NLS2015112, Response to NRC Requests for Additional Information for Relief Request No. RR5-01, Alternate Weld Overlay Repair for a Dissimilar Metal Weld Joining Nozzle to Control Rod Drive End Cap

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Response to NRC Requests for Additional Information for Relief Request No. RR5-01, Alternate Weld Overlay Repair for a Dissimilar Metal Weld Joining Nozzle to Control Rod Drive End Cap
ML15310A059
Person / Time
Site: Cooper Entergy icon.png
Issue date: 10/29/2015
From: Limpias O
Nebraska Public Power District (NPPD)
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2015112
Download: ML15310A059 (7)
v • d • e


Text

N Nebraska Public Power District Always there when you need us NLS2015112 October 29, 2015 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Response to Nuclear Regulatory Commission Requests for Additional Information for Relief Request No. RR5-01, Alternate Weld Overlay Repair for a Dissimilar Metal Weld Joining Nozzle to Control Rod Drive End Cap Cooper Nuclear Station, Docket No. 50-298, DPR-46

References:

1. Email from Lois M. James, Nuclear Regulatory Commission, to Jim Shaw, Nebraska Public Power District, dated September 15, 2015, "Request for Additional Information - Relief Request for RR5-01, Cooper - TAC No.

MF63 32"

2. Letter from Oscar A. Limpias, Nebraska Public Power District, to the U.S.

Nuclear Regulatory Commission, dated June 9, 2015, "10 CFR 50.55a Requests for Fifth Ten-Year Inservice Inspection Interval" (ML15167A066)

Dear Sir or Madam:

The purpose of this letter is for the Nebraska Public Power District to respond to the Nuclear Regulatory Commission's Requests for Additional Information (RAI) (Reference 1) related to the Cooper Nuclear Station "10 CFR 50.55a Requests for Fifth Ten-Year Inservice Inspection Interval" (Reference 2).

The response to the specific RAI questions is provided in the attachment to this letter.

This letter does not contain any new regulatory commitments.

If you have any questions concerning this matter, please contact Jim Shaw, Licensing Manager, at (402) 825-2788.

Sincerely Vice President - Nuclear and Chief Nuclear Officer

/dv COOPER NUCLEAR STATION 0(7 Telephone: (402) 825-3811 / Fax:. (402) 825-5211 ww.nppd.com

NLS20151 12 Page 2 of 2

Attachment:

Response to Nuclear Regulatory Commission Requests for Additional Information for 10 CFR 50.55a Requests for Fifth Ten-Year Inservice Inspection Interval cc: Regional Administrator w/ attachment USNRC - Region IV Cooper Project Manager w/ attachment USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachment USNRC - CNS NPG Distribution w/o attachment CNS Records w/ attachment

/

NLS20151 12 Attachment Page 1 of 5 Attachment Response to Nuclear Regulatory Commission Requests for Additional Information for 10 CFR 50.55a Requests for Fifth Ten-Year Inservice Inspection Interval Cooper Nuclear Station, Docket No. 50-298, DPR-46 The Nuclear Regulatory Commission (NRC) requests for additional information (RAI) regarding the 10 CFR 50.55a Request for Fifth Ten-Year Inservice Inspection Interval are shown in italics.

The Nebraska Public Power District (NPPD) response to the requests is shown in normal font.

RAI-]

In the submittal dated June 9, 2015, the licensee proposedto use Code Case N-638-4 with the Conditions that "Demonstrationfor ultrasonicexamination of the repairedvolume is required using representative samples which contain construction type flaws" and that "CNS will comply with 3(e)(1) of the code case."' NRC staff notes that paragraph2.10) of the code case states:

"The average lateralexpansion value of the three HAZ Charpy V-notch specimens shall be no less than the average lateralexpansion value of the three unaffected base metal specimens. However, if the average lateral expansion value of the HAZ Charpy V-notch specimens is less than the average value for the unaffected base metal specimens and the procedure qualification meets all the other requirements of this Case, either of the following shall be performed."

