W3F1-2017-0004, Responses to Request for Additional Information from Sets 4, 5 and 6 Regarding the License Renewal Application

Responses to Request for Additional Information from Sets 4, 5 and 6 Regarding the License Renewal Application

ML17037D400
Person / Time
Site:	Waterford
Issue date:	02/06/2017
From:	Chisum M Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2017-0004
Download: ML17037D400 (11)

Text

Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504-739-6660 Fax 504-739-6698 mchisum@entergy.com Michael R. Chisum Site Vice President Waterford 3 W3F1-2017-0004 February 6, 2017 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Responses to Request for Additional Information from Sets 4, 5 and 6 Regarding the License Renewal Application for Waterford Steam Electric Station, Unit 3 (Waterford 3)

Docket No. 50-382 License No. NPF-38

REFERENCES:

1. Entergy letter W3F1-2016-0012 License Renewal Application, Waterford Steam Electric Station, Unit 3 dated March 23, 2016.

2. NRC letter to Entergy Requests for Additional Information for the Review of the Waterford Steam Electric Station, Unit 3, License Renewal Application - Set 3 dated October 12, 2016.

3. NRC letter to Entergy Requests for Additional Information for the Review of the Waterford Steam Electric Station, Unit 3, License Renewal Application - Set 5 dated November 7, 2016.

3. NRC letter to Entergy Requests for Additional Information for the Review of the Waterford Steam Electric Station, Unit 3, License Renewal Application - Set 6 dated November 7, 2016.

Dear Sir or Madam:

By letter dated March 23, 2016, Entergy Operations, Inc. (Entergy) submitted a license renewal application (Reference 1).

In References 2, 3 and 4, the NRC staff made Requests for Additional Information (RAI) Sets 3, 5 and 6 needed to complete its review. Three RAIs contained in these requests required longer response times than the others. Enclosure 1 provides the responses to these RAIs.

There are no new regulatory commitments contained in this submittal. If you require additional information, please contact the Regulatory Assurance Manager, John Jarrell, at 504-739-6685.

W3F1-2017-0004 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on February 6, 2017.

Sincerely, MRC/AJH

Enclosure:

1. RAI Responses - Waterford 3 License Renewal Application cc: Kriss Kennedy RidsRgn4MailCenter@nrc.gov Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 E. Lamar Blvd.

Arlington, TX 76011-4511 NRC Senior Resident Inspector Frances.Ramirez@nrc.gov Waterford Steam Electric Station Unit 3 Chris.Speer@nrc.gov P.O. Box 822 Killona, LA 70066-0751 U. S. Nuclear Regulatory Commission Phyllis.Clark@nrc.gov Attn: Phyllis Clark Division of License Renewal Washington, DC 20555-0001 U. S. Nuclear Regulatory Commission April.Pulvirenti@nrc.gov Attn: Dr. April Pulvirenti Washington, DC 20555-0001 Louisiana Department of Environmental Ji.Wiley@LA.gov Quality Office of Environmental Compliance Surveillance Division P.O. Box 4312 Baton Rouge, LA 70821-4312

Enclosure 1 to W3F1-2017-0004 RAI Responses Waterford 3 License Renewal Application to W3F1-2017-0004 Page 1 of 8 RAI 4.2.1-1 (Set 3)

Background:

LRA Section 4.2.1 describes Waterford Unit 3 reactor vessel neutron fluence calculations and that the methods used satisfy the criteria set forth in Regulatory Guide (RG) 1.190. The LRA also states that these methods have been approved by the NRC and are described in detail in WCAP-14040-A, Revision 4, and WCAP-16083-NP-A, Revision 0.

The staff noted that WCAP-18002-NP, Revision 0 describes neutron embrittlement TLAAs related to Waterford Unit 3 reactor vessel integrity. Specifically, Section 2 of WCAP-18002-NP, Revision 0 indicates the following:

• WCAP-14040-A, Revision 4, and WCAP-16083-NP-A, Revision 0 describe NRC-approved fluence methods, which include the one-dimensional/two-dimensional (1D/2D) flux synthesis technique to obtain a three-dimensional (3D) neutron flux. These WCAP reports also mention the 3D neutron transport calculation code, TORT.

• The neutron fluence values of Waterford Unit 3 reactor vessel were calculated using a Westinghouse-developed code, RAPTOR-M3G similar to TORT.

