RS-15-318, Stations - Submittal of Relief Requests I3R-10, 1ISI-004, and 2ISI-013 Concerning Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements

Stations - Submittal of Relief Requests I3R-10, 1ISI-004, and 2ISI-013 Concerning Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements

ML15328A514
Person / Time
Site:	Nine Mile Point, Clinton
Issue date:	11/24/2015
From:	David Gudger Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-15-318, TAC MF6115, TAC MF6116, TAC MF6117
Download: ML15328A514 (14)

Text

Exelon Generation ~,

200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.55a RS-15-318 November 24, 2015 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461 Nine Mile Point Nuclear Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-63 and NPF-69 NRC Docket Nos. 50-220 and 50-41 O

Subject:

Submittal of Relief Requests 13R-1 O, 1ISl-004, and 21Sl-013 Concerning Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements

References:

1) Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Submittal of Relief Requests 13R-1 O, 1ISl-004, and 21Sl-013 Concerning Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements," dated April 1O, 2015

2) E-mail from B. Purnell (U.S. Nuclear Regulatory Commission) to T. Loomis (Exelon Generation Company, LLC), "Clinton Power Station, Unit 1 -

Request for Additional Information Regarding Relief Request l3R-10 (TAC No. MF6115)," dated September 2, 2015

3) Letter from J. Barstow (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Submittal of Relief Requests 13R-10, 1ISl-004, and 21Sl-013 Concerning Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements," dated September 29, 2015

4) E-mail from 8. Purnell (U.S. Nuclear Regulatory Commission) to T. Loomis (Exelon Generation Company, LLC), "Clinton Power Station, Unit 1, and Nine Mile Point Nuclear Station, Units 1 and 2 - Request for Additional Information Regarding Relief Requests 13R-10, 1ISl-005, and 21Sl-013 (CAC Nos. MF6115-MF6117)," dated October 15, 2015 In the Reference 1 letter, Exelon Generation Company, LLC (Exelon) submitted for your review Relief Requests l3R-10, 1ISl-004, and 21Sl-013 associated with the Clinton Power Station, Unit 1, and the Nine Mile Point Nuclear Station, Units 1 and 2, respectively. These relief requests propose to utilize BWRVIP guidelines in lieu of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code requirements.

Relief Requests 13R-10, 1ISl-004, and 21Sl-013 Concerning Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements November 24, 2015 Page2 In the Reference 2 e-mail, the U.S. Nuclear Regulatory Commission requested additional information. Reference 3 provided our response. In the Reference 4 e-mail, the U.S.

Nuclear Regulatory Commission requested additional information. Attached is our response.

No commitments are contained in this letter.

If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Respectfully,

~)r..J.J~

David T. Gudger Manager - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Response to Request for Additional Information cc: Regional Administrator, Region I, USNRC Regional Administrator, Region Ill, USNRC USNRC Senior Resident Inspector, Nine Mile Point, Clinton Power Station USNRC Project Manager, Nine Mile Point, Clinton Power Station Illinois Emergency Management Agency- Division of Nuclear Safety A. L. Peterson, NYSERDA

Attachment Response to Request for Additional Information

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 1 By application dated April 10, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 151 OOA228), Exelon Generation Company, LLC (the licensee) submitted relief requests (RRs) for its Clinton Power Station (CPS), Unit 1, and Nine Mile Point Nuclear Station, Units 1 and 2 (NMP 1 and 2). The RRs propose to use various Boiling Water Reactor (BWR) Vessel and Internals Project (BWRVIP) guidelines as an alternative to certain requirements of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (ASME Code) for inservice inspection of reactor vessel internal (RVI) and core support structure components.

The Nuclear Regulatory Commission (NRC) staff is reviewing the submittal and has determined that the additional information below is needed to complete its review.

NMP-RAl-1:

The application references the BWR Vessel and Internals Inspection Summaries for Spring 2013 Outages dated April 11, 2014 (ADAMS Accession No. ML14125A303), and Spring 2012 Outages dated June 19, 2013 (ADAMS Accession No. ML13176A003), as they relate to NMP 1 and 2, respectively. Based on the NRC staff's review of the report, additional information is needed.

