AEP-NRC-2009-58, Notification of MRP-139 Deviation

Notification of MRP-139 Deviation

ML092230352
Person / Time
Site:	Cook
Issue date:	07/24/2009
From:	Hruby R Indiana Michigan Power Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
AEP-NRC-2009-58
Download: ML092230352 (14)

Text

Indiana Michigan Power Company Nuclear Generation Group IMDIAMA One Cook Place MICHIGAN Bridgman, Mt 49106 POWER aep.com July 24, 2009 AEP-NRC-2009-58 10 CFR 50.55a Docket No.: 50-315 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Donald C. Cook Nuclear Plant Unit 1 MRP-139 DEVIATION NOTIFICATION In accordance with the Nuclear Energy. Institute (NEI) Guideline for the Management of Materials Issues (NEI 03-08, Revision 1), Indiana Michigan Power Company is providing notification of a deviation from the Electric Power Research Institute (EPRI) Materials Reliability Program (MRP): Primary System Piping Butt Weld Inspection and Evaluation Guidelines (MRP-139). This notification is being sent consistent with industry initiatives to provide timely communications to U. S. Nuclear Regulatory Commission staff regarding conformance with industry guidance. This notification is for information and no response is requested.

This is a schedule deviation request from the implementation requirements of MRP document MRP-139, "Primary System Piping Butt Weld Inspection and Evaluation Guideline." The specific deviation being requested from EPRI is the deferral of ultrasonic (UT) inspections of Donald C. Cook Nuclear Plant (CNP) Unit 1 Reactor Vessel Hot Leg Nozzle to Safe End welds until the U1C23 refueling outage; now scheduled for March of 2010. The schedule deferral (for a March 2010 refueling outage) due to a turbine-related forced outage represents an approximate three-month deviation for the reactor vessel hot legs and no required deviation for the reactor vessel cold legs.

The technical justification for the acceptability of the deviation is that Unit 1 would have had 23.9 Effective Full Power Years (EFPY) of operation through October 2009, the originally scheduled refueling outage. Due to turbine repair work, the EFPY is now projected to be 22.9 assuming a March 2010 refueling. Unit 1 has been operating at reduced temperature and pressure since June 1989 as part of the program to preserve the steam generators. This reduced operating temperature would slow any potential crack growth.

Concurrence on the technical justification for the deviation by an independent third party materials expert has been obtained in accordance with NEI 03-08 requirements. The deviation has been processed through the CNP corrective action program and is documented in plant records.

Awlf

U. S. Nuclear Regulatory Commission AEP-NRC-2009-58 Page 2 There are no new or revised commitments in this letter. Should you 'have any questions, please contact Mr. James M. Petro, Jr., Regulatory Affairs Manager, at (269) 466-2489.

Sincerely, Raymond A. Hruby, Jr.

Vice President - Site Support Services RSP/rdw

Attachment:

Letter from Joseph N. Jensen, Indiana Michigan Power Company, to Electric Power Research. Institute (EPRI) Materials Reliability Program, "Technical Justification for Deviation from EPRI MRP-139 Inspection Requirements for Reactor Vessel Alloy 600/82/182 Welds at DC Cook Nuclear Plant."

U. S. Nuclear Regulatory Commission AEP-NRC-2009-58 Page 3 c: R. Aben - DLEG/BCCFS/BD T. A. Beltz - NRC Washington DC K. D. Curry - AEP Ft. Wayne J. T. King - MPSC MDEQ - WHMD/RPS NRC Resident Inspector M. A. Satorius - NRC Region III

Attachment to AEP-NRC-2009-58 Letter from Joseph N. Jensen, Indiana Michigan Power Company, to Electric Power Research Institute (EPRI) Materials Reliability Program, "Technical Justification for Deviation from EPRI MRP-139 Inspection Requirements for Reactor Vessel Alloy 600/82/182 Welds at DC Cook Nuclear Plant."

American Electric Power AMERICAAIO Cook Nuclear Plant One Cook Place ELECTRIC Bridgman, Ml 49106 POWER wvvwv.aep.com Electric Power Research Institute Materials Reliability Program Attn: Christine King, Program Manager 3420 Hiliview Avenue Palo Alto, CA 94304 July 13, 2009

Subject:

Technical Justification for Deviation from EPR1 MRP-139 Inspection Requirements for Reactor Vessel Alloy 600/82/182 Welds at DC Cook Nuclear Plant, Indiana Michigan Power Company (I&M) is submitting this report pursuant to the requirements of NEI 03-08, "Materials Guidelines Implementation Protocol," for a deviation to the requirements of EPRI MRP-139, "Material Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline."

