NMP1L3051, Submittal of Relief Requests I3R-10, 1ISI-004, and 2ISI-013 Concerning Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements

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

ML15272A029
Person / Time
Site:	Nine Mile Point, Clinton
Issue date:	09/29/2015
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NMP1L30510, RS-15-262, TAC MF6115
Download: ML15272A029 (8)

Text

200 Exelon Way Kennett Square. PA 19348 Exelon Generation www 'xeloncorp.corT 10 CFR 50.55a NMP1L3051 RS-15-262 September 29, 2015 U.S. Nuclear Regulatory Commission ATTN: 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 50-410

Subject:

Submittal of Relief Requests 13R-10, 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-10, 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 13R-10 (TAC No. MF6115)," dated September 2, 2015 In the Reference 1 letter, Exelon Generation Company, LLC (Exelon) submitted for your review relief requests 13R-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.

In the Reference 2 e-mail, the U.S. Nuclear Regulatory Commission requested additional information. Attached is our response. No commitments are contained in this letter.

Relief Requests 13R-10, 1ISl-004, and 21Sl-013 Concerning Use of the BWRVIP Guidelines in Lieu of Specific ASME Code Requirements September 29, 2015 Page2 If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Respectl~, ~ µ-

James Barstow Director - 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 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. 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 (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 below is needed to complete its review.

Clinton Power Station. Unit 1 CPS-RAl-1:

For CPS, the application references the BWR Vessel and Internals Inspection Summaries for Fall 2013 Outages dated August 7, 2014 (ADAMS Accession No. ML14241A014). Based on the NRC staff's review of the report, the following information is needed:

(a) Identify if the following type of welds have been inspected at CPS, Unit 1: (1) furnace sensitized stainless steel vessel attachment welds or (2) vessel attachment welds made with nickel base alloy 182 welding electrodes. If these welds were inspected in the past, provide the number of welds that were identified with cracks and the corrective actions taken.

(b) Welds made with nickel base alloy 182 welding electrodes are more susceptible to intergranular stress corrosion cracking than stainless steel welds. Identify if any of the alloy 182 welds (both ASME Section XI welds and non-ASME Section XI welds) of the reactor vessel internal components have been inspected at CPS, Unit 1. If these welds were inspected in the past, provide the number of welds that were identified with cracks and the corrective actions taken.

Response

(a) Furnace sensitized stainless steel vessel attachment welds and/or vessel attachment welds made with nickel base alloy 182 welding electrodes were inspected as required by ASME Section XI and BWRVIP applicable inspection guidelines. No weld indications were identified.

(b) Vessel attachment welds made with nickel base alloy 182 welding electrodes were inspected as required by ASME Section XI and BWRVIP applicable inspection guidelines. No weld indications were identified. As discussed in the September 2, 2015 clarification call with the U.S. Nuclear Regulatory Commission Staff, this question pertains to vessel attachment welds.

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

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 CPS. 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 CPS, provide the latest measured values for parameters (a) and (b) above.

Describe the availability of HWC/HWC+NMCA during the normal operation of CPS, Unit 1.

Identify when HWC/NMCA or HWC/OLNC was implemented at CPS, Unit 1.

Response

CPS, Unit 1 currently implements 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.

CPS, Unit 1 applies 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. CPS, Unit 1 has measured a total noble metal loading of 1.34 µg/cm 2 and a Platinum loading of 0.02 µg/cm 2 after being exposed to a single OLNC application. 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 monthly average for August 2015 was -498 mV(SHE), with the current Cycle 16 average at -496 mV(SHE). HWC availability for Cycle 16 is currently 99.0%.

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 3 HWC, NMC, and OLNC were implemented at CPS, Unit 1 as follows:

a) NMCA, 4/2002 b) HWC, 7/2002 c) OLNC, 4/2014 CPS-RAl-3:

(a) On page 5 of the August 7, 2014, inspection summary, it states that during the 2011 refueling outage a flaw was discovered in vertical weld V11. Provide the following information related to this flaw: (1) length of the flaw; (2) location of the flaw (i.e., specify if it is in the weld or heat affected zone or near the junction of the horizontal weld/heat affected zone), (3) corrective actions taken with respect to this flaw, and (4) when would this flaw be inspected in the future.

(b) Confirm whether radial ring welds in repaired shrouds were inspected at CPS, Unit 1.

This inspection is recommended in Section 3.4 of the NRG-approved BWRVIP-76-A report.

Response

(a) The length of the core shroud flaw in vertical weld V11 is 0.8 inch. The location of the flaw is between horizontal welds H3 and H4 at azimuth 174 degrees. The flaw is located in the heat affected zone 10.4 inches from top edge of H4 weld and 69.6 inches from the lower edge of weld H3.

This flaw was evaluated per BWRVIP-76, Section 3.3, and found to be acceptable as-is for 1O years of operation. This weld, including the flaw, is scheduled to be inspected in 2021.

(b) Radial ring welds in the CPS, Unit 1 repaired shroud are not inspected. BWRVIP-76 states that no inspections of the ring welds are required if the repair designer can demonstrate that the repair hardware does not rely on the integrity of these welds in order for it to function properly. General Electric, the designer of the core shroud repair, does not require performing any inspection because the repair design does not rely on the integrity of the ring welds.

CPS-RAl-4:

Confirm that a plant-specific 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 CPS, Unit 1, 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

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 4 (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.

Identify if a plant-specific integrated leakage assessment was performed at CPS, Unit 1, for the RVI components associated with the BWRVIP documents listed above.

Response

A plant specific leakage assessment was performed as required by BWRVIP-18 (core spray),

since CPS, Unit 1 has identified cracks in the core spray piping. A leakage assessment was not performed for the jet pump assembly, Low Pressure Coolant Injection (LPCI) coupling, and core shroud vertical welds since CPS, Unit 1 has not identified any through wall cracking in these locations.

The following Tables show:

• The number and length of all cracks detected in past examinations

• The number and length of all cracks evaluated in the leakage assessment

• The calculated leak rate from each crack evaluated in the leakage assessment including postulated leakage through hidden/inaccessible welds Low Pressure Core Spray (LPCS)

Weld Number Reported Flaw Size Projected Flaw Size Projected Leak Rate (Inches) in May 2015 at 48 Months (Inches) at 48 Months (1wm)

A-BP2 2.43 6.1 24.9 Inaccessible or N/A N/A 25.3 Hidden Weld LPCS Total Leak Rate: 50.2 gpm High Pressure Core Spray (HPCS)

Weld Number Reported Flaw Size Projected Flaw Size Projected Leak Rate (Inches) in May 2015 at 48 Months (Inches) at 48 Months (aom)

B-CP2 2.53 6.2 25.3 B-DP2 2.43 6.1 24.9 Inaccessible or N/A N/A 25.3 Hidden Weld HPCS Total Leak Rate: 75.5 gpm

Response to Request for Additional Information Use of the BWRVIP Guidelines in Lieu of Attachment Specific ASME Code Requirements Page 5 For the repaired core shroud, the leakage assessment was addressed in the CPS, Unit 1 core shroud repair design document as follows:

• The horizontal circumferential welds, H1 thru H7, are assumed leak tight based on compressive loads provided by the repair hardware. This is in accordance with BWRVIP-02, concerning the core shroud repair design criteria.

• Four holes were machined in the shroud support plate as anchor points to install four tie rods in the annulus area. Each of the holes in the shroud support plate will have some leakage paths. A toggle assembly is installed through each hole to anchor tie rods installed as part of the repair. A small gap exists between the tie rods and the holes.

The calculated combined leakage thru this gap for all four holes is 100 gpm.

The current CPS, Unit 1 LOCA analysis includes leakage from the core shroud repair. The LOCA analysis assumes a lower ECCS flow than the actual design ECCS flow. The difference between these flows, that is, margin, is used to determine the acceptance criteria for allowed leakage from reactor internal flaws in the ECCS. The current core spray leakage is below the acceptance criteria for the high and low pressure core spray. All leakage is accounted for and/or is bounded by the CPS, Unit 1 LOCA analysis. The CPS, Unit 1 LOCA analysis is essentially the plant specific integrated leakage assessment.

CPS-RAl-5:

In the application, the licensee identified BWRVIP-25, "BWR Core Plate Inspection and Flaw Evaluation Guidelines," as part of its inspection program for the RVI components at CPS, Unit 1.

The August 7, 2014, inspection summary does not include the inspection results of the core plate at CPS, Unit 1.

Identify whether the core plate is part of its ASME Code,Section XI, core support structure. If this is not the case, explain why augmented inspections (non-ASME Code,Section XI) of the core plate, consistent with the inspection guidelines of BWRVIP-25, were not performed.

Provide a summary of previous core plate inspection results, if conducted, so the NRC staff can assess the extent of any aging degradation in the core plate at CPS, Unit 1. The summary should specify the number of welds that were identified with cracks or any other aging degradation and the corrective actions taken by the licensee. Furthermore, identify whether wedges on the core plate have been installed at CPS, Unit 1.

Response

BWRVIP-25, Section 3.1, does not require any inspection of core plate welds or rim bolts.

CPS, Unit 1 installed wedges on the core plate during construction.