 6, Table VI shows the fluence received by B&WOG CapsuleA5 as 0.637-1.042 x 10 19 n/cm 2

• Explain how B&WOG Capsule A5 is meeting the requirement for the last capsule since the fluence received is less than one times the projected EOl vessel fluence.

December 7, 2011 Mr. Jon A. Franke, Vice President Crystal River Nuclear Plant (NA2C)

ATTN: Supervisor, Licensing & Regulatory Programs 15760 W. Power Line Street Crystal River, Florida 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 (two letters), August 18 and 25, 2011, and October 11 and 25, 2011, 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 (MWt) to 3014 MWt for Crystal River Unit 3 Nuclear Generating Plant.

The 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. Phil Rose of your staff on November 22, 2011, and it was agreed that a response to the enclosed request for additional information would be provided within 45 days 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:

PUBLIC LPL2-2 RIF RidsNrrDorlLpl2*2 RidsNrrPMCrystalRiver RidsNrrLACSola RidsOgcMailCenter RidsNrrDorlDpr RidsAcrsAcnw_MailCTR S. Ungam, NRR RidsRgn2MailCenter RidsNrrDraAfpb RidsNrrDraAhpb RidsNrrDeEpnb RidsNrrDeEptb RidsNrrDeEsgb RidsNrrDeEvib N. Iqbal, NRR M. Jimenez, NRR J. Tsao, NRR M. Orenak, NRR A. Obodoako, NRR C. Hunt, NRR E. Andruszkiewicz, NRR RidsNrrDeEicb ADAMS Accession No: ML11326A231 OFFICE LPL2*2IPM LPL2-21LA AFPB/BC EPNB/BC NAME SUngam CSoia (BClayton for) AKlein* TLupold*

DATE 11/23/11 11/23/11 10/26/11 7/12/11 OFFICE EPTB/BC ESGB/BC EVIB/BC NAME AMcMurtray' VCusumano (At HGonzalez (A DATE 11/1/11 11/1/11 11/3/11