ML110420241

EPU Vessels and Internals Integrity (Cvib) Requests for Additional Information - Round 1

ML110420241
Person / Time
Site:	Turkey Point
Issue date:	02/11/2011
From:	Jason Paige Plant Licensing Branch II
To:	Abbatiello T Florida Power & Light Co
Paige, Jason C, NRR/DORL,301-415-5888
References
Download: ML110420241 (5)

Text

From: Paige, Jason Sent: Friday, February 11, 2011 11:13 AM To: tom.abbatiello@fpl.com Cc: Abbott, Liz; Tiemann, Philip; Tomonto, Bob

Subject:

Turkey Point EPU - Vessels and Internals Integrity (CVIB) Requests for Additional Information - Round 1

Tom, Below are requests for additional information (RAIs) regarding the Turkey Point Extended Power Uprate license amendment request. On January 28, 2011, the Nuclear Regulatory Commission (NRC) staff and Florida Power & Light Company (FPL) discussed draft RAIs to gain a common understanding of the questions. During the call, it was concluded that question CVIB-1.1 needed clarification if a range of the EFPY value is acceptable and to delete question CVIB-1.5. Also, the NRC staff revised question CVIB-1.4 based from discussions during the call. As follow-up to the January 28, 2011, call, the NRC staff and FPL held another call on February 3, 2011, to clarify question CVIB-1.1. It was concluded that a specific EFPY value is needed by the NRC staff as opposed to a range. The below RAIs reflect the questions discussed during the January 28 and February 3, 2011, calls. FPL agreed upon providing its responses within 30 days of the date of this email. If you have any questions, feel free to contact me.

CVIB-1.1 The revised surveillance capsule withdrawal schedule for Turkey Point, Units 3 and 4 allows the last capsule, X4, to be withdrawn between 31.4 and 47.8 effective full power years (EFPY), which by letter dated October 21, 2010, footnote (3) in Table 2.1.1-6 of the EPU Licensing Report states will yield a capsule fluence not less than once or greater than twice the 48 EFPY peak vessel fluence of 6.377 x 1019 n/cm2 (E > 1.0 MeV). Withdrawal of the last capsule at any fluence of not less than once or greater than twice the end-of-life (EOL) peak vessel fluence would meet the recommendation of American Society for Testing and Materials (ASTM) E 185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, for the last capsule withdrawn from a reactor with five surveillance capsules installed. However, the NRC staff requests the licensee provide a single estimated EFPY value and neutron fluence value at which the capsule will be withdrawn rather than a range, so the NRC staff can confirm the capsule withdrawal schedule meets the recommendation of ASTM E 185-82.

CVIB-1.2 The revised equivalent margins analysis (EMA) forwarded by letter dated December 21, 2010, stated that the low-upper shelf fracture mechanics evaluation is performed according to the acceptance criteria and evaluation procedures contained in Appendix K to Section XI of the American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME Code), and references the ASME Code,Section XI, 1998 Edition through 2000 Addenda.

Title 10 of Code of Federal Regulations (10 CFR) Part 50, Appendix G, IV.A.1.a, requires that such analyses use the latest edition and addenda of the ASME Code incorporated by reference into 10 CFR 50.55a(b)(2) at the time the analysis is submitted. The latest edition of the ASME Code,Section XI (Division 1) incorporated by reference into 10 CFR 50.55a at the time of the submittal is the

2004 edition. The NRC staff therefore requests the licensee reconcile the differences between the 1998 through 2000 Addenda, and 2004 editions of the ASME Code,Section XI, specifically as the differences affect the low-upper shelf toughness evaluation.

CVIB-1.3 In Section 7 of Reference 1, the licensee indicated that the applied J-integral was calculated using the following equation:

Japplied (a) = 1000K2 Itotal(a)(1-2)/E This is essentially the same as the ASME Code,Section XI K-5210 equation:

J = 1000(KI)2/E where:

E = E/(1-2)

KI = stress intensity factor adjusted for small scale yielding.

Article K-5000, subparagraph K-5210 of the ASME Code,Section XI, Appendix K (2004 Edition), provides an adjustment of the effective flaw depth for small-scale yielding as follows:

ae = a + [1/(6)](KI/y)2 Where:

a = actual flaw depth, ae = effective flaw depth, KI = applied stress intensity, y = yield strength.

Paragraph K-5210 further states that the stress intensity factor for small scale yielding, KI, shall be calculated by substituting ae for a.

The licensee did not discuss whether the effects of small scale yielding were included in the KItotal term. The NRC staff therefore requests that the licensee discuss how the effects of small scale yielding were accounted for in the KItotal term.

