3F0112-04, Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR

Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR

ML12011A035
Person / Time
Site:	Crystal River
Issue date:	01/05/2012
From:	Franke J Progress Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F0112-04, TAC ME6527
Download: ML12011A035 (11)

Text

Progress Energy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 January 5, 2012 3F01 12-04 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)

References:

1. CR-3 to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License Amendment Request #309, Revision 0, Extended Power Uprate" (Accession No. ML112070659)

2. NRC to CR-3 letter dated December 7, 2011, "Crystal River Unit 3 Nuclear Generating Plant - Request for Additional Information for Extended Power Uprate License Amendment Request (TAC No. ME6527)" (Accession No. ML11326A231)

Dear Sir:

By letter dated June 15, 2011, Florida Power Corporation, doing business as Progress Energy Florida, Inc., requested a license amendment to increase the rated thermal power level of Crystal River Unit 3 (CR-3) from 2609 megawatts (MWt) to 3014 MWt. On December 7, 2011, the NRC provided a request for additional information (RAI) required to complete its evaluation of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR).

The attachment, "Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 EPU LAR," provides the CR-3 formal response to the RAI needed to support the Steam Generator Tube Integrity and Chemical Engineering Branch (ESGB) technical review of the CR-3 EPU LAR.

In support of the ESGB technical review RAI response, an enclosure, "Wear Rate Analysis:

Combined Summary Report," is being provided which contains a sample list of components for which wall thinning is predicted and measured by ultrasonic testing.

This correspondence contains no new regulatory commitments.

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 3F0112-04 If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Superintendent, Licensing and Regulatory Programs at (352) 563-4796.

Sincerel Jon . a e e President rystal River Nuclear Plant JAF/gwe

Attachment:

Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 EPU LAR

Enclosure:

Wear Rate Analysis: Combined Summary Report xc: NRR Project Manager Regional Administrator, Region II Senior Resident Inspector State Contact

U.S. Nuclear Regulatory Commission Page 3 of 3 3F0112-04 STATE OF FLORIDA COUNTY OF CITRUS Jon A. Franke 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 on A. Franke Vice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this _-_,_ day of

'ja* t4,- 2012, by Jon A. Franke.

-- Signature of Notary Public

ýilýCHARLENE yMILLER PState of Florida t Notary Public S State of Florida My Comm. Expires Nov 12, 20121

" Sandd Through National Notary Assn.

(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 ATTACHMENT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TO SUPPORT NRC STEAM GENERATOR TUBE INTEGRITY AND CHEMICAL ENGINEERING BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR

U. S. Nuclear Regulatory Commission Attachment 3F01 12-04 Page 1 of 4 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TO SUPPORT NRC STEAM GENERATOR TUBE INTEGRITY AND CHEMICAL ENGINEERING BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR By letter dated June 15, 2011, Florida Power Corporation (FPC), doing business as Progress Energy Florida, Inc., requested a license amendment to increase the rated thermal power level of Crystal River Unit 3 (CR-3) from 2609 megawatts (MWt) to 3014 MWt. On December 7, 2011, the NRC provided a request for additional information (RAI) required to complete its evaluation of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR). The following provides the CR-3 formal response to the RAI needed to support the Steam Generator Tube Integrity and Chemical Engineering Branch (ESGB) technical review of the CR-3 EPU LAR.

For tracking purposes, each item related to this RAI is uniquely identified as ESGB X-Y, with X indicating the RAI set and Y indicating the sequential item number.

Steam Generator Tube Integrity and Chemical Engineering Branch (ESGB)

26. (ESGB 1-1)

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.

Response

As noted in Section 2.1.7, "Protective Coating Systems (Paints) - Organic Materials," of the EPU Technical Report (TR) (Reference 1, Attachments 5 and 7) the 1990 DBA test was performed for previous coating products approved for application in the CR-3 Reactor Building.

Afterward, following the CR-3 response to Generic Letter 98-04, "Potential for Degradation of the Emergency Core Cooling System and the Containment Spray System After a Loss-Of-Coolant Accident Because of Construction and Protective Coating Deficiencies and Foreign Material in Containment," FPC qualified an additional coating system; Carboguard 2011 SN surfacer topcoated with Carboguard 890N.

