ML14253A317

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Enclosure 2: I&M CAP Document AR 2010-1804-10, Root Cause Evaluation Attachment, Rx Vessel Core Support Lug Bolting Anomalies
ML14253A317
Person / Time
Site: Cook  
Issue date: 09/04/2014
From:
Indiana Michigan Power Co
To:
Office of Nuclear Reactor Regulation
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ML14253A310 List:
References
AEP-NRC-2014-59
Download: ML14253A317 (97)
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ENCLOSURE 2 TO AEP-NRC-2014-59 I&M CAP Document AR 2010-1804-10, Root Cause Evaluation Attachment, "Rx Vessel Core Support Lug Bolting Anomalies"

INDIANA MICHIGAN POVIERe A w*0fAMWt~n CftW PREVENTION DETECTION CORRECTION Root Cause Evaluation of Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10)

March 21, 2010 Root Cause Evaluation Team April Lloyd Lloyd Burton Alex Olp Team Members:

Kevin Kalchik Ben Blumka James Greendonner Mathew White Kevin Neubert Gary Thompson

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10)

Table of Contents 1.0 EXECUTIVE

SUMMARY

4 1.1 PROBLEM STATEMENT.............................................................................................................

4 1.2 E V EN T D E SC R IP T IO N........................................................................................................................

4 1.3 ROOT CAUSE(S) & CORRECTIVE ACTION(S) TO PRECLUDE REPETITION........................

4 1.3.1 CORRECTIVE ACTIONS TO PRECLUDE REPETITION............................................................

4 1.4 EXTENT OF CAUSE & CORRECTIVE ACTIONS.............................................................................

5 1.4.1 EXTENT OF CAUSE CORRECTIVE ACTIONS.................................................................................

5 1.5 CONTRIBUTING CAUSE(S) AND CORRECTIVE ACTION(S)...................................................

6 1.5.1 CORRECTIVE ACTIONS..........................................................................................

6 1.6 OTHER CORRECTIVE ACTION(S)...............................................................................................

6 1.6.1 INTERIM CORRECTIVE ACTION(S).............................................................................................

6 1.6.2 ADDITIONAL CORRECTIVE ACTION(S)....................................................................................

7 1.6.3 ADDITIONAL ACTION(S).........................................................................................................

7 2.0 D E T A ILE D R E PO R T............................................................................................................................

7 2.1 B A C K G R O U N D....................................................................................................................................

7 2.2 DETAILED EVENT DESCRIPTION.............................................................................................

12 2.3 EXTENT OF CONDITION..........................................................................................................

14 2.4 E V E N T A N A L Y SIS............................................................................................................................

15 2.4.1 FA IL U R E A N A LY SIS.........................................................................................................................

15 2.4.2 E V E N T T IM E L IN E.............................................................................................................................

16 2.4.3 FAILURE MODES AND EFFECTS ANALYSIS..........................................................................

16 2.4.4 SUPPORT REFUTE MATRIX......................................................................................................

16 2.4.5 W H Y ST A IR C A SE..............................................................................................................................

17 2.4.6 LOW TEMPERATURE CRACK PROPAGATION (LTCP)..........................................................

20 2.4.7 D O W E L P IN........................................................................................................................................

2 1 2.4.8 INDUSTRY GUIDANCE....................................................................................................................

24 2.4.9 EVALUATION OF INSPECTION METHOD AND FREQUENCY..................................................

24 2.5 ORGANIZATIONAL AND PROGRAMMATIC FAILURE MODE REVIEW............................

24 2.6 SAFETY CULTURE IMPACT REVIEW........................................................................................

25 2.7 EQUIPMENT RELIABILITY REVIEW........................................................................................

25 2.8 SAFETY SIGNIFICANCE..................................................................................................................

25 2.9 EVENT CONSEQUENCES.................................................................................................................

26 3.0 OPERATING EXPERIENCE........................................................................................................

26 3.1 INTERNAL OPERATING EXPERIENCE......................................................................................

27 3.2 EXTERNAL OPERATING EXPERIENCE...................................................................................

28 3.3 REPEAT EVENT OR FAILURE TO LEARN FROM PREVIOUS OPERATING EXPERIENCE.... 28 3.4 L E SSO N S LE A R N E D.........................................................................................................................

29 4.0 EFFECTIVENESS REVIEW PLAN...............................................................................................

29 5.0 A T T A C H M E N T S................................................................................................................................

30 5.1 P R E -A N A L Y SIS.................................................................................................................................

3 1 5.2 EXTENT OF CAUSE EVALUATION..........................................................................................

32 5.3 E V EN T T IM E L IN E.............................................................................................................................

39 5.4 FAILURE MODES AND EFFECTS ANALYSIS..........................................................................

40 5.5 SUPPORT REFUTE MATRIX......................................................................................................

41 5.6 W H Y ST A IR C A SE..............................................................................................................................

49 5.7 INDUSTRY GUIDANCE REVIEW FOR CLEVIS INSERT BOLTS............................................

50 5.8 ORGANIZATIONAL AND PROGRAMMATIC FAILURE MODE REVIEW............................

53 5.9 SAFETY CULTURE IMPACT REVIEW......................................................................................

56 5.10 EQUIPMENT RELIABILITY REVIEW.............................................................................................

63 5.11 A PPL IC A B L E O E................................................................................................................................

66 5.12 DOCUMENTS REVIEWED................................................................................................................

69 5.13 C A R B C O M M E N T S............................................................................................................................

96 2

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10)

LIST OF FIGURES F igure 1: P lan view of L R S S.........................................................................................................................................

9 F igure 2: E levation V iew of L R SS..............................................................................................................................

10 Figure 3: RSK and C levis Insert Interference.........................................................................................................

II Figure 4: A s-designed C levis Insert Installation......................................................................................................

12 Figure 5: Typical condition of broken bolt (left). Broken dowel pin tack welds (right).......................................

13 Figure 6: Synergistic Effects Required for Stress Corrosion in Metals..................................................................

17 Figure 7: LR SS C levis Insert and L ugs.......................................................................................................................

22 Figure 8: Bolts Intact on Load Side Section A-A...................................................................................................

22 Figure 9: Bolts Broken on Load.Side Section A-A.................................................................................................

23 LIST OF TABLES Table 1: Materials and Sizes of Bolts, Lock Bar and Dowel Pin............................................................................

12 Table 2: As-Found Unit 1 Cap Screw and Dowel Pin Conditions from March 2010..............................................

14 Table 3: Summary of nickel based components considered in development of RVI AMP guidance documents....... 36 Table 4: Summary of nickel based components inside the reactor vessel considered in the Alloy 600 Program....... 36 Table 5: List of Orphan Locations in Westinghouse-Designed PWRs from MRP-274........................................

37 Table 6: Comments and Recommendations for Reviewed Components................................................................

38 3

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 1.0 EXECUTIVE

SUMMARY

1.1 PROBLEM STATEMENT Lower Radial Support System (LRSS) clevis insert bolts and dowel pin failed unexpectedly resulting in the organization addressing emergent concerns of structural margin and loose parts in the reactor coolant system (RCS).

1.2 EVENT DESCRIPTION During the 10-year in-service inspection (ISI) of the Donald C. Cook Nuclear Plant Unit 1 (D. C. Cook Unit 1) reactor vessel in March 2010, an anomaly was noted concerning the LRSS clevis inserts. Seven (of 48) bolt locations had wear observed on the lock bar, indicating that the cap screw head had detached from the shank and that flow was causing the head to vibrate against the lock bar. One (of 12) dowel pins had broken tack welds and it was rotated and displaced into the clevis insert.

1.3 ROOT CAUSE(S) & CORRECTIVE ACTION(S) TO PRECLUDE REPETITION Primary water stress corrosion cracking (PWSCC) of the clevis insert bolts due to the use of Alloy X-750 with a susceptible heat treatment.

1.3.1 CORRECTIVE ACTIONS TO PRECLUDE REPETITION Developed and implemented a minimum bolting pattern under EC-51640, '"RX Vessel Lower Radial Support System (LRSS) Clevis Replacement Bolting for Unit I," which incorporated an improved heat treatment and an under head radius that reduced peak stresses.

I WO# 55399712 Owner: TPG Completion Date: 9/28/2013 Basis:

The issues reviewed in this evaluation showed the failure of clevis insert bolts does not have a safety or operational impact on the plant. The observed condition was repaired with a qualified minimum bolt pattern. Any additional inspections, repairs, pre-emptive replacements, or other actions are performed per management discretion.

A meeting was held on January 20, 2014 to discuss corrective actions. The attendees came to agreement on items entered into the CAP system as a result of this evaluation, including absence of corrective actions driving commercial issues. The following is a list of attendees:

Engineering Vice President Site Vice President Performance Improvement Manager 4

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10)

Regulatory Affairs Manager

  • Design Engineering Director
  • Mech. and Struct. Design Engineering Manager
  • Design Engineering Mechanical Supervisor Reactor Vessel Internals Engineer Plant Engineering Director Operations Refueling 1.4 EXTENT OF CAUSE & CORRECTIVE ACTIONS The cause of the observed failures was PWSCC of the clevis insert bolts due to tile use of Alloy X-750 with a susceptible heat treatment. The extent of cause evaluation reviewed components that met all of the following criteria as these could be subjected to a similar cause and result in loose parts in the RCS.

Alloy X-750 or other nickel based alloys Inside the reactor vessel Exposed to primary water The extent of cause also focused on components that meet the criteria above that are not covered under a specific program, or where programmatic elements may overlap as these could contribute to a similar consequence of having to address emergent concerns upon failure. The identified components that warranted corrective actions are listed below. A full evaluation of the extent of cause is provided in Attachment 5.2.

Clevis Insert Bolts Clevis Dowel Pins LRSS Lug Weld BMI Nozzles and Welds Additionally, a rigorous component review of the reactor vessel internals is being performed to support the use of MRP-227-A in the Reactor Vessel Internals Aging Management Program. CA# 2010-1804-31 has been created to review the results of the component evaluation and create actions as necessary based on the causes of the LRSS bolt failure.

1.4.1 EXTENT OF CAUSE CORRECTIVE ACTIONS Determine if the Alloy 600 LRSS clevis insert dowel pins should be added to the Alloy 600 program/procedure since they are currently not identified in EHI-5070-ALLOY600, "Alloy 600 Material Management Program".

CA# 2010-1804-37 1 Owner: ENU Due Date: 4/14/2014 5

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10)

Evaluate if potential mitigating actions exist, such as peening, to prevent PWSCC of the LRSS lug weld and generate corrective actions to implement mitigating techniques if warranted.

I CA# 2010-1804-29 Owner: ENU Due Date: 4/14/2014 Evaluate if potential mitigating actions exist, such as peening, to prevent PWSCC of the BMIs and generate corrective actions to implement mitigating techniques if warranted.

ICA# 2010-1804-30 1Owner: ENU IDue Date: 4/14/2014 Review the results from contract 1500016-156 and create actions to address PWSCC aging effects for any newly identified nickel based alloys based on the root cause of the LRSS bolt failures as necessary.

I CA# 2010-1804-31 1 Owner: DEM Due Date: 6/18/2014 1.5 CONTRIBUTING CAUSE(S) AND CORRECTIVE ACTION(S)

Localized peak stresses at the LRSS clevis insert bolt head to shank radius due to inadequate bolt design dimensions.

1.5.1 CORRECTIVE ACTIONS See corrective actions to preclude repetition.

1.6 OTHER CORRECTIVE ACTION(S) 1.6.1 INTERIM CORRECTIVE ACTION(S)

A rigorous justification for continued operation of Unit 1 was performed to allow for operation for one cycle following the clevis insert bolt failures.

IAR# 2010-1804-18 1Owner: ESY ICompletion Date: 4/03/2010 A more rigorous justification for continued operation of Unit 1 was performed to allow for operation up to two cycles following the clevis insert bolt failures to allow development and implementation of a repair.

AR# 2010-1804-23 1 Owner: ESY I Completion Date: 8/19/2011 Y

D.C. Cook Unit 2 Engineering Evaluation of the Radial Support System Clevis Insert Bolts and Operation through Spring of 2012.

1AR# 2010-1804-22 1 Owner: DEM Completion Date: 9/16/2010 6

Unit I Rx Vessel Core Support Lue Bolting Anomalies (AR 2010-1804-10)

Unit 2 LRSS Clevis Insert Inspection WO# 55371767-39 1 Owner: EISI I Completion Date: 11/1/2010 Developed and implemented a minimum bolting pattern under EC-51640, "RX Vessel Lower Radial Support System (LRSS) Clevis Replacement Bolting for Unit 1," which incorporated an improved heat treatment and an under head radius that reduced peak stresses.

WO# 55399712 Owner: TPG Completion Date: 9/28/20 13 1.6.2 ADDITIONAL CORRECTIVE ACTION(S)

Provide root cause findings to the EPRI MRP and PWROG MSC to allow for evaluation and disposition of results for use in industry guidance.