CVIB-1.4 Provide the basis, such as a report or calculation, for the pressure-temperature (P-T) limits for Turkey Point, Units 3 and 4, that are given in proposed revised Technical Specification Figures 3.4-2 and 3.4-3. If the report or calculation does not contain the following items, then the following items should be provided separately:

a. Provide a tabulation of the thermal stress intensity factors (KIt) used to generate the heatup and cooldown curves for each coolant temperature for heatup and cooldown.

b. Provide a tabulation or graph of the temperature differential from the coolant to the crack tip used to generate the P-T limits, and describe the methodology used to determine this differential, unless Figure G-2214-1 and Figure G-2214-2 of the ASME Code,Section XI, Appendix G, were used to determine the temperature differential.

c. Provide the numerical temperature versus pressure values corresponding to the heatup and cooldown curves, and the hydrotest curve, in Technical Specification Figures 3.4-2 and 3.4-3.

d. The P-T curves provided in EPU Licensing Report Figures 2.1.2-1 and 2.1.2-2 and TS Figures 3.4-2 and 3.4-3 do not indicate whether there is any pressure difference between the reactor vessel (RV) pressure and pressure at the measurement location. If such a pressure difference exists, provide the correction factors used to correct between the actual reactor vessel (RV) pressure and the indicated pressure at the measurement location.

e. In the technical specification (TS) bases markups provided with the EPU application, the licensee provided a revised Table B 3/4.4-1 that shows the closure flange RTNDT has been changed from 44 °F to -50 °F. Therefore, the

NRC staff requests the licensee provide the basis for changing the highest RTNDT of the closure flange region that is highly stressed by bolt preload from 44 °F to -50 °F.

f. The EPU Licensing Report Figures and the marked up TS bases 3/4.4.9 indicate that the revised P-T limits are based on the KIa methodology of the 1996 Edition of ASME Code,Section XI, Appendix G. Since 1996 is an addenda rather than an edition of the ASME Code, the NRC staff requests the licensee confirm that the revised P-T limits are based on the 1995 Edition through 1996 Addenda of the ASME Code,Section XI, Appendix G, and requests the licensee modify the TS bases accordingly.

g. Provide the following information related to the determination of the adjusted reference temperature (ART) for the limiting RV beltline materials:

1. supporting data for, and the calculation of, the chemistry factors for those reactor vessel (RV) materials that have surveillance data;

2. the copper and nickel values for the surveillance materials;

3. the credibility evaluation of the surveillance data; and

4. whether the ratio procedure of Regulatory Guide 1.99, Rev. 2, Position 2.1 was used.

CVIB-1.5 Reference 2, Section 2.1.4.2.5 concludes that the new EPU environmental conditions (chemistry, temperature, and neutron fluence) will not introduce any new aging effects on the RVI components during 60 years of operation, nor will the EPU change the manner in which the component aging will be managed by the aging management program credited in the topical report WCAP-14577, Rev.

1-A, License Renewal Evaluation: Aging Management of Reactor Internals, and accepted by the NRC in the Safety Evaluation Report (SER). The susceptibility of the Turkey Point, Units 3 and 4 RVI components to these aging effects was also assessed for license renewal as documented in the License Renewal Application (LRA) for Turkey Point Units 3 and 4 and the associated SER, NUREG-1759.

Although the licensee stated that there will be no new aging effects, Reference 2 does not address whether particular RVI components will become susceptible to additional aging effects due to higher neutron fluences, temperatures, or stresses introduced by the EPU. The NRC staff therefore requests the following information:

a. Describe the method of determining if additional RVI components become susceptible to the aging effects of 1) cracking due to stress corrosion cracking (SCC), irradiation assisted stress corrosion cracking (IASCC), or primary water stress corrosion cracking (PWSCC); 2) reduction of fracture toughness due to irradiation embrittlement (IE); 3) loss of material due to wear; 4) loss of mechanical closure integrity due to IASCC, IE, irradiation creep, or stress relaxation (SR); and 5) loss of preload due to SR, or

dimensional change due to void swelling. The discussion should address whether a detailed fluence and temperature map was used, and whether stresses in individual components were reevaluated.

b. Confirm whether the design projections of gamma heating bound the projected amount of gamma heating of the RVI under EPU conditions.

Discuss the acceptability of the effects of gamma heating on the RVI components under EPU conditions.

c. Clarify whether any additional RVI components were determined to be susceptible to the aging effects listed in part a of this question as a result of EPU, compared to those listed as susceptible to these mechanisms in Table 3.2-1 of the LRA for Turkey Point, Units 3 and 4.

CVIB-1.6 Several aging effects identified for RVI in the LRA for Turkey Point, Units 3 and 4, are not evaluated in the EPU evaluation of RVI materials. The SER related to the Turkey Point, Units 3 and 4 LRA, NUREG-1759, concurred with the aging effects requiring management for the RVI. The NRC staff requests the licensee provide an evaluation of the effects of EPU on the following aging effects requiring management, or explain why the aging effect did not require reevaluation.

a. LRA Section 3.2.5.2.3 stated that loss of material due to mechanical wear is an aging effect requiring management for the period of extended operation.

Loss of material due to wear can occur on the lower core plate fuel pins, core barrel flanges, guide tubes and guide pins, upper core plate alignment pins, and radial keys and clevis inserts.

b. The LRA indicates loss of mechanical closure integrity due to SCC and SR is an aging effect for upper support column, guide tube, and clevis insert bolting. For baffle-former bolting and barrel-former bolting, loss of mechanical closure integrity can be caused by IASCC, IE, irradiation creep, and irradiation-assisted SR.

c. The LRA indicates loss of preload due to SR can occur for the RVI hold-down spring.

Jason Paige, Turkey Point Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Phone: (301) 415-5888