ML12264A649

Summary of Telephone Conference Call Held on August 29, 2012, Between the NRC and Entergy, Concerning RAI Pertaining to the Grand Gulf Nuclear Station, LRA

ML12264A649
Person / Time
Site:	Grand Gulf
Issue date:	11/15/2012
From:	Ferrer N License Renewal Projects Branch 1
To:	Entergy Operations
Ferrer N, 415-1045
References
TAC ME7493
Download: ML12264A649 (9)

Text

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LICENSEE: Entergy Operations, Inc.

FACILITY: Grand Gulf Nuclear Station

SUMMARY

OF TELEPHONE CONFERENCE CALL HELD ON AUGUST 29, 2012, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND ENTERGY OPERATIONS, INC., CONCERNING REQUESTS FOR ADDITIONAL INFORMATION PERTAINING TO THE GRAND GULF NUCLEAR STATION, LICENSE RENEWAL APPLICATION (TAC. NO. ME7493)

The U.S. Nuclear Regulatory Commission (NRC or the staff) and representatives of Entergy Operations, Inc., (Entergy) held a telephone conference call on August 29, 2012, to discuss and clarify the staffs requests for additional information (RAls) concerning the Grand Gulf Nuclear Station, license renewal application. The telephone conference call was useful in clarifying the intent of the staff's RAls.

Enclosure 1 provides a listing of the participants and Enclosure 2 contains a listing of the RAls discussed with the applicant, including a brief description on the status of the items.

The applicant had an opportunity to comment on this summary.

Nathaniel Ferrer, Project ManagerLicense Renewal Branch, RPB1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-416

Enclosures:

As stated cc w/encls: See next page

TELEPHONE CONFERENCE CALL GRAND GULF NUCLEAR STATION LICENSE RENEWAL APPLICATION LIST OF PARTICIPANTS AUGUST 29,2012 PARTICIPANTS AFFILIATIONS Nate Ferrer U.S. Nuclear Regulatory Commission (NRC)

Seung Min NRC Ben Parks NRC Matt Homiack NRC Ted Ivy Entergy Operations, Inc. (Entergy)

Andy Taylor Entergy Alan Cox Entergy Stan Batch Entergy ENCLOSURE 1

REQUESTS FOR ADDITIONAL INFORMATION (SET 36)

LICENSE RENEWAL APPLICATION AUGUST 29,2012 The U.S. Nuclear Regulatory Commission (NRC or the staff) and representatives of Entergy Operations, Inc., held a telephone conference call on August 29, 2012, to discuss and clarify the following requests for additional information (RAls) concerning the license renewal application (LRA).

Draft RAI 4.2.1-1a Background. By letter dated July 25,2012, the applicant responded to RAI 4.2.1-1, which addresses why the applicant did not identify the reactor vessel neutron fluence calculation as a time-limited aging analysis (TLAA). The applicant stated that the neutron fluence calculation is not a TLAA since, as a stand-alone analysis, it does not meet the definition in 10 CFR 54.3(a).

The applicant also stated that specifically, a neutron fluence calculation does not "consider the effects of aging," which is the second element of the six-element definition of a TLAA in 10 CFR 54.3(a).

Issue. Since the neutron fluence analysis considers embrittlement of the reactor vessel due to fast neutron fluence (E > 1 MeV), the neutron fluence analysis considers the effects of aging (Le., neutron embrittlement of the reactor vessel). The reactor vessel neutron fluence analysis with time-limited assumptions is also integral to the other neutron embrittlement TLAAs for the reactor vessel (e.g., upper-shelf energy analysis and P-T limits analysis). Therefore, the staff finds that the neutron fluence analysis should be identified as a TLAA with an adequate TLAA disposition as addressed in 10 CFR Part 54.21 (c)(1 )(i), (ii) and (iii).

Request.

a. Identify the reactor vessel neutron fluence analysis as a TLAA, based on the fact that the neutron fluence analysis considers the reactor vessel neutron embrittlement and is also integral to the other neutron embrittlement TLAAs for the reactor vessel. Alternatively, provide additional justification for why the reactor vessel fluence analysis is not a TLAA.

b. If the reactor vessel fluence analysis is a TLAA, describe the applicant's TLAA disposition of the reactor vessel neutron fluence analysis in terms of the dispositions described in 10 CFR Part 54.21 (c)(1)(i), (ii) and (iii). In addition, ensure that LRA Section 4.2.1, Table 4.1-1 and Section A.2.1.1 are revised to include the adequate TLAA disposition.