Identify whether there are any furnace-sensitized stainless steel reactor pressure vessel (RPV) attachment welds at NMP 1 and 2, and identify the previous inspections performed for these weld materials including the inspection results. The response may reference the applicable line items in the above inspection summaries.

Response

Nine Mile Point (NMP), Unit 1:

The following is a listing of the furnace-sensitized stainless steel RPV attachment welds at NMP, Unit 1, and the current inspection results:

• Control Rod Drive (CAD) stub tubes are furnace-sensitized Type 304 stainless steel with Alloy 82/182 attachment welds. See ML14125A303 for the previous inspections. VT-2 (visual) examinations are performed each outage. No new leakage has been identified since 2001. The NMP, Unit 1 stub tube cracking is discussed in Reference 1.

• Steam Dryer support brackets are Type 304 stainless steel and are conservatively considered to be furnace sensitized. The attachment weld is Alloy 82/182. An inspection performed in 2011 identified IGSCC cracking of the lug associated with the Heat Affected Zone (HAZ). No cracking was identified in the attachment welds. The inspection results are summarized in the NMP, Unit 1 90 day lnservice Inspection report (Reference 2).

• The Feedwater Sparger bracket attachment welds (non-beltline region) are conservatively considered furnace sensitized. The inspection history is found in ML14125A303. No cracking was identified in the attachment welds.

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 2

• The Track Guide (Guide Rod) bracket attachment welds are conservatively considered furnace sensitized. The inspection history is found in ML14125A303. No cracking was identified in the attachment welds.

• The upper and lower surveillance specimen holder bracket attachment and weld (filler metal 308) are conservatively considered furnace sensitized. The inspection history is found in ML14125A303. No cracking was identified in the attachment welds.

NMP, Unit 2:

There are no furnace-sensitized RPV attachment welds at NMP, Unit 2.

NMP-RAl-2:

Welds fabricated with nickel base alloy lnconel 182 welding electrodes (alloy 182 welds) are more susceptible to intergranular stress corrosion cracking (IGSCC) than the austenitic stainless steel welds. Provide the following information concerning Alloy 182 welds:

(a) Identify any Alloy 182 welds for the RPV internal and core support structure that are within the scope of the BWRVIP guidelines at NMP 1 and 2 (both ASME Code Section XI welds and non-ASME Code Section XI welds).

(b) Identify the inspections performed for these welds including the inspection results.

The response may reference the applicable line items in the above inspection summaries.

Response

NMP, Unit 1:

The NMP, Unit 1 attachment welds use Alloy 82/182 with the exception of the surveillance brackets which used 308 stainless steel filler metal. The following is a listing of Alloy 82/182 welds with the inspection results:

• Cracking has been identified in the stub tubes. The stub tubes at NMP, Unit 1 are 304 stainless steel attached to the reactor vessel using Alloy 82/182 material. The NMP, Unit 1 stub tube cracking is in the base metal 304 stainless steel materials. No cracking has been detected in the Alloy 82/182 weld metal.

• The H-9 shroud support attachment welds are Alloy 82/182. Inspections performed under the BWRVIP program have documented cracking in the Alloy 182 attachment weld. The NMP, Unit 1 H-9 weld baseline inspections are documented in the Reference 3 Safety Evaluation Report. The last inspection performed in 2011 showed no significant change in the condition compared to the baseline inspection performed in 2001.

• Steam Dryer support bracket attachment welds are Alloy 82/182. See response to NMP-RAl-1.

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 3

• The Feedwater Sparger bracket attachment welds use Alloy 82/182 weld material. See response to NMP-RAl-1.

• The Track Guide (Guide Rod) bracket attachment welds use Alloy 82/182. See response to NMP-RAl-1.

NMP, Unit 2:

• Vessel attachment welds made with nickel based Alloy 82/182 welding electrodes were inspected as required by ASME Section XI and BWRVIP applicable inspection guidelines.

