NL-16-027, Response to Request for Additional Information Related to License Amendment Request Regarding Conditional Exemption from End-of-Life (Eol) Moderator Temperature Coefficient (Mtc) Measurement

Response to Request for Additional Information Related to License Amendment Request Regarding Conditional Exemption from End-of-Life (Eol) Moderator Temperature Coefficient (Mtc) Measurement

ML16069A311
Person / Time
Site:	Indian Point
Issue date:	03/02/2016
From:	Coyle L Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF7192, CAC MF7193, CAC MF7194, NL-16-027
Download: ML16069A311 (9)

Text

~Enterov

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Nuclear 450 Broadway, Northeast GSB Point Energy Center

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- P.O. Box 249 Buchanan, NY 10511-0249 Tel 914 254 6700 Lawrence Coyle Site Vice President NL-1 6-027 March 2, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville, MD 20852-2738 ,

SUBJECT Response to Request for Additional Information Related to License Amendment Request Regarding Conditional Exemption from End-of-Life (EOL) Moderator Temperature Coefficient (MTC) Measurement (CAC Nos. MF-7192 and MF-7194)

Indian Point Unit Numbers 2, and 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64

REFERENCES:

1. Entergy Letter NL-15-144 to NRC Regarding License Amendment Request

- Conditional Exemption from End-of-Life (EOL) Moderator Temperature Coefficient (MTC) Measurement, dated December 10, 2015.

2. NRC Letter to Entergy Request For Additional Information Related to License Amendment Request Regarding Moderator Temperature Coefficient Measurement Change (CAC NOS. MF7193 and MF7194), dated February 18, 2016

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) requested a License Amendment for Indian Point N o.2 (IP2), Operating License DPR-26, Docket No. 50-247, and for Indian Point No.3 (IP3), Operating License DPR-64, Docket No. 50-286 in Reference 1.

The purpose of this letter is to submit additional information in response to an NRC request for additional information in Reference 2.

The proposed License Amendment Request was to change the near-end of life (EOL) Moderator Temperature Coefficient (MTC) Surveillance Requirement (SR) 3.1.3.2 and Specification 5.6.5 for Indian Point Units 2 and 3 by placing a set of conditions on reactor core operation, which if met, would allow exemption from the required MTC measurement. Attachment 1 provides the response to the request for additional information regarding this TS change request. Attachment 2 provides the marked up TS pages showing the additional changes from those submitted in Reference 1.

In accordance with 10 CFR 50.91{b)(1), "State Consultation," a copy of this letter and its associated Attachments is being provided to the designated New York State officials. _ '

jIL/

NL-1 6-027 Docket Nos. 50-247 and 50-286 Page 2 of 2 This letter contains no new NRC commitments. Should you have any questions concerning this letter, or require additional information, please contact Robert Walpole, Manager, Regulatory Assurance at (914) 254-6710.

I declare under penalty of perjury that the foregoing is true and correct. Executed on March * ,2016.

Sincerely, LC/mm Attachments: 1. Response to Request for Additional Information regarding Conditional Exemption from End-of-Life (EOL) Moderator Temperature Coefficient (MTC) Measurement

2. Additional Technical Specification Changes (mark-up) [both units are included in this attachment]

cc: Mr. Douglas Picket, Senior Project Manager, NRC NRR DORL Ms. Kimberly A. Conway, Project Manager, NRC FSME DWMEP DURLD Mr. Daniel H. Dorman, RegionalAdministrator, NRC Region 1 NRC Resident Inspector's Office Mr. Francis J. Murray, Jr., President and CEO, NYSERDA Ms.

Bridget Frymire, New York State Dept. of Public Service

ATTACHMENT 1 TO NL-16-027 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING CONDITIONAL EXEMPTION FROM END-OF-LIFE (EOL)

MODERATOR TEMPERATURE COEFFICIENT (MTC) MEASUREMENT ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286

Attachment 1 NL-1 6-027 Docket Nos. 50-247 and 50-286 Page 1 of 2 Entergy Nuclear Operations, Inc. (Entergy) Letter to NRC (NL-15-144), dated December 10, 2015, requested an amendment to Operating License DPR-26, Docket 50-247 and Operating License DPR-64, Docket 50-286 for Indian Point Nuclear Generating Unit Nos. 2 (IP2) and 3 (IP3), respectively.