I CA# 2010-1804-34 1 Owner: DEM Due Date: 4/2/2014 Update OE30993 based on results of Root Cause of failed bolts.

CA# 2010-1804-38 1 Owner: DEM Due Date: 4/2/2014 1.6.3 ADDITIONAL ACTION(S)

WO package CREM comments for the U IC23 ISI inspection made no mention to the failure of the clevis insert bolts. Specifically CREM comments in WO 55343766-14 for the Lower Internal Visual Examination perform in U 1 C23 indicate inspection SAT despite failures observed on 7 LRSS clevis insert bolts and one dowel pin.

AR# 2013-19412 1 Owner: ENU I Due date: 2/20/2014 Consider the cost benefit of the enhancement of zinc addition to the RCS with respect to the potential to inhibit PWSCC.

GT# 2013-19422 Owner: CHM Due Date: 1/29/2014 2.0 DETAILED REPORT

2.1 BACKGROUND

The design and function of the LRSS is described in Exhibit 8b and Exhibit 9 as follows:

In Westinghouse-style pressurized water reactors (PWRs), the LRSS represents the interface between the lower reactor internals and the reactor vessel (RV). The LRSS 7

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) consists of six support locations equally spaced around the bottom circumference of the RV (60-degree spacing), see Figure 1 and Figure 2. Each support location consists of the following:

A Radial Support Key (RSK) attached to the core barrel A clevis insert A vessel clevis lug, which is welded to the wall of the RV See Figure 3 for the interface between the RSK and the clevis insert. For the D.C. Cook design, each core barrel clevis insert contains eight bolts (cap screws), for a total of 48 bolts. These bolts are restrained from rotation by a welded lock bar at each bolt location.

In the event that a bolt fails, the lock bar design captures the bolt and keeps it in place to prevent a loose parts condition. Each insert location also contains 2 dowel pins, for a total of 12 pins. The clevis insert dowel pins provide added, possibly redundant, retention of the insert from long term vibratory motion. The interference fit of the pins keeps the insert from having small displacement slippage in the upward direction over time. They also help to prevent the clevis from sliding upward during core barrel removal. This prevents production of bending stresses in the bolt shanks if the clevis should shift. The clevis insert design and bolts prevent motion in other directions. Table 1 describes the bolt, lock bar, and dowel pin materials and sizes. Figure 4 illustrates the as-designed installation of the clevis inserts at D.C. Cook.

8

Unit 1 Rx Vessel Core SUpport Lug Bolting Anomalies (AR 2010-1804-10) 90"

, /

_,00/

-IGO*

270" Figure 1: Plan view of LRSS 9

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10)

Figure 2: Elevation View of LRSS 10

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10)

Cap Screw C

'ent Bwe arrel

'Plan View Figure 3: RSK and Clevis Insert Interference I I

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10)

Clearance Fit Between the Vessel 0"evs Lup " *le Ce~s Insert Mt thtese Interraes Reactor Vesset Vessel Clevis Lug Clevis Insert Interference Fit Betfwen the Vessel Clevis LUgS and tho COwvis Instil at View A-A me. interfaues Figure 4: As-designed Clevis Insert Installation Table 1: Materials and Sizes of Bolts, Lock Bar and Dowel Pin Component Material Dimension Clevis Bolts Alloy X-750 Shank 2.50" 1-8 Thread Head 1.0" x 1.50" dia.

Lock Bar ASTM B-166 (Ni-Cr-Fe) Annealed Bar 1.56 x 0.25 x 0.35 Dowel Pin ASTM B-166 (Ni-Cr-Fe) Annealed Rod 3.75 x 1.375 dia.

2.2 DETAILED EVENT DESCRIPTION On March 20, 2010, during the 10-year ISI of the D. C. Cook Unit 1 reactor vessel, an anomaly was noted concerning the LRSS clevis inserts. The 10-year ISI program includes a remotely operated visual inspection of the vessel after the core barrel is removed. This visual inspection examines the condition of the six clevis inserts, including general integrity of bolted and welded connections.

Damage was observed in the LRSS clevis insert bolts. Wear was observed on the lock bars and cap screw heads at a total of seven bolt locations; at five of these locations, the bolt heads were dislodged from their expected locations. Additionally, at one location, 12

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) the three tack welds holding the dowel pin in place were noted to be broken. Figure 5 shows an image of a typical broken bolt (left) and the broken dowel pin tack welds (right). Table 2 shows the as-found condition of the cap screw and dowel pin with respect to each clevis location [Exhibit 9].

ed Lock Bar Clevis Insert Bolt Welded Dowel Pin 5: Typical condition of broken bolt (left). Broken dowel pin tack welds (right)

Following the discovery of the degraded clevis bolting, Westinghouse performed operability assessments allowing D.C. Cook Unit 1 to run until spring 2013 prior to replacement of the damaged clevis insert bolting. During Unit I Cycle 25 refueling outage in March 2013, a total of 29 clevis insert bolts were removed from the reactor vessel. Westinghouse performed a detailed evaluation defining a minimum acceptable bolting pattern on the clevis inserts for continued operation of Unit 1. The minimum acceptable bolting pattern, totaling 28 new clevis insert bolts, was installed during Unit 1 Cycle 25 refueling outage.

In September of 2010, Westinghouse performed an engineering evaluation of the Unit 2 LRSS clevis insert bolts, justifying continued operation of Unit 2 until inspections could be performed in spring 2012. During the Unit 2 Cycle 19 refueling outage October 2010, the core barrel was removed from the reactor vessel due to the discovery of baffle-former bolt degradation. With the core barrel removed, the decision was made to perform a VT-3 examination on the Unit 2 clevis insert bolts. No indication of degradation was observed on Unit 2 clevis insert bolting at the time of inspection.

13

Unit I Rx Vessel Core Suppori Lug Bolting Anomalies (AR 2010-1804-10)

Table 2: As-Found Unit I Cap Screw and Dowel Pin Conditions fromn March 2010 As-Found Cap Screw and Dowel Pin Conditions.

00 Location Left Side Rikht Side

1 R0o T1eatkbn T..ft ~4Ae Rtchti SAN Top Bolt

()

Second Bolt (1)

)

Dowel Pin

()

0_

Third Bolt (1)

()

Fourth Bolt

_I 60P Location Left Side Right Side Top Bolt Second Bolt (1) o)

Dowel Pin (1)

()

Third Bolt (1)

Fou*th Bolt 1200 Location Left Side Right Side Top Bolt Wear/Dislodged

)

Second Bolt

(_)__)

Cracked Welds Dowel Pin and rotated 20-30° CCW )

Third Bolt Wear/Dislodged Top Bolt I

i_

Second Bolt J

{)

(I).

Dowel Pin Third Bolt

().

WearDislodged Fourth Bolt

.0) 1 Wear/Dislodged 2400 Location Left Side Right Side Top Bolt (3)

Second Bolt

71)

()

Dowel Pin (0-(3 ThirdBolt

( O) ci)

Fouit Bolt 3000 Location Left Side Right Side Top Boit

()

Wear Second Bolt (1)

Wear Dowel Pin (1)

(1 ThirdBolt

3)

(x)

Fourth Bolt Wear/Dilodged I (3)

Fourth Bolt (3)

(1)

Now*

1. No visilt ii
2. Cota lo&ukw.

2.3 EXTENT OF CONDITION The extent of condition review was performed to identify where else the station was vulnerable to the same or similar condition. The condition was that LRSS clevis insert bolts and dowel pin failed unexpectedly. The immediate extent of condition includes the 19 LRSS clevis insert bolts in Unit 1 which were not replaced during the minimum bolt replacement in the Unit 1 Cycle 25 refueling outage and the 48 LRSS clevis insert bolts in Unit 2. This bounds the extent of condition.

The observed condition of bolt failures in Unit I was identified by the existing actions in place to perform visual inspections during each 10 year in-service inspection. The indications were documented and corrected through the Corrective Action Program. The actions taken were adequate to identify and correct the condition in Unit 1. There are no safety or operability concerns resulting from failure of LRSS clevis insert bolts or dowel pins. A rigorous loose parts generation and transport evaluation was performed to justify continued operation for two cycles following discovery. Therefore, no additional corrective actions are being suggested for Unit 1 or Unit 2 resulting from failure of the LRSS clevis insert bolts.

14

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 2.4 EVENT ANALYSIS 2.4.1 FAILURE ANALYSIS Metallurgical Analysis: "Babcock & Wilcox Technical Services Group (B&W) Report" S-1473-002 During the Unit 1 Cycle 25 refueling outage in March/April 2013, a total of 29 clevis insert bolts were removed from the reactor vessel. Of the 29 bolts that were removed, 13 were considered intact (later found to be cracked) and 16 were fractured (head separated from shank). These bolts were sent to B&W for failure analysis. The failure analysis concluded the following [Exhibit 5e]:

All of the 29 submitted bolts contained cracking in the head-to-shank transition; no cracking was identified in the threaded region of any bolts.

There was a generally uniform fracture pattern observed in the bolts, which consisted of crack initiation at two diametrically opposing sides of the bolt in the head-to-shank transition region and crack growth that extended upward into the bolt head at a -35' angle relative to horizontal.

Fractographic SEM analysis and cross section metallographic examinations confirmed the fracture mode was essentially 100% intergranular on all of the bolts. Very minor mixed mode cracking consisting of transgranular cleavage and ductile fracture was noted near the center of one bolt which would have been the final failure region.

There was no evidence that the bolts failed due to fatigue cracking or mechanical overload.

The chemical analysis results for all four bolts were consistent with Alloy X-750 material.

The mechanical properties and microstructure of the bolts were consistent with those published for Alloy X-750 material.

  • No unexpected characteristics in the material properties, microstructures, or form of the bolts were identified.

The laboratory data indicated the bolts failed by intergranular stress corrosion cracking (IGSCC).

15

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 2.4.2 EVENT TIMELINE A Timeline was created to depict the chronology of relevant events since initial construction of DC Cook Units 1 and 2 to support the detailed event description provided in Section 2.2. The Timeline focused on the following:

  • Inspections and results Dates of published industry guidance for reactor vessel internals The event itself
  • The interim actions that were put into place following the event
  • Time periods with conditions that could promote IGSCC See Attachment 5.3.

2.4.3 FAILURE MODES AND EFFECTS ANALYSIS A Failure Modes and Effects Analysis was performed to identify all possible failure modes of the LRSS clevis insert bolts and dowel pin failures through a collegial review of the identified problem and third party evaluation, as seen in Attachment 5.4. The effects from the identified failures were then used to build conclusions in the Support Refute Matrix discussed below and shown in Attachment 5.5.

2.4.4 SUPPORT REFUTE MATRIX A Support Refute Matrix was created to support or refute the failure modes identified in the failure modes and effects analysis and determine which of them most likely caused or contributed to the LRSS clevis bolts and dowel pin failure (See Attachment 5.5). The following were supported in the sLupport/refute matrix for the clevis bolt failures:

Intergranular Stress Corrosion Cracking (IGSCC)/Primary Water SCC (PWSCC)

  • Improper material heat treatment specification
  • Inadequate bolt design dimensions at the bolt head to shank radius Low Temperature Crack Propagation (LTCP) - Potential contributor IGSCC/PWSCC requires a combination of tensile stresses (both applied and/or residual),

a corrosive environment, and a susceptible material to be present. IGSCC/PWSCC will not occur if any of these three factors is eliminated. The support refute found that the stresses produced at the head to shank radius were not sufficient to cause overload failure of the LRSS clevis insert bolts. However, due to the tight dimensions at this location the stresses produced were found to be a contributing cause to PWSCC. The support refute also found that the heat treatment used on the LRSS clevis insert bolts is now well known to be susceptible to SCC in a PWR environment. Without the use of this susceptible heat treatment SCC would not have occurred.

These potential failure modes are discussed in more detail in subsequent sections of this evaluation.

16

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 2.4.5 WHY STAIRCASE A Why Staircase was constructed to validate the causal factors leading to failure of Unit 1 Clevis Insert Bolts. (See Attachment 5.6) Completion of the Why Staircase validated the following root cause:

Root Cause: Primary water stress corrosion cracking (PWSCC) of the clevis insert bolts due to the use ofAlloy X-750 with a susceptible heat treatment.