No weld indications were identified. Ultrasonic Testing of the shroud support H-9 weld using UT from the vessel outer diameter did not identify indications.

NMP-RAl-3:

NUREG-0619, Revision 1, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking, dated November 1980 (ADAMS Accession No. ML031600712) describes the NRC staff position regarding BWR feedwater nozzle and related component inspections. Confirm that NUREG-0619, Revision 1, is currently used for the inspection of feedwater sparger tee welds and piping brackets and will continued to be used if the RRs are approved. If NUREG-0619, Revision 1, is not used for inspection of these components, describe how the proposed alternative will ensure that the integrity of these components will be maintained and identify any plant-specific authorization from the NRC to relax these inspection criteria for these components.

Response

NMP, Unit 1:

Visual inspections of the Control Rod Drive return nozzle and Feedwater Spargers are performed in accordance with NUREG-0619, Revision 1 and will continue to be inspected per the lnservice Inspection program.

NMP, Unit 2:

The original NMP, Unit 2 lnservice Inspection program Safety Evaluation Report (Reference 4) accepted the NMP, Unit 2 position that the NUREG-0619, Revision 1 visual inspections of the Feedwater Spargers were not required. NMP, Unit 2 has included inspection of the Feedwater Spargers as part of the ASME B-N-1 VT-3 examinations of all accessible surfaces of the Feedwater Spargers. The VT-3 examinations include coverage of the end brackets and Sparger tee locations. This examination has been performed once each period.

NUREG-0619, Revision 1, will be used for the inspection of Feedwater Sparger tee welds and piping brackets if the relief request is approved.

NMP-RAl-4:

Based on the review of the April 11, 2014, inspection summary, the staff determined that additional information is needed concerning the NMP 1 BWRVIP inspection findings and the

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 4 status of relevant indications for several RVI components. Provide the following information on the BWRVIP inspection findings and the status of relevant indications for the RVI components:

(a) Core Spray Piping

• Provide the status of the 1.5 inch linear indication in the shroud support ring above the pipe support bracket found during March 2007 EVT-1 exam.

• Identify whether any relevant indications were found as a result the EVT-1 exams of the annulus pipe support bracket attached to the shroud during March 2013, April 2011, and April 2009 inspections.

(b) Top Guide

• Briefly summarize the status of the 79 nongeometric indications that were identified during UT exams performed in March 2005 or, as appropriate, state that there was no change. The status of a few of these indications are identified in the inspections summary, but most are not identified.

(c) Standby Liquid Control Nozzle

• State whether any leakage was noted during the April 2011 EVT-2 exam.

(d) Control Rod Drive (CRD) Guide Tubes (Lower Plenum)

• Define "OFS."

• Describe the corrective action that was taken for the loose alignment pin that was identified during installation of the "OFS" in April 2009.

(e) CAD Stub Tubes

• Provide the status of indications found in the stub tube base metal at two core locations during the April 2009 EVT-1 exams.

• Identify if earlier UT exams of stub tube J-weld areas performed in March 2001, March 2003, and March 2005 outages identified any relevant indications.

(f) Vessel ID Brackets (Interior Attachments)

• Provide the status for the indications recorded on three of four steam dryer support brackets during the May 2015 EVT-1 exam, and state whether any change in the indications was identified compared to the April 2011 recordings for these indications.

• Summarize the flaw evaluation results for the indications detected on all four steam dryer lugs during May 2013 UT exams.

Response

(a) Core Spray Piping The 1.5 inch linear indication in the shroud support ring found during the March 2007 exam has not been re-inspected. The indication is circumferential oriented. With the shroud tie rod repair in place, the consequence of a circumferential oriented indication on the support ring is structurally insignificant and re-inspections are not necessary. The 1.5 inch indication is not associated with the HAZ of the PB-104 core spray bracket and does

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 5 not impact the integrity of the bracket. The EVT-1 inspection of the PB-104 location in 2015 did not identify any indications in the core spray attachment bracket examination area. The 2015 inspection of PB-104 did not identify the previously identified 1.5 inch indication in the ring.