Discussion: The applicant stated that the question was unclear because of the description of fluence considering embrittlement. The staff was referring to the accrual of neutrons on the vessel surface and will reword the issue and request sections as follows:

Issue. Since the neutron fluence analysis considers the accrual of neutrons on the vessel surface as a function of the reactor operating power level, the neutron fluence analysis considers the effects of aging (Le., neutron embrittlement of the reactor vessel). The reactor vessel neutron fluence analysis with time-limited assumptions is also integral to the neutron embrittlement TLAAs for the reactor vessel (e.g., upper-shelf energy analysis and P-T limits analysis). Therefore, the ENCLOSURE 2

-2 staff finds that the neutron fluence analysis should be identified as a TLAA with an adequate TLAA disposition as addressed in 10 CFR Part 54.21 (c)(1)(i), (ii) and (iii).

Request.

a. Identify the reactor vessel neutron f1uence analysis as a TLAA, based on the fact that the neutron fluence analysis considers the accrual of neutrons on the vessel surface as a function of the reactor operating power level and is also integral to the neutron embrittlement TLAAs for the reactor vessel. Alternatively, provide additional justification for why the reactor vessel fluence analysis is not a TLAA.

b. If the reactor vessel fluence analysis is a TLAA, describe the applicant's TLAA disposition of the reactor vessel neutron fluence analysis in terms of the dispositions described in 10 CFR Part 54.21 (c)(1)(i), (ii) and (iii).

In addition, ensure that LRA Section 4.2.1, Table 4.1-1 and Section A.2.1.1 are revised to include the adequate TLAA disposition.

The staff will issue the reworded question as a formal RAI.

Draft RAI 4.2.1-2a Background. By letter dated July 25, 2012, the applicant responded to RAI4.2.1-2, which addresses the adequacy of combining two neutron fluence calculation methods in its neutron f1uence analysis (I.e., combination of the pre-EPU MPM method and the post-EPU GEH method in the analysis). As part of its response, the applicant provided the following information:

• The post-EPU peak neutron flux values for the welds H1, V1, V2, V3, and V4 are less than the corresponding pre-EPU peak flux values approximately by an order of magnitude of 3 (i.e., approximately by a thousand times; the post-EPU peak neutron flux in the order of 1E7 n/cm 2-s in contrast with the pre-EPU peak neutron flux in the order of 1E10 n/cm 2-s, for E > 1 MeV).

• Welds H1, V1, V2, V3, and V4 are the welds on the reactor vessel internal top guide that sits above the core shroud.

• The applicant entered this discrepancy between the post-EPU neutron flux and the pre EPU neutron flux into the corrective action program.

• No locations evaluated in the post-EPU GEH fluence evaluation except for welds H1, V1, V2, V3, and V4 were found to have flux values lower than pre-EPU flux values.

In terms of the weld locations, Figures 3-1 and 3-9 of BWRVIP-02-A, "BWR Vessel and Internals Project BWR Core Shroud Repair Design Criteria, Revision 2," indicate that welds H1, V1, V2, V3, and V4 are core shroud horizontal (H) and vertical (V) welds, which are located in the top portion of the core shroud cylindrical shells. Figure 2-10 of BWRVI P-26-A, "BWR Vessel and Internals Project BWR Top Guide Inspection and Flaw Evaluation Guidelines," indicates that these welds are above and adjacent to the top guide.

- 3 By letter dated July 25, 2012, the applicant also provided the 40-year and 60-year neutron fluence values for the reactor vessel internals as part of its response to RAI 4.7.3-1. The fast neutron fluence data (E > 1 MeV) include the neutron fluence for the core spray spargers that are adjacent to the top portion of the core shroud cylindrical shells. The 60-year fast neutron fluence (9.04E18 n/cm 2 ) of the core spray spargers is less than the 40-year fast neutron fluence (1.50E21 n/cm 2) approximately by an order of magnitude of 2.

In addition, the core shroud head dome and core shroud head stud adjacent to the core spray spargers have similar fluence discrepancy between the 60-year and 40-year neutron fluence values (E > 1 MeV). The core shroud head dome and shroud head stud have a 60-year fluence value less than 9.04E18 n/cm 2 and a 40-year fluence value less than 1.50E21 n/cm 2 , based on the f1uence calculations for the nearest available f1uence calculation node.

