IR 05000316/2016010

NRC Post-Approval Site Inspection for License Renewal Inspection Report 05000316/2016010

ML16319A115
Person / Time
Site:	Cook
Issue date:	11/10/2016
From:	Jeffers M NRC/RGN-III
To:	Gebbie J Nuclear Generation Group
References
IR 2016010
Download: ML16319A115 (14)

Text

UNITED STATES ber 10, 2016

SUBJECT:

DONALD C. COOK NUCLEAR POWER PLANT, UNIT 2 NRC POST-APPROVAL SITE INSPECTION FOR LICENSE RENEWAL INSPECTION REPORT 05000316/2016010

Dear Mr. Gebbie:

On October 14, 2016, the U. S. Nuclear Regulatory Commission (NRC) completed the outage segment of the Post-Approval Site Inspection for License Renewal at your Donald C. Cook Nuclear Power Plant, Unit 2. The enclosed report documents the inspection activities, which were discussed on October 14, 2016, with you and other members of your staff.

This inspection was an examination of activities conducted under your renewed license as they relate to the completion of commitments made during the renewed license application process and compliance with the Commissions rules and regulations and the conditions of your operating license. Within these areas, the inspection involved examination of selected procedures and representative records, observations of activities, and interviews with personnel. On the basis of the sample selected for review, there were no findings of significance identified during this inspection. The NRC staff did not identify any instances of incomplete commitments with respect to timeliness or adequacy.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRCs Public Document Room or from the Publicly Available Records (PARS)

J Gebbie -2-component of the NRC's Agencywide Documents Access and Management System (ADAMS).

ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/

Mark Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Docket No. 50-316 License No. DPR-74

Enclosure:

Inspection Report 05000316/2016010

REGION III==

Docket Nos: 05000316 License Nos: DPR-74 Report No: 05000316/2016010 Licensee: Indiana Michigan Power Company Facility: D. C. Cook Nuclear Power Plant, Unit 2 Location: Bridgman, MI Dates: October 11 - 14, 2016 Inspectors: B. Jose, Senior Reactor Engineer (Lead)

G. ODwyer, Reactor Engineer Approved by: Mark Jeffers, Chief Engineering Branch 2 Division of Reactor Safety Enclosure

SUMMARY

Inspection Report 05000316/2016010; 10/11/16 - 10/14/2016; Donald C. Cook Nuclear Power

Plant, Unit 2; Post-Approval Site Inspection for License Renewal.

The inspection was conducted by two regional based inspectors. No instances were noted of incomplete license renewal commitments with respect to timeliness or adequacy. No findings were identified by the inspectors. The significance of inspection findings are indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) and determined using Inspection Manual Chapter 0609, Significance Determination Process, dated April 29, 2015.

Cross-cutting aspects are determined using Inspection Manual Chapter 0310, Components Within the Cross-Cutting Areas, dated December 4, 2014. All violations of U.S. Nuclear Regulatory Commission (NRC) requirements are dispositioned in accordance with the NRCs Enforcement Policy dated August 1, 2016. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 6, dated July 201

NRC-Identified

and Self-Revealed Findings No findings were identified.

Licensee-Identified Violations

No violations were identified.

REPORT DETAILS

Summary of Plant Status

Donald C. Cook Nuclear Power Plant, Unit 2, was in a refueling outage during the period of this inspection.

4OA5 Other Activities

.1 Post-Approval Site Inspection for License Renewal (Phase I) - Inspection

Procedure 71003

a. Inspection Scope

(1) Review of Newly Identified Systems, Structures and Components (SSCs)

The inspectors discussed the identification of new structures, systems, and components (SSCs), under the purview of Title 10 of the Code of Federal Regulations (CFR),

Part 54.37(b), with the licensees license renewal staff. No new SSCs were reviewed during Phase I of Inspection Procedure (IP) 71003 License Renewal Program Inspection.

Any new SSCs that are identified will be discussed and reviewed during the Phase II portion of IP 71003.

(2) Review of Revised Commitments As part of reviewing the Aging Management Programs (AMPs) associated with the commitments, the inspectors identified the licensee performed a number of commitment revisions related to license renewal. These commitment changes and any others that may be generated will be reviewed during the Phase II portion of IP 71003.