Stress corrosion cracking is a known phenomenon that occurs in Alloy X-750 nickel alloy components under specific conditions. Figure 6 below shows the conditions that need to be present for SCC to occur. SCC will not occur if any of these three factors is out of range of susceptibility. In the case of the D.C. Cook Unit I clevis insert bolts, the bolts failed intergranularly (at the grain boundaries of the metal); thus the failure mechanism being IGSCC. Alloy X-750 bolts have historically failed by IGSCC in PWR primary water environments; a phenomenon also referred to as Primary Water Stress Corrosion Cracking (PWSCC). [ Exhibit 421 rl-71 t T f-,

1; 1

Lwýý Figure 6: Synergistic Effects Required for Stress Corrosion in Metals Material and Heat Treatment Selection of Alloy X-750 [Ref: Exhibit 121 Nickel based Alloy X-750 was originally developed for use in high temperature applications where good corrosion resistance and high strength are required, such as in gas turbines. Alloy X-750 is typically used in nuclear power plants where high corrosion resistance similar to that of Alloy 600 is required and higher strength and fatigue resistance is needed, such as in control rod guide tube support pins. Alloy X-750 is now 17

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) considered a 'mature' alloy in that there has been extensive research, testing, and improvements of the alloy. As discussed in the EPRI Materials Handbook for Pressure Boundary Applications, perhaps the most important result of industry research on Alloy X-750 was the determination that the heat treatment condition strongly affects its susceptibility to SCC in PWR and BWR environments. It was found that heat treatments used in early industry specifications, while suitable for non-Light Water Reactor environments such as gas turbines, resulted in high susceptibility to SCC in high temperature water environments.

Heat treatment of Alloy X-750 typically starts with a hot worked material that is solution heat treated at a high temperature to ensure a single uniform microstructure is present prior to aging. Typical heat treatments for Alloy X-750 components used in early nuclear plant construction included the AH and BH heat treatments which call for solution treating at 1625 OF and 1800 OF respectively. There is a significant amount of operating experience showing failures of Alloy X-750 bolts in this heat treatment when exposed to primary water at high temperatures [Exhibit 12]. Current industry guidance shows a significant improvement in resistance to stress corrosion cracking when the solution treating temperature is raised to 2000 OF such as with the HTH heat treatment. The LRSS clevis insert bolts that failed at CNP Unit 1 were manufactured in accordance with Westinghouse Materials Specification 70041 EJ, which calls for an equalization heat treatment of 1625 OF for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> followed by a solution treatment of 1775 OF for one hour. Therefore, the D.C. Cook Unit 1 clevis insert bolts were heat treated using a process that is known to make Alloy X-750 susceptible PWSCC, which was identified as the root cause to the observed PWSCC.

Root Cause: Primary water stress corrosion cracking (PWSCC) of the clevis insert bolts due to the use ofAlloy X-750 with a susceptible heat treatment.

High Stress at Head to Shank Radius The B&W metallurgical analysis determined that all removed LRSS clevis insert bolts failed or exhibited cracking in the head to shank transition region [Exhibit 5e]. The bolt head to shank transition or radius acts as a stress riser and is one of the areas of highest localized stresses on the bolts. A high stress concentration can result in high peak stresses depending on the preload and any thermal or operating stresses.

MRP-175, "PWR Internals Material Aging Degradation Mechanism Screening and Threshold Values," gives screening criteria for aging degradation mechanisms. In the case of Alloy X-750, the document sets a stress corrosion cracking screening criterion of 100 ksi or greater for Alloy X-750 in the HTH condition. An Alloy X-750 component loaded beyond this level has some finite probability of experiencing SCC. MIRP-175 also notes that Alloy X-750 in the AH or BH heat treated condition (similar heat treatments to that of the CNP LRSS clevis insert bolts) would be even more susceptible than material in the HTH condition and would thus have an even lower screening criterion. According to a Westinghouse stress evaluation, the bolts were subject to as much as 95 ksi in the head to shank transition due to preload and thermal stresses while all of the bolts were 18

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) intact [Exhibit 58 and Exhibit 8b]. This stress is high enough for stress corrosion cracking initiation of the more susceptible Alloy X-750 material at D.C. Cook. Hence the following was identified as a contributing cause to the observed PWSCC.

Contributing Cause: Localized peak stresses at the LRSS clevis insert bolt head to shank radius due to inadequate bolt dimensions.

PWR Primary Water Environment There is extensive operating experience for Alloy X-750 bolts with a susceptible heat treatment failing in primary water by PWSCC [Exhibit 121. EPRI guidelines for stress criteria described above are based on maintaining a water chemistry condition in accordance with EPRI PWR Primary Water Chemistry Guideline. Any anomalies in the Primary Water Chemistry at D.C. Cook Unit I could explain why the bolts in Unit I failed when no other failures have been reported throughout the industry [Exhibit 421.

D.C. Cook Chemistry Department personnel indicate that the CNP primary chemistry program complies with EPRI's Pressurized Water Reactor Primary Water Chemistry Guidelines.

Technical Requirements Manual limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion cracking. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant.

[TRM 8.4] Previous Technical Specification primary chemistry limits and bases were similar.

Historical Licensee Event Reports for out-of-specification primary chemistry were reviewed with chemistry personnel and determined to be not applicable to PWSCC

[Exhibits 118 and Exhibit 1191.

Review of PWR chemistry recommendations found that RCS zinc addition could be a possible inhibitor to PWSCC initiation and propagation in nickel-based alloys.

Approximately 30 percent of PWRs currently utilize zinc addition [Exhibit 132]. RCS zinc addition was most recently evaluated at CNP for dose reduction (GT 00844243-35) and rejected. As an enhancement, a general tracker (GT 2013-19422) has been created to consider the cost benefit of zinc addition to the RCS with respect to the potential to inhibit PWSCC.

19

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 2.4.6 LOW TEMPERATURE CRACK PROPAGATION (LTCP)

LTCP is a driving mechanism for crack growth that has been seen in Alloy X-750 materials. Industry literature has shown LTCP occurs under the following conditions

[Exhibit 431:

Temperatures between 122 and 302 degrees F

  • Hydrogen concentrations > 20cc H2/kg H20 Existing crack like defect to act as an initiation site When these conditions are present, Alloy X-750 can experience rapid cracking (rates of crack propagation on the order of millimeters per minute) due to a hydrogen embrittlement effect. Hydrogen can come externally from the coolant in contact with the metal, from corrosion of the metal surface or it can be present internally due to previous exposure to hydrogen. Chemistry records fi'om the Unit I extended shutdowns in 1998-2000 and 2008-2009 were reviewed and it was determined that periods existed in which temperature and hydrogen conditions would promote LTCP during the first approximate six days of the 1998-2000 extended shutdown. For the 2008-2009 extended shutdown, hydrogen levels were reduced to below 20cc H2/kg H20 on the first day of the turbine event and coming out of this shutdown period temperatures were raised above 302 degrees F prior to hydrogen levels elevating back above 20cc H2/kg H20. [Exhibit 8b]

The failure analysis performed by B&W found no indications of existing crack-like defects or evidence of machining or fabrication issues that could have led to LTCP of the clevis insert bolts. Therefore, if LTCP played a role in the failure of the LRSS clevis insert bolts, it would have only been a driver once cracking initiated from PWSCC.

Preventing LTCP would not have prevented crack initiation or the eventual fracture of the bolts. Due to the rapid crack growth rates associated with LTCP, and without knowledge of when the bolt heads failed, LTCP cannot be validated as a contributing cause.

Further review of the CNP chemistry controls was conducted by the team to determine if additional corrective actions would be warranted to reduce the time spent in LTCP conditions. During normal operation LTCP is not of concern due to the RCS temperature being above the temperatures at which LTCP would occur. 12-THP-6020-CHM-110, "RCS Chemistry - Shutdown and Refueling" Figure 1 provides the RCS dissolved hydrogen target bands. Review of this figure shows that actions are already in place to ensure that hydrogen levels stay below 20cc H2/kg in the temperature range of concern for LTCP. It appears that these controls were put in place during revision 8 of 12-THP-6020-CHfM-i 10 in 2002 [Exhibit 125]. No additional actions are needed. See Exhibit 8b

& Exhibit 43 for further information on LTCP.

20

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 2.4.7 DOWEL PIN The following description of the dowel pin is modified from Exhibit 8b:

The Alloy 600 dowel pins were interference fit into the clevis insert and lug during construction. The tack welds were intended to provide capture of the pin in case of failure and provide no structural purpose. The welds are small and offer relatively little resistance to pin motion as compared to the resistance resulting from the interference fit.

There are no obvious indications of deformation in the welds as observed from inspection images. Therefore, the most likely cause of failure is either fatigue or SCC. The forces required to rotate and translate the pin, as observed in the ISI inspection, would be more than sufficient to drive either failure mechanism.

Normal loading of the interference fit dowel pin would not be expected to break tack welds and cause rotation or translation of the pin. It is significant that the dowel pin with broken tack welds and movement was located on the clevis insert where the most bolt damage was observed. All of the bolts on this clevis had completely fractured. The loss of clamping load on this side of the clevis with the displaced dowel pin may have resulted in vibration associated with the cyclic loading of the insert during operation. The dowel pin would then be subject to vibration due to loss of clamping load.

Normal loading of the clevis insert following bolt failure was likely sufficient to drive rotation and axial translation of the dowel pin. The dowel pin must not have been fully seated in the reamed hole at installation which allowed the pin to be driven deeper into the lug. The dowel pin holes were reamed after installation of the clevis insert, so any Poisson burr I remnant from machining would tend to ratchet movement of the dowel pin deeper into the hole. Therefore, the failure of the dowel pin tack welds and dowel pin motion is a consequence of the bolt failure at that location.

Figure 7 shows an overview of the clevis insert and the location of cross section for the following two figures. Figure 8 shows an exaggeration of how the clevis insert would deflect under radial loading with intact bolts. Figure 9 shows an exaggeration of how the clevis insert would deflect under radial loading with failed bolts [Exhibit 94].

Plastic deformation of material which may be caused by machining operations such as edge features at the exit face of material after drilling or reaming.

21

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 7 1~

r"I,\\

\\

/

AV--0--

0 G

O- - A E L/

II le i............

4 Figure 7: LRSS Clevis Insert and Lugs

- Dowel Pin (Interference fit in Clevis Insert and Vessel Clevis Lug)

Dowel Pin tack welds Exaggerated deflection of the Clevis Insert under load

'Bolts restrict axial load and deformation at dowel pin location.

Figure 8: Bolts Intact on Load Side Section A-A 22

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10)

Load can be developed in the dowel pin tack welds if dowel pin fit is tighter in the Vessel Clevis Lug -

than in the Clevis Insert.....

Why this might happen:

- Hole in Vessel Clevis Lug is a blind hole versus a thru hole in the Clevis Insert

- Thermal response of Vessel Clevis Lug is different than that of the Clevis Insert during a transient condition Exaggerated deflection of the Clevis Insert under load

  • Deformation at the dowel pin not restricted Figure 9: Bolts Broken on Load Side Section A-A 23

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 2.4.8 INDUSTRY GUIDANCE Because CNP Units 1 and 2 have Alloy X-750 LRSS clevis insert bolts with a heat treatment that has extensive operating experience showing it produces components susceptible to PWSCC, the root cause team reviewed the aging management plan and industry guidance for managing the aging of LRSS clevis insert bolts. The conclusions of this review found that industry guidance indicated a medium likelihood of stress corrosion cracking of clevis insert bolts. However, no actions beyond inspecting per the 10-year IS1 were recommended due to overall effect of PWSCC of the clevis insert bolts on the clevis insert not being significant. The industry guidance documents reviewed found that the existing 10 year in-service inspection or other in-place aging management plans are sufficient to preclude a safety, reliability, or financial concern. A closer review of the guidance document attributes and their relation to the Reactor Vessel Internals Aging Management Program is provided in Attachment 5.7. Furthermore, the event and causal factor chart (Timeline) shown in Attachment 5.3 lists the industry guidance documents that the CNP reactor vessel internals aging management program is based on and when they were published.

2.4.9 EVALUATION OF INSPECTION METHOD AND FREQUENCY The issues reviewed in this evaluation supported the 10 year ISI VT-3 examination as the appropriate periodicity and inspection based on the low safety and operability significance of clevis bolt failure. The reactor vessel internals aging management guidance document MRP-227-A, shows the clevis insert bolts are considered to have a low likelihood of damage, medium likelihood of failure, and a consequence of failure being only a significant economic impact. This guidance document has been updated since the observed LRSS clevis bolt failures and the event is included in the OE.

The condition reviewed in this evaluation did not result in the loss of the ability of the LRSS to perform its intended design function. Even with the postulated failure of all bolts, all dowel pins, and clevis-to-lug interference fit, the clevis insert will remain in place due to the capturing geometry of the vessel lugs, core barrel radial keys, and the bolt and dowel pin remnants. The LRSS will maintain its design function in all operational and faulted conditions. [Exhibits 6 8b, 2 55 74 85a, 87b, 88 113 and 121]

2.5 ORGANIZATIONAL AND PROGRAMMATIC FAILURE MODE REVIEW This evaluation found no organizational or programmatic issues. All evidence shows that the clevis insert bolts were known to have a potential for failure, but because of the low significance of failure from a safety and operational standpoint, and no history of failure in the industry, no contingencies were in place in the event of a failure. Existing industry guidance found in MRP-227 relied on the ASME Section XI 10-year ISI inspection for ensuring the integrity of the clevis insert bolts, which properly identified the failure. See.8 for Organizational And Programmatic Failure Mode Review.