Issue:

It is not clear whether the applicants fluence method, which uses the RAPTOR-M3G code, has been incorporated into the current licensing basis including staffs review and approval.

Request:

1. Clarify whether the applicants fluence method, which uses the RAPTOR-M3G code, has been incorporated into the current licensing basis.

2. If RAPTOR-M3G is not part of the current licensing basis:

a. Provide justification for the use of the code.

b. Clarify how the plant-specific dosimetry data of Waterford Unit 3 were used in measurement benchmarks to confirm the adequacy of use of the RAPTOR-M3G code for Waterford Unit 3 reactor vessel fluence calculations.

Waterford 3 Response

1. The use of RAPTOR-M3G for reactor vessel neutron fluence calculations has been incorporated into the WF3 current license basis by EC 68581. The guidance in NEI 96-07 Section 4.3.8.2 was used to determine that the transition to RAPTOR-M3G is not a departure from a method of evaluation, and was implemented via a 50.59 evaluation.

2. The response to parts 2a and 2b are not applicable based on response to part 1.

to W3F1-2017-0004 Page 2 of 8 RAI 4.2.3-1 (Set 5)

Background:

LRA [License Renewal Application] Section 4.2.3 describes the applicants time-limited aging analysis on pressurized thermal shock (PTS). During the audit, the staff noted that the following report describes more detailed information on the PTS analysis: WCAP-18002-NP, Revision 0, Waterford Unit 3 Time-Limited Aging Analysis on Reactor Vessel Integrity, dated July 2015.

WCAP-18002-NP, Revision 0, indicates that the initial unirradiated reference temperature (called RTNDT(U) or initial RTNDT) of lower shell plate M-1004-2 is updated from 22 °F to 0 °F. The WCAP report also indicates that this update is based on drop-weight and transverse-orientation Charpy V-notch test data per ASME Code Section III, NB-2300 in comparison with the previously determined value (22 °F) based on NRC Branch Technical Position (BTP) MTEB 5-2, which is comparable to the current BTP 5-3 in NUREG-0800, 2007.

The staff also noted that the previously determined initial RTNDT value (22 °F) is described in Section 5 of WCAP-16088-NP, Revision 1, Waterford Unit 3 Reactor Vessel Heatup and Cooldown Limit Curves for Normal Operation, dated September 2003 (ADAMS ML041620063).

Issue:

The LRA, including LRA Table 4.2-3, does not describe a specific provision of ASME Code Section III, NB-2331 that the applicant used in updating the initial RTNDT of lower shell plate M-1004-2. In a similar manner, the staff noted that additional information is necessary to clarify whether the applicants test data were adequately used in updating the initial RTNDT values for the following beltline materials: (a) intermediate shell plates M-1003-1, M-1003-2 and M-1003-3; and (b) lower shell plates M-1004-1 and M-1004-3.

Request:

In order to demonstrate that the applicants test data were adequately used in updating the initial RTNDT of the beltline materials discussed above, describe the specific provision of ASME Code Section III, NB-2331 that the applicant used. As part of the response, provide the temperature (TCv) representing a minimum of 50 ft-lb absorbed energy and 35 mil lateral expansion as obtained in transverse-orientation Charpy V-notch tests for each material if such temperature was determined in the evaluation of material properties.

Waterford 3 Response The initial RTNDT values for the six Waterford Unit 3 beltline plates have been historically documented as shown in Table 2-1 of WCAP-16088-NP, Revision 2. Per Waterford Unit 3 FSAR, Revision 308, Table 5.3-13, these initial RTNDT values were determined using Branch Technical Position (BTP) MTEB 5-2. The title and revision of the BTP series have changed over time. The older BTP MTEB 5-2 is comparable to the newer NUREG-0800, BTP 5-3. Therefore, the initial to W3F1-2017-0004 Page 3 of 8 RTNDT and initial upper-shelf energy (USE) values for all of the beltline and extended beltline plate or forging materials were reevaluated. For those materials that already have documented values, the basis for those values was reconsidered, and new values were assigned as appropriate. Per Table 2-10 of WCAP-18002-NP, Revision 0, only the three cylindrical shell courses of plate material (3 upper shell plates, 3 intermediate shell plates, and 3 lower shell plates) reach fluence greater than or equal to 1 x 1017 n/cm2 (E > 1.0 MeV) at 55 EFPY. Thus, reactor vessel nozzles are not considered herein.