(1) The welding procedure shall be requalified.

(2) An Adjustment Temperaturefor the procedure qualification shall be determined in accordancewith the applicableprovisions of NB-4335.2 of Section III, 2001 Edition with 2002 Addenda. The RTNDT or lowest service temperature of the materialsfor which the weldingprocedure will be used shall be increased by a temperature equivalent to that of the Adjusted Temperature."

Based on the above information, the staffrequests that the licensee provide the following information:

(a) Identify whether aforementioned option (1) or option (2) was used in the temperbead weld qua4lfcationfor Cooper weld overlays.

(b) If the welding procedure specification (WPS) was qualified with option (2), provide the value for the new Adjusted Temperaturefor the vessel component (where temperbeadstructural overlay is to be applied). Identify'if this Adjusted Temperature value was considered in the evaluation of the vessel integrity analyses (e.g., pressure-temperaturecurves-ifapplicable).

NLS20151 12 Attachment Page 2 of 5 NPPD Response Until a vendor is selected to install the overlay and the WPS with their procedure qualification records are reviewed, Cooper Nuclear Station (CNS) will not know how the requirements of 2.1(j) have been satisfied by the chosen vendor. As CNS understands the NRC's approval of Code Case N-638-4 in Regulatory Guide 1.147, Revision 17, the NRC has not placed any conditions on meeting 2.1 (j), including its options. In the event the selected vendor's WPS qualification is based on average lateral expansion values of the three Heat Affected Zone (HAZ)

Charpy V-notch specimens being less than the average lateral expansion value of the three unaffected base metal specimens, CNS will require requalification (option 1). In the unlikely event the WPS cannot be requalified demonstrating average lateral expansion in the weld area equal to or greater than the unaffected area, option 2 would be selected or another vendor chosen.

If option 2 is determined to be the only available solution, the effects of the Adjusted Temperature would be determined before the Full Structural Weld Overlay (FSWOL) is installed. If the new Adjusted Temperature is determined to affect the pressure-temperature curves, the curves would be revised before plant startup from refuel outage 29. However, because the location of the FSWOL is outside of the beltline region [fluence values greater than 1 x 1017 n/cm2 (E>I1 MeV)], it is not expected that minor changes to the Adjusted Temperature would affect the pressure-temperature curves.

RAI-2

In the submittal dated June 9, 2015, the licensee proposed to use Code Case N-638-4 with the Conditions that "Demonstrationfor ultrasonicexamination of the repairedvolume is required using representative samples which contain construction type flaws" and that "CNS will comply with 3(e)(1) of the code case." NRC staffnotes that paragraph2.1(c) of the code case states:

"Considerationshall be given to the effects of irradiationon the properties of material, includingweld materialfor applications in the core belt line region of the reactor vessel.

Special material requirementsin the Design Specification shall also apply to the test assembly materialsfor these applications."

NRC staff notes that Title 10 of the Code of FederalRegulations (10 CFR) Part50, Appendix G,Section II. G defines the beltline region of reactorvessel as:

"the region of the reactor vessel (shell material including welds, heat affected zones, and plates orforgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that arepredicted to experience sufficient neutron radiation damage to be consideredin the selection of the most limiting material with regardto radiationdamage."

NLS20151 12 Attachment Page 3 of 5 NRC staff notes that Regulatory Issue Summary (RIS) 2014-11 defines the beltline asfollows:

"the be/tline definition in 10 CFR Part50, Appendix G is applicable to all reactorvessel ferritic materials with projected neutronfluence values greater than 1 x 10I7 n/cm 2 (E>1 MeV), and this fluence threshold remains applicablefor the design life as well as throughout the licensed operatingperiod."

Based on the above information, the staff requests that the licensee provide the following information:

(a) Identfify whether or not the subject weldjoint is considered to be in the beltline region per the definitions of 10 CFR Part50 Appendix G and RIS 2 014-11.