This report provides the engineering evaluation establishing the basis to deviate from the MRP-139 ultrasonic (UT) inspection schedule requirement for the D.C.Cook Plant Alloy 82/182 Reactor Vessel Nozzle welds.

Should you have any questions, please call Carl Lane at (269) 466-2894 at your convenience.

Sincerely, Chief Nuclear Officer and Senior Vice President PD/adg 1 Enclosure NDM Correspondence Control #2009-444

CR Number: 00844056 Technical Justification for Deviation from EPRI MRP-139 Mandatory Schedule for inspection Requirements for Alloy 6001821182 Welds for Cook Unit I Date Prepared By (Preparer) Date 702A Mviewed By (Reviewer) Date 07D tez Approve By (Supe Asor) Date 1 / 4 4?

Engineering Programs Manager Date Date

)ice President- Site Support Seices ni Viresident& Date eear Officer ne ent Reviewer Date Page I of 9

CR Number: 00844056 Technical Justification for Deviation from EPRI MRP-139 Mandatory Schedule for Inspection Requirements for Alloy 6001821182 Welds for Cook Unit I Reason for Evaluation/ Scope:

This document provides the basis for schedule deviation from the implementation requirements of Material Reliability Program (MRP) document MRP-139, "Primary System Piping Butt Weld Inspection and Evaluation Guideline."

(Reference 1) The specific deviation being requested is the deferral of ultrasonic (UT) inspections of Cook Nuclear Plant Unit 1 Reactor Vessel Hot Leg Nozzle to Safe End welds until the U 1C23 refueling outage. Unit 2 Reactor Vessel Hot and Cold Leg Nozzle to Safe End welds are not in this population, because they are fabricated with non-susceptible materials (Stainless Steel) (Reference 16 & 21).

Also no other Reactor Coolant System butt welds are in the population because they are stainless steel (Reference 21). The U1C23 refueling outage has been rescheduled due to the failure of the Unit 1 Main Turbine. Repair scenarios indicate a return to service date of Fall 2009. The UIC23 refueling outage is now scheduled for March of 2010. All 8 legs (Hot and Cold Legs) will be ultrasonically inspected and repaired as necessary from primary water stress corrosion cracking (PWSCC). The schedule deferral (for a March 2010 refueling outage) represents an approximate three month deviation for the reactor vessel hot legs and no required deviation for the reactor vessel cold legs.

Detailed Evaluation MRP-1 39 Section 1.2 states, "By December 31, 2009, Alloy 82/182 butt welds that are greater than 14" NPS and exposed to temperatures equivalent to the hot leg will be volumetrically examined per this guideline. By December 31, 2010, Alloy 82/182 butt welds that are greater than 4" NPS and exposed to temperatures equivalent to the cold leg will be volumetrically examined per this guideline." Section 6.8.2 states "Owners who know that their welds cannot be volumetrically inspected are not required to perform a best effort NDE; however, by the time the examination is due, they are required to have an approved Deviation in place including a plan to address either the susceptibility of the weld or the inspectability of the weld. Actions identified in this plan will be performed at the earliest possible RFO."

Due to the Cook Unit 1 turbine failure related forced outage and subsequent rescheduling U1C23 refueling outage, the mandated requirements of MRP-139 will not be met (Reference 3).

The unit is in cold shutdown and since Primary Water Stress Corrosion Cracking is temperature dependent, cracking is not postulated to grow or initiate (Reference 13). There is no physical impact on the plant as a result of not meeting the mandatory requirements.

