3F1212-08, Response to Third Request for Additional Information to Support NRC Vessels and Internals Integrity Branch (Evib) Technical Review of the CR-3 Extended Power Uprate LAR

Response to Third Request for Additional Information to Support NRC Vessels and Internals Integrity Branch (Evib) Technical Review of the CR-3 Extended Power Uprate LAR

ML12361A010
Person / Time
Site:	Crystal River
Issue date:	12/18/2012
From:	Franke J Duke Energy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F1212-08, TAC ME6527
Download: ML12361A010 (12)

Text

Duke Energy@

Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 December 18, 2012 3F1212-08 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - Response to Third Request for Additional Information to Support NRC Vessels and Internals Integrity Branch (EVIB) Technical Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)

References:

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

2. FPC to NRC letter dated September 27, 2012, "Crystal River Unit 3 -

Response to Second Request for Additional Information to Support NRC Vessels and Internals Integrity Branch (EIVB) Technical Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)" (ADAMS Accession No. ML12272A344)

3. Email from S. Lingam (NRC) to D. Westcott (FPC) dated October 2, 2012, "RE: Crystal River, Unit 3 EPU LAR - Draft RAIs from EVIB (TAC No.

ME6527)"

4. Email from S. Lingam (NRC) to D. Westcott (FPC) dated October 9, 2012, "RE: PHONE CALL WITH FPL FOR CRYSTAL RIVER EPU RAI response submitted on 9/27/12 (EVIB) - ME6527"

5. Email from S. Lingam (NRC) to D. Westcott (FPC) dated November 1, 2012, "RE: EIVB RAI follow-up questions" (ADAMS Accession No. ML12313A112)

Dear Sir:

By letter dated June 15, 2011, Florida Power Corporation (FPC) 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 (Reference 1). On September 27, 2012, FPC provided a response to the NRC request for additional information (RAI) regarding the CR-3 pressure and temperature limits and the CR-3 Materials Reliability Program inspection requirements (Reference 2). On October 2, 2012 and October 9, 2012, via electronic mail, the NRC provided clarifying questions regarding the FPC RAI responses (References 3 and 4) and a teleconference was conducted between the NRC and FPC on October 10, 2012, to discuss the clarifications. On October 11, 2012, following the teleconference, additional clarifications were provided via electronic mail Crystal River Nuclear Plant 15760 W. Powerline Street Crystal River, FL 34428

U.S. Nuclear Regulatory Commission Page 2 of 3 3F1212-08 needed to complete the EVIB evaluation of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR) and confirmed by the NRC staff via electronic mail on November 1, 2012 (Reference 5).

The attachment to this correspondence, "Response to Third Request for Additional Information -

Vessels and Internals Integrity Branch Technical Review of the CR-3 EPU LAR," provides the formal response to the RAI.

This correspondence contains no new regulatory commitments.

If you have any questions regarding this submittal, please contact Mr. Dan Westcott.

Superintendent, Licensing and Regulatory Programs at (352) 563-4796.

Since~O. ,

Jon YA ranke Pyig President (Crystal River Nuclear Plant JAF/gwe

Attachment:

Response to Third Request for Additional Information -Vessels and Internals Integrity Branch Technical Review of the CR-3 EPU LAR xc: NRR Project Manager Regional Administrator, Region II Senior Resident Inspector State Contact

U.S. Nuclear Regulatory Commission Page 3 of 3 3F1212-08 STATE OF FLORIDA COUNTY OF CITRUS Jon A. Franke states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation; 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. ----

JO,?{A. Franke

/Vice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this I day of

. 2012, by Jon A. Franke.