(b) If the subject weld joint is consideredto be in the beltline region as defined by 10 CFR Part 50 Appendix G and RIS 2 014-11, identify what considerationwas given to the effect of irradiationon the properties of material, as required by paragraph2.1l(c) of Code Case N-63 8-4.

NPPD Response (a) The weld subject to the FSWOL is the Control Rod Drive Return Line Nozzle-to-Cap weld and is located outside the beitline region of the reactor vessel as defined by 10 CFR 50 Appendix G and RIS 2014-11. The nozzle is at elevation 452 inches above vessel 0 and the beitline region is located between elevations 195.35 inches and 372.5 inches assumed at 54 Effective Full Power Years.

(b) N/A

RAI-3

In the submittal dated June 9, 2015, the licensee states that the current configuration of the subject weld joint is an "A-508, Class 2 low alloy steel nozzle "joined to an "SB -166, Alloy 600 nickel alloy cap" with "Alloy 182/82 materials". The licensee proposed to perform the FSWOL welding using "ERNiCrFe-7A (Alloy 52M) filler metal." Later in the submittal, the licensee states that the "overlay will completely cover the area of the flaw and other Alloy 182 or susceptible austenitic stainlesssteel material with the highly resistantAlloy 52M weldfiller material."

Based on the above information, the staff requests that the licensee provide the following clarification:

(a) Identify whether alloy 82 or 182 is used in the final layer of the existing weld joint to which the FSWOL is proposed to be applied.

NLS20151 12 Attachment Page 4 of 5 (b) Clarify whether or not austeniticstainless steel materialis used in the existing weld joint to which the FSWOL is proposed to be applied. If austeniticstainless steel materialis used in the exiting weld joint, provide a detailed weld sketch identifying all the following materials:

A-508 Class 2, Stainless Steel (specify type), SB -1 66 Alloy 600, Alloy 82 filler metal, and Alloy 182 filler metal.

NPPD Response (a) When the cap was installed, the existing 182 open butt weld was cut and machined back to approximately 0.350" remaining at the bevel. CNS installed the cap using a full open butt 82 (ER-NI-CR-3) weld. Stainless steel cladding originally covered the 182 weld but was removed to facilitate fit up of the SB-166 Alloy 600 cap exposing a small portion of the 182 at the inside diameter. See Figure 1.

(b) As described on page 57 of the Attachment to NLS2015025, the configuration to receive the FSWOL consists of an SA-508, Class 2 low alloy steel nozzle, Alloy 182/82 weld materials, and an SB- 166, Alloy 600 cap. There are no stainless steel weld filler materials contained in the weld joint. See Figure 1.

Figure 1 Existina Ng Weld Can)

,-m ............

Confiouration

  • ....... l'lm........

SA 508 CI2 Nozzle

NLS20151 12 Attachment Page 5 of 5

RAI-4

Provide a sketch or diagram of each subject weld that is applied with a FSWOL demonstrating that the ultrasonic examination of the overlaidweld will achieve 100 percent coverage of the required volume per the ISI alternative. The sketch/diagram should include lines that represent the ultrasonic beam angles and signalpath that cover the requiredvolume.

NPPD Response The FSWOL will be designed in accordance with the requirements of Nonmandatory Appendix Q, Article Q-3000 and will be examined in accordance with Article Q-4000. The examination volume specified by Q-4100-1 will be ultrasonically examined to assure adequate fusion with the base metal and to detect welding flaws such as interbead lack of fusion, inclusions, or cracks.

For preservice and inservice examination, the volume specified by Q-4300-1 will be ultrasonically examined to meet the requirements of Q-4300. The design of the FSWOL will ensure that an adequate configuration is provided to facilitate examination of the required volumes. The requested diagram/sketch without specific dimensions would not provide meaningful information. While there may be conceptual designs developed in advance of refuel outage 29, the specific design would not be finalized until the actual flaws are characterized.