Unit 1 would have had 23.9 Effective Full Power Years (EFPY) through October 2009, the originally scheduled refueling outage. The EFPY is now projected to be Page 2 of 9

CR Number: 00844056 Technical Justification for Deviation from EPRi MRP-139 Mandatory Schedule for Inspection Requirements for Alloy 6001821182 Welds for Cook Unit 1 22.9 assuming a March 2010 refueling. Unit 1 has been operating at reduced temperature and pressure since June 1989 as part of the program to preserve the steam generators. Amendment No. 126 to Operating License No. DPR-58 for CNP Unit 1 permits the operation of the plant at reduced primary system temperature and pressure conditions. This reduced operating temperature would slow any potential crack growth (Reference 13). The MRP equation 4-5 is exponential as a function of temperature. As the temperature is lowered the crack growth rate slows exponentially (Reference 13).

Unit 1 has been operating at reduced temperature and pressure since June 1989 as part of the program to preserve the steam generators. The conclusion is based on several MRP documents that show the crack growth slows significantly as the temperature gets below 600'F. CNP normal operating conditions are Cold Leg temperatures at 520°F and Hot Leg temperatures at 590 0 F. These relatively low operating temperatures provide additional margin.

Reactor Vessel Hot and Cold Leg Nozzle to Safe End Welds Or0ginalSchedule (pre-September2008): The Unit 1 examinations will be completed in Ul C23 (Fall 2009). The procurement of services, outage planning, engineering and mockup activities are in progress. The project plan had the examinations conducted first to meet the commitment, followed by the mitigation activities that included baseline UT examination. Several sets of contingency actions are planned to ensure the commitment is satisfied.

Revised Schedule: Unit 1 has been in a forced outage since September 20, 2008. This was caused by a main turbine failure (Reference 3). The reactor has been and is currently operating in Mode 5 with the reactor coolant system depressurized. There are currently no forced outage scenarios that would require CNP to defuel the reactor. Repair scenarios include repairing the turbine rotors which would support a unit restart in the fall of 2009. Based on these developments, the previously scheduled U1C23 outage date has been moved from the fall of 2009 to spring of 2010. Inspection and contingency repair activities (Reference 5) are continuing to be pursued for completion during the U1C23 refueling outage in March 2010.

The Unit 1 Reactor Vessel Hot Leg Nozzle to Safe End welds have also been evaluated for leak before break (LBB) (Reference 17 &18). PWSCC was not.

considered during the licensing of the main loop piping for LBB. These welds have been examined ultrasonically from the ID as part of the 10 year vessel examination using qualified personnel and procedures and no flaws were reported. This examination was conducted in October 1995 (Reference 14). The next inspection will be conducted in U1C23 currently scheduled for March 2010 (Reference 4).

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CR Number: 00844056 Technical Justification for Deviation from EPRI MRP-139 Mandatory Schedule for Inspection Requirements for Alloy 6001821182 Welds for Cook Unit 1 In summary, CNP's aggressive approach to mitigation of this material issue has resulted in an industry leading program. The events surrounding the turbine failure of CNP Unit 1 has caused a delay in the completion of the inspection and contingency repair activities associated with the Unit 1 Reactor Vessel Hot Leg Nozzle to Safe End welds. CNP Unit 1 has been and continues to be operating in Mode 5 with the reactor coolant system depressurized. This operating condition does not provide the temperature requirements to support PWSCC as an active degradation mechanism (Reference 13). Therefore, CNP is pursuing a deviation from the requirement that calls for all Alloy 82/182 butt welds that are greater than 14" NPS and exposed to temperatures equivalent to the hot leg will be volumetrically inspected per this guideline by December 31, 2009.

Justification Basis:

Guidance:

Nuclear Energy Institute (NEI) document NEI 03-08 (Reference 2), "Guideline for the Management of Materials Issues" allows deviations with the appropriate justification and documentation. Addendum E, Section 7, "Deviations" of this document states, "When a utility determines that "Mandatory" or "Needed" work product elements will not be fully implemented or will not be implemented in a manner consistent with their intent, or when a work product will not be implemented in the timeframe specified by the responsible (Industry Materials Issue Program) IP, a technical justification for deviation shall be developed and retained with the owner's program documentation or owner-controlled tracking systems. In addition, deviations from "Mandatory" and "Needed" work product elements will be entered into corrective action programs (CAP). The technical justification shall provide the basis for determining that the proposed deviation meets the same objective and intent, or level of conservatism exhibited by the original work product, and shall clearly state how long the deviation will be in effect. Deviations from "Mandatory" and "Needed" elements shall receive final concurrence by knowledgeable materials expert independent of the utility "justifying the deviation" must be obtained.