The NRC staff also identified an issue with the applicant's neutron fluence calculations for the period of extended operation. As addressed above, the pre-EPU fluences determined using the flux values from the Manahan method (MPM method) were added to the post-EPU fluences determined using the flux values from the GEH method. The NRC staff requested, in RAI 4.2.1-2, request d.4, that the applicant address the analytic uncertainty associated with combining f1uences in this fashion. The applicant stated, in its response dated July 25, 2012, that" ... it is expected that the combination of these values is acceptable with respect to the uncertainty treatment specifications of RG 1.190."

Issue. Based on the staff's review as summarized above, the staff identified the following items that need additional information:

a. It is not clear whether the welds H1, V1, V2, V3, and V4 are core shroud welds in the top portion of the shroud cylindrical shells as indicated in BWRVIP-02-A, or welds in the top guide as indicted in the applicant's response.

b. The staff needs to confirm whether adequate corrective actions were taken for the fluence calculations on the welds H1, V1, V2, V3, and V4 so that the applicant's corrective actions resolved the significant difference between the pre-EPU and post-EPU fast neutron flux values of these welds.

c. The staff needs justification for why the 60-year fluence (E > 1 MeV) of the core spray sparger, core shroud dome, and core shroud head stud components are less than their 40-year fluence.

d. Given that the welds H1, V1, V2, V3, and V4 have post-EPU neutron flux (GEH method) significantly less than the pre-EPU neutron flux (MPM method), the staff needs additional information regarding the reactor vessel neutron flux (E > 1 MeV) to confirm that the reactor vessel plates, welds and nozzles have post-EPU neutron flux values that are reasonably greater than the pre-EPU neutron flux values.

e. The applicant did not provide its criteria, in terms of the difference between the pre-EPU and post-EPU neutron flux values (E > 1 MeV), to initiate a corrective action for the reactor vessel and reactor vessel internal neutron f1uence analyses (e.g., a corrective action is initiated to evaluate neutron flux differences if a post-EPU neutron flux is not greater than X percent of the corresponding pre-EPU neutron flux, in view that the EPU is planned to implement approximately Y percent thermal power increase).

-4

f. Regulatory Guide (RG) 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," Regulatory Position 1.4.1, provides guidance for analytic uncertainty analysis to support methodology qualification and uncertainty estimates (including combination of uncertainties). The applicant's response, addressing an expectation of acceptability, does not provide adequate information to determine how the new calculational method, which is based on adding the fluence values, obtained using different calculational methods, together, adheres to the guidance contained in RG 1.190.

Request.

a. Clarify whether the welds H1, V1, V2, V3, and V4 are core shroud welds in the top portion of the core shroud cylindrical shells, or welds in the top guide.

b. Provide additional information to confirm that adequate corrective actions were taken for the fluence calculations on the welds H1, V1, V2, V3, and V4 so that the applicant's corrective actions resolved the significant difference between the pre-EPU and post-EPU fast neutron flux values.

c. Provide justification for why the 60-year fluence (E > 1 MeV) of the core spray sparger, core shroud dome, and core shroud head stud components is less than the 40-year fluence of these components.

d. Provide the pre-EPU and post-EPU reactor vessel inner surface neutron flux values (E >

1 MeV) of the reactor vessel plates, welds and nozzles in order to confirm that these reactor vessel materials have post-EPU neutron flux values that are reasonably greater than the pre-EPU neutron flux values. These neutron flux comparisons should include the reactor vessel plates, welds and nozzles listed in LRA Table 4.2-2.

e. As part of the response, confirm whether the reactor vessel inner surfaces near the welds H1, V1, V2, V3, and V4 have post-EPU neutron flux values (E > 1 MeV) that are reasonably greater than the pre-EPU neutron flux values (E > 1 MeV).

f. Provide the applicant's criteria, in terms of the difference between the pre-EPU and post EPU neutron flux values (E > 1 MeV), to initiate a corrective action for the reactor vessel and reactor vessel internal neutron fluence analyses (e.g., a corrective action is initiated to evaluate neutron flux differences if a post-EPU neutron flux is not greater than X percent of the corresponding pre-EPU neutron flux, in view that the EPU is planned to implement approximately Y percent thermal power increase).

g. Demonstrate that the combined calculational uncertainty associated with both fluence methodologies remains within RG 1.190 guidance, or provide an alternative justification for the acceptability of this method that demonstrates that it satisfies the regulations discussed in the Introduction section of RG 1.190.