Request No. RR5-01 does not propose an alternative to the required examination volumes specified by ASME Section XI, Nonmandatory Appendix Q. If the required examination volume cannot be obtained, a revision to RR5-01 would be submitted for NRC approval before restart from the refueling outage.

N Nebraska Public Power District Always there when you need us NLS2015112 October 29, 2015 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Response to Nuclear Regulatory Commission Requests for Additional Information for Relief Request No. RR5-01, Alternate Weld Overlay Repair for a Dissimilar Metal Weld Joining Nozzle to Control Rod Drive End Cap Cooper Nuclear Station, Docket No. 50-298, DPR-46

References:

1. Email from Lois M. James, Nuclear Regulatory Commission, to Jim Shaw, Nebraska Public Power District, dated September 15, 2015, "Request for Additional Information - Relief Request for RR5-01, Cooper - TAC No.

MF63 32"

2. Letter from Oscar A. Limpias, Nebraska Public Power District, to the U.S.

Nuclear Regulatory Commission, dated June 9, 2015, "10 CFR 50.55a Requests for Fifth Ten-Year Inservice Inspection Interval" (ML15167A066)

Dear Sir or Madam:

The purpose of this letter is for the Nebraska Public Power District to respond to the Nuclear Regulatory Commission's Requests for Additional Information (RAI) (Reference 1) related to the Cooper Nuclear Station "10 CFR 50.55a Requests for Fifth Ten-Year Inservice Inspection Interval" (Reference 2).

The response to the specific RAI questions is provided in the attachment to this letter.

This letter does not contain any new regulatory commitments.

If you have any questions concerning this matter, please contact Jim Shaw, Licensing Manager, at (402) 825-2788.

Sincerely Vice President - Nuclear and Chief Nuclear Officer

/dv COOPER NUCLEAR STATION 0(7 Telephone: (402) 825-3811 / Fax:. (402) 825-5211 ww.nppd.com

NLS20151 12 Page 2 of 2

Attachment:

Response to Nuclear Regulatory Commission Requests for Additional Information for 10 CFR 50.55a Requests for Fifth Ten-Year Inservice Inspection Interval cc: Regional Administrator w/ attachment USNRC - Region IV Cooper Project Manager w/ attachment USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ attachment USNRC - CNS NPG Distribution w/o attachment CNS Records w/ attachment

/

NLS20151 12 Attachment Page 1 of 5 Attachment Response to Nuclear Regulatory Commission Requests for Additional Information for 10 CFR 50.55a Requests for Fifth Ten-Year Inservice Inspection Interval Cooper Nuclear Station, Docket No. 50-298, DPR-46 The Nuclear Regulatory Commission (NRC) requests for additional information (RAI) regarding the 10 CFR 50.55a Request for Fifth Ten-Year Inservice Inspection Interval are shown in italics.

The Nebraska Public Power District (NPPD) response to the requests is shown in normal font.

RAI-]

In the submittal dated June 9, 2015, the licensee proposedto use Code Case N-638-4 with the Conditions that "Demonstrationfor ultrasonicexamination of the repairedvolume is required using representative samples which contain construction type flaws" and that "CNS will comply with 3(e)(1) of the code case."' NRC staff notes that paragraph2.10) of the code case states:

"The average lateralexpansion value of the three HAZ Charpy V-notch specimens shall be no less than the average lateralexpansion value of the three unaffected base metal specimens. However, if the average lateral expansion value of the HAZ Charpy V-notch specimens is less than the average value for the unaffected base metal specimens and the procedure qualification meets all the other requirements of this Case, either of the following shall be performed."

(1) The welding procedure shall be requalified.

(2) An Adjustment Temperaturefor the procedure qualification shall be determined in accordancewith the applicableprovisions of NB-4335.2 of Section III, 2001 Edition with 2002 Addenda. The RTNDT or lowest service temperature of the materialsfor which the weldingprocedure will be used shall be increased by a temperature equivalent to that of the Adjusted Temperature."