As mentioned in MRP-139 Section 6.8.2, owners are required to have a plan in place to address either the susceptibility or the inspectability of the Alloy 82/182 welds. The Inservice Inspection Interval (Reference 20) was extended for three years due to a shutdown for configuration control issues. During this time, the plant was in cold shutdown. The outage has been moved from the fall of 2009 to spring of 2010 due to an unrelated turbine forced outage (Reference 3).

Industry Safety Assessment:

The objective of the MRP project was to evaluate the viability of detection of leakage from a through wall flaw in an operating plant to preclude the potential Page 4 of 9

CR Number: 00844056 Technical Justification for Deviation from EPRI MRP-139 Mandatory Schedule for Inspection Requirements for Alloy 6001821182 Welds for Cook Unit I for rupture. CNP's original plans included operating-Unit 1 for the entire fuel cycle (18 months). The current plan will result in Unit 1 operating for some period less than a full operating cycle (approximately 11 months) (Reference 19).

In addition, while Unit 1 is in a forced outage, due to the turbine failure, the unit is in cold shutdown with the RCS depressurized. Thus, removing the temperature requirements that PWSCC depends on for cracking to either grow or initiate (Reference 13).

During the Salem Unit 1 Fall 2008 refueling outage (1R19), a significant flaw was identified in the 14 hot leg reactor pressure vessel (RPV) nozzle-to-safe end Alloy 600, Inconel 182 dissimilar metal (DM) weld (Reference 9). The circumferentially oriented, ID connected flaw was discovered during the pre-Mechanical Stress Improvement Process (MSIP) Ultrasonic (UT) examination.

The flaw was analyzed and determined to be acceptable for applying MSIP as a form of stress improvement. The flaw was also analyzed in accordance with ASME Section Xl, IWB-3640 and would have allowed for 36 months additional plant operation without repair or mitigation.

H.B. Robinson performed an ID inspection of their three hot leg and three cold leg RV nozzles this week and shared the official results on an EPRI phone call (October 21, 2008). Axial flaws were detected in all three hot legs and one cold leg. None of the flaws were determined to be ID connected (PWSCC initiated).

Eddy current was required to confirm the UT results. The largest flaw was on the "B" cold leg (0.6 to 1.3 "deep and 1 to 1.5" in length). Using their flaw handbook, they calculated they can operate for at least three years without repair. Nine (9) flaws were found on the hot legs and varied in depth from 0.36 to 0.51" and from 0.25 to 0.60" in length. A number of circumferential flaws were detected in the cladding.

They did not experience any issues with obtaining 100% coverage of the area to be inspected as their weld IDs were machined after application of the cladding during fabrication.

The extent of the "B" cold leg flaw would necessitate re-inspection for the next three outage cycles. They plan on pulling up their mitigation plans in lieu of constant re-inspections. They will pursue an ID mitigation since they are limited to three inches of OD clearance.

Details of the H.B. Robinson plant reactor vessel nozzle inspections are:

1. H.B. Robinson is a 3 Loop Westinghouse Unit, Reactor Vessel fabricated by Combustion Engineering. The 10 year ISI was scheduled for 2011, but to meet the inspection requirements of Page 5 of 9

CR Number: 00844056 Technical Justification for Deviation from EPRI MRP-139 Mandatory Schedule for Inspection Requirements for Alloy 6001821182 Welds for Cook Unit 1 MRP-1 39, they decided to perform the RV Nozzle Dissimilar Welds now. These are RV Nozzle to Safe-End welds.

2 Discovered flaws in all 3 hot legs (HL) and in one cold leg (CL).

Thickness of pipe, HL= 2.6 inches and CL = 2.5 inches.

3 Hot Legs, Loop A - 1 flaw; Loop B had 4 flaws; and Loop C had 4 flaws. Sizes vary Length= 0.25 in to 0.51 inch, and depth= 0.36 to 0.6 inches. The Hot leg welds were not buffered.

4 Cold leg Loop B- One flaw, Length 0.6 inch to 1.5 inches, depending on the angle of view. Cold Leg Welds are buttered. Cold leg weld has gone thru several repairs during initial fabrication as well as several post weld heat treatments. Utility considers all flaws are fabrication flaws. Last inspection of these welds was in 2001.