24

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 2.6 SAFETY CULTURE IMPACT REVIEW A Safety Culture Impact Review was performed looking at Nuclear Regulatory Commission (NRC) Components of Safety Culture and Institute of Nuclear Power Operations (INPO) 8 Principles of a Strong Nuclear Safety Culture. This evaluation found no applicable cross-cutting area components or nuclear safety culture issues related to the failure of the LRSS clevis bolts. See Attachment 5.9 for Safety Culture Impact Review.

2.7 EQUIPMENT RELIABILITY REVIEW Incorrect Classification (AP-913 INCORT) was applied due to the clevis bolts being incorrectly classified; the classification of the reactor vessel, 1-OME-1, which is a critical component, was applied. The LRSS clevis bolts are piece parts, not a piece of equipment. Therefore, they are not classified as critical, non-critical, or run-to-failure.

Original Design Less than Adequate (AP-913 ORIG) was applied due to the bolts failing by PWSCC. The original design used an Alloy X-750 bolt in a heat treatment unknowingly susceptible to PWSCC. Additionally, all of the bolts failed in the head to shank transition due to the localized high stresses in this region of the bolt. The replacement bolts utilized an Alloy X-750 bolt with a less susceptible heat treatment and an improved head-to-shank design transition to reduce stress in this region.

See Attachment 5.10 for Equipment Reliability Review.

2.8 SAFETY SIGNIFICANCE The condition described in this action was reviewed for its actual and potential impact on nuclear, radiological, personnel/industrial safety, and plant operation. Based on this review, the condition described is of low safety significance.

Nuclear Safety The condition reviewed in this evaluation did not result in the loss of the ability of the LRSS to perform its intended design function. Even with the postulated failure of all bolts and dowel pins, clevis inserts are predicted to remain in place. [Exhibit 55 and Exhibit 8b]

Although bolt failures do not challenge the ability of the LRSS to perform its design function, it is important to the plant from a commercial perspective. The clevis inserts have several redundant means of attachment including bolts, pins, an interference fit into lugs, and the geometry of the assembled system. However, if the first three listed means of attachment are defeated, this creates a commercial concern. When the core barrel is removed for maintenance, a clevis insert may then be allowed to displace from the lugs 25

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) and fall to the bottom of the reactor vessel. LRSS clevis replacement following displacement of a clevis insert from its lugs may challenge the economic viability of the plant. This significant clevis insert displacement cannot occur with the core barrel installed. The core barrel is not removed unless the reactor is completely defueled.

Radiological Safety There was some impact on radiological safety. Dose for Unit 1 LRSS repairs was 4635 mR and there was one, level one personnel contamination.

Personnel/Industrial Safety There was no impact on personnel or industrial safety from the condition reviewed in this evaluation.

Operational Impact Plant operation with the condition described in this action had no adverse effect on the design functions of any UFSAR-described SSCs.

Probabilistic Risk Assessment (PRA)

The PRA group was contacted regarding a PRA model for the LRSS clevis insert bolts.

The LRSS clevis insert bolts do not have a PRA model because PRA only considers SSCs where failures could cause a safety or operability concern. Therefore it is not appropriate to include them in a PRA model.

2.9 EVENT CONSEQUENCES This event did not adversely impact operability or safety of plant operation. The event reviewed in this evaluation resulted in financial consequences due to operability evaluations, LRSS repair project, and impact on the UIC25 refueling outage schedule, duration, and dose. This event resulted in an ongoing high level of interest in the issue and our response to the issue from other utilities, industry groups, and the NRC.

3.0 OPERATING EXPERIENCE The purpose of Operating Experience (OE) is to reduce the possibility of events occurring that have been previously identified by taking actions to improve the barriers to prevent occurrence, and benchmarking other utilities to identify actions taken that have been successful in preventing similar events.

A review of internal and external OE was conducted to identify events that had similar causes to the events that were investigated in this root cause evaluation. These OE were reviewed to identify potential corrective actions, contributing causes and lessons learned that could prevent future occurrences or assist in the analysis section of this report. The details of this review are documented in Attachment 5.11 of this report.

26

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 3.1 INTERNAL OPERATING EXPERIENCE Searches of the ActionWay and the Cook OE database were conducted using the following key words:

Split Pins

  • Alloy X-750
  • Clevis Bolts Lower Radial Support System (LRSS)

Primary Water Stress Corrosion Cracking (PWSCC) & Bolt Low Temperature Crack Propagation (LTCP)

The majority of OE found in ActionWay was related to Alloy X-750 valve discs which were determined to be non-applicable to this evaluation. Additionally, there were several actions related to the control rod guide tube support pings (split pins) as discussed below.

Searches also found actions related to PWSCC susceptibility of RCS piping which has previously been mitigated by the Cook Plant. The following actions were found to be most applicable to this evaluation.

GT 2012-1808/2012-1809 This GT LRP is to replace the Control Rod Guide Tube (CRGT) Support Pins (split pins) no later than U2C26 in Spring 2021 and UIC28 in Fall 2017.

The support pins align CRGTs to the Lipper core plate. The original material of the split pins was Alloy X-750 with a low temperature solution anneal heat treat. These pins were susceptible to PWSCC and failed. Split pins were replaced by Babcock & Wilcox Company in both D.C Cook Unit 1 and Unit 2 in the mid 1980's with new Alloy X-750 split pins with improved design to reduce stress risers and a high temperature solution anneal heat treat. These improved features reduced susceptibility to PWSCC, but did not eliminate the problem. The comparison of the split pin failure to the clevis bolts was discussed by Westinghouse in WCAP-14577 Rev I-A, March 2001. While it was known that the Alloy X-750 clevis bolt material was susceptible to PWSCC, the failure was not considered likely due to differences in fluence, temperature, and stresses between the two bolts.

AR 2010-10940 This was the root cause evaluation for the core baffle bolt failures in the Unit 1 reactor vessel. The root cause of failed baffle-former bolts was Irradiation Assisted Stress Corrosion Cracking (IASCC) in conjunction with thermal and irradiation induced loss of preload in several baffle-former bolts. IASCC was considered in the Support/Refute analysis for this evaluation.

27

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 3.2 EXTERNAL OPERATING EXPERIENCE Searches of the INPO OE database were conducted using the following key words:

0 Split Pins

" Alloy X-750

" Clevis Bolts

  • Lower Radial Support System (LRSS)
  • Low Temperature Crack Propagation (LTCP)

The results of the searches varied from no results to numerous results, depending on the key word search and format. A significant amount of OE was reviewed related to CRGT split pin failures of Alloy X-750 bolts by PWSCC. External OE related to split pin failures has not been included in this report considering the extent of internal OE at Cook.

See internal operating experience in Section 3.1 of this report for D.C Cook related OE on split pin failures. The most applicable remaining results to this evaluation are detailed below.

NUREG 1801, Revision 2 The Generic Aging Lessons Learned (GALL) Report from the NRC identifies the clevis insert bolts as having a possible aging effect/mechanism as loss of material due to wear.

This report supports the use of EPRI MRP-227 for managing the aging effects of clevis insert bolts.

OE99 - 009452 - Nine Mile Point Unit ]

During the 1999 refueling outage fifteen (RFO 15), one of the Alloy X-750 3/8" cap screws was found to be broken during a visual inspection of a core shroud repair tie rod assembly. The root cause was determined to be IGSCC in conjunction with large, sustained differential thermal expansion stress due to fastening of dissimilar materials with the cap screw. The difference in thermal expansion between dissimilar materials being bolted together has been evaluated in the support/refute section of this report.

3.3 REPEAT EVENT OR FAILURE TO LEARN FROM PREVIOUS OPERATING EXPERIENCE Review of the external and internal OE determined that the issues found in this root cause did not represent a repeat event and could not have been prevented by lessons learned in the previous events cited in this section.

Station personnel were aware of the X-750 LRSS clevis insert bolts, and their possible susceptibility to PWSCC, prior to observing bolt failures. The station followed inspection and 28

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) aging management guidance provided in ASME Code, EPRI-MRP, and PWROG documents which indicated low risk and low consequence of failure.

3.4 LESSONS LEARNED Due to the low significance of a failure and no history of failure in the industry, no contingencies were in place in the event of a failure of the clevis insert bolts. Existing industry guidance in MRP-227 relied on the ASME Section XI 10-year ISI inspection for ensuring the integrity of the clevis insert bolts. Failure mechanisms identified during the OE review have been added to the support/refute matrix for further evaluation.

4.0 EFFECTIVENESS REVIEW PLAN No additional corrective actions to correct or preclude repetition of the observed condition were generated beyond those already performed. The ISI program continues to monitor the LRSS according to ASME Code. Therefore, no effectiveness review is required. This was discussed and resolved in the same meeting on January 20, 2014 that corrective actions were discussed. Attendees are listed in Section 1.3.1.

29

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) 5.0 ATTACHMENTS 5.1 PRE-ANALYSIS 5.2 EXTENT OF CAUSE EVALUATION 5.3 EVENT TIMELINE 5.4 FAILURE MODES AND EFFECTS ANALYSIS 5.5 SUPPORT REFUTE MATRIX 5.6 WHY STAIRCSE 5.7 INDUSTRY GUIDANCE REVIEW FOR CLEVIS INSERT BOLTS 5.8 ORGANIZATIONAL AND PROGRAMMATIC FAILURE MODE REVIEW 5.9 SAFTETY CULTURE IMPACT REVIEW 5.10 EQUIPMENT RELIABLITY REVIEW 5.11 APPLICALBE OE 5.12 DOCUMENTS REVIEWED 5.13 CARB COMMENTS 30

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.1 5.1 PRE-ANALYSIS Event Description During the 10-year in-service inspection of the Donald C. Cook Nuclear Plant Unit 1 reactor vessel in March 2010, an anomaly was noted concerning the Lower Radial Support System (LRSS) clevis inserts. Seven (of 48) bolt locations had wear observed on the lock bar, indicating that the cap screw head had detached from the shank and that flow was causing the head to vibrate against the lock bar. One (of 12) dowel pin had broken tack welds and it was rotated and displaced into the clevis insert.

Problem Statement Lower Radial Support System (LRSS) clevis insert bolts and dowel pin failed unexpectedly resulting in the organization addressing emergent concerns of structural margin and loose parts in the reactor coolant system.

Prompt and Interim Actions A rigorous justification for continued operation was generated for two cycles to allow for repair development. A minimum bolt pattern was installed per EC-51640, "RX Vessel Lower Radial Support System (LRSS) Clevis Replacement Bolting for Unit 1."

Scope of Evaluation The evaluation will investigate the causes and potential organizational and programmatic issues regarding the LRSS bolt and dowel pin failure through the use of metallurgical testing, support/refute analysis, failure modes and effects analysis, and why staircase.

Management Sponsor, Team Lead, and Team Members Management Sponsor:

April Lloyd (AEP/DE)

Team Lead:

Alex Olp (AEP/DEM)

Cause Evaluator:

Lloyd Burton (AEP/WCA)

PID Mentor:

Jessica Huycke (AEP/PID)

Team Members:

Kevin Kalchik (AEP/DEM)

Matthew White (AEP/ENU)

James Greendonner (AEP Operations)

Ben Blumka (AEP Maintenance)

Kevin Neubert (Westinghouse)

Gary Thompson (MPR)

Schedule Present Pre-analysis to CARB:

10/25/13 Team Pre-job Brief:

11/13/13 CARB Update:

12/6/13 Evaluation Approval:

12/20/13 31

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.2 5.2 EXTENT OF CAUSE EVALUATION The bolts failed because of PWSCC of the clevis inset bolts due to the use of Alloy X-750 with a susceptible heat treatment. Other nickel based alloys could be susceptible to stress corrosion cracking in primary water under tensile stress.

This evaluation is bounded based on several factors. The initial volume of interest was unisolable portions of the RCS. There are many programs which cover inspection and aging management of the system and specific components within the system. The clevis bolt failures occurred in the reactor vessel, so the review is limited to this volume because this presents a higher consequence for non-pressure boundary component failures for safety, operability, and economic impact due to difficulties associated with repair/replacement activities.

The scope of the evaluation does not include consumable items, such as fuel assemblies, reactivity control assemblies, or nuclear instrumentation, because these components are not typically within the scope of the components that are required to be subject to an aging management review, as defined by the criteria set in 10 CFR 54.21(a)(1).

Therefore, the investigation is bounded to non-consumable components manufactured with nickel based alloys in the volume of the reactor vessel and internals exposed to primary water that do not have a pressure boundary function.

The reactor vessel is governed by multiple programs which include, but are not limited to, the RVI AMP and the Reactor Vessel Integrity Program. The RVI AMP typically governs reactor vessel internal removable structures, not all non-pressure boundary features and components inside the reactor vessel. The Reactor Vessel Integrity Program manages the reduction of fracture toughness due to irradiation embrittlement of reactor vessel beltline materials to assure that the pressure boundary function of the reactor vessel beltline is maintained. The LRSS clevis inserts, bolts, dowel pins, and lugs seem to fall at the interface between these two particular major component groups and programs.