The amendment request proposed to change the near-end-of-life moderator temperature coefficient (MTC) Surveillance Requirement 3.1.3.2 and Technical Specification (TS) 5.6.5. In order to perform their review, the NRC staff requested additional information in letter dated February 18, 2016. This responds to that request.

Reactor Systems Branch (SRXB) 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/ANC 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/ANC code package when demonstrating that the predictive correction terms are equivalent between PHOENIX-P and PARAGON. Please clarify which code (PARAGON/ANC or NEXUS/ANC) is performing the design calculation of core MTC.

Response

The PARAGON/ANC code system is performing the design calculation of core MTC for the current Indian Point Unit 2 and Indian Point Unit 3 fuel cycles as well as for future fuel cycles. Indian Point anticipates that the NEXUS cross-section generation code, which uses PARAGON as the lattice transport code, will be used to confirm reload parameters, including MTC, in future cycles, subject to review and approval of a separate license amendment request.

Reactor Systems Branch (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-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.

Response

Entergy has reviewed the BVPS, Farley and Vogtle TS COLR sections where the PARAGON and NEXUS neutronic codes are referenced as approved methodology. Since Indian Point only proposes to apply the PARAGON neutronic code to confirm reload parameters, including MTC, and to be

Attachment 1 NL-1 6-027 Docket Nos. 50-247 and 50-286 Page 2 of 2 consistent with the previous precedents, Entergy agrees that PARAGON should be included in the TS COLR section. Indian Point anticipates that the NEXUS cross-section generation code, which uses PARAGON as the lattice transport code, will be used to confirm reload parameters, including MTC, in future cycles, subject to review and approval of a separate license amendment request. Indian Point also uses the ANC neutronic code to confirm reload parameters, including MTC. Although Farley and Vogtle do not reference ANC in their respective TS COLR sections, Entergy agrees to include the ANC neutronic code methodology in the TS COLR section. The TS markups previously supplied with NL-15-144 have been modified to include WCAP-1 6045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2014 and WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.

Reactor Systems Branch (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/ANC, or NEXUS/ANC) is used to predict the BOL HZP ITC.

Response

To clarify, a mix of the PHOENIX-P/ANC and PARAGON/ANC code systems were used to predict the BOL HZP ITC in Attachment 5 Tables 1 and 2. The predicted BOL HZP ITC values for Indian Point Unit 2 Cycles 20 and 21 and Indian Point Unit 3 Cycle 17 were generated using the PHOENIX-P/ANC code system. The predicted BOL HZP ITC values for Indian Point Unit 2 Cycle 22 and Indian Point Unit 3 Cycles 18 and 19 predicted BOL HZP ITC values were generated using the PARAGON/ANC code system.

ATTACHMENT 2 TO NL-16-027 ADDITIONAL TECHNICAL SPECIFICATION CHANGES (MARK-UP)

[BOTH UNITS ARE INCLUDED IN THIS ATTACHMENT]

Text changes indicated by lineout for deletion and Bold/Italics for additions Unit 2 Actual Affected Pages (revised pages are indicated below):

Page 5.6-4 Unit 3 Actual Affected Pages (revised pages are indicated below):

Page 5.0-35 ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. WCAP-1 2610-P-A, "VANTAGE+' Fuel Assembly Reference Core Report", April 1995;

9. WCAP-1 0079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code", August 1985;

10. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", August 1985; d

11. WCAP-1 0054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and Cosi Condensation Model", July 1997-1 c Thoe core operating limits shall be determined such that all applicable limits

(.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Cre Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient

~analysis limits, and accident analysis limits) of the safety analysis are met.