Discussion: The applicant stated that it was not clear what additional information request (e) was seeking in addition to request (d). The staff considers the information requested (e) a further clarification related to request (d) and will reword the request section as follows:

- 5 Request.

a. Clarify whether the welds H1, V1, V2, V3, and V4 are core shroud welds in the top portion of the core shroud cylindrical shells, or welds in the top guide.

b. Provide additional information to confirm that adequate corrective actions were taken for the fluence calculations on the welds H 1, V1, V2, V3, and V4 so that the applicant's corrective actions resolved the significant difference between the pre-EPU and post-EPU fast neutron flux values.

c. Provide justification for why the 60-year fluence (E > 1 MeV) of the core spray sparger, core shroud dome, and core shroud head stud components is less than the 40-year fluence of these components.

d. Provide the pre-EPU and post-EPU reactor vessel inner surface neutron flux values (E > 1 MeV) of the reactor vessel plates, welds and nozzles in order to confirm that these reactor vessel materials have post-EPU neutron flux values that are reasonably greater than the pre-EPU neutron flux values. These neutron flux comparisons should include the reactor vessel plates, welds and nozzles listed in LRA Table 4.2-2. As part of the response, include a discussion of the reactor vessel inner surfaces near the welds H1, V1, V2, V3, and V4.

e. Provide the applicant's criteria, in terms of the difference between the pre EPU and post-EPU neutron flux values (E > 1 MeV), to initiate a corrective action for the reactor vessel and reactor vessel internal neutron fluence analyses (e.g., a corrective action is initiated to evaluate neutron flux differences if a post-EPU neutron flux is not greater than X percent of the corresponding pre-EPU neutron flux, in view that the EPU is planned to implement approximately Y percent thermal power increase).

f. Demonstrate that the combined calculational uncertainty associated with both fluence methodologies remains within RG 1.190 guidance, or provide an alternative justification for the acceptability of this method that demonstrates that it satisfies the regulations discussed in the Introduction section of RG 1.190.

The staff will issue the reworded question as a formal RAI.

November 15, 2012 LICENSEE: Entergy Operations, Inc.

FACILITY: Grand Gulf Nuclear Station SUB.JECT:

SUMMARY

OF TELEPHONE CONFERENCE CALL HELD ON AUGUST 29, 2012, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND ENTERGY OPERATIONS, INC., CONCERNING REQUESTS FOR ADDITIONAL INFORMATION PERTAINING TO THE GRAND GULF NUCLEAR STATION, LICENSE RENEWAL APPLICATION (TAC. NO. ME7493)

The U.S. Nuclear Regulatory Commission (NRC or the staff) and representatives of Entergy Operations, Inc., (Entergy) held a telephone conference call on August 29,2012, to discuss and clarify the staffs requests for additional information (RAls) concerning the Grand Gulf Nuclear Station, license renewal application. The telephone conference call was useful in clarifying the intent of the staff's RAls. provides a listing of the participants and Enclosure 2 contains a listing of the RAls discussed with the applicant, including a brief description on the status of the items.

The applicant had an opportunity to comment on this summary.

IRA!

Nathaniel Ferrer, Project Manager License Renewal Branch, RPB1 Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-416

Enclosures:

As stated cc w/encls: See next page ADAMS Accession No.: ML12264A649 *concurred via email OFFICE LA:DLR* PM:RPB1:DLR BC:RPB1:DLR PM:RPB1 :DLR NAME YEdmonds NFerrer DMorey NFerrer DATE 10/2/12 10/27112 10/26/12 11/15/12 OFFICIAL RECORD COpy

Memorandum to Entergy Operations Inc. from N. Ferrer dated November 15, 2012

SUMMARY

OF TELEPHONE CONFERENCE CALL HELD ON AUGUST 29, 2012 BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND ENTERGY OPERATIONS, INC., CONCERNING REQUESTS FOR ADDITIONAL INFORMATION PERTAINING TO THE GRAND GULF NUCLEAR STATION, LICENSE RENEWAL APPLICATION (TAC. NO. ME7493)

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