(3) Review of Commitments The inspectors reviewed supporting documents including completed surveillance records, conducted interviews, observed non-destructive examination activities, performed visual inspection of structures and components, including those not accessible during power operation, and observed the activities described below to verity the licensee completed the necessary actions to comply with the license conditions that are a part of the renewed operating license. The inspectors verified the licensee implemented the outage related AMPs included in NUREG-1831, Safety Evaluation Report Related to the License Renewal of the Donald C. Cook Nuclear Plant, Unit 1 and Unit 2, in accordance with 10 CFR Part 54, Requirements for the Renewal of Operating Licenses for Nuclear Power Plants.

b. Results of Detailed Reviews The inspectors reviewed portions of the commitments below, which are listed by AMP number and referenced to Commitment Item number from Appendix A of the Safety Evaluation Report. Activities observed related to these commitments are also listed.

Specific documents reviewed are listed in the Attachment.

(1) (B.1.1) Alloy 600 Aging Management Program (Commitment Items 1, 2, and 3)

The Alloy 600 AMP is a new plant specific program and will be implemented prior to the period of extended operation. This program manages aging effects of Alloy 600/690 components and Alloy 52/152 and 82/182 welds in the reactor coolant system that are not addressed by other AMPs. This program detects primary water stress corrosion cracking prior to the loss of component intended function by using the discussion related to examination and inspection requirements specified in American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI.

Additionally, an inspection plan was submitted for staff review and approval 3 years prior to the period of extended operation (via letter AEP-NRC-2011-39, dated August 17, 2011)to determine if the program demonstrates an ability to manage the effects of aging per 10 CFR 54.21 (a) (3).

Indiana & Michigan Power Company (I&M) will continue to participate in industry initiatives, such as the Westinghouse Owners Group and the Electric Power Research Institute (EPRI) Material Reliability Program (MRP). Susceptibility rankings and program inspection requirements regarding Alloy 82/182 pipe butt welds will be consistent with the later version of the EPRI MRP safety assessment or its successors. This Program was developed utilizing EPRI MRP-126, Material Reliability Program: Generic Guidance for Alloy 600 Management, November 2004, which specifies the objectives and requirements for an Alloy 600 management plan. The Alloy 600 Material Management Program is a living document and will be revised periodically to reflect the latest plant configurations and regulatory requirements.

The inspectors reviewed the procedures and documentation related to the Alloy 600 AMP. The inspectors had no concerns with the reviewed activities.

(2) (B.1.4) Boric Acid Corrosion Program (Commitment Item 4)

The Boric Acid Corrosion Program is an existing program that is consistent with NUREG-1801, Generic Aging Lessons Learned Report, July 2001,Section XI.M10, Boric Acid Corrosion, with an enhancement. The program manages aging effects for structures and components, and supports as a result of borated water leakage. The program requires periodic visual inspection of adjacent structures, components, and supports for evidence of leakage and corrosion. The program includes provisions for:

(a) determination of principal location of leakage,

(b) examination requirements and procedures for locating small leaks, and

(c) engineering evaluations and corrective actions to ensure that boric acid corrosion does not lead to degradation at the leakage source or adjacent structures or components. This program complies with the requirements of U.S. Nuclear Regulatory Commission (NRC) Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in Pressurized Water Reactor Plants.

Commitment Item 4 states, the following enhancements to the Boric Acid Corrosion Prevention Program will be implemented prior to the period of extended operation:

(a) the program scope will be revised to address electrical components in addition to ferrite steel, and

(b) the program acceptance criteria will be revised to address electrical components in addition to ferrite steel.

The inspectors reviewed the procedures and documentation related to the Boric Acid Corrosion Control Program. The inspectors also reviewed a sample of corrective action records from the previous and current refueling outages pertaining to identified boric acid leakage or buildup, and the boric acid corrosion evaluations associated with these identified boric acid leaks/deposits. The inspectors had no concerns with the reviewed activities.

(3) (B.1.6) Buried Pipe Inspection Program (Commitment Items 5, and 42)

The Buried Pipe Inspection Program is consistent with, but includes an exception to, the program described in NUREG-1801, July 2001,Section XI.M34, Buried Pipe Inspection Program. The program manages the loss of material caused by corrosion of the external surface of buried carbon steel piping and tanks. The Buried Pipe Inspection Program includes

(a) preventive measures to mitigate corrosion, and

(b) periodic inspections to manage the effects of corrosion on the pressure retaining capability of buried carbon steel piping and tanks.

Commitment Item 5 states, the buried Piping Inspection Program will be implemented prior to the period of extended operation. Commitment Item 42 states, in response to an Advisory Committee on Reactor Safety License Renewal Subcommittee comment regarding scheduling of buried piping inspections, the licensee commits to enhance the new Buried Piping Inspection Program to require performance of an inspection of a sample of buried piping included in the scope of this program within 10 years after entering the period of extended operation, unless an opportunistic inspection of similar underground piping has occurred within this 10-year period. Before the end of the 10 year of extended operation, I&M [the licensee] will perform an engineering evaluation to determine if sufficient inspections have been conducted to draw a conclusion regarding the ability of the underground coatings to protect the underground piping from degradation. If not, I&M will conduct an inspection of a sample of buried piping to allow that conclusion to be reached.