Since then some of the flaws that were observed in 2001 have appeared to have grown in size.

5 Inspection was by both UT and Eddy Current from the inside wall.

UT inspection was unable to tell if these were inside surface connected. Therefore Eddy Current was performed. Inside surface is smooth and no coverage issues were encountered.

6 Several axial and circumferential indications have been identified in the cladding.

All flaws are embedded flaws. However, when measurement uncertainty (0.39 inches) is added to both ends of the flaw, they have to be treated as surface flaws (inside surface).

At the Mihama (Reference 10) plant (Japan) shot-peening work to reduce residual stress in the surface of the steam generator (S/G) reactor coolant inlet and outlet nozzle welds (total four welds) where 600 series Ni base alloy had been used was to be performed as a preventive maintenance task. To check the surface conditions of the nozzle weld portions prior to the work, eddy current test (ECT) was'performed for the S/G nozzle weld portion surfaces and significant indications were identified for 13 locations on the inlet nozzle weld portion of S/G-A. Also, visual checking of these locations having significant signals was performed and a flaw was identified at one location. For the outlet nozzle of S/G-A and the inlet and outlet nozzles of S/G-B, no significant signal was identified.

Further, for the locations found with significant ECT indication, a penetrant test (PT) was performed and as a result significant penetration indicating patterns (maximum length: approx. 17 mm) were identified. In addition an ultrasonic test (UT) was performed and thereby the maximum flaw depth was evaluated at approx. 13 mm (for the location of the maximum length), resulting in the residual Page 6 of 9

CR Number: 00844056 Technical Justification for Deviation from EPRI MRP-139 Mandatory Schedule for Inspection Requirements for Alloy 6001821182 Welds for Cook Unit I thickness of the affected location (approx. 68 mm) evaluated to be below the thickness described (75 mm) in the application for construction permit based on the Electricity Utilities Industry Law (Reference 10). The significance of the Japanese experience is that it shows typical cracking in the axial direction. The axial cracks run along the pipe and are not safety significant. The cracks arrest at the nozzle and safe end wall. The overall pipe integrity is maintained while minor leakage is possible, but well within the plant equipments capability to safely shutdown.

Based on the above, the low operating temperatures, industry operating experience, and the net reduction in the EFPY support the extension of the required examination date by approximately three months.

Leak Detection Capability:

Cook Surveillance Procedure 1-OHP-4030-102-016 "Reactor Coolant System Leak Rate Test" (Reference 8) meets the latest Westinghouse guidance (References 11 & 12) for conducting reactor coolant system (RCS) leak rate testing. The procedure requires that a small increase in the RCS leak rate (0.05 GPM) be reported to plant management for action. This low threshold is designed to alert plant management of a potential PWSCC through wall leak.

Inspection Limitations:

Profiling of the weld and adjoining base metal contours of the Hot Leg Nozzles shows 100% inspectable surface from the interior (Reference 14). These welds were made in a factory (Reference 7). The welding controls in a factory or shop setting generally produce higher quality welds than field welds. These welds have no recorded repairs in the area of interest. This is based on a review of receiving documentation. The materials of construction were documented in Westinghouse Letter Report (Reference 7). No abnormalities in construction of the reactor vessel were reported in the krea of interest.

The ice condenser plant design includes many missile and divider blocks that divide the upper containment from the lower containment volume (Reference 15).

To access the reactor cavity, the missile blocks weighing several tons have to be moved. The current forced outage does not provide the plant configuration for removing the core barrel and performing the examination of the DMW from the interior or ID of the reactor nozzles. The missile block, head lift, fuel removal, and core barrel lift are all high risk activities. The hot and cold leg welds are not configured for an exterior examination. Significant machining and surface preparation would be necessary for an exterior or OD examination (Reference 7).

The nozzle configuration especially the overall thickness would make an OD examination difficult. The exam would have to be single sided which does not meet the greater than 90% coverage requirement. The ability to perform Page 7 of 9

CR Number: 00844056 Technical Justification for Deviation from EPRI MRP-139 Mandatory Schedule for Inspection Requirements for Alloy 600/821182 Welds for Cook Unit 1 machining would also impact minimum reinforcement of the nozzle. The increase to plant risk for these extra activities is not justified for the compliance with the MRP-1 39 examination requirement to inspect the hot legs, when considering the actual EFPY.