The sentence in subsection, "Description of Analyses and Evaluations," of Section 2.1.7: "Based on the higher pressure, temperature, and accumulated dose used for the Carboguard2011SN DBA test, this test is considered to be the most limiting DBA test, " infers, based on the grammatical reading, that the Carboguard 2011 SN DBA test is the most limiting test described in the remainder of Section 2.1.7. This was not the intended meaning. To clarify, the more limiting test denotes the 1990 DBA test and is the limiting test described in the remainder of Section 2.1.7. This administrative error was entered into the vendor's corrective action program in December 2011 and does not affect the conclusions regarding the qualification of the CR-3 containment protective coatings at EPU conditions.

The DBA test profiles identified in Figures 2.1.7-1 and 2.1.7-2, and the accumulated radiation dose of 1.80 E+08 rads cited in Section 2.1.7, were obtained from the 1990 DBA test report. The

U. S. Nuclear Regulatory Commission Attachment 3F01 12-04 Page 2 of 4 1990 DBA protective coating test is considered a more severe test of pressure and temperature conditions than those predicted in the containment during a design basis Loss-of-Coolant Accident (LOCA) at EPU conditions. Additionally, as noted in Section 2.1.7, the accumulated radiation exposure of the 1990 DBA protective coating test is greater than the 40-year predicted accumulated radiation exposure for the EPU condition. Thus, the Service Level 1 coatings at CR-3 are qualified to withstand the containment temperature, pressure, and radiation conditions during a design basis LOCA at EPU conditions.

27. (ESGB 1-2)

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.

Response

The Fuel Pool Rack Neutron Absorber Monitoring Program is an existing CR-3 program that manages the effects of aging on the Carborundum (B4 C) panels located in the high density spent fuel storage racks in Spent Fuel Pool A and Boral panels located in the high density spent fuel storage racks in Spent Fuel Pool B. No change is proposed regarding a CR-3 Fuel Pool Rack Neutron Absorber Monitoring Program as part of the EPU LAR.

The details of this monitoring program, including the methods of neutron attenuation testing, frequency of inspection, sample size, data collection, and acceptance criteria, have been provided to the NRC in a CR-3 letter dated January 27, 2010 (Reference 2). Also, FPC has committed to enhance the administrative controls for the CR-3 Fuel Pool Rack Neutron Absorber Monitoring Program as part of the License Renewal LAR. To avoid duplication of NRC reviews regarding a Fuel Pool Rack Neutron Absorber Monitoring Program, FPC proposes to not address it further as part of the CR-3 EPU LAR review.

In addition, to ensure compliance with 10 CFR 50.68(b)(4) at EPU conditions, CR-3 proposes a change to the Applicability of Improved Technical Specification (ITS) 3.7.14, "Spent Fuel Pool Boron Concentration." As described in Table 1, "CR-3 Operating License and Technical Specification Technical Changes," of the CR-3 EPU LAR Attachment 1 (Reference 1), this change is made to require spent fuel pool boron concentration to be maintained > 1925 ppm at all times while fuel assemblies are stored in the spent fuel pool to ensure both CR-3 fuel storage pools remain subcritical under CR-3 licensing basis conditions. The amount of soluble boron required to maintain the spent fuel storage rack multiplication factor, keff, < 0.95 with the worst case misloaded fuel assembly is > 198 ppm in Pool A and > 571 ppm in Pool B. As such, the limit of 1925 ppm specified in ITS 3.7.14 provides adequate margin to assure kff is maintained within 10 CFR 50.68(b)(4) limits significantly reducing reliance on neutron absorbing materials within the spent fuel racks.

28. (ESGB 1-3)

U. S. Nuclear Regulatory Commission Attachment 3F0112-04 Page 3 of 4 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 FAC but cannot be inspected, it is analytically evaluated using the CHECKWORKS Pass 2 results. The analytical predictions are then compared to actual wear rate resultsfor actually inspected, usually adjacent, components which have the same fluid conditions. These results are used to trend the un-inspected component and ifpossible, 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.

Response

The purpose of a Pass 2 analysis is to adjust the values predicted by the empirical model to more closely correlate to plant inspection data. This adjustment is made by the application of the Line Correction Factor (LCF). The LCF is established by comparing the value of measured wear with the value of predicted wear within a run definition. Once the LCF for a run definition has been determined, the predicted values for inspected and un-inspected components are adjusted by the LCF. By adjusting the Pass 1 predictions to more closely approximate plant inspection data, the Pass 2 analysis provides analytical results that can be used to trend remaining life for un-inspected components.