Signature of Notary Public State of Florida (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 THIRD REQUEST FOR ADDITIONAL INFORMATION -VESSELS AND INTERNALS INTEGRITY BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR

U. S. Nuclear Regulatory Commission Attachment 3F1212-08 Page 1 of 8 RESPONSE TO THIRD REQUEST FOR ADDITIONAL INFORMATION

- VESSELS AND INTERNALS INTEGRITY BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR By letter dated June 15, 2011, Florida Power Corporation (FPC) 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 (Reference 1). On September 27, 2012, FPC provided a response to the NRC request for additional information (RAI) regarding the CR 3 pressure and temperature limits and the CR 3 Materials Reliability Program inspection requirements (Reference 2). On October 2, 2012 and October 9, 2012, via electronic mail; the NRC provided clarifying questions regarding the FPC RAI responses and a teleconference was conducted between the NRC and FPC on October 10, 2012, to discuss the clarifications. On October 11, 2012, following the teleconference, additional clarifications were provided to complete the Vessels and Internals Integrity Branch (EVIB) evaluation of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR) and confirmed by the NRC staff on November 1, 2012.

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

1. (EVIB 3-1)

Regarding the last paragraph of FPC's response to RAI EVIB-1 (EVIB 2-1), please provide clarification relative to areas of geometric discontinuities in the RPV considering irradiation embrittlement under EPU. Specifically:

(a) Provide the initial RTNDT, the copper (Cu) and nickel (Ni) content values, and estimated fluence for the outlet nozzle. If these values are not available, please provide the generic.

values that were used in generating the current P-T limits. For estimated fluence, interpolation is permissible.

Response

Per 10 CFR 50, Appendix G, the initial RTNDT is to be determined in accordance with Section III, NB-2331 "Material for Vessels," of the American Society of Mechanical Engineers (ASME) Boiler Code. However, NRC Branch Technical Position (BTP) 5-3, "Fracture Toughness Requirements," provides acceptable alternative methods of determining the initial RTNDT for plants with construction permits prior to August 15, 1973. The two CR-3 reactor pressure vessel (RPV) outlet nozzle forgings were fabricated on September 22, 1970, and the construction permit for CR-3 is dated prior to August 15, 1973. The certified material test reports (CMTRs) for the CR-3 RPV outlet nozzle forgings provide data for Charpy V-notch tests performed at 100 F, but the sample orientation, transverse or longitudinal, is not specified. However, since transverse is the weak orientation, it is assumed that the test data provided in the CMTRs corresponds to samples tested in the more conservative longitudinal orientation. Per Section B. 1.1(4) of BTP 5-3, the test temperature of 10°F can be used for the initial RTNDT because all Charpy values were above 45 ft-lbs for both outlet nozzle forgings at this temperature.

The standard deviation for the initial RTNDT used for the adjusted reference temperature (ART) calculation is 00 F. The two outlet nozzle forgings were tested a total of four times, twice for each material heat, at a temperature of 100 F. Since the four tests meet the 45 ft-lb

U. S. Nuclear Regulatory Commission Attachment 3F1212-08 Page 2 of 8 requirement of BTP 5-3, each test results in an initial RTNDT of 10°F. As a result, the standard deviation of the four tests is 0°F because the four tests have the same initial RTNDT-value.

The Cu content in the the CR-3 specific outlet nozzle forgings is not available. A Cu content of 0.11 weight-percentage (wt%) was selected based on a review of available Cu content values from the specific forging supplier that provided the CR-3 outlet nozzle forgings. A statistical determination was performed to obtain a lower tolerance generic value with 95 percent confidence that at least 95 percent of the population is greater than the tolerance limit. The mean was 0.06 wt%, the standard deviation (o) was 0.01 wt%, and the one-sided factor (k) was 3.187. The generically calculated Cu content value, using the mean plus ka, was 0.11 wt%.

The Ni content for the CR-3 outlet nozzle forgings was obtained from the CMTRs. From the two material heats, the highest Ni content value was 0.80 wt%.

The current P-T limits at 32 effective full power years (EFPY) are valid through 27.5 EFPY considering EPU conditions. At 27.5 EFPY, the 1/4T fluence at the outlet nozzle comer region was conservatively assumed to be 7.6E+16 n/cm2 , which is the peak wetted surface fluence of the outlet nozzle forgings. The fluence factor equation from Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials," Revision 2, was used to estimate any shift in the P-T limits due to a change in RPV material embrittlement. Since this equation is valid at fluence values > 1E+ 17 n/cm 2 , the equation is also considered acceptable to Frovide a reasonable approximation of the fluence factor at a fluence value of 7.6E+ 16 n/cm.