Based on the above information, the staffrequests that the licensee provide the following information:

(a) Identify whether aforementioned option (1) or option (2) was used in the temperbead weld qua4lfcationfor Cooper weld overlays.

(b) If the welding procedure specification (WPS) was qualified with option (2), provide the value for the new Adjusted Temperaturefor the vessel component (where temperbeadstructural overlay is to be applied). Identify'if this Adjusted Temperature value was considered in the evaluation of the vessel integrity analyses (e.g., pressure-temperaturecurves-ifapplicable).

NLS20151 12 Attachment Page 2 of 5 NPPD Response Until a vendor is selected to install the overlay and the WPS with their procedure qualification records are reviewed, Cooper Nuclear Station (CNS) will not know how the requirements of 2.1(j) have been satisfied by the chosen vendor. As CNS understands the NRC's approval of Code Case N-638-4 in Regulatory Guide 1.147, Revision 17, the NRC has not placed any conditions on meeting 2.1 (j), including its options. In the event the selected vendor's WPS qualification is based on average lateral expansion values of the three Heat Affected Zone (HAZ)

Charpy V-notch specimens being less than the average lateral expansion value of the three unaffected base metal specimens, CNS will require requalification (option 1). In the unlikely event the WPS cannot be requalified demonstrating average lateral expansion in the weld area equal to or greater than the unaffected area, option 2 would be selected or another vendor chosen.

If option 2 is determined to be the only available solution, the effects of the Adjusted Temperature would be determined before the Full Structural Weld Overlay (FSWOL) is installed. If the new Adjusted Temperature is determined to affect the pressure-temperature curves, the curves would be revised before plant startup from refuel outage 29. However, because the location of the FSWOL is outside of the beltline region [fluence values greater than 1 x 1017 n/cm2 (E>I1 MeV)], it is not expected that minor changes to the Adjusted Temperature would affect the pressure-temperature curves.

RAI-2

In the submittal dated June 9, 2015, the licensee proposed to use Code Case N-638-4 with the Conditions that "Demonstrationfor ultrasonicexamination of the repairedvolume is required using representative samples which contain construction type flaws" and that "CNS will comply with 3(e)(1) of the code case." NRC staffnotes that paragraph2.1(c) of the code case states:

"Considerationshall be given to the effects of irradiationon the properties of material, includingweld materialfor applications in the core belt line region of the reactor vessel.

Special material requirementsin the Design Specification shall also apply to the test assembly materialsfor these applications."

NRC staff notes that Title 10 of the Code of FederalRegulations (10 CFR) Part50, Appendix G,Section II. G defines the beltline region of reactorvessel as:

"the region of the reactor vessel (shell material including welds, heat affected zones, and plates orforgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that arepredicted to experience sufficient neutron radiation damage to be consideredin the selection of the most limiting material with regardto radiationdamage."

NLS20151 12 Attachment Page 3 of 5 NRC staff notes that Regulatory Issue Summary (RIS) 2014-11 defines the beltline asfollows:

"the be/tline definition in 10 CFR Part50, Appendix G is applicable to all reactorvessel ferritic materials with projected neutronfluence values greater than 1 x 10I7 n/cm 2 (E>1 MeV), and this fluence threshold remains applicablefor the design life as well as throughout the licensed operatingperiod."

Based on the above information, the staff requests that the licensee provide the following information:

(a) Identfify whether or not the subject weldjoint is considered to be in the beltline region per the definitions of 10 CFR Part50 Appendix G and RIS 2 014-11.

(b) If the subject weld joint is consideredto be in the beltline region as defined by 10 CFR Part 50 Appendix G and RIS 2 014-11, identify what considerationwas given to the effect of irradiationon the properties of material, as required by paragraph2.1l(c) of Code Case N-63 8-4.

