ML14323A666

Request for Additional Information Concerning Changes to Pressure Temperature Curves (TAC No. MF4351)(L-14-150)

ML14323A666
Person / Time
Site:	Perry
Issue date:	12/03/2014
From:	Ellen Brown Plant Licensing Branch III
To:	Harkness E FirstEnergy Nuclear Operating Co
Eva Brown, NRR/DORL 415-2315
References
L-14-150, TAC MF4351
Download: ML14323A666 (6)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 3, 2014 Mr. Ernest J. Harkness Site Vice President FirstEnergy Nuclear Operating Company Mail Stop A-PY -A290 P.O. Box 97, 10 Center Road Perry, OH 44081-0097

SUBJECT:

PERRY NUCLEAR POWER PLANT, UNIT NO. 1 -REQUEST FOR ADDITIONAL INFORMATION CONCERNING CHANGES TO PRESSURE TEMPERATURE CURVES (TAC NO. MF4351 )(L-14-150)

Dear Mr. Harkness:

By application dated June 23, 2014, Agencywide Documents Access and Management System (ADAMS) Accession No. ML14174A663), FirstEnergy Nuclear Operating Company (FENOC) submitted a license amendment request for the Perry Nuclear Power Plant. The proposed amendment would update the technical specification pressure and temperature (P-T) figures and makes editorial changes related to the P-T figures including clarifications and updates to the associated titles, labeling, and notes. The U.S. Nuclear Regulatory Commission (NRC) staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with Mr. Phil Lashley of your staff on November 10, 2014, it was agreed that FENOC would provide a response within 30 days from the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources.

E. Harkness If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2315.

Sincerely, IRA!

Eva A. Brown, Senior Project Manager Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-440

Enclosure:

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

REQUEST FOR ADDITIONAL INFORMATION CHANGES TO PRESSURE TEMPERATURE CURVES FIRSTENERGY NUCLEAR OPERATING COMPANY PERRY NUCLEAR POWER PLANT, UNIT NO. 1 DOCKET NO. 50-440

1. On April 29, 2003, the U.S. Nuclear Regulatory Commission (NRC) issued Amendment 127, which established the current pressure-temperature (P-T) limits. The P-T limits depend not only on the selection of controlling material, but also the adjusted reference temperature (ART) of the controlling material. Attachment 4, to the submittal dated June 23, 2014, states on page 7 that:

However, it was noted that another seam weld material was controlling and therefore the PIT curves were not revised.

Since the issuance of Amendment No. 127, identify whether Reference 19 to Attachment 4 and/or any other reactor pressure vessel (RPV) structural integrity documents, consistently reference the same ART values of the controlling material and the materials listed in Table 1 of Attachment 4. Provide the basis for any revisions of these ART values made since issuance of Amendment 127.

2. Table 1 of Attachment 4 identified that Plate C2557-1 controls ART for the water level instrument nozzle (WLIN). Address whether the initial nil-ductility transition reference temperature (RTNoT) for all plates was determined in accordance with American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME Code), Section Ill, NB-2331 , using the test data from transverse Charpy specimens.

3. The licensee indicates that the proposed P-T limits are based on the Boiling Water Reactor Owners Group (BWROG) report BWROG-TP-11-023-A, "Linear Elastic Fracture Mechanics Evaluation of General Electric Boiling Water Reactor Water Level Instrument Nozzles for Pressure-Temperature Curve Evaluations." As indicated in the safety Evaluation for BWROG-TP-11-023-A, the BWROG clarified that the loads from the piping attached to the safe end of the instrument nozzle were considered in the evaluation. Attachment 4 states; however, on page 9 that:

Pipe reactions from the instrument line were not included as the resulting stress was determined to be negligible.

Attachment 4 states that the unit pressure stress intensity factor is 68.4 ksi-in°* 5 . Given that the value of 69.4 ksi-in° 5 used in the subsequent calculations on page 10, address the discrepancy in the stress intensity factors.

Enclosure

4. Sample bottom head curve calculations in Appendix 8 (pages 8-20, 8-23, and 8-27) to SIR-05-044, Revision 1, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors," indicated that the calculated pressure is reduced by the static head to arrive at the adjusted pressure. Attachment 4 states on page 12 that:

=

The corresponding static head is 853.13 inches x 0.0361 psi/inch 30.8 psig. This value is used to decrease the allowable pressure for the upper vessel and beltline curves. The bottom head curves were not changed in this update and the data source for these curves [is] referenced in the attachments.

Discuss whether the bottom head curves were updated to account for the static head, considering that the bottom head will experience the full effect of the static head, if not explain why.

5. Attachment 4 states on page 14 that:

RT NDT values for the closure flange materials are found in DIN

[Document Index] 19, Attachment 3, Table 4-1, and is equal to 10 degrees F.

Address whether this value is based on transverse Charpy data. If not, discuss how this value was determined.

6. Attachment 4 states on page 16 that:

The leak test data from the last three refuels was plotted against the revised WLIN curve for 22 EFPY [effective full-power year]. Test pressure slightly exceeded the new 22 EFPY curve. A review of the 32 EFPY WLIN curve indicated that test temperature will have to increase above 125 degrees F[ahrenheit] to achieve similar test pressures.

a. Discuss why an evaluation in accordance with ASME Code, Appendix E, "Evaluation of Unanticipated Operating Events," is not necessary.

b. Provide the test pressure and temperature line for future operation to demonstrate that this line is to the right and below the proposed pressure test curves to 32 EFPY in Figure 3.4.11-1 (a).

7. Attachment 4 states on page 18 regarding historical review of heatup and cooldown data that:

The historical data indicates no violation of the 22 EFPY revised curves was observed. The 32 EFPY curves were satisfied with the exception of the leak test curves. The minimum temperature of the leak test needs to be increased to satisfy the impact of the WLIN Leak Test curve at 32 EFPY, Attachment 2.

a. This statement from page 18 of Attachment 4 appears to contradict the statement highlighted in request for additional information-6 from page 16 of the same document. Clarify whether the "no violation" from page 18 means an "insignificant violation" was observed. Identify whether there are any other "insignificant violations" for the 22 EFPY curves during the last 3 outage cycles.

b. The last sentence of the quote indicated that the minimum temperature of the leak test needs to be increased to satisfy the impact of the WLIN Leak Test curve at 32 EFPY. Explain how the minimum temperature increase will be accomplished (i.e.

test procedure revision, etc ... ).

8. The NRC staff understands that, beginning with the next fuel cycle, Global Nuclear Fuel (GNF) 2 fuel will be loaded into the reactor core replacing General Electric (GE) 14 fuel.

Explain whether the fluence calculations supporting the proposed P-T limits are bounding for both fuel designs.

9. The P-T limit curves being updated by the licensee remove the 22 EFPY curves while updating the 32 EFPY curves. The updated 32 EFPY curves show differences compared to the current 32 EFPY curves currently in Technical Specification (TS) 3.4.11. Please provide a detailed description of the fluence calculations used to determine the proposed 32 EFPY P-T limit curves for implementation into TS 3.4.11.

E. Harkness If circumstances result in the need to revise the requested response date, please contact me at (301) 415-2315.

Sincerely, IRA/

Eva A. Brown, Senior Project Manager Plant Licensing 111-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-440

Enclosure:

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

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