(b) Estimate the 1/4T ART value of the outlet nozzle forging comer region using the values identified in (a).

Response

For the current CR-3 P-T limits, an adjusted reference temperature (ART) of 60'F was used for the CR-3 RPV outlet nozzle forgings. The ART value for the comer region of the CR-3 RPV outlet nozzle forgings at the 1/4T wall depth has been estimated using the methodology of Regulatory Guide 1.99, Revision 2, and the values cited herein for the initial RTNDT, the Cu and Ni content values, and an estimated fluence of 7.6E+16 n/cm 2 .

The following assumptions were made:

" The fluence factor equation from Regulatory Guide 1.99, Revision 2 is considered a reasonable approximation at a fluence value of 7.6E+16 n/cm2;

" The 1/4T fluence at the outlet nozzle comer region is conservatively assumed to be the peak wetted surface fluence of the RPV outlet nozzle forging, i.e., the bottom of the outlet nozzle forging to lower nozzle belt forging weld; and

• A chemistry factor of 77°F was determined based on a Cu content of 0.11 wt% and a Ni content of 0.80wt%.

U. S. Nuclear Regulatory Commission Attachment 3F1212-08 Page 3 of 8 The 1/4T ART value for the comer region of both CR-3 outlet nozzle forgings was estimated to be 24.0'F at 27.5 EFPY, which is well below the 60'F value used for the current CR-3 P-T limits.

(c) Generate P-T limits (heatup, cooldown, and hydrostatic/leak test) for the outlet nozzle using information in (b) and demonstrate that the current P-T limits remain bounding for EPU conditions. Adjust the EPFY value for the current P-T limits if necessary.

Response

The P-T limits for the CR-3 RPV outlet nozzle have been determined based on the 1/4T ART value of 247F. The outlet nozzle P-T limits have been generated for heatup, cooldown, and inservice leak and hydrostatic (ISLH) test. These limits were compared against the current respective P-T limits at 32 EFPY as shown in Figure 1, "Location Corrected Normal Heatup P-T Limits at 32 EFPY," Figure 2, "Location Corrected Normal Cooldown P-T Limits at 32 EFPY," and Figure 3, "Location Corrected ISLH P-T Limits at 32 EFPY." The results show that the RPV outlet nozzle P-T limits are conservative with respect to the current CR-3 P-T limits for each of the three transient conditions; heatup, cooldown, and ISLH test. FPC concludes that the outlet nozzle material is not the limiting material for the CR-3 P-T limits at EPU conditions. As such, no further adjustment to the EFPY of the current P-T limit curves is required.

Figure 1: Location Corrected Normal Heatup P-T Limits at 32 EFPY 2800 2600 2400 2200 C 2000 0

1800 0

" 1600 E

o 1400 C

o 1200 a

,0 1000 800 0

400 200 0

0 50 100 150 200 250 300 350 400 450 500 550 600 RV Downcomer Temperature, F

U. S. Nuclear Regulatory Commission Attachment 3F1212-08 Page 4 of 8 Figure 2: Location Corrected Normal Cooldown P-T Limits at 32 EFPY 2800 2600 2400 2200

'. 2000

'A 1800 0

. 1600 E

o 1400 1200 a

, 1000 800

< 600 400 200 0

200 250 300 350 RV Downcomer Temperature, F Figure 3: Location Corrected ISLH P-T Limits at 32 EFPY 3800 3600 3400 3200 3000 20

'F 2800 260 0 2400 0

2200 E 2000 0

0 E 1800 0 1600

> 1400

. 1200 o 1000

< 800 600 400 200 0

0 50 100 150 200 250 300 350 400 450 500 550 600 RV Downcomer Temperature, F

U. S. Nuclear Regulatory Commission Attachment 3F1212-08 Page 5 of 8 (d) Demonstrate that closure head is not the limiting material in any portion of the composite P-T limits.