Table 3 of Attachment 5.2 is a summary of nickel based alloy components in the reactor vessels contained in MRP-191, a document created during the development of MRP-227-A to screen, categorize, and rank RVI components. The last two columns discuss any additional relevant details or management strategies specific to DC Cook for these components.

The ASME Section XI ISI Program includes, but is not limited to, inspections of the reactor vessel internals, and the reactor vessels including its internal components and attachments. This program is credited in other programs, including the RVI AMP, for VT-3 visual inspections of LRSS components, including clevis insert bolts.

The Alloy 600 Material Management Program manages aging effects of both pressure and non-pressure boundary components made of Alloy 600/690, 82/182, and 52/152 materials in the RCS. The clevis inserts, lugs, and lug welds are considered and 32

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.2 evaluated in this program. However, the clevis bolts are not included in the Alloy 600 Program because they are fabricated from Alloy X-750 which is not a screened material.

Table 4 is a summary of nickel based alloy components in the reactor vessels contained in the Alloy 600 Program and supporting documents. It is recognized that this screening may not contain a complete list of nickel based components in the reactor vessels due to the scope limitations of the programs. This seems to create the possibility of 'orphan' components which may be covered only by the ASME B&PV Code.

The review of applicable programs shows that all accessible components in the reactor vessel are monitored by at least one existing program. It also revealed that some components are evaluated and monitored more rigorously based on function, safety significance, operational significance, material, susceptibility, and OE.

The EPRI created MRP-274, "Materials Reliability Program: Assessment of Westinghouse and Combustion Engineering Nickel-Based Alloy Orphan Locations,"

which discusses nickel-based orphan locations in the RCS. The discussion focuses on locations in the reactor vessel, fabricated primarily from Alloy 600 and 82/182, which have not been addressed in previous evaluations. Alloys 690 and 52/152 were excluded due to existing programs or studies. Alloy X-750 was excluded due to its limited and specific uses inside the reactor vessel. Upper and lower head penetrations, and inlet and outlet nozzles have been excluded due to existing programs or studies.

MRP-274 mentions applications of Alloy X-750 in Westinghouse plants including CRGT support pins (split pins) and clevis insert bolts. CRGT support pin OE was discussed and the LRSS clevis bolt condition observed at DC Cook Unit 1 was discussed. The report states the following:

"Recent experience has confirmed that there is a need to inspect this component for potential aging degradation and has also confirmed that the ASME Code Section X1 inspection is adequate to detect degradation of these bolts."

The list of austenitic nickel-based material orphan locations in Westinghouse plants from MRP-274 has been recreated in Table 5.

The components in Tables 3, 4, and 5 were compiled and reviewed with additional comments and recommendations in Table 6. The review found the following components which may be susceptible to PWSCC, and have component specific actions to consider:

  • Remaining original Unit 1 LRSS clevis insert bolts

" LRSS lug welds BMI Nozzles and Welds 33

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.2 Several mitigating techniques were mentioned in the report. RCS zinc addition may be a generic method to inhibit PWSCC initiation and propagation in nickel-based alloys.

Surface stress improvements are component specific methods in development or available including, but not limited to, peening and burnishing. Mitigation, repair, or replacement with improved materials is a common option as well.

A rigorous component review is being performed to support NRC commitments related to the use of MRP-227-A in the RVI AMP. This review will include a list of each component in the reactor vessel internals including material. The results of this review from contract 1500016-156 should be reviewed to determine if any additional actions need to be taken for reactor internals aging management.

Risk of Operating with Degraded Clevis Insert Hardware This evaluation determined that there were no safety or operational impacts to D.C. Cook Unit 1 from the degraded clevis insert bolting. It has been documented that continued operation with degraded bolting has potentially significant risk involved. The risk of continued operation with degraded clevis insert hardware is described in Exhibit 25 as follows:

In the case of Unit 2, increased operation with degraded bolting can lead to increased wear on surrounding components. With increased operating times, it is possible that a clevis insert to clevis lug interference fit could be lost, and a new clevis insert would be required to be installed. This would require a significant repair operation that has never been attempted before. Even if the interference fit is not lost, continued operation with degraded bolting could lead to the clevis insert becoming loose and potentially damaging surrounding components. This could lead to future repairs being more extensive than would otherwise be necessary.

While the scenario discussed above would no longer apply to D.C. Cook Unit 1, the risk of loose parts migration must still be considered for any remaining original design bolting in either Unit 1 or 2. As operation continues with degraded clevis insert bolts, the probability that a loose part becomes small enough to enter the fuel assembly would increase. This is primarily because continued operation in the degraded condition will increase the potential for a larger number of loose parts. Additionally, the increased operating time will allow any loose parts that are created to tumble within the reactor internals and potentially break or wear into smaller parts. Thus, the loose parts, which in their nominal condition would typically be trapped by the fuel assembly debris filter bottom nozzles, could be reduced in size enough to pass the nozzle and enter the fuel assembly.

In addition to the risks of potential fuel leaks that continued operation presents, there are also risks associated with core reload activities. As previously discussed, the potential for loose parts to reduce in size and increase in quantity increases with continued operation.

The smaller pieces and larger quantity of loose parts could make foreign object search and retrieval (FOSAR) activities more challenging. As a result, more care and time would 34

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.2 be required to ensure that all parts are retrieved. Should pieces be missed during the FOSAR activities, fuel misalignment issues and additional steps to remove loose parts could result.

Each CNP unit has a Digital Metal Impact Monitoring System (DMIMS) to identify loose parts. The first place that loose parts may be detected is at the reactor vessel lower plenum where there are two sensors. The DMIMS did not detect the Unit 1 LRSS clevis bolt degradation with broken bolt heads vibrating within their installation counterbores.

These sensors are calibrated to detect a loose part as small as 0.25 Ibs, but the largest estimated loose part generated and released into the RCS flow from failed LRSS clevis bolts during operation is 0.039 lbs. Therefore, DMIMS is not expected to detect a degraded LRSS clevis bolt condition, or loose parts generated from such a condition during operation.

A more detailed evaluation on the risks associated with operating with degraded clevis insert hardware can be found in Exhibit 25.

See Sections 1.3 and 1.4 for actions generated as a result of this review.

35

Unit I lRx Vessel Core Support Lug Bohting Anomalies (AR 2010-1804-10) ATTACHMENT 5.2 Table 3: Summary of nickel based components considered in development of R'l AMP guidance documents.

MRP-191' Assembly Sub-Component Material None SCC Wear Fatigue IMT Likelihood Likelihood Unit I Detnils/ MNanagement Unit 2 Details/ Management Assemnbly Conseq. Of ofFailure of Damage Failure L. M, H L. M. H Upper Internals Assembly Control Rod Flexures Component is fabricated from 304 Component is fabricated from 304 Guide Tube SST, screening from this line no SST, screening from this line no Assemblies A X-750 SCC G

H M

applicable applicable and Flow Guide Tube Support Pins A X-750 SCC Wear Fat NONE H

M GT-LRP 2012-1808, Replace in 2016 GT-LRP 2012-1809, Replace in 2016 Downcomers Support pin nuts A X-750 X

NONE GT-LRP 2012-1808, Replace in 2016 GT-LRP 2012-1809, Replace in 2016 Interfacing Components Interfacing Clevis insert bolts EC-51640, Replaced minimum pattern Components A X-750 SCC Wear G

Nt L

in 2010 Clevis insert lock keys (lock bars)

A 600 X

G EC-51640, Partial removal in 2010 Clevis inserts A 600 Wear G

L L

G - Causes significant economic impact Table 4: Summary ofnickel based components inside the reactor vessel considered in the Alloy 600 Program Description Inspection Susceptibility Component Unit Location Material Description Method Requirement Frequency Acceptance Unit Description Alloy Effective Service Susceptibility Criteria Stress (MPaI Temp (F) laden Reactor Vessel 1,2 Core Support Pads Alloy 600 The core support VT-3 ASME Once per Per I

Core Support Pad (at weld) 600 306.8 528.2 2.72E-1 I (LRSS Clevis Lugs and Alloy pads are fabricated Section Xl, interval (per inspection (LRSS Lugs)

Insert) 82/182 from Alloy 600 and Category B-inspection requirement Core Support Pad Weld 82/182 348.1 528.2 2.71E-1 I they are attached to N-2, item requirement)

(LRSS Luo Welds) the reactor vessel B13.60 Core Support Pad 600 29.2 528.2 2.24E-15 using Alloy 82/182 (LRSS Lugs) welds.

2 Core Support Pad (at weld) 600 306.8 544 5.56E-1 I I LRSS Luos)

Core Support Pad Weld 82/182 348.1 544 5.53E-1 I (LRSS Lug Welds)

Core Support Pad 600 29.2 544 4.56E-15 (LRSS Luas)

Reactor Vessel 1,2 Reactor Vessel Internals Alloy 600 The clevis inserts are VT-3 ASME Once per Per WCAP-16198-P states "The clevis inserts (R 6) are bolted within the reactor vessel internals and are fabricated from Alloy Section XI, interval (per inspection at approximately cold-leg temperatures. The only known incident where cracking of this component 600 material.

Category B-inspection requirement was suspected was subsequently found by destructive examination to be limited to cracking of the N-2, item requirement) hard facing on the Alloy 600 insert and not cracking of the insert itself Since this component does BI 3.60 not support tensile loads and does not have a history of failure related to PWSCC, it was not assigned a susce*tibility index."

Reactor Vessel 1,2 BMI Nozzles Alloy 600 The Alloy 600 Bare Metal ASME Code Each Per I

BMI Nozzles 600 455.1 528.2 2.64E-10 Alloy nozzles are connected Visual Case N-722-refueling inspection BMI Nozzle to Guide Tube 82/182 393 528.2 4.39E-1 I 82/182 to the reactor vessel 1, Item outage (only requirement Welds using Alloy 82/182 BR15.80 required 2

BNII Nozzles 600 455.1 544 5.38E-10 partial penetration every other BMI Nozzle to Vessel Welds 82/182 410.9 544 1.07E-10 swelds. The welds refueling connecting the BMI outage)

BMI Nozzle to Guide Tube 82/182 393 544 8.96E-1 I nozzles to the Welds stainless steel guide tubes are also Alloy

_82/182 2 Columns with relevant information from MRP-191. Tables 4-4, 5-1, and 6-5. columns with no information for these components have been omitted.

Susceptibility Index value is used to compare susceptibility of components, more detail can be found in the Alloy 600 Program and supporting documentation.

36

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.2 Table 5: List of Orphan Locations in Westinghouse-Designed PWRs from MRP-274 Component/Subcomponent Material Temperature Core Support Lugs:

Core SLIpport lugs Alloy 600 Tcold Core support lug weld Alloy 182/82 Tcold Clevis Inserts:

Clevis Inserts Alloy 600 Tcold Leak-Off Monitor Tubes:

Leak-off monitor tubes Alloy 600 Mean head temp.

Leak-off monitor tube welds Alloy 182/82 Mean head temp.

37

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.2 Table 6: Comments and Recommendations for Reviewed Components Item Material Comments and Recommendations Clevis Insert Bolts Alloy X-See CATPR - section 1.3 750 Clevis Insert Bolt Lock A 600 Failure of this item is expected to be secondary to bolt Bars failures and do not provide a structural function, therefore no additional actions recommended Clevis Dowel Pins A 600 Failure of this item is expected to be secondary to bolt failures. EHI-5070-ALLOY600 which provides a PWSCC susceptibility screening of alloy 600 components does not list clevis dowel pins, hence an action is recommended to update this procedure to include the dowel pins.

Clevis Inserts A 600 Susceptibility of this component to PWSCC is low as observed in a basis document to the Alloy 600 Program

[Exhibit 99], no known economically viable mitigation solution is available, therefore, no additional actions recommended LRSS Lug A 600 WCAP-16 198-P states "Except for the location immediately adjacent to the weld, the residual stress in the core support pad [LRSS Lug] is negligible."

Therefore, management of the lug would be bounded by management of the lug weld, no additional actions recommended LRSS Lug Weld 82/182 Mitigating actions for this weld may include remote underwater peening, recommend investigating need and availability of options CA# 2010-1804-29 Guide Tube Support Alloy X-Current strategy of replacement is adequate, no Pins 750 additional actions recommended Support Pin Nuts Alloy X-Current strategy of replacement is adequate, no 750 additional actions recommended BMI Nozzles and Welds A 600 Code case N-722-1 calls for a bare metal visual and examination every other refueling outage. CNP 82/182 performs this examination every outage to identify leakage. Leakage was identified on a BMI at Palo Verde in October of 2013. CNP is actively monitoring the industry response to the event at Palo Verde through participation in the EPRI MRP. Recommend CNP evaluate if potential mitigating actions exist to prevent PWSCC of the BMIs. CA# 2010-1804-30 38

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.3 5.3 EVENT TIMELINE I

TknrIne for LRSI Unit I Unit 2 Wdus Gu~am Docmnwm mumbo gh~d IC mwm~dUOllh6 39

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.4 5.4 FAILURE MODES AND EFFECTS ANALYSIS This chart was built from an equipment failure analysis technique. See Potential Failure Mode Evidence Matrix for details.