No relevant indications were found as a result of the May 2013, April 2011, and April 2009 inspections of the three other annulus pipe support brackets attached to the shroud.

(b) Top Guide A UT re-inspection of the top guide was completed in 2015. The 79 indications were confirmed to be essentially unchanged. The flaw evaluation had concluded the potential crack growth was low.

The 2015 UT inspection identified two additional indications associated with the grid beam notch region:

• One indication was determined to have been present in 2005 and that the 2015 UT signature is essentially the same.

• One indication shows a change from the 2005 baseline. In this case, the 2005 UT signature at the notch was below the threshold for detection. The 2015 UT signature resulted in an indication length (assumed through thickness) of 0.9 inch (this includes measured length plus uncertainty).

(c) Standby Liquid Control (SLC) Nozzle No leakage was identified in the April 2011 exam.

(d) Control Rod Drive (CRD) Guide Tubes (Lower Plenum)

"OFS" refers to Orifice Fuel Support. This component is described in BWRVIP-47-A.

The alignment pin at location 22-23 is monitored to confirm the pin function is retained.

Inspection in 2015 has confirmed the pin remains functional. No corrective action is required.

(e) CRD Stub Tubes The NMP, Unit 1 stub tube indications identified in 2009 are consistent with the stub tube HAZ cracking seen at all other stub tubes first identified in the 1984 baseline inspections.

The locations identified in the 2009 bottom head visual show no leakage. VT-2 exams are performed each outage and no leakage has been identified at these locations.

The J-weld UT performed in 2001, 2003, 2005 did not identify any relevant indications.

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 6 (f) Vessel ID Brackets (Interior Attachments)

The 2015 UT data showed a small change in depth on all four Steam Dryer brackets, within the UT sizing uncertainty of 0.394 inch. The EVT-1 data showed some increase in length; however, the change was small when considering the corresponding reported UT lengths and small relative to the crack growth rates required to be used in the flaw evaluation. The evaluation concluded the flaw sizes remained within the analysis supplied in Reference 2.

NMP, Unit 1 considers the flaws to be within the jurisdiction of the ASME code boundary.

As such, NMP, Unit 1 followed the rules of IWB-3142.4, "Acceptance by Analytical Evaluation," and submitted the flaw evaluations to the NRG in accordance with IWB-3144(b).

NMP-RAl-5:

Identify whether the latest BWRVIP RVI inspections for NMP 2 for calendar year 2014 resulted in any new relevant indications or significant change to previous indications. For any new relevant indications or changes to indications, briefly summarize the results of the evaluation.

If the calendar year 2015 BWRVI P inspections have been performed for NMP 1, identify whether the latest BWRVIP RVI inspections for NMP 1 during calendar year 2015 resulted in any new relevant indications or significant change to previous indications. For any new relevant indications or changes to indications, briefly summarize the results of the evaluation.

Response

NMP, Unit 1:

The Steam Dryer inspection of the tie bar SD-TB-N-1 location had a previously characterized and evaluated crack in the attachment weld and associated hood base metal. In N1 R23 (2015) preemptive stop drill repairs were performed for the cracks in the hood associated with the tie bar. The N1 R23 inspection identified additional horizontal cracking in the tie bar attachment weld that was not previously identified. Review of the previous outage inspection video confirmed the indication was present but not identified as relevant. The 2015 indication shows no change from the 2013 inspection and the indication remains within the flaw acceptance criteria. This location is planned for inspection in 2017. The 2015 inspection of other previously identified Steam Dryer crack locations showed no significant change.

Top Guide indications were discussed in NMP-RAl-4. No other new relevant indications or significant changes to previous indications were identified.

NMP, Unit 2:

The 2015 NMP, Unit 2 BWRVIP reactor vessel inspections, which include the Steam Dryer, identified new indications on the Steam Dryer which required evaluation. The Steam Dryer results are described in Reference 5. No other new relevant indications or significant changes to previous indications were identified.