NPPD Response (a) The weld subject to the FSWOL is the Control Rod Drive Return Line Nozzle-to-Cap weld and is located outside the beitline region of the reactor vessel as defined by 10 CFR 50 Appendix G and RIS 2014-11. The nozzle is at elevation 452 inches above vessel 0 and the beitline region is located between elevations 195.35 inches and 372.5 inches assumed at 54 Effective Full Power Years.

(b) N/A

RAI-3

In the submittal dated June 9, 2015, the licensee states that the current configuration of the subject weld joint is an "A-508, Class 2 low alloy steel nozzle "joined to an "SB -166, Alloy 600 nickel alloy cap" with "Alloy 182/82 materials". The licensee proposed to perform the FSWOL welding using "ERNiCrFe-7A (Alloy 52M) filler metal." Later in the submittal, the licensee states that the "overlay will completely cover the area of the flaw and other Alloy 182 or susceptible austenitic stainlesssteel material with the highly resistantAlloy 52M weldfiller material."

Based on the above information, the staff requests that the licensee provide the following clarification:

(a) Identify whether alloy 82 or 182 is used in the final layer of the existing weld joint to which the FSWOL is proposed to be applied.

NLS20151 12 Attachment Page 4 of 5 (b) Clarify whether or not austeniticstainless steel materialis used in the existing weld joint to which the FSWOL is proposed to be applied. If austeniticstainless steel materialis used in the exiting weld joint, provide a detailed weld sketch identifying all the following materials:

A-508 Class 2, Stainless Steel (specify type), SB -1 66 Alloy 600, Alloy 82 filler metal, and Alloy 182 filler metal.

NPPD Response (a) When the cap was installed, the existing 182 open butt weld was cut and machined back to approximately 0.350" remaining at the bevel. CNS installed the cap using a full open butt 82 (ER-NI-CR-3) weld. Stainless steel cladding originally covered the 182 weld but was removed to facilitate fit up of the SB-166 Alloy 600 cap exposing a small portion of the 182 at the inside diameter. See Figure 1.

(b) As described on page 57 of the Attachment to NLS2015025, the configuration to receive the FSWOL consists of an SA-508, Class 2 low alloy steel nozzle, Alloy 182/82 weld materials, and an SB- 166, Alloy 600 cap. There are no stainless steel weld filler materials contained in the weld joint. See Figure 1.

Figure 1 Existina Ng Weld Can)

,-m ............

Confiouration

  • ....... l'lm........

SA 508 CI2 Nozzle

NLS20151 12 Attachment Page 5 of 5

RAI-4

Provide a sketch or diagram of each subject weld that is applied with a FSWOL demonstrating that the ultrasonic examination of the overlaidweld will achieve 100 percent coverage of the required volume per the ISI alternative. The sketch/diagram should include lines that represent the ultrasonic beam angles and signalpath that cover the requiredvolume.

NPPD Response The FSWOL will be designed in accordance with the requirements of Nonmandatory Appendix Q, Article Q-3000 and will be examined in accordance with Article Q-4000. The examination volume specified by Q-4100-1 will be ultrasonically examined to assure adequate fusion with the base metal and to detect welding flaws such as interbead lack of fusion, inclusions, or cracks.

For preservice and inservice examination, the volume specified by Q-4300-1 will be ultrasonically examined to meet the requirements of Q-4300. The design of the FSWOL will ensure that an adequate configuration is provided to facilitate examination of the required volumes. The requested diagram/sketch without specific dimensions would not provide meaningful information. While there may be conceptual designs developed in advance of refuel outage 29, the specific design would not be finalized until the actual flaws are characterized.

Request No. RR5-01 does not propose an alternative to the required examination volumes specified by ASME Section XI, Nonmandatory Appendix Q. If the required examination volume cannot be obtained, a revision to RR5-01 would be submitted for NRC approval before restart from the refueling outage.

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