Response

The original CR-3 RPV closure head limits used the generic ART value of 60'F and were compared against the composite P-T limits shown in Figure 4, "Location Corrected Composite P-T Limits at 32 EFPY." During refueling outage R13, the RPV closure head was replaced at CR-3. The replacement RPV closure head has an ART value of minus

(-) 40'F. As illustrated in Figure 4, the replacement RPV closure head (RVCH) limits are conservative with respect to the current CR-3 composite P-T limits. Hence, the replacement RPV closure head material is not the limiting material in any portion of the composite P-T limits at EPU conditions.

Figure 4: Location Corrected Composite P-T Limits at 32 EFPY 2800 2600 2400 -- e- Composite P-T Umits

--- Odinal RVCHU 2200 - - Repkacement RVCH 2000

• 1800

-I-C- 1600 II E

0 1400

_ 4- 4---

C _ __ _ I I _

0 1200

' 1000 I- - -

~800 _________!_________

R 600 1!

400 200 0

0 50 100 150 200 250 300 350 400 450 500 550 600 RV Downcomer Temperature, F

2. (EVIB 3-2)

With regard to ferritic RCPB materials other than those in the RPV, the licensee's current 32 EFPY P-T limits allow the RCS temperature as low as 60 'F, but the licensee indicates in their response to EVIB-1 (EVIB 2-1) that the lowest service temperature cannot be lower than 150 'F for the piping. ASME Code,Section III, NB-2332 (b) states, "The lowest service temperature shall not be lower than RTNDT + 100 'F unless a lower temperature is justified by following methods similar to those contained in Appendix G." Please clarify how you meet this requirement.

U. S. Nuclear Regulatory Commission Attachment 3F1212-08 Page 6 of 8

Response

The lowest service temperature is defined in Footnote 1 of Subsection NB-2332(a)(1) in Section.

III of the ASME Code, 1974 Edition. Footnote 1 states, "Lowest service temperature is the minimum temperature of the fluid retained by the component, or alternatively, the calculated volumetric average metal temperature expected during normal operation, whenever the pressure within the component exceeds 20% of the preoperational system hydrostatic test pressure." For CR-3, the preoperational Reactor Coolant System (RCS) hydrostatic test pressure is 3125 psig.

Thus, at CR-3, the calculated volumetric average metal temperature expected during normal operation is based on a preoperational system hydrostatic test pressure of 625 psig. At temperatures ranging from 60'F up to the lowest service temperature of 150'F, the allowable pressures reported in the CR-3 P-T limit curves are demonstrated to remain below 625 psig.

Hence, the lowest service temperature requirement for the RCS piping has been demonstrated to have been met.

3. (EVIB 3-3)

Please provide the edition (year) of the ASME Code,Section III (Construction Code) based on which that the RCPB (e.g., piping, pumps, and valves) was constructed and identify whether the NB-2332 (b) cited above was already in that edition of the Construction Code.

Response

Subsection NB-2332, "Material for Piping, Pumps, and Valves, Excluding Bolting Material,"

was included in Section III of the Summer 1972 edition of the ASME Code. The construction code and code case interpretations for the respective CR-3 RCS components are listed in Table 4-2, "Reactor Coolant System Component Codes," of the CR-3 Final Safety Analysis Report.

As noted, most of the RCS components (e.g., the RPV, control rod drive nozzles, etc...) were constructed to ASME Section III Class A, Summer 1967 edition. The RCS piping was constructed to USA Standard (USAS) B31.7-1968. The Summer 1967 edition of the ASME Section III and USAS B31.7-1968 did not include a requirement consistent with the requirement cited in Subsection NB-2332(b), rather Subsection N-332 of the ASME Section III stated that impact tests results are considered suitable for materials of vessels to be pressure tested at a temperature of not less than 60'F higher than RTNDT.