29. (ESGB 1-4)

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 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.

Response

The enclosure to this submittal, "Wear Rate Analysis: Combined Summary Report," provides a sample list of Condensate System components for which wall thinning is predicted and measured by ultrasonic testing. This list includes the initial (nominal) wall thickness, current (measured) wall thickness, and the thickness predicted by the CHECWORKS FAC model. Specifically, the enclosure provides a combined summary of the wear rate analysis for Condensate System Train A and Train B heat exchanger piping. As noted, the summary report identifies wall thicknesses for various Condensate System piping segments; example, for Component Name 111-01 OP (P =

piping), the initial (nominal) wall thickness is 0.375 in., the current (measured) wall thickness is 0.330 in., and the thickness predicted by the CHECWORKS FAC model is 0.309 in.

References

1. CR-3 to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License Amendment Request #309, Revision 0, Extended Power Uprate." (Accession No. ML112070659)

U. S. Nuclear Regulatory Commission Attachment 3F01 12-04 Page 4 of 4

2. CR-3 to NRC letter dated January 27, 2010, "Crystal River Unit 3 - Response to Request for Additional Information for the Review of the Crystal River Unit 3, Nuclear Generating Plant, License Renewal Application (TAC NO. ME0274) and Amendment #9." (Accession No. ML100290366)

PROGRESS ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/LICENSE NUMBER DPR-72 ENCLOSURE SAMPLE OF WEAR RATE ANALYSIS: COMBINED

SUMMARY

REPORT

Company: PROGRESS ENERGY SERVICE COMPANY Report Date/Time : 16-Nov-2011 09:53 am Plant: CRYSTAL RIVER Analysis Date/Time : 16-Nov-2011 9:53 am Unit: 3 DB Name: CR3SFA_CURRENT(v3) CHECWORKS SFA Version: 3.0 SP-2 (build 200)

Wear Rate Analysis: Combined Summary Report Run Name: CD CDHE-2 TO CDHE-3 Ending Period: 17A OPERATING Total Plant Operating Hours:216776 Duty Factor (Global) : 1.000 WRA Data Option: NFA->ARD->HBD->COMP Exclude Measure Wear: NO Line Correction Factor: 0.879 Average Current Comp Predict [1] Total Lifetime In-Service Comp In-Service Comp Time (hrs)

Component Geom Wear Rate Wear Rate ------- ss ..... Time to Tcrit (hrs) Wear (mils) Wear (mils) r Method, Time Last Name Code (mils/vr) (mils/vr) fiinit), (Pt:il Thkoop Tcrit Inspected Prd.[21 Meas. Prd.[21 Meas. (in)[41 1`31 (hrs)[41 Inspected

===> Grouped by Line: CD-100 CDHE-2A to CDHE-3A, Sorted by: Flow Order 111-001N 31 1.765 0.735 0.500 0.476 0.235 0.235 2873850 Yes 39.4 50.0 39.4 50.0 0.480 MT 169064 169064 11 1-002RE 16 2.384 0.993 0.375 0.316 0.235 0.235 718450 No 0.0 0.0 0.0 0.0 0.375 0 0 11 1-002RE (D/S) 16 3.452 1.438 0.375 0.290 0.208 0.208 493938 No 0.0 0.0 0.0 0.0 0.375 0 0 111-003P 66 2.227 0.928 0.375 0.320 0.208 0.208 1051774 No 0.0 0.0 0.0 0.0 0.375 0 0 111-004E 2 4.120 1.717 0.375 0.273 0.208 0.208 329465 No 0.0 0.0 0.0 0.0 0.375 0 0 111-005P 52 2.784 1.160 0.375 0.306 0.208 0.208 737357 No 0.0 0.0 0.0 0.0 0.375 0 0 111-006E 2 4.120 1.717 0.375 0.273 0.208 0.208 329465 No 0.0 0.0 0.0 0.0 0.375 0 0 111-007P 52 2.784 1.160 0.375 0.306 0.208 0.208 737357 No 0.0 0.0 0.0 0.0 0.375 0 0 111-008E 2 4.120 1.717 0.375 0.273 0.208 0.208 329465 No 0.0 0.0 0.0 0.0 0.375 0 0 111-009E 4 4.120 1.717 0.375 0.273 0.208 0.208 329465 No 0.0 0.0 0.0 0.0 0.375 0 0 54 3.564 1.485 0.375 0.208 0.208 594068 Yes 67.3 44.0 67.3 44.0 MT 103510 103510 111-011E 2 4.120 1.717 0.375 0.273 0.208 0.208 329465 No 0.0 0.0 0.0 0.0 0.375 0 0 111-012EE 19 4.455 1.856 0.375 0.332 0.208 0.208 582641 Yes 103.1 61.0 103.1 61.0 0.339 MT 185384 185384 111-012EE (D/S) 19 3.815 1.589 0.303 0.235 0.235 376980 Yes 88.3 102.0 88.3 102.0 0.309 MT 185384 185384 111-013N 30 3.886 1.619 0.500 0.404 0.235 0.235 916247 No 0.0 0.0 0.0 0.0 0.500 0 0