Initial RTNDT and initial USE values are defined for each of the nine reactor vessel plates. Although the historical values made use of BTP MTEB 5-2 and longitudinal test data, transverse data is available for the nine reactor vessel plates. The transverse data will be used per the provisions of ASME Code Section III, Subarticle NB-2331. For completeness, this response identifies the methodologies used for determination of the initial RTNDT values for the extended beltline plate materials and the initial USE values, per ASTM E185-82 for all nine reactor vessel plates discussed herein.

Subarticle NB-2331 of Section III of the ASME Code Summer of 1972 Addenda to the 1971 Edition or later requires both drop-weight test data as well as Charpy V-notch test data from transverse specimens for determination of initial RTNDT values.

ASME Code Section III, NB-2331, Material for Vessels Pressure-retaining materials for vessels, other than bolting, shall be tested as follows:

(a) Establish a reference temperature RTNDT; this shall be done as follows:

(1) Determine a temperature TNDT that is at or above the nil-ductility transition temperature by drop weight tests.

(2) At a temperature not greater than TNDT + 60°F (TNDT + 33°C), each specimen of the Cv test (NB-2321.2) shall exhibit at least 35 mils (0.89 mm) lateral expansion and not less than 50 ft-lb (68 J) absorbed energy. Retesting in accordance with NB-2350 is permitted.

When these requirements are met, TNDT is the reference temperature RTNDT.

(3) In the event that the requirements of (2) above are not met, conduct additional Cv tests in groups of three specimens (NB-2321.2) to determine the temperature TCv at which they are met. In this case the reference temperature RTNDT = TCv - 60°F (TCv - 33°C). Thus, the reference temperature RTNDT is the higher of TNDT and [TCv - 60°F (TCv - 33°C)].

(4) When a Cv test has not been performed at TNDT + 60°F (TNDT + 33°C), or when the Cv test at TNDT + 60°F (TNDT + 33°C) does not exhibit a minimum of 50 ft-lb (68 J) and 35 mils (0.89 mm) lateral expansion, a temperature representing a minimum of 50 ft-lb (68 J) and 35 mils (0.89 mm) lateral expansion may be obtained from a full Cv impact curve developed from the minimum data points of all the Cv tests performed.

(b) Apply the procedures of NB-2331(a) to NB-2331(b)(1), (2), and (3):

(1) the base material; to W3F1-2017-0004 Page 4 of 8 (2) the base material, the heat affected zone, and weld metal from the weld procedure qualification tests in accordance with NB-4330; (3) the weld metal of NB-2431.

The test data necessary to use these Code provisions for determination of the initial RTNDT values are documented in the Waterford Unit 3 Certified Material Test Reports (CMTRs) or C-PENG-ER-004 for each of the nine Waterford Unit 3 reactor vessel cylindrical shell course plate materials (3 upper shell plates, 3 intermediate shell plates, and 3 lower shell plates). Using this test data, the initial RTNDT values for these materials were determined directly from the data or by using a hyperbolic tangent curve fit through the minimum data points, in accordance with ASME Code Section III, Subarticle NB-2331, Paragraphs (a)(2) and (a)(4), respectively. The differences between the initial RTNDT values summarized in the various Waterford Unit 3 analyses of record (e.g. WCAP-16088-NP, Revision 2) and the new values are due to use of the hyperbolic tangent curve-fitting method with the minimum data points as well as a closer look at all available data and testing information (e.g., transverse Charpy V-notch results).

In some cases, hyperbolic tangent curve fits are not needed since the available data is sufficient to determine the initial RTNDT value. The orientation of the test specimen (transverse vs. longitudinal) is considered since it is vital in terms of the methodology used for initial RTNDT determination. NUREG-0800, BTP 5-3, provides guidance for instances where the available data is not sufficient to meet ASME Code Section III, Subarticle NB-2331 provisions for determination of initial RTNDT. However, for this reevaluation, NUREG-0800, BTP 5-3 is no longer required or used for the nine Waterford Unit 3 reactor vessel cylindrical shell course plates.

The initial RTNDT values for Waterford Unit 3 that have been historically used for the beltline plate materials are documented in Table 2-1 of WCAP-16088-NP, Revision 2. The values from WCAP-16088-NP, Revision 2 are used in the prior reactor vessel integrity analyses of record and were considered part of the licensing basis for Waterford Unit 3. Table 1 below compares the prior licensing basis values with the updated values. As shown in Table 1, the initial RTNDT values included in the original licensing basis are more conservative than the updated values for the three lower shell plates and are less conservative than the updated values for the three intermediate shell plates.