Unit 1 LRSS Clevis Bolt and Dowel Pin Failure

_DsgnFiue:.>

2.:,aea Seecio 15.talr~a~i~K T: ~itutn.alr

, Design Failure 2

3Material Selection 3-Manufacturing or.

4 nstallation or

5.

Operatio Falure 6

Maintenance Failure

7.

External Condition 1 8.

Sabotage

i Failure,.:. :

F

FaricationFailure;_.

]iC tiction.Fii r :

Induced Failure 1.1. Overload Failure 1.1.1. Overstress 1.1.2. BRoting Sizing 1.1.3. High peak strse due to appied and residual stress 1.2. Low Cycle Fatigue (LCF) 1.2.1. Thermal Cycling 1.2.2. Refuel Cycle 1.2.3. FRow Transients 1.2.4. Pump Issues 1.2.5. Pressure Overload 1.2.6. Irrodation Induced Loss of Preload 1.3. High Cycle Fatigue (HCF) ra3.1. Flow Induced Vibration

.3.1.1.

Pump Pulsation 1.3.1.2.

Lower Interrnas St*rodure Vibration r.3,r.3.

Reactor Vessel Vibraton 1.3.1.4.

Vortex Sheddng 1.3.2. Mechanically induced Vibration 1.3.3. Electrically Induced Vibration (ex. Motor) 1.3.4. Magnetcaly Induced 1.4. Corrosion Failure 1.4.1. Intergranrlar Stress Corronion Cracking (IGSCC) 1.4.1.1. PdmaryWaterSCC (PWSCC) 1.4.2. Irradation Assisted 5CC (IASCC) 1.4.3. Low Temperature Crack Propagation (LTCP) 1.4.4. RPow Assisted Corrosion

(.5. Erosion Failure 1.5.1.

Tow Turbulence 1.5.2. High Plow Velocity 1.5.3. Direct Plow Impingement 1.6. Inadequate Soft Dimensions 1.7. Inadequate Locking Mechanism 1.0. Inadequate Torque Design Value 1.9. Inadequate Tolerance Specifications 1.9.1. Bolt head to shank perpenrdcularity out of square 1.9.2. Core Barrel Hang-Up or Li)

Cocking 1.!.3. ClevisdLug Misalignment 2.1. Bolt Selection of Alloy X-750 2.2. Loading Strength Selection 2.3. Fatigue Strength Selection 2.4. Creep Strength Selection 2.5. Embdlttrement 2.5.1. Irraduaon Ernntitement 2.5.2. Thermal Embritteerent 2.6. Machinabillty 2.7. Void Swelling 2.7.1. Loss of Pretoad tee to Void Swelling 2.7.2. Overload due to Void Swelling 2.8. Improper Material Heat Treatment Specification 3.1. Inadequate Clearance Specifications between Key and Clevis 3.2. Inadequate Machining (Outside of drawings or specifications) 3.2.1. Irprroper Surface Treowent lincludong aer EDM) 3.2.2. Machine/Too]

Marks Creat NotoidDeftctiess Riser (Within drawing specification) 3.3. Inadequate Performance of Material Heat Treatment 3.3.1. Excessive Residual stess In Clams 3.3.2. Excessive Residual Stresw In Lugs 3.3.3. Excessive Residual Stress in bolts 3.4. Inadequate Welding 3.4.1. Inadequate Weld Preheat -

Locking Bar 3.4.2. Inadequate PWHT - Locking Bar 3.4.3. Improper filer Metal - Locking Bar 3.4.4. Inadequate Weld Process Control - Looting Bar 3.4.5. Residual Stress in Lugs from Welding 3.4.6. Residual Stess in Clevis Insert rnm Welting 3.4.7. Distortion in Lugs hrom Welding 3.4.B. Distortion rn Clevis Insert from Weldn 3.5. Bottomed Out Bolting 3.8. Improper Internal Thread Form 3.7. Improper External Thread Form 4.1. Inadequate Torqueing of Bolts

4. 1.1. Over.Torqueing 4.1.2. Under-Tornupeng 4.1.3. Inadequate Torque Paremn 4.1.4.

Torqueing sequence when installing inses with respect to thermro expansion from low temperature exposure 4.2. Inadequate Odentation of Parts 4.3. Inadequate Parts/Wrong Batch of Bolts Used

44. Missing Parts 4.5. Clevis Separation at Installation 4.8. Clevis Deformation at Installation 4.7. Lug Deformation at Installation 4.8. Mechanical Scratch/NickrDefect Caused Dudng Assembly 5.1. Temperature Operating Condition out of Specification 5.2. Mishandling of Internals Due to Human Errom 5.3. Clevies Insert Distortion 5.4. Thermal Expansion Differentials 5.5. Overload on DefuelfRefuel 5.0. Overload on Installation Lower Internats 5.7. Overload on Removal of Lower Intemals 5.9. Overload Across Lower Intemals at Start-Up of Individual Loop RCPs 5.9. Mis-Operation Due to Failure of Another Component 5.9.1. Failure in Thermal Shield 5.9.2. Failure in Core Barrel 5.9.3. Failure of Hold Doon Spdng 5.9.4. Flow Disturbances due to Dislodged Thennal Sleeves 5.9.5. Hold Down Spring Relaxation 5.10. Random Turbulence Excitation 5.11. Thermally Induced Bolt Loss of Preloed 5.12. Irradiation Induced Bolt Loss of Preload 5.13. Inadequate Water Chemistry with Respect to PWSCC 5.14. Inadequate Water Chemistry with Respect to LTCP 5.15. Foreign material Bound in Clevis/Key Clearance 5.16. Configuration of Cold Leg Impingement Inducing Mechanical Loads 5.17. Individual Coolant Loop Flow Imbalance 6.1. Preventive Maintenance 6.1.1. Inadequate Inspection Technique 6.1.2. Inadequate Inspection Frequency 6.1.3. Fafure to Apply OE 6.2. Corrective Maintenance 6.2.1. Inadequate Repair of Barrel Former Boilts 6.3. Distortion from Hydraulic Testing 7.1. Earthquake 7.2. 2009 Turbine Failure Excitation 7.2.1. System Tranrients 7.22. Mechanical Excitation 8.1 Inadequate Security and Surveillance un r im so

-r, -

40

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.5 5.5 SUPPORT REFUTE MATRIX POTENTIAL FAILURE MODE EVIDENCE SUPPORT/REFUTE MATRIX POTENTIAL FAILURE MODE SUPPORTING INFORMATION REFUTING INFORMATION SIR

1. Design Failure No signs of overload at initiation sites of fracture 1.1.

Overload Failure surfaces as seen in Hot Cell Report.

1.1.1. Overstress Stresses met code allowables as seen in the 1.1.2. Bolt Sizing None Minimum Bolting Pattern Analysis 1.1.3. High peak stress due to applied Per Design met codes allowables as seen in the and residual stress Minimum Bolting Pattern Analysis.

No signs of overload at initiation location as seen in Hot Cell Report.

1.2.

Low Cycle Fatigue (LCF) 1.2.1. Thermal Cycling 1.2.2. Refuel Cycle 1.2.3. Flow Transients None No signs of Low Cycle Fatigue on fracture 1.2.4. Pump Issues surfaces as seen in Hot Cell Report.

1.2.5. Pressure Overload 1.2.6. Irradiation Induced Loss of Preload 1.3.

High Cycle Fatigue (HCF) 1.3.1. Flow Induced Vibration 1.3.1.1. Pump Pulsation 1.3.1.2. Lower Internals Structure Vibration None No signs of High Cycle Fatigue on fracture R

1.3.1.3. Reactor Vessel Vibration surfaces as seen in Hot Cell Report.

1.3.1.4. Vortex Shedding 1.3.2. Mechanically Induced Vibration 1.3.3. Electrically Induced (ex. Motor) 1.3.4. Magnetically Induced 4 Support (S), Refute (R), Root Cause (RC), Contributing Cause (C), Potential Contributing Cause (PC) 41

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.5 POTENTIAL FAILURE MODE EVIDENCE SUPPORT/REFUTE MATRIX POTENTIAL FAILURE MODE SUPPORTING INFORMATION REFUTING INFORMATION S/R 1.4.

Corrosion Failure 1.4.1. Intergranular Stress Corrosion Cracking (IGSCC) 1.4.1.1. Primary Water SCC (PWSCC)

IGSCCIPWSCC CN-RIDA-10-27 Appendix A shows localized peak stresses due to stress risers at the head to shank radius far greater then yield stress and recommended stresses to stay below to prevent SCC.

Literature shows Alloy X-750 solution annealed below 1800 degrees F are susceptible to SCC in a PWR environment Exhibit 12 Fracture faces showed signs of IGSCC at initiation site and for nearly full propagation distance (some overload at final failure).

Local Environment for bolts was Primary Water (PW)

IASCC CN-RIDA-10-27 Appendix A shows localized peak stresses far greater then yield stress and recommended stresses to stay below to prevent SCC Literature shows Alloy X-750 solution annealed below 1800 degrees F are susceptible to SCC.

LTCP LTCP can cause existing cracks and sharp surface defects to grow at a rapid rate.

Flow Assisted Corrosion None IGSCC/PWSCC 0

None IGSCC/PWSCC S-RC 1.4.2. Irradiation Assisted SCC (IASCC)

IASCC Not in an area where you would expect IASCC Fluence level significantly lower than threshold for IASCC LTCP LTCP may have made cracks propagate faster, but did not initiate them.

12-thp-6020-chm-100 fig. 1 shows RCS dissolved Hydrogen target bands met recommendations for prevent LTCP starting in 2002

    • Note: These refute LTCP from being a root cause, but these do not refute LTCP from being a potential contributing cause IASCC R

1.4.3. Low Temperature Crack Propagation (LTCP)

LTCP S-PC Flow Assisted Corrosion R

R 1.4.4. Flow Assisted Corrosion Flow Assisted Corrosion Initial Fracture is in a non-flow area 1.5.

Erosion Failure 1.5.1. Flow Turbulence 1.5.2. High Flow Velocity 1.5.3. Direct Flow Impingement 0

Area of high flow 0

Initial Fracture is in a non-flow area Location of cracks would not have seen flow impingement No signs of erosion in metallurgical report.

42

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.5 POTENTIAL FAILURE MODE EVIDENCE SUPPORT/REFUTE MATRIX POTENTIAL FAILURE MODE SUPPORTING INFORMATION REFUTING INFORMATION S/R 0

Radius causes a stress riser that can be reduced by design but not be designed out.

Note: The stresses produced at the head to 0

None for bolt head to shank transition 1.6.

Inadequate Bolt Dimensions shank radius is not sufficient to cause overload Remaining bolt dimensions are adequate to S-C failure on their own, but this was found to be a ensure proper bolt strength.

contributing cause to PWSCC as shown in line 1.4.1.

1.7.

Inadequate Locking 0

None No Locking Bar failures R

Mechanism 1.8.

Inadequate Torque Design a

None 0

Minimum Bolt Pattern supported same preload R

value_______________________

No inadequate tolerances seen in Hot Cell report.

Core Barrel Hang up would only be a concern for overload All 29 bolts failed 0

We saw no overload that would have yielded the 1.9.

Inadequate Tolerance All bolts were cracked therefore measurements bolt.

Specifications could not be accurately taken.

0 The method of manufacturing bolts machines 1.9.1. Bolt head to shank Bolt crack orientations show no particular most external features in one setup on a lathe perpendicularity out of square pattern which suggests that if there was a making misalignment difficult.

R 1.9.2. Core Barrel Hang-Up or LID dimensional defect causing an A-symmetric 6

No signs of overload Cocking load that it would be on the bolts, not on other Pictures from bolt replacement show concentricity 1.9.3. Clevis/Lug Misalignment joint features.

between clevis holes and lug holes 1.9.4. Counter bore Misalignment No pattern to the orientation of fracture surface lines of symmetry.

0 This feature of the clevis inserts was manufactured in a shop and therefore would expect to have a consistent effect on the symmetry of fracture surfaces.

2. Material Selection Failure 0

Replacement material had same chemistry to support thermal expansion coefficient and 2.1.

Bolt Selection of Alloy X-750 0

None strength considerations R

a Alloy X-750 was/is considered the proper bolting material because of strength and general corrosion properties 2.2.

Loading Strength Selection None Alloy X-750 typically selected for high strength R

properties 2.3.

Fatigue Strength Selection None No signs of fatigue on fracture surfaces R

2.4.

Creep Strength Selection None 0

Temperature is too low.

R 2

No evidence of creep in Hot Cell Report.