===> Grouped by Line: CD-101 CDHE-2B to CDHE-3B, Sorted by: Flow Order 108-001N 31 4.857 2.024 0.500 0.380 0.235 0.235 628935 No 0.0 0.0 0.0 0.0 0.500 0 0 108-001P 61 2.575 1.073 0.375 0.442 0.235 0.235 1697418 Yes 51.1 76.0 51.1 76.0 0.455 MT 119830 119830 108-002RE 16 2.384 0.993 0.375 0.371 0.235 0.235 1206493 Yes 47.3 54.0 47.3 54.0 0.383 MT 119830 119830 108-002RE (D/S) 16 3.452 1.438 0.375 0.387 0.208 0.208 1088056 Yes 68.5 51.0 68.5 51.0 0.404 MT 119830 119830 108-003P US 66 2.227 0.928 0.375 0.333 0.208 0.208 1176618 Yes 44.2 63.0 44.2 63.0 0.344 MT 119830 119830 108-003P DS 66 2.227 0.928 0.375 0.320 0.208 0.208 1051774 No 0.0 0.0 0.0 0.0 0.375 0 0 108-004E 2 4.120 1.717 0.375 0.273 0.208 0.208 329465 No 0.0 0.0 0.0 0.0 0.375 0 0 108-005P 52 2.784 1.160 0.375 0.306 0.208 0.208 737357 No 0.0 0.0 0.0 0.0 0.375 0 0 108-006E 2 4.120 1.717 0.375 0.273 0.208 0.208 329465 No 0.0 0.0 0.0 0.0 0.375 0 0 108-007P US 52 2.784 1.160 0.375 0.306 0.208 0.208 737357 No 0.0 0.0 0.0 0.0 0.375 0 0 108-007P DS 52 2.784 1.160 0.375 0.349 0.208 0.208 1058328 No 0.0 0.0 0.0 0.0 0.360 MT 135675 0 108-008E 2 4.120 1.717 0.375 0.347 0.208 0.208 707587 Yes 85.1 67.0 85.1 67.0 0.364 MT 135675 135675 108-009EE 19 4.455 1.856 0.375 0.371 0.208 0.208 766067 Yes 92.0 92.0 92.0 92.0 0.389 MT 135675 135675 Page 1

Average Current Comp Predict [1] Total Lifetime In-Service Comp In-Service Comp Time (hrs)

Component Geom Wear Rate Wear Rate ...... Thickness (in) Time to Tcrit (hrs) Wear (mils) Wear (mils) Tmeas, Method, Time Last Name Code (mils/vr) (mils/vr! Init. Prd.rll Thooo Tcrit Inspected Prd.i21 Meas. Prd.121 Meas. (in)[41 [31 (hrs)[41 Inspected 108-009EE (D/S) 19 3.815 1.589 0.375 0.363 0.235 0.235 710215 Yes 78.8 65.0 78.8 65.0 0.379 MT 135675 135675 108-01 ON 30 3.886 1.619 0.500 0.404 0.235 0.235 916247 No 0.0 0.0 0.0 0.0 0.500 0 0 Notes:

[1] Predictions are based on last Tmeas to analysis ending period.

[2] Predictions are for the time of last known meas. wear. Can be P-to-P value depending on meas. wear method.

[3] GW = Tmeas is minimum thickness from Band, Blanket or Area Method of greatest wear.

MT = Tmeas is component minimum thickness.

PW = Tmeas is Tinit - predicted wear.

US = Tmeas is user specified.

[4] If no Tmeas has been determined from measured data, then Tmeas = Tinit and Time = current component installation time.

Tmeas is used to determine Predicted Thickness and Component Predicted Time to Tcrit.

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