The new initial RTNDT values for the six traditional beltline intermediate and lower shell plate materials supersede those that have been historically used. The initial RTNDT values for the three extended beltline upper shell plate materials were used in reactor vessel integrity evaluations for the first time in WCAP-18002-NP, Revision 0. Thus, historical values are not listed in Table 1. In summary, the values utilized in the Waterford Unit 3 evaluation of the reactor vessel integrity TLAAs (WCAP-18002-NP, Revision 0) should now be considered the current licensing basis. Values for the nine reactor vessel plates, as documented in Table 1, are relevant to the 20-year license renewal activities for the plant. Table 2 shows the specific provision of ASME Code Section III, NB-2331 used and the temperature (TCv) representing a minimum of 50 ft-lb absorbed energy and 35 mil lateral expansion as obtained in transverse-orientation Charpy V-notch tests for each of the Waterford Unit 3 plate materials.

to W3F1-2017-0004 Page 5 of 8 Table 1 Comparison of Waterford Unit 3 Reactor Vessel Plate Initial RTNDT and Initial USE RTNDT(U)Values (a)

Updated Initial USE (a) Updated(b)

Material Description (°F) RTNDT(U) (ft-lb) Initial USE

(°F) (ft-lb)

Intermediate Shell Plate M-1003-1 -30 -25.1 94 108 Intermediate Shell Plate M-1003-2 -50 -20 97 132 Intermediate Shell Plate M-1003-3 -42 -20 90 111 Lower Shell Plate M-1004-1 -15 -37.6 106 135 Lower Shell Plate M-1004-2 22 0 141 141 Lower Shell Plate M-1004-3 -10 -20 94 118 Upper Shell Plate M-1002-1 --- -15.4 --- 104 Upper Shell Plate M-1002-2 --- -1.4 --- 95 Upper Shell Plate M-1002-3 --- -20 --- 120 Notes for Table 1:

(a) Per WCAP-16088-NP, Rev. 2.

(b) Initial USE values for the six reactor vessel traditional beltline plate materials were updated in WCAP-17969-NP as part of the Capsule 83° analysis.

Enclosure 1 to W3F1-2017-0004 Page 6 of 8 Table 2 Summary of Waterford Unit 3 Reactor Vessel Plate Initial RT NDT Determination RTNDT(U) TCV Charpy TNDT Overall ASME Section III, Material Description Limiting Limiting Subarticle NB-2331,

(°F) (°F) Parameter (°F) Parameter Paragraph (a)(2), (a)(3) or (a)(4) Applicability Impact (a)

Intermediate Shell Plate M-1003-1 -25.1 34.9 -30 TCV (a)(4) energy Intermediate Shell Plate M-1003-2 -20 40 N/A(c) -40 TCV(a) (a)(3)

Intermediate Shell Plate M-1003-3 -20 40 N/A(c) -30 TCV(a) (a)(3)

Impact (a)

Lower Shell Plate M-1004-1 -37.6 22.4 -40 TCV (a)(4) energy Impact (b)

Lower Shell Plate M-1004-2 0 47.0 0 TNDT (a)(3) energy Lower Shell Plate M-1004-3 -20 10 N/A(c) -20 TNDT(b) (a)(2)

Impact (a)

Upper Shell Plate M-1002-1 -15.4 44.6 -20 TCV (a)(4) energy Impact (a)

Upper Shell Plate M-1002-2 -1.4 58.6 -20 TCV (a)(4) energy (c) (b)

Upper Shell Plate M-1002-3 -20 40 N/A -20 TNDT (a)(2)

Notes for Table 2 (a) RTNDT(U) = TCV - 60°F (Charpy Limited)

(b) RTNDT(U) = TNDT (Drop Weight Limited)

(c) Both the impact energy (ft-lb) and the lateral expansion (mils) exceeded the minimum values of 50 ft-lb and 35 mils, respectively, at the same tested temperature to W3F1-2017-0004 Page 7 of 8 RAI 4.7.4-1 (Set 6)

Background:

LRA Section 4.7.4 provides the applicant TLAA for the aging evaluation of reactor vessel internals (RVI), other than those associated with applicants metal fatigue TLAA for these components. The applicant identifies that the aging evaluations of irradiation-assisted stress corrosion cracking and loss of fracture toughness due to thermal aging and neutron irradiation embrittlement in its 2003 extended power uprate (EPU) license amendment request are analyses that conform to the definition of a TLAA in 10 CFR 54.3(a). The applicant stated that the implementation of LRA AMP B.1.33, Reactor Vessel Internals Program, will ensure that these TLAAs are acceptable in accordance with 10 CFR 54.21(c)(1))(iii).