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 7 NMP-RAl-6:

Hydrogen water chemistry (HWC) and/or HWC plus noble metal chemical addition (NMCA) are methods used to mitigate intergranular stress corrosion cracking. Specify whether HWC or HWC+NMCA are currently implemented at NMP 1 and 2. Provide details on the methods for determining the effectiveness of HWC/NMCA using the latest measured values of the following parameters:

(a) Electro-chemical potential applicable when HWC or HWC+NMCA is implemented, (b) Hydrogen/oxygen molar ratio applicable when HWC+NMCA method is implemented, and (c) Catalyst loading (platinum) applicable when HWC+NMCA is implemented.

Many BWR units have implemented the newly developed on-line noble chemical (OLNC) addition to their reactor vessels. If OLNC has been implemented at NMP 1 and 2, provide the latest measured values for parameters (a) and (b) above.

Describe the availability of HWC/HWC+NMCA during the normal operation of NMP 1 and 2.

Identify when HWC/NMCA or HWC/OLNC was implemented at NMP 1 and 2.

Response

NMP, Units 1 and 2 currently implement Hydrogen Water Chemistry (HWC) and On-Line NobleChem TM (OLNC) to mitigate lntergranular Stress Corrosion Cracking (IGSCC).

HWC and OLNC mitigate IGSCC when the Electrochemical Corrosion Potential (ECP) is reduced to protective levels. When ECP values are reduced below -230 mV(SHE) by the addition of excess hydrogen, the surfaces are considered mitigated against IGSCC. ECP readings that are more negative than the threshold value of -230 mV(SHE) provide additional margin for reduction of crack growth rates and IGSCC mitigation.

NMP, Units 1 and 2 apply the EPRI Boiling Water Reactor Vessels and Internals Assessment (BWRVIA) model to determine the hydrogen injection rate required for molar ratio > 3 at the upper downcomer location throughout the fuel cycle.

For Noble Metal Chemical Application (NMCA) and OLNC, the catalyst loading on system surfaces must be sufficient to reduce the ECP to s -230 mV(SHE) when a molar ratio of hydrogen to oxidants of ~ 2 is established.

For NMP, Unit 1, a measured ECP < -230 mV(SHE) at conditions representative of components of interest when the coolant molar ratio is ~ 2 indicates sufficient catalyst loading. The ECP for the current Cycle 22 average is -425 mV(SHE) (end of September 2015). HWC availability for Cycle 22 is currently 99% (end of September 2015). The average molar ratio for Cycle 22 is 248 (end of September 2015).

For NMP, Unit 2, a measured ECP < -230 mV(SHE) at conditions representative of components of interest when the coolant molar ratio is ~ 2 indicates sufficient catalyst loading. The ECP for the current Cycle 15 average is -495 mV(SHE) (end of September 2015). HWC availability for

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 8 Cycle 15 is currently 99% (end of September 2015). The average molar ratio for Cycle 15 is 105 (end of September 2015).

HWC, NMCA, and OLNC were implemented as follows:

NMP, Unit 1:

HWC, 4/2000 NMCA, 5/2000 OLNC, 12/2006 NMP, Unit 2:

HWC 2/2001 N MCA, 9/2000 OLNC 12/2007 NMP-RAl-7:

Confirm that a plant-specific integrated leakage assessment was performed, as required by BWRVIP-18 (core spray), BWRVIP-41 Uet pump assembly), BWRVIP-42 (low pressure coolant injection system), and BWRVIP-76 (core shroud), for the internals at NMP 1 and 2 which accounts for the leakage from all internals that impact the ability to cool the core and maintain peak clad temperature within allowed limits during postulated loss of coolant accidents. Provide a summary of all internal components included in the leakage assessment along with a summary of the following for each component:

(a) the number and length of all cracks detected in past examinations for the component (b) the number and length of all cracks evaluated in the leakage assessment (c) the calculated leak rate from each crack evaluated in the leakage assessment.

Response

NMP, Unit 1:

The NMP, Unit 1 core shroud tie rod and vertical weld repair established the shroud leakage criteria associated with the shroud vertical and tie rod repairs coupled with through-wall leakage assumptions associated with the core shroud vertical and horizontal weld through-wall cracking.