The RPV closure head was constructed to ASME Section III Class I, Subsections NCA and NB, 1989 edition. Also, the replacement steam generators were constructed to ASME Section III Class I, 2000 edition. However, ASME Code,Section III, Subsection NB-2332 does not apply to the RPV closure head or the replacement steam generators, rather Subsection NB-2331 applies to these components.

A portion of the upper RCS hot leg and flowmeter were replaced during the CR-3 steam generator replacement activity. This portion of the RCS piping conforms to the requirements of ASME Code,Section III, Subsection NB for Class 1 piping, 1998 Edition through 2000 Addenda, including NB-2332.

U. S. Nuclear Regulatory Commission Attachment 3F1212-08 Page 7 of 8

4. (EVIB 3-4)

Regarding the first paragraph of FPC's response to EVIB-2 (EVIB 2-2), please clarify whether the program was developed and implemented.

Response

An aging management program (AMP) has been developed for the CR-3 reactor vessel internals; WCAP-17113-NP, "PWR Vessels Internal Program Plan for Aging Management for Reactor Internals at Crystal River Unit 3." The CR-3 reactor vessel internals AMP has been implemented with one deviation as cited in RAI Response EVIB 2-2 in the FPC to NRC letter dated September 27, 2012 (Reference 2) and further clarified in RAI Response EVIB 3-5 of this attachment.

5. (EVIB 3-5)

Regarding the last paragraph of FPC's response to EVIB-2 (EVIB 2-2), please clarify your response to Item (2) of the RAI in relation to the deferred ultrasonic testing of the accessible RPV core barrel bolts. Would such deferral require relief from the ASME Code?

Response

The deviation of the ultrasonic examination of the accessible RPV lower core barrel bolts and their locking devices was processed in accordance with the guidelines of Nuclear Energy Institute (NEI) 03-08, "Guideline for the Management of Materials Issues," to delay the ultrasonic examination until the next scheduled removal of the core support assembly (CSA).

FPC plans to perform the initial ultrasonic examination of the accessible RPV lower core barrel bolts and their locking devices during refueling outage R17. Refueling outage R17 has been further delayed as a result of the CR-3 containment repair activities. In addition, the examination cannot be readily performed during the current extended outage period because the polar crane, which is required for this examination, is unavailable due to the physical condition of the CR-3 containment structure. This further delay in the performance of the ultrasonic examination is considered acceptable since the material properties and conditions of the RPV lower core barrel bolts and their locking devices are unlikely to change during the current extended outage period with the CR-3 RPV in the defueled condition.

A relief from the ASME Code, as allowed by 10 CFR 50.55a(a)(3), is not required for the MRP-227 deferment of an ultrasonic examination of the lower core barrel bolts and their locking devices. Ultrasonic examination of the lower core barrel bolts and their locking devices is not required by the ASME Code. The RPV lower core barrel bolts are part of the CSA. ASME Section XI, Subsection IWB-2000, Category B-N-3, Item No. B13.70, requires a visual examination of accessible areas of the CSA every inservice inspection (ISI) interval, typically during the 10-year ISI. The accessible areas of the lower core barrel bolts and their locking devices are included in this visual examination of the CSA. The previous visual examination of the accessible areas of the CSA was completed in 2007 during refueling outage R15 and the next visual examination is not required per ASME Section XI until the end of the fourth CR-3 10-year ISI interval. Therefore, relief from ASME Code compliance is not required for the MRP-227 deferment of the lower core barrel bolts ultrasonic examination.

U. S. Nuclear Regulatory Commission Attachment 3F1212-08 Page 8 of 8 References

1. FPC to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License Amendment Request #309, Revision 0, Extended Power Uprate." (ADAMS Accession No. MLl 12070659)

2. FPC to NRC letter dated September 27, 2012, "Crystal River Unit 3 - Response to Second Request for Additional Information to Support NRC Vessels and Internals Integrity Branch (EVIB) Technical Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)."

(ADAMS Accession No. ML12272A344)