The license amendment request for the EPU was submitted on November 3, 2003, and approved in an NRC-issued safety evaluation (SE) dated April 15, 2005 (ML051030068). Section 2.1.4 of the SE identifies that the projected neutron fluences for RVI components in the vicinity of the reactor core will range from 3.0 - 5.0 X 1022 n/cm2 (E > 0.1 MeV) through 40 years of licensed operations.

Issue:

EPRI Report MRP-191 estimates that RVI components in the core shroud would generally have neutron fluences ranging from 1.0 - 5.0 X 1022 n/cm2 through 60 years of licensed operations. The staff needs additional demonstration that the neutron fluence values for these types of RVI components through 60 years of licensed operation will not exceed the fluence estimates for the components in Table 4-7 of the MRP-191 report. Otherwise, the staff will need further assessment of the inspection bases for core shroud assembly components if the 60-year projected fluences for these components will exceed those specified for the components in MRP-191.

Request:

Justify (with a technical explanation) why the projected neutron fluences for RVI core shroud components through 60 years of operations are considered bounded by the fluence estimates for these components in Table 4-7 of the MRP-191 report. Otherwise, clarify what the impact will be on the FMECA assessment for these components and the inspection plan for RVI components if the 60-year neutron fluence value for any RVI core shroud component will exceed the neutron fluence estimate for the component in Table 4-7 of the MRP-191 report.

to W3F1-2017-0004 Page 8 of 8 Waterford 3 Response MRP-227-A (Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines) provides inspection and evaluation guidelines for managing aging effects in pressurized water reactor vessel internal components. Specifically, the guidelines are applicable to reactor vessel internal structural components, including core shroud components. MRP-191 describes the process and results of categorizing Westinghouse and Combustion Engineering (CE) designed pressurized water reactor (PWR) internals components according to age-related degradation and significance. Its results are a key element in developing the inspection and evaluation guidelines of MRP-227-A.

By letter dated December 16, 2013, as supplemented by letters dated January 19, June 18, July 9, 2015, Entergy submitted an aging management program for the reactor vessel internals at Waterford

3. MRP-227-A and its supporting reports were used as the technical bases for developing WF3's aging management program.

The fluence values for RVI components in EPRI Report MRP-191 are estimates. They are not bounding values above which the MRP-191 evaluations would be invalid.

EPRI issued letter MRP-2013-025 (MRP-227-A Applicability Template Guideline, October 14, 2013 (NRC ADAMS Accession No. ML13322A454) to establish a range of conditions for which the MRP-227-A inspection and evaluation guidelines are applicable. For a CE-designed reactor, MRP-2013-025 identified that neutron fluence and heat generation rates are acceptable for applicability of the inspection and evaluation guidance of MRP-227-A if reactor parameters meet the following threshold values.

• Active fuel to fuel alignment plate (FAP) distance > 12.4 inches

• Average core power density < 110 Watts/cm3

• Heat generation figure of merit, F 68 Watts/cm3 These threshold values address the plant-specific applicability of MRP-227-A in the axial and the radial direction, which includes the core shroud component locations.

In 2015, Westinghouse evaluated the WF3 reactor against the above threshold values. As indicated in the NRC safety evaluation for the WF3 aging management program for reactor vessel internals, WF3 provided the plant-specific values of the heat generation rate, the maximum average core power density, and the FAP distance. The NRC staff reviewed these values and determined that they comply with the values in MRP-2013-025. Therefore, the staff concluded that the licensee satisfied the guidelines related to fuel management issues addressed in MRP-2013-025. As discussed in MRP-2013-025, those guidelines were established to ensure that plant-specific fluence levels remained within acceptable values to ensure continuing applicability of MRP-227-A.

Therefore, there is no impact on the inspection plan for RVI components if the 60-year neutron fluence value for any RVI core shroud component exceeds the neutron fluence estimate for the component in Table 4-7 of the MRP-191 report.