12. WCAP-1 3749-P-A, "Safety Evaluation NRC upon issuance for each reload cycle.

Supporting the Conditional Exemption of the Most ;nt Monitoringq Report Negative EOL Moderator Temperature Coefficient .=port is required by Condition B or F of LCO 3.3.3, "Post Accident Measuremen It", March 1997; I(PAM) Instrumentation," a report shall be submitted within the following

• k]4ay.ih report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the 1* Wr.AP.t F

• nA.*.A )rtion channels of the Function to OPERABLE status.

'Qualification of the Two-Dimensional Transport Code *erator Tube Inspection Report PAIRA(*Nh" AIInlJ.*f 2A'0&4 and t "'= ......

!l1 be submitted within 180 days after the initial entry into MODE 4 J.....n*,Jpletion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include:

14. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer IIdation mechanisms found, rope of inspections performed on each SG, Code", September 1986.

c.

J Nondestructive examination techniques utilized for each degradation mechanism, INDIAN POINT 2 5.6 - 4 INDIA POIT 2 .6-4Amendment No. 281

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

Position CPB 4.3-1, Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981. (Specification 3.2.3, Axial Flux Difference (AFD) (Constant Axial Offset Control));

• 4*a-*WCAP-12945-P-A, Volume 1 (Revision 2) and Volumes 2 through 5 (Revision 1), "Code Qualification Document for Best-Estimate Loss-of-Coolant-Accident Analysis,"

March 1998 (Westinghouse Proprietary);

[ 4*~ WCAP-II397-P-A, "Revised Thermal Design Procedure,"

April 1989 (Specification 2.1, Safety Limits (SL)) and Specification 3.4.1, (RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits);

• WCAP-8745-P-A, "Design Bases for the ThermalOvroe AT and Thermal Overtemperature AT Trip Functions,"

September 1986 (Specification 2.1, Safety Limits (SL));

6a. WCAP-10054-P-A, "SMALL BREAK ECCS EVALUATION MODEL USING NOTRUMP CODE," (W Proprietary). (Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z));

6b 4- WCAP-10054-P-A, Adedm2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code; Safety Injection into the Broken Loop and Cosi Condensation Model," July 1997 (Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)));

6c *4--WA-00 PA "NOTRUMP NODAL TRANSIENT SMALLBRAAN GENERAL NETWORK CODE," (W Proprietary) . (Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z))); and

[Y *- WCAP-12610, "VANTAGE+ Fuel Assembly Report,"

(W Proprietary). (Sp~ecification 3.2.1, Heat Flux Hot Channel Factor),-U cal8. WCAP-13749-P-A, "Safety Evaluation Supporting such that all

a. 1the Conditional Exemption of the Most Negative EOL ical limits, c Moderator Temperature Coefficient ICooling Sy tMeasurement," March 1997. (Specification 3.1.3, ISDM, tr nModerator Temperature Coefficient); and isis limits) of d.

S The COL, Insert9.including and 10. from next page.

any midcyc e revisions or supplements, shall be provided for each reload cycle to the NRC.

5.6.6 NOT USED (continued)

INDIAN POINT 3 5.0 - 35 50-3 Amendment mnmn 225 2

9. WCAP-1 6045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004. (Specification 3.1.1, Shutdown Margin, Specification 3.1.3, Moderator Temperature Coefficient, Specification 3.1.5, Shutdown Bank Insertion Limits, Specification 3.1.6, Control Bank Insertion Limits, Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)), Specification 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor, Specification 3.2.3, Axial Flux Difference (AFD),

Specification 3.9.1, Boron Concentration); and

10. WCAP-1 0965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986. (Specification 3.1.1, Shutdown Margin, Specification 3.1.3, Moderator Temperature Coefficient, Specification 3.1.5, Shutdown Bank Insertion Limits, Specification 3.1.6, Control Bank Insertion Limits, Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)), Specification 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor, Specification 3.2.3, Axial Flux Difference (AFD),

Specification 3.9.1, Boron Concentration).