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NRR E-mail Capture - Clinton Power Station, Unit 1, and Nine Mile Point Nuclear Station, Units 1 and 2 - Request for Additional Information Regarding Relief Requests I3R-10, 1ISI-005, and 2ISI-013 (CAC Nos. MF6115-MF6117)
	ML15288A565
Person / Time
Site:	Nine Mile Point, Clinton
Issue date:	10/15/2015
From:	Blake Purnell; Plant Licensing Branch III
To:	Tom Loomis; Exelon Corp
References
	MF6115, MF6116, MF6117
	Download: ML15288A565 (6)
	v • d • e

1 NRR-PMDAPEm Resource From:

Purnell, Blake Sent:

Thursday, October 15, 2015 10:25 AM To:

Loomis, Thomas R:(GenCo-Nuc) (thomas.loomis@exeloncorp.com)

Cc:

david.helker@exeloncorp.com; Tate, Travis

Subject:

Clinton Power Station, Unit 1, and Nine Mile Point Nuclear Station, Units 1 and 2 - Request for Additional Information Regarding Relief Requests I3R-10, 1ISI-005, and 2ISI-013 (CAC Nos. MF6115-MF6117)

Attachments:

MF6115 BWRVIP RAI2 Rev 1.docx Mr. Thomas Loomis:

By application dated April 10, 2015 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML15100A228),as supplemented by letter dated September 29, 2015 (ADAMS Accession No. ML15272A029), Exelon Generation Company, LLC (EGC, the licensee) submitted relief requests (RRs) for its Clinton Power Station (CPS), Unit 1, and Nine Mile Point Nuclear Station (NMP), Units 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 for inservice inspection of reactor vessel internal (RVI) components.

The Nuclear Regulatory Commission (NRC) staff is reviewing the submittal and has determined that the additional information attached is needed to complete its review of the RRs. The staff requests that a response to this request be provided by November 25, 2015. This request was discussed with EGC personnel on October 14, 2015.

If you have any questions please contact me at 301-415-1380.

Sincerely, Blake Purnell, Project Manager Plant Licensing Branch III-2 and Planning and Analysis Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Docket Nos. 50-461, 50-220, and 50-410

Hearing Identifier:

NRR_PMDA Email Number:

2455 Mail Envelope Properties (Blake.Purnell@nrc.gov20151015102500)

Subject:

Clinton Power Station, Unit 1, and Nine Mile Point Nuclear Station, Units 1 and 2

- Request for Additional Information Regarding Relief Requests I3R-10, 1ISI-005, and 2ISI-013 (CAC Nos. MF6115-MF6117)

Sent Date:

10/15/2015 10:25:21 AM Received Date:

10/15/2015 10:25:00 AM From:

Purnell, Blake Created By:

Blake.Purnell@nrc.gov Recipients:

"david.helker@exeloncorp.com" <david.helker@exeloncorp.com>

Tracking Status: None "Tate, Travis" <Travis.Tate@nrc.gov>

Tracking Status: None "Loomis, Thomas R:(GenCo-Nuc) (thomas.loomis@exeloncorp.com)" <thomas.loomis@exeloncorp.com>

Tracking Status: None Post Office:

Files Size Date & Time MESSAGE 1418 10/15/2015 10:25:00 AM MF6115 BWRVIP RAI2 Rev 1.docx 29744 Options Priority:

Standard Return Notification:

No Reply Requested:

No Sensitivity:

Normal Expiration Date:

Recipients Received:

REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST TO USE BWRVIP GUIDELINES CLINTON POWER STATION, UNIT 1 NINE MILE POINT NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-461, 50-220, AND 50-410 By application dated April 10, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15100A228), 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 1 and 2 NMP-RAI-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 staffs 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.

NMP-RAI-2 Welds fabricated with nickel base alloy Inconel 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.

NMP-RAI-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.

NMP-RAI-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 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) CRD 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.

NMP-RAI-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 BWRVIP inspection 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.

NMP-RAI-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.

NMP-RAI-7 Confirm that a plant-specific integrated leakage assessment was performed, as required by BWRVIP-18 (core spray), BWRVIP-41 (jet 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.

NMP-RAI-8 The April 10, 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 exams) of the core plate hold-down bolts at NMP 2 since calendar year 2000.

CPS and NMP CPS/NMP-RAI-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.

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.