0 Neutron fluence is 2.2X1OA-3 DPA at the clevis 2.5.

Embrittlement insert bolts, this is below the threshold of 5 DPA 2.5.1. Irradiation Embrittlement None for void swelling R

2.5.2. Thermal Embrittlement a

Alloy X-750 not susceptible to Thermal Embrittlement 43

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.5 POTENTIAL FAILURE MODE EVIDENCE SUPPORT/REFUTE MATRIX POTENTIAL FAILURE MODE SUPPORTING INFORMATION REFUTING INFORMATION S/R 2.6.

Machinability

& None Machinable R

2.7.

Void Swelling 0

Neutron fluence is 2.2X1OA-3 DPA at the clevis 2.7.1. Loss of Preload due to Void insert bolts, this is below the threshold of 5 DPA Swelling

& None for void swelling R

2.7.2. Overload due to Void Swelling 0

No evidence of overload in the initiation sites of fracture faces in Hot Cell Report.

0 Literature shows Alloy X-750 solution annealed 2.8.

Improper Material Heat below 1800 degrees F are susceptible to SCC Treatment Specification in a PWR environment. Note: This support was None S-RC found to be a contributor to PWSCC as shown in line 1.4.1.

3. Manufacturing or Fabrication Failure 3.1.

Inadequate Clearance Specifications between Key None No evidence of overload on bolting R

and Clevis 3.2.

Inadequate Machining a

Hot Cell Report shows no evidence of improper (Outside of drawings or surface treatment specifications) a Hot Cell report has been updated to show the 3.2.1. Improper Surface Treatment radius is within specification.

(including after EDM)

N The bolt head-to-shank interface was within size 3.2.2. Machine/Tool Marks Creat tolerance, but did not follow the intended surface Notch/Defect/Stress Riser profile. However, fractures were not observed to (Within drawing specification) initiate at any suspect areas.

3.3.

Inadequate Performance of Material Heat Treatment No signs of cold working in the microstructure of 3.3.1. Excessive Residual stress in the head to shank radius as seen in Hot Cell Clevis None Report.

R 3.3.2. Excessive Residual Stress in Residual stress is only a concern for the part in Lugs which the residual stress is in.

3.3.3. Excessive Residual Stress in 0

Test showed material behaved per specification.

bolts 3.4.

Inadequate Welding Not a structural weld 3.4.1. Inadequate Weld Preheat -

No preheat Locking Bar NoPWHT 3.4.2. Inadequate PWHT - Locking Bar Residual stress is only a concern for the part in 3.4.3. Improper filler Metal - Locking which the residual stress is in.

3.4.4. Inadequate Weld Process r

No welding on clevis insert beyond locking bar.

Controle-Locking Bar W

None Tacking welding of lock bar does not create R

3.4.5. Residual Stress in Lugs from enough heat to induce distortion.

Welding S Lock bar weld is a sufficient distance from hole 3.4.6. Residual Stress in Clevis Insert and seating surface that there is no concern.

from Welding Holes in the lugs were field drilled following 3.4.7. Distortion in Lugs from Welding welding of the lugs, hence if distortion in the lugs 3.4.8. Distortion in Clevis Insert from existed this would not have an effect on the Welding alignment of holes between the clevis and lugs.

44

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.5 POTENTIAL FAILURE MODE EVIDENCE SUPPORT/REFUTE MATRIX POTENTIAL FAILURE MODE SUPPORTING INFORMATION REFUTING INFORMATION S/R 0

Each bolt hole was inspected during bolt 3.5.

Bottomed Out Bolting 0

None replacement campaign, all bolt holes had R

adequate depth.

0 Improper thread form would lead to failures in Replacement bolts experience galling at threaded region not the under head radius of bolt.

3.6.

Improper Internal Thread external thread crest during installation, major Galling in threaded region would reduce stresses R

Form diameter turned down to LMC on remaining at the head to shank radius if anything.

replacement bolts which solved problem a

No indication of failure in the threads as indicated by the Hot Cell Report.

3.7.

Improper External Thread N

No indication of failure in the threads as indicated Form None by the Hot Cell Report.

4. Installation or Construction Failure All bolts are behaving the same so torque was consistent and following quality inspection procedure 4.1.

Inadequate Torqueing of Bolts 0

Under torqueing would result in fatigue which was 4.1.1. Over-Torqueing not evident in the Hot Cell Report.

4.1.2. Under-Torqueing Insert was immersed in liquid nitrogen for no yiendin the Hot Cell R

eport h

4.12. nde-Toquengshrinking, clevis bolts were torqued to 270-290 0

No yielding of the bolt or overload seen at the 4.1.3. Inadequate Torque Pattern ft.inkin, clevis ware toroom tom270-290, initiation site of the fracture faces as seen in Hot 4.1.4. Torqueing sequence when ft. lbs., the clevis warmed to room temperature, Cell Report.

installing insert with respect to the bolts torqued to 555-575 ft. b OEM implemented a rigorous quality program thermal expansion from low aduring construction to ensure critical parameters temperature exposure were met.

0 Quality review trip reports were provided which showed that deviations were documented and dispositioned appropriately.

4.2.

Inadequate Orientation of None Parts appeared to be in the correct orientation R

Parts 4.3.

Inadequate Parts/Wrong Batch N

Metallurgy matched that expected for heat treat R

of Bolts Used None 4.4.

Missing Parts 0

None 6

No parts were missing R

4.5.

Clevis Separation at 0

The Minimum Bolt Pattern Analysis analyzed the Installation None pre load on the bolts and determined the clevis R

would be pulled up to lug eliminating gap 4.6.

Clevis Deformation at 0

Liquid nitrogen bath to shrink clevis insert for 0

If this was a dimensional installation issue then Installation installation, lock bar tack welding highly unlikely we would be seeing a generic bolt R

failure issues throughout all clevises.

0 If this was a dimensional installation issue then 4.7.

Lug Deformation at Installation 0

Lugs are welded to the RX wall highly unlikely we would be seeing a generic bolt R

failure issues throughout all clevises.

0 The bolt head-to-shank interface was within size 4.8.

Mechanical tolerance, but did not follow the intended surface ScratchlNick/Defect Caused 0

None profile. However, fractures were not observed to R

During Assembly initiate at any suspect areas.

0 No cold working was observed on any bolts or microstructures.

45

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.5 POTENTIAL FAILURE MODE EVIDENCE SUPPORTIREFUTE MATRIX POTENTIAL FAILURE MODE SUPPORTING INFORMATION REFUTING INFORMATION SIR

5. Operation Failures 5.1.

Temperature Operating None No discrepancies found for cold leg temperature R

Condition out of Specification via R time review.

0 Installation of the lower internals is performed in accordance with 12-OHP-4050-FHP-045. The internals lift rig engagement screws, guide bushings, reactor vessel guide studs, vessel 5.2.

Mishandling of Internals Due to alignment keys, lower radial support system keys R

Human Errors Nand clevis inserts all work together to ensure alignment and proper installation of the lower internals into the reactor vessel.

0 No signs of overload at initiation sites of fracture surfaces as seen in Hot Cell Report.

5.3.

Clevis Insert Distortion None 0

Temperature and Operational loads are not R

sufficient to distort the clevis 5.4.

Thermal Expansion N

Thermal expansion differential were analyzed in Differentials Nthe MBPA and was found to be adequate 5.5.

Overload on Defuel/Refuel 0

None a

No signs of overload on fracture faces R

0 No signs of overload on fracture faces Installation of the lower internals is performed in accordance with 12-OHP-4050-FHP-045. The 5.6.

Overload on Installation Lower

& Procedure allows installation of lower internals internals lift rig engagement screws, guide Internals w/o Microdrive with plant manager approval bushings, reactor vessel guide studs, vessel R

alignment keys, lower radial support system keys and clevis inserts all work together to ensure alignment and proper installation of the lower internals into the reactor vessel.

0 No signs of overload on fracture faces a

Removal of the lower internals is performed IAW 12-OHP-4050-FHP-044. To prevent misalignment during disassembly, the difference between 5.7.

Overload on Removal of Lower internals lift rig attachment points is limited to one Internals None or less turns of the engagement screw, which is R

1/4 inch or less of misalignment. For the Unit 1 lower internals removal in 2010, the maximum difference between screw engagement was recorded to be 1/4 turn (10 turns - 9 and 3/4 turns) which is 1/16 of an inch. [Exhibit 1081.

5.8.

Overload Across Lower Internals at Start-Up of None

=

No signs of overload on fracture faces R

Individual Loop RCPs 46

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.5 POTENTIAL FAILURE MODE EVIDENCE SUPPORT/REFUTE MATRIX POTENTIAL FAILURE MODE SUPPORTING INFORMATION REFUTING INFORMATION S/R a

Hot Cell Report indicated an aging degradation 5.9.

Mis-Operating Due to Failure mechanism of Another Component h

No related link between the 3 Former Bolts and 5.9.1. Failure in Thermal Shield

  • U1 has a 304 SS hold-down spring which is Clevis 5.9.2. Failure in Core Barrel susceptible to thermal "ratcheting", leading to 5.9.2. Failure in Hore Ba ring permanent deformation which leads to reduced 0

Per Roy Hall's Inspections 5.9.3. Failure of Hold Down Spring hold down force over time. This has a One indication during 1985 inspection - Indication 5.9.4. Flow Disturbances due to appeared to be a tooling gouge, no other Dislodged Thermal Sleeves moderate likelihood of occurrence indication 5.9.5. Hold Down Spring Relaxation 0

Thermal sleeve OE evaluation performed by K.

Kalchik in Away 0

Turbulence and flow induced vibration was 5.10. Random Turbulence Excitation 0

None analyzed in the Minimum Bolting Pattern and R

found acceptable Bolt preload at the as-installed and hot preload is 5.11. Thermally Induced Bolt Loss described in the Minimum Bolting Pattern Analysis R

of Preload None at the original design conditions and found to be adequate 5.12. Irradiation Induced Bolt Loss Loss of preload is minimal and it does not have of Preload a

None significant adverse effects as described in the R

Minimum Bolting Pattern Analysis 5.13. Inadequate Water Chemistry a

None 0

LER and A-Way searches for chlorides, fluorides, R

with Respect to PWSCC and sulfates returned no relevant results.

5.14. Inadequate Water Chemistry

& See 1.4.4 0

See 1.4.4 with Respect to LTCP 5.15. Foreign Material Bound in Clevis/Key Clearance None No indication of Foreign Material found there R

5.16. Configuration of Cold Leg Impingement Inducing None 0

No evidence of overload fatigue R

Mechanical Loads 5.17. Individual Coolant Loop Flow Imbalance None No A-symmetric failure identified R

6. Maintenance Failure 47

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.5 POTENTIAL FAILURE MODE EVIDENCE SUPPORTIREFUTE MATRIX POTENTIAL FAILURE MODE SUPPORTING INFORMATION REFUTING INFORMATION S/R Inspection would only tell you current condition of bolts A different inspection technique would not decrease the probability of failure and no 6.1.

Preventive Maintenance a

Currently Inspections are only visual and this consequence of LRSS bolt failures on clevis 6.1.1. Inadequate Inspection type of inspection will not find a failure in a bolt system design function occurred Technique unless the head is broke off.

0 No Catastrophic failure X-5 No other plants have experienced LRSS clevis 6.1.2. Inadequate Inspection 0

Significant industry experience with Alloy X-750 bolt failures Frequency failure in RVI components such as split pins bolt failures 6.1.3. Failure to Apply OE (support pins) a Industry guidance indicates that bolt failures do not result in an immediate operability or safety concern.

0 CNP followed industry guidance which was approved by the NRC to perform VT-3 of clevis insert bolts.

No Maintenance has ever been done on Clevis 6.2.

Corrective Maintenance 0

Method of Former Bolt failure inadequate to Insert bolts 6.2.1. Inadequate Repair of Barrel determine all failures 0

A-symmetric failure for former bolts and symmetric R

Former Bolts 0

No clear way to determine bolt failure failure for clevis bolts 0

No related link between the 3 Former Bolts and Clevis 6.3.

Distortion from Hydraulic None 0

No deformation or overload R

Testing

7. External Condition Induced Failure Cook has had 3 small earthquakes on record. All had a severity considerably less than the design OBE The Unit 1 turbine failure caused a transient that 7.1.

Earthquake would be similar to an earthquake, but the loads 7.2.