The leakage criteria was established based on ensuring the minimum subcooling to ensure proper recirculation pump inlet subcooling. The Reference 6 Safety Evaluation Report concluded that the leakage criteria was acceptable for all operating conditions including ECCS performance.

BWRVIP-41 and BWRVIP-42 do not apply to NMP, Unit 1. No BWRVIP-18 core spray piping flaw indications exist that require a core spray leakage assessment.

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 9 NMP, Unit 2:

No cracking associated with BWRVIP-18 (Core Spray), BWRVIP-41 (Jet Pump assembly), BWRVIP-42 (Low Pressure Coolant Injection system), and BWRVIP-76 (Core Shroud), exists that require leakage assessment for impact on the ability to cool the core and maintain the peak clad temperature within allowed limits during postulated loss of coolant accidents.

NMP-RAl-8:

The April 1O, 2015, application references a deviation for NMP 2 from the inspection guidelines of BWRVIP-25. The NRC staff notes that this same deviation has been submitted for a number of other BWRs, based on the lack of practicable UT or EVT-1 examination methods for the core plate hold-down bolts.

Identify whether lateral restraint wedges are installed to prevent lateral displacement of the core plate if there is a loss of preload for the core plate hold-down bolts. If lateral restraint wedges are not installed, provide justification for not performing other inspections (e.g., VT-3 exam) of the core plate hold-down bolts at NMP 2 since calendar year 2000.

Response

No lateral restraint wedges are installed at NMP, Unit 2. NMP, Unit 2 has a Deviation Disposition that establishes the basis for not performing the BWRVIP-25 specified core plate bolt inspections. A summary of the technical justification is provided in Reference 7. A baseline VT-3 sample of the upper portion of the core plate bolts was performed in accordance with vendor recommendations in 1998 (N2R06) with no anomalies noted. During the 2012 refueling outage, access to the below core plate region was provided at selected locations. The inspection scope included verification that the core plate hex nuts remain in place. The VT-3 examination was able to identify that the bolt hex nuts remained in place at the locations where access existed.

CPS/NMP-RAl-9:

The following RAI applies to the RRs for CPS and NMP 1 and 2.

The NRC staff is aware of several GE-Hitachi (GEH) Safety Communications (SCs) that were issued to BWR licensees informing them of potential non-conservatisms in the analysis of loading conditions on the RPV and internals. In June 2013, the BWRVIP informed the NRC about the following four SCs:

• SC 09-01, "Annulus Pressurization Loads Evaluation,"

• SC 09-03, "Shroud Screening Criteria Reports," including Revision 2,

• SC 11-07, "Impact of Inertial Loading and Potential New Load Combination from Recirculation Suction Line Break Acoustic Loads," and

• SC 12-20, "Error in Method of Characteristics Boundary Conditions Affecting Acoustic Loads Analyses."

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 10 The NRC staff is concerned that the issues raised in the SCs concerning the methodologies used for analyzing applied loads may lead to increased loads on the RPV and internals components, which could potentially invalidate the inspection and evaluation criteria of some of the BWRVIP guidelines. In light of this concern, describe how the latest revisions of the above SCs and other related SCs have been or will be addressed for CPS and NMP 1 and 2. Identify whether the loads on the RPV internals have been revised based on these SCs.

Response

Revision 1 of SC 09-03, "Shroud Screening Criteria Reports," is the latest revision of this SC.

Additionally, SC 13-08, "Shroud Support Plate-to-Vessel Evaluation for AC Loads," was issued in December 2014 and SC 14-03, "Acoustic Load Pressure Difference on Access Hole Cover,"

Revision O, was issued in May 2015. SC 12-20 listed in the RAI has been revised to Revision 1 in December of 2014.

The following is a discussion of the impact of the SCs on Clinton Power Station (CPS), Unit 1, NMP, Unit 1, and NMP, Unit 2:

CPS, Unit 1:

• SC 09 An evaluation has been performed to address the Annulus Pressurized (AP) loads. No loads have been revised as a result of this evaluation.

