ML16041A578

Request for Additional Information Related to License Amendment Request Regarding Moderator Temperature Coefficient Measurement Change

ML16041A578
Person / Time
Site:	Indian Point
Issue date:	02/18/2016
From:	Pickett D Plant Licensing Branch 1
To:	Entergy Nuclear Operations
Pickett D, NRR/DORL/LPLI-1
References
CAC MF7193, CAC MF7194
Download: ML16041A578 (4)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 18, 2016 Vice President, Operations Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3- REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST REGARDING MODERATOR TEMPERATURE COEFFICIENT MEASUREMENT CHANGE (CAC NOS. MF7193 AND MF7194)

Dear Sir or Madam:

By application dated December 10, 2015 (Agencywide Documents Access and Management System Accession No. ML15350A011), Entergy Nuclear Operations, Inc., the licensee, submitted a license amendment request that would change the near-end-of-life moderator temperature coefficient Surveillance Requirement 3.1.3.2 and Technical Specification 5.6.5 for Indian Point Nuclear Generating Unit Nos. 2 and 3.

The U.S. Nuclear Regulatory Commission staff is reviewing the amendment request and has determined that additional information is needed to complete its review. The specific questions are found in the enclosed request for additional information (RAI). Based on our discussions, we understand that a response to the RAI will be provided within 30 days of the date of this letter.

Please contact me at (301) 415-1364 or Douglas.Pickett@nrc.gov if you have any questions.

Sincerely, Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-247 and 50-286

Enclosure:

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

REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST REGARDING MODERATOR TEMPERATURE COEFFICIENT MEASUREMENT CHANGE ENTERGY NUCLEAR OPERATIONS. INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286 By letter dated December 10, 2015 (Agencywide Documents Access and Management System Accession No. ML15350A011 ), Entergy Nuclear Operations, Inc., the licensee, submitted a license amendment request (LAR) that proposes to change the near-end-of-life moderator temperature coefficient (MTC) Surveillance Requirement 3.1.3.2 and Technical Specification (TS) 5.6.5 for Indian Point Nuclear Generating (Indian Point) Unit Nos. 2 and 3.

The U.S. Nuclear Regulatory Commission staff has reviewed the amendment request and determined that the following additional information is required in order to complete the evaluation.

Reactor Systems Branch (SRXB) - Request for Additional Information (RAI) 1 The licensee's application states that in order to replace performing the MTC measurement at end-of-life, the surveillance will be met by using a design calculation, provided that predefined requirements are met as outlined in WCAP-137 49-P-A. The licensee further states that core design calculations are performed using a PARAGON/ANG system, which is considered equivalent to the PHOENIX-P/ANC, based on the approval of Topical Report WCAP-16045-P-A.

However, in Attachment 5 of the submittal, in its response to the Beaver Valley Power Station (BVPS) RAI, Question 2, the licensee mentions the NEXUS/ANG code package when demonstrating that the predictive correction terms are equivalent between PHOENIX-P and PARAGON. Please clarify which code (PARAGON/ANG or NEXUS/ANG) is performing the design calculation of core MTC.

SRXB-RAI 2 In the LAR, Attachment 5, response to the Joseph M. Farley Nuclear Plant and Vogtle Electric Generating Plant (Vogtle) RAI, Question 2, it is stated that Indian Point does not propose to add PHOENIX-P, PARAGON, or NEXUS to the listed core operating limit report (COLR) references in the TSs and cites Vogtle as a precedent. However, the NRC staff has reviewed the Vogtle COLR references in its TSs and has found that Vogtle does include PARAGON and NEXUS in the TS COLR references. Additionally, the NRC staff reviewed other similar precedents and found that the neutronics methods are included in the TS COLR section. Since the neutronics codes are used to confirm reload parameters, including MTC, the NRC staff views these codes as an integral part of establishing the COLR parameters for each cycle. However, the staff acknowledges that this may not be clear in NRC Generic Letter 88-16, "Removal of Cycle-Enclosure

specific Parameter Limits from Technical Specifications." Therefore, the staff requests that the neutronics codes that will be used to confirm reload parameters be included in the TS COLR section, to be consistent with the previous precedents.

SRXB-RAI 3 The licensee's submittal provides BOL HZP ITC data in Attachment 5 in response to BVPS RAI, Question 1. Tables 1 and 2 detail measured and predicted BOL HZP ITC values. Please clarify what code system (PHOENIX-P/ANC, PARAGON/ANG, or NEXUS/ANG) is used to predict the BOL HZP ITC.

February 18, 2016 Vice President, Operations Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST REGARDING MODERATOR TEMPERATURE COEFFICIENT MEASUREMENT CHANGE (CAC NOS. MF7193 AND MF7194)

Dear Sir or Madam:

By application dated December 10, 2015 (Agencywide Documents Access and Management System Accession No. ML15350A011 ), Entergy Nuclear Operations, Inc., the licensee, submitted a license amendment request that would change the near-end-of-life moderator temperature coefficient Surveillance Requirement 3.1.3.2 and Technical Specification 5.6.5 for Indian Point Nuclear Generating Unit Nos. 2 and 3.

The U.S. Nuclear Regulatory Commission staff is reviewing the amendment request and has determined that additional information is needed to complete its review. The specific questions are found in the enclosed request for additional information (RAI). Based on our discussions, we understand that a response to the RAI will be provided within 30 days of the date of this letter.

Please contact me at (301) 415-1364 or Douglas.Pickett@nrc.gov if you have any questions.

Sincerely, IRA!

Douglas V. Pickett, Senior Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-247 and 50-286

Enclosure:

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

PUBLIC RidsNrrDorlLpl1-1 RidsNrrPMlndianPoint LPL 1-1 Reading File RidsNrrLAKGoldstein RidsACRS_MailCTR RidsNrrDorlDpr RidsRgn1 MailCenter GDentel, R-1 RidsNrrDssSrxb WMacFee, NRR ADAMS A ccess1on No.: ML16041A578 *b1y e-ma1*1 OFFICE DORL/LPL 1-1/PM DORL/LPL 1-1/LA DSS/SRXB/BC(A}* DORL/LPL 1-1/BC NAME DPickett KGoldstein EOesterle TTate (LRonewicz for)

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