2008 Turbine Failure would have been considerably less than the Excitation None design OBE R

7.2.1. System Transients A missile block was dropped during a Unit 2 7.2.2. Mechanical Excitation refueling outage, but this would not have caused adverse effects on reactor vessel internals based on information found in AR 00124109 No large-scale flooding of equipment has occurred at Cook Ul

8. Sabotage 8.1.

Inadequate Security No access to component Surveillance Seuiyen*

Bolts were manufactured to Westinghouse R

None

_specifications which was confirmed by Hot Cell 48

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.6 5.6 WHY STAIRCASE No observed CNP UMk I EDY' fakwee of LRSS below whero OE chrAs k"On bob showed fakne globogy to dift for same meletief ping ri"n"fnem, I

0 IDY-Effective Dep*4*tion Years is the calculated vaue to nornalize operational time and temperature for expected degradation compariwso of cono*pn-t between tests and plant data 49

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.7 5.7 INDUSTRY GUIDANCE REVIEW FOR CLEVIS INSERT BOLTS A summary of information gained from a review of industry guidance documents on the aging management of reactor vessel internals components with respect to clevis insert bolts is provided below. The EPRI MRP (Materials Reliability Program) documents were developed in a hierarchy fashion building off of each other to obtain a final inspection and evaluation guidelines for managing long-tenn aging of reactor vessel internal components of pressurized water reactors as documented in MRP-227-A. It should be noted that beyond the summary statement for each document, specifics provided below are based off statements in each document related to Alloy X-750 clevis insert bolts only.

1. March 2001: WCAP-14577, "License Renewal Evaluation: Aging Management for Reactor Internals"
a. Summary: Evaluates aging of the reactor internals components to ensure that intended functions will be maintained during an extended period of operation. Endorsed by tile NRC.
b. Heat treatment employed could result in a material susceptible to PWSCC.
c. Fluence, temperature, and stresses are lower for clevis insert bolts than support pins
d. No clevis bolt degradation or cracking reported in any Westinghouse plants to date
e. Beyond bolt failures additional components would have to fail to compromise clevis integrity
f. Effects of PWSCC of the clevis insert bolts are not significant
g. Credits continuation of the ISI VT-3 inspection
h. Degradation can be detected before function is compromised - pg 129
2. December 2005: MRP (EPRI Materials Reliability Program)-156, "Pressurized Water Reactor Issue Management Table, PWR-IMT Consequence of Failure"
a. Summary: Provides initial input to address the consequences of failure for the identified components in reactor coolant systems for operating US PWRs designed by Babcock & Wilcox, Combustion Engineering, and Westinghouse.
b. Consequence of failure for clevis insert bolts and lock keys were identified as level G likely due to potential effect on clevis insert.
c. Level G: Causes a significant economic impact. Significant events are those for which we do not have a proven fix and would result in significant regulatory and/or public scrutiny, such as first-of a-kind consideration would be a suitable test. It can be 50

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.7 considered that "non-significant" events are those for which it is expected that a proven fix exists that will require minimal regulatory and/or public scrutiny.

3. November 2006: MRP-191, "Screening, Categorization, and Ranking of Reactor Internals Components for Westinghouse and Combustion Engineering PWR Design
a. Summary: Screened, categorized, and ranked Westinghouse and CE designed PWR internals components based on probability of occurrence of a degradation mechanism, probability of failure, and the severity of any consequences with respect to operation, safety, financial impact and plant reliability with respect to 60 years of operation.
b. Provided FMECA (Failure modes, effects, and criticality analysis) rankings based on failure likelihood of occurrence and consequence of failure with respect to safety, reliability, and economic risk. Clevis insert bolts received a I out of a 1-3 ranking with one being the lowest.
c.

Provided an initial categorization of components to either category A, B, or C:

i. Category A: Component items for which aging effects are below the screening criteria. Additional components may be ultimately categorized as A as discussed in B below.

ii. Category B: Defined as those component items that are above screening levels but not 'lead' components. Aging degradation significance is moderate. May require additional evaluations to be shown tolerant of the aging effects with no loss of functionality. If it is further concluded that the existing 10 year in-service inspection or other in-place aging management plans are sufficient to preclude a safety, reliability, or financial concern, such components can be reassigned as Category A

1. Clevis insert bolts initially categorized as B based on SCC and Wear.

iii. Category C: 'Lead' component items for which aging effects are above screening levels. Aging degradation significance is high or moderate. Enhanced/augmented inspections and/or surveillance sampling typically may be warranted to asses aging affects and verify component item functionality.

d. Components that were a FMECA group. I were binned into either A or B components. Components with a moderate probability of failure were placed in Category B. (page 131)
e. Consequence of Failure: Category G = Significant economic impact
4. December 2008: MRP-232, "Aging Management Strategies for Westinghouse and Combustion Engineering PWR Internals" [Exhibit 88]

51

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.7

a. Summary: Summarizes the aging management strategy development for Westinghouse and Combustion Engineering (CE) reactor internals. Provides the technical basis for the aging management requirements of Westinghouse and CE reactor internals in MRP-227.
b. Final disposition on LRSS clevis insert bolt was a re-categorization from B to A which required a determination that there was 'no credible damage issue' for the component (page 36). Reclassification appeared to be based on low consequence of failure, not based on probability of stress corrosion cracking.
c. All B and C components then defined into a program group. Clevis insert bolts were binned into existing programs in which generic or plant-specific programs are capable of managing aging effects. Clevis insert bolts are currently inspected as part of the ten year ISI inspection.
5. December 2011: MRP 227-A, "Pressurized Water Reactor Internals Inspection and Evaluation Guidelines"
a. Summary: Provides inspection and evaluation guidelines for managing long-term aging reactor vessel internal components of pressurized water reactors reactor internals. Endorsed by the NRC.
b. Reiterates category A for stress corrosion cracking and management by existing programs.
c. Recommends inspections of the clevis insert for loss of material or wear per the ASME Code Section XI, In Service Inspection utilizing a visual (VT-3) examination looking at all accessible surfaces at specified frequency.

52

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.8 5.8 ORGANIZATIONAL AND PROGRAMMATIC FAILURE MODE REVIEW Organizational & Programmatic Failure Modes Review O&P Did the evaluation identify any Organizational & Programmatic Applicable Failure Failure Mode Failure Mode Definition Failure Modes? Yes/No (Y/N)

Corrective Mode Y/N Basis Actions Organizational Structure (01)

Inadequate organizational structure, span of control, or levels within the organization.

No The issues reviewed in this evaluation did not provide any evidence regarding inadequate organizational structure, span of control, or levels within the organization.

N/A Organizational Inadequate teamwork or communication The issues reviewed in this evaluation did not provide any Teamwork No evidence regarding inadequate teamwork or communication N/A (06) within the organization, within the organization.

Rewards /No rewards or sanctions were reviewed as a part of this Sanctions Inappropriate rewards or sanctions.

No evaluation.

N/A (07)

Accountability The issues reviewed in this report found no lack of commitment N/A (0)Lack of commitment or accountability.

No N/Acontbliy (08.)

or accountability.

Engagement Lack of employee engagement.

No The issues reviewed in this report found no lack of employee N/A (09) engagement.

Empowerment Lack of employee empowerment.

No The issues reviewed in this report found no lack of employee N/A (010) empowennent.

Employee The issues reviewed in this report found no lack of employee Development Lack of employee development.

No development.N/A (011)

The issues reviewed in this report found evidence of knowledge management weaknesses. All evidence shows that the clevis Knowledge Knowledge management weaknesses; insert bolts were known to have a potential for failure, but N/A Management including insufficient management of key No because of the low significance of failure and no history of (012) organization knowledge or skills.

failure in the industry, no contingencies were in place in the event of a failure. This was the industry standard.

Staffing Insufficient staffing or resources.

No The issues reviewed in this report found no evidence of N/A (OP3a)

In c

s n

r insufficient staffing or resources.

a)

C) 0)

C) 0)

C Decision Making (O5a)

Weak organizational decision making.

No The issues reviewed in this report found no evidence of weak organizational decision making.

N/A 53

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.8 Organizational & Programmatic Failure Modes Review O&P Did the evaluation identify any Organizational & Programmatic Applicable Failure Failure Mode Failure Mode Definition Failure Modes? Yes/No (Y/N)

Corrective Mode Y/N Basis Actions Program Definition (P2)

Insufficient program definition or scope.

No The issues reviewed in this report found no evidence of insufficient program definition or scope.

N/A Program Insufficient, excessive, or conflicting The issues reviewed in this report found no evidence of Requirements program requirements.

No insufficient, excessive, or conflicting program requirements.

N/A (P3)

Program Lack of commitment to program The issues reviewed in this report found no evidence of a lack Commitment implementation.

No of commitment to program implementation.

N/A (OPI)imlmnao.

The issues reviewed in this evaluation supported the 10 year ISI VT-3 examination as the appropriate periodicity and inspection based on the low safety significance of a clevis bolt failure.

WCAP-14577 Rev 1-A, March 2001, states, "the effects of Monitoring Inadequate program monitoring, evaluation, PWSCC of the clevis insert bolts are not significant". The (OP2) or management.

recently revised (since the D.C. Cook clevis insert bolt failure)

MRP-227-A, MRP-227 Roadmap, shows the clevis insert bolts are considered to have a low likelihood of damage, medium likelihood of failure, and a consequence of failure being only a significant economic impact.

Program to Program Weaknesses or lack of interface between No The issues reviewed in this report found no evidence of N/A Interface programs.

weaknesses or lack of interface between programs.

(PP4)

Organization to The issues reviewed in this report found no evidence of lack of Program Lack of organizational authority or No organizational authority or engagement for program N/A Interface engagement for program implementation.

irplmenation.

(OP4) implementation.

E)

E) 2)

MS Organization to Organization Interface (003)

Lack of interface between two organizational groups for a program.

No The issues reviewed in this report found no evidence of a lack of interface between two organization groups for a program.

N/A 54

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.8 Organizational & Programmatic Failure Modes Review O&P Did the evaluation identify any Organizational & Programmatic Applicable Failure Failure Mode Failure Mode Definition Failure Modes? Yes/No (Y/N)

Corrective Mode Y/N Basis Actions Change Weaknesses in implementation of people, The issues reviewed in this report found no weaknesses in Management processes, or equipment changes.

No implementation of people, processes, or equipment changes.

N/A (J)II 55

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.9 5.9 SAFETY CULTURE IMPACT REVIEW Safety Culture Impact Review NRC Components of Safety Culture Did the evaluation identify any of the Reference corrective Safety Safety following Cross-Cutting Area Components?

actions addressing any Culture:

Culture:

Safety Culture:

Yes/No (Y/N) acty address-Cross-Cross-Cross-Cutting Area Document the basis for the safety culture Cutting Area C

ros Cutting Area Component Definition Cross-Cutting Area Component as related or Component noted as Cutting Area Cmoetnot.

Cmoetntda Y/N Basis yes.

Decision CNP decisions demonstrate This evaluation found no crosscutting Makion that nuclear safety is an No issevaluat foend nucrosafety.

N/A Making overriding priority.

issues that affected nuclear safety.

CNP ensures that personnel, The evaluation did not identify any CNPa equmensetht poedresonn challenges to nuclear safety. No Human equipment, procedures, and personnel, equipment, procedure, or N/A Performance Resources other resources are available No other resource issues that would impact and adequate to assure nuclear safety were identified by this nuclear safety.

evaluation.

CNP plans and coordinates No work control issues were identified in Work Control work activities, consistent No this evaluation that had any impact on N/A with nuclear safety.

nuclear safety.

No work practice issues were identified Work Personnel work practices No by this evaluation that had any impact on N/A Practices support human performance.

nuclear safety.

56

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.9 Safety Culture Impact Review NRC Components of Safety Culture Did the evaluation identify any of the Reference corrective Safety following Cross-Cutting Area Components?

Safely Culture:

Safety Culture:

Yes/No (Y/N) actions addressing any Cre:

Cross-Cross-Cutting Area Document the basis for the safety culture safety culture Cross-Cros-Aea Cmpoent s rlate orCutting Area C

rea Cutting Area Component Definition Cross-Cutting Area Component as related or C

ingnoted as Cutting Area Component not.

Y/N Basis yes.

CNP ensures that issues potentially impacting nuclear safety are promptly Corrective identified, fully evaluated, No issues were identified by this Action and that actions are taken to No evaluation regarding inadequate us of the N/A Program address safety issues in a Corrective Action Program.

timely manner, Problem commensurate with their Identification significance.

and CNP uses OE information, Resolution icuigvno Oper g

including vendor No issues were identified by this intrating rernlmenerates lsn s

No evaluation regarding the use of Operating N/A Experience internally generated lessonsExperience.

learned, to support plant safety.

CNP conducts self-and Self&

independent assessments of No issues were identified by this Independent their activities and practices, No evaluation regarding self and N/A as appropriate, to assess Assessment performance and identify independent assessments.

areas for improvement.

57

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.9 Safety Culture Impact Review NRC Components of Safety Culture Did the evaluation identify any of the Reference corrective Safety Safety following Cross-Cutting Area Components?

actions addressing any Culture:

Culture:

Safety Culture:

Yes/No (Y/N) acty address Cross-Cross-Cross-Cutting Area Document the basis for the safety culture safety culture Cross-Cross-Cutting Area Cutting Area Cutting Area Component Definition Cross-Cutting Area Component as related or Component noted as Component not.

Y/N Basis yes.

An environment exists in which employees feel free to This evaluation found no crosscutting raise concerns both to their Environment management and/or the NRC issues that affected nuclear safety. There for Raising without fear of retaliation, No were no indications that employees have N/