• SC 09-03, Revision 0 and Revision 1 - These SCs are not applicable to CPS, Unit 1. The shroud tie rod repair has been implemented.

• SC 11-07, SC 12-20 including Revision 1, SC 13-08, and SC 14 Resolution of the issues associated with these SCs is being addressed through application of Finite Break Opening Time (FBOT) through BWROG efforts.

NMP, Unit 1:

• SC 09 This SC is not applicable to NMP, Unit 1 because the AP load is not part of the NMP, Unit 1 design or licensing basis.

• SC 09-03, Revision O and Revision 1 - These SCs are not applicable to NMP, Unit 1 because the shroud tie rod and vertical weld repairs have been implemented.

• SC 11 Resolution of this potential issue is addressed through application of FBOT through BWROG efforts.

• SC 12 This SC is not applicable to NMP, Unit 1. The Method of Characteristics (MOC) code was not used for NMP, Unit 1.

• SC 13-08 -This SC is not applicable to NMP, Unit 1 because TRACG computer code loads were used to define shroud support loads.

• SC 14 This SC is not applicable to NMP, Unit 1. NMP, Unit 1 does not have an access hole cover.

NMP, Unit 2:

• SC 09 NMP, Unit 2 corrected the AP loads per this SC for the Extended Power Uprate (EPU) AP load definition.

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 11

• SC 09-03, Revision O and Revision 1 - The correct AC loads were applied and the normal upset condition for the core shroud remains the limiting condition. Therefore, the SC has no impact.

• SC 11 Resolution of this potential issue is addressed through application of FBOT through BWROG efforts.

• SC 12 NMP, Unit 2 addressed SC 12-20 during the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) project review. A corrected MOC AC load was derived for NMP, Unit2.

• SC 13 The MELLLA+ project applied the SC 12-20 corrected load to the core shroud support plate to vessel weld.

• SC 14 The MELLLA+ project correctly applied the AC load to the access hole cover.

References:

1) Letter from N. Salgado (U.S. Nuclear Regulatory Commission) to S. Belcher (Nine Mile Point Nuclear Station, LLC), " Nine Mile Point Nuclear Station, Unit No. 1, Request to Utilize the Alternative of Applying ASME Code Case N-730 for the Repair and lnservice Inspection of Control Rod Drive Bottom Head Penetrations for the License Renewal Period of Extended Operations (TAC No. MD9604)," dated August 3, 2009 (ML091980454)

2) Letter from P. Orphanos (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Fourth lnservice Inspection Interval, Second lnservice Inspection Period 2015 Owner's Activity Report for RF0-23 lnservice Examinations," dated July 9, 2015 (ML15191A385)

3) Letter from P. Tam (U.S. Nuclear Regulatory Commission) to J. Conway (Nine Mile Point Nuclear Station, LLC), "Nine Mile Point Nuclear Station, Unit No. 1 - Inspection of Core Shroud Support Weld H9 {TAC NO. MB6893)," dated March 20, 2003 (ML030790512)

4) Letter from R. Capra (U.S. Nuclear Regulatory Commission) to L. Burkhardt (Niagara Mohawk Power Corporation), "Safety Evaluation of the First Ten-Year Interval lnservice Inspection Program Plan, Nine Mile Point Unit 2 {TAC Nos. 66071 and 75152)," dated November 1, 1990

5) Letter from A. Sterio (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information, Nine Mile Point Nuclear Station, Unit 2 - Request for Additional Information Regarding Post Extended Power Uprate Steam Dryer Inspection Results," dated October 8, 2015

6) Letter from S. Bajwa (U.S. Nuclear Regulatory Commission) to J. Mueller (Niagara Mohawk Power Corporation), "Supplemental Safety Evaluation Regarding Alternative Repair of the Core Shroud Vertical Welds, Nine Mile Point Nuclear Station, Unit No. 1 (TAC NO. MA4701)," date May 24, 1999

7) Letter from J. Pacher (Constellation Energy Nuclear Group) to U.S. Nuclear Regulatory Commission, "Deviation from BWRVIP-25 Inspection Requirements," dated March 30, 2011 (ML110960415)