Duration of Deviation:

CNP is pursuing a deviation from the requirement that calls for all Alloy 82/182 butt welds that are greater than 14" NPS and exposed to temperatures equivalent to the hot leg to be volumetrically inspected per the MRP-1 39 guideline by December 31, 2009. Cook Unit 1 Turbine repairs are scheduled to be completed by September 2009. Inspection and contingency repair activities are continuing to be pursued for completion during the U1C23 refueling outage in March 2010.

Conclusions/Findings:

Due to Cook Unit 1 operating in cold shutdown with the RCS system depressurized, the temperature dependency of PWSCC to initiate or grow is eliminated. Additionally, the overall length of operational time is reduced in comparison to a normal operational cycle. Therefore, deferral of the MRP-139 ultrasonic (UT) inspections of Cook Nuclear Plant Unit 1 Reactor Vessel Hot and Cold Leg Nozzle to Safe End welds until the U1C23 refueling outage is technically justified. The deviation is to allow the hot legs welds in Unit 1 to be examined in March 2010 vice December 2009.

References:

1. Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guidelines (MRP-139), EPRI, Palo Alto, CA: 2005.

1010087.

2. Issue of Revision 3 to Materials Guideline Implementation Protocol, NEI 03-08 Addendum E.

3. CR 00838732, Perform ROE on Unit 1 turbine failure

4. RFP 10014 Reactor Vessel Ten Year Examination

5. RFP 10015 RPV Hot & Cold Leg Nozzle Mitigation

6. MRP 2007-0040, Spring 07 Lessons Learned for Ultrasonic Examinations of Dissimilar Metal Weld

7. Westinghouse Letter Report LTR-PCAM-07-74, D.C.Cook Units 1&2 RV Nozzle Welds, Task 2 & 3 Deliverables

8. Operating Department Procedure, Reactor Coolant System Leak Rate Test, 1-OH P-4030-102-016, Rev. 19

9. OE27894 - Flaw Detected in Reactor Pressure Vessel Nozzle-to-Safe End Weld Prior to Mechanical Stress Improvement Process Application (Salem)

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CR Number: 00844056 Technical Justification for Deviation from EPRI MRP-139 Mandatory Schedule for Inspection Requirements for Alloy 6001821182 Welds for Cook Unit I

10. EAR TYO 08-005; Flaws Found in Steam Generator A Reactor Coolant Inlet Nozzle Weld (25.September.2007, Mihama 2, KANSAI) 11 .WCAP-16423-NP, Pressurized Water Reactor Owners Group, Standard Process and Methods for Calculating RCS

12. WCAP-16465-NP, Pressurized Water Reactor Owners Group, Standard RCS Leakage Action Levels and Response Guidelines for Pressurized Water Reactors Leak Rate for Pressurized Water Reactors

13. Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004, 1006696.

14.1995 Inservice Examination of Selected Components at the Donald C.

Cook Nuclear Plant, Unit 1, Final Report, SwRl Project 7184, October 1995,

15. Indiana And Michigan Power D. C. Cook Nuclear Plant Updated Final Safety Analysis Report, Revision 22, Chapter 5, Containment System

16. Indiana And Michigan Power D. C. Cook Nuclear Plant Updated Final Safety Analysis Report, Revision 22, Chapter 4, Table 4.2-1, Materials of Construction of The Reactor Coolant System Components

17. Safety Evaluation Report, Amendment 126, June 9, 1989

18. Design Basis Document For The Reactor Coolant System, DB- 12-RCS, Rev.3

19. Design Information Transmittal DIT-S-00705-10, Unit 1 & 2 Burn-up Data, March 26, 2008.

20. ISI Program Basis Document, Third Ten Year Inservice Inspection Interval, D.C. Cook, Units 1& 2

21. WCAP-16198-P, July 2004, Revision 1 PWSCC Susceptibility Assessment of the Alloy 600 and Alloy 82/182 Components in D. C. Cook Units 1 and 2 Extent of Condition:

This applies to the Unit 1 Reactor vessel nozzles only. Unit 2 reactor vessel nozzles are not fabricated with susceptible materials and are not in the scope of MRP-139. The pressurizer nozzles on both units have been mitigated and comply with MRP-1 39.

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