The inspectors conducted interviews, reviewed program, work, and implementing documents. The inspectors had no concerns with the observed activities.

(4) (B.1.7) Cast Austenitic Stainless Steel Evaluation Program (Commitment Item 6)

The Case Austenitic Stainless Steel Evaluation Program is a new plant specific program and is consistent with the program described in NUREG-1801, July 2001, Section Xl.M12. The program includes a determination of the susceptibility of the cast austenitic stainless steel components to thermal aging embrittlement based on casting method, molybdenum content, and percent ferrite. Prior to the period of extended operation, Donald C. Cook Nuclear Plant will develop aging management program details (for example, plans for additional volumetric inspections or flaw tolerance evaluations) for the Reactor Coolant System piping heats of material that are susceptible to reduction of fracture toughness.

The inspectors conducted interviews, reviewed program, work, and implementing documents. The inspectors had no concerns with the observed activities.

(5) (B.1.11.1) Fire Protection Program (Commitment Item 7)

The Fire Protection Program is an existing program and is consistent with, but includes exceptions to, the program described in NUREG-1801, July 2001, Section Xl.M26, as documented in License Renewal Application (LRA), Section B.1.11

.1. The program will

be enhanced to include the attributes documented in LRA, Section B.1

.11 .1. The

following enhancements will be implemented prior to the period of extended operation:

In the C02 and halon procedures, ensure that conditions that may affect the performance of the system (such as corrosion, mechanical damage, or damage to dampers) are observed and degraded conditions are addressed via the Corrective Action Program.

Enhance procedures to ensure the diesel fuel supply line is monitored for degradation during performance testing.

The inspectors conducted interviews, reviewed program, work, and implementing documents. The inspectors had no concerns with the observed activities.

(6) (B.1.11.2) Fire Water System Program (Commitment Item 8)

The Fire Water System Program is an existing program and is consistent with, but includes exceptions to, the program described in NUREG-1801, July 2001, Section Xl.M27, as documented in License Renewal Application, Section B.1.11

.2. The

program will be enhanced to include the attributes documented in License Renewal Application, Section 8.1.11.2.

The following enhancements will be implemented prior to the period of extended operation:

A sample of sprinkler heads will be inspected using the guidance of National Fire Protection Association 25, Section 2-3.1.1.

The Fire Water System Program will be enhanced to perform non-intrusive measurement of pipe wall thickness per the NRC interim staff guidance (ISG-04, Accession Number ML023440137).

The inspectors conducted interviews, reviewed program, work, and implementing documents. The inspectors had no concerns with the observed activities.

(7) (B.1.13) Heat Exchanger Monitoring Program (Commitment Item 9)

The Heat Exchanger Monitoring Program is a new plant specific program. The program is credited with inspecting the heat exchangers for age related degradation that includes loss of material and cracking of the tubes. The Heat Exchanger Monitoring Program includes provisions to:

(1) develop testing techniques based on industry operating experience;

(2) determine the sample population of heat exchangers based on operating experience before the inspections; and

(3) perform an eddy current inspection of the tubes every 10 years, or more frequently if inspection results indicate a need for more frequent inspections.

Commitment Item 9 states the Heat Exchanger Monitoring Program will be initiated prior to the period of extended operation.

The inspectors observed eddy current testing and visual inspection performed on Unit 2, Component Cooling Water Heat Exchanger 2-HE-15W. In addition, inspectors reviewed heat exchanger test data associated with past visual testing performed on piping and end caps after 2-HE-70 was removed.

The inspectors conducted interviews with plant personnel, and reviewed work orders and procedures associated with the heat exchanger monitoring program. The inspectors had no concerns with the observed activities.

(8) (B.1.18) Inservice Inspection - American Society of Mechanical EngineersSection XI, Augmented Inspections (Commitment Item 10)

The Inservice Inspection - ASME Section XI, Augmented Inspections Program is an existing plant specific program. The program manages the effects of aging on selected components outside the jurisdiction of ASME Section XI. The program description states, Augmented Inspection shall be consistent, to the extent practical, with the appropriate ASME requirements of ASME Section XI (i.e., selection of inspection methods, inspection frequency, percentage of components examined within a population, and acceptance criteria). The program is implemented in accordance with the applicable requirements of the 1989 Edition of ASME Section XI, approved NRC alternatives and relief requests, and other requirements specified in 10 CFR 50.55a for the third Inservice Inspection interval.

Commitment Item 10 states, the following enhancements to the Inservice Inspection (ISI) - ASME Section XI, Augmented Inspections Program will be implemented prior to the period of extended operation:

(a) An augmented ISI volumetric inspection of the spray additive tanks (SATs) and the portions of the containment spray system that are wetted by sodium hydroxide;

(b) An augmented ISI volumetric inspection of the portions of the discharge header in containment that may contain water with concentrated contaminants.

The inspectors reviewed procedures and documentation related to the augmented ISI Program. The inspectors also reviewed Ultrasonic Test Examination results on two welds in a containment spray line. The inspectors had no concerns with the reviewed activities.

(9) (B.1.29) Service Water System Reliability Program (Commitment Items 21 and 41)

The Service Water System Reliability Program is consistent with, but includes exceptions to, the program described in NUREG 1801, July 2001,Section XI.M20, as documented in LRA, Section B.1.29. The program will be enhanced to include the attributes documented in LRA, Section B.1.29.

The following enhancements to the Service Water System Reliability Program will be implemented prior to the period of extended operation:

The Service Water System Reliability Program will be enhanced to check for selective leaching during visual inspections.

Develop new preventative maintenance activity or revise existing PM activity to ensure the 8-inch expansion joints in the essential service water supply lines to the Emergency Diesel Generator (EDG) heat exchangers are inspected for evidence of loss of material, change in material properties and cracking.

Commitment Item 41 was documented as an enhancement to the Service Water Reliability Program described in LRA, Section B.1.29 as:

I&M will enhance the Service Water System Reliability Program to manage loss of material due to selective leaching of susceptible materials by visual inspections and hardness testing or an equivalent physical test Commitment 21 had been revised since issuance of NUREG 1831 by the deletion of the second enhancement to inspect expansion joints because the applicant had determined that the expansion joints were already being replaced on a periodic basis by a Periodic Maintenance Task which removed the joints from the license renewal AMP. The inspectors verified that the change to Commitment 21 (deletion of the enhancement to inspect the expansion joints)had been properly accomplished.

The inspectors verified that some inspections to detect selective leaching were accomplished, e.g., the inspectors verified that Heat Exchangers had been inspected per Heat Exchanger Inspection Procedure 12-EHP-8913-001-002, Revision 11. The inspection procedure required that heat exchangers susceptible to selective leaching were identified and were inspected visually and by a hardness test or an equivalent mechanical test, e.g., scratch test; and results documented and corrective actions taken if necessary.

The inspectors verified that Procedure 12-EHP-8913-001-002 was used appropriately by observing the performance of Work Order 55487554-04 which directed the inspection of 2-HE-15W (Unit 2 Component Cooling Water heat exchanger 15W) per Procedure 12-EHP-8913-001-002 with a selective leaching hardness test. In addition, inspectors reviewed heat exchanger test data associated with past eddy current testing performed on the Unit 2 AB EDG jacket water cooler (2-QT-131-AB), and Unit 2 AB EDG lube oil cooler (2-QT-110-AB).

The inspectors conducted interviews, reviewed procedures, implementing documents, observed visual inspections during the performance of Work Orders, and reviewed additional program inspection results and documents. The inspectors had no concerns with the observed activities. The inspectors verified procedural enhancements were performed to ensure inspections for selective leaching. The inspectors had no concerns with the updated and revised documentation. The inspectors concluded that Commitment Items 21 and 41 had been appropriately implemented and would manage aging effects throughout the period of extended operation.

4OA6 Management Meetings

.1 Exit Meeting Summary

On October 14, 2016, the inspectors presented the inspection results to Mr. J. Gebbie and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary.

ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

J. Gebbie, Senior Vice President and Chief Nuclear Officer

L. Baun, Performance Assurance Director

H. Kish, Regulatory Affairs

T. Gottlieb, License Renewal/Training

R. Kalinowski, License Renewal

S. Mitchell, Regulatory Affairs

LIST OF ITEMS

OPENED, CLOSED AND DISCUSSED

Opened, Closed and

Discussed

None LIST OF ACRONYMS USED ADAMS Agencywide Document Access Management System AMP Aging Management Program ASME American Society of Mechanical Engineers CFR Code of Federal Regulations EDG Emergency Diesel Generator EPRI Electric Power Research Institute I&M Indiana Michigan Power Company IMC Inspection Manual Chapter IP Inspection Procedure LRA License Renewal Application MRP Material Reliability Program NRC U.S. Nuclear Regulatory Commission PARS Publicly Available Records System SSCs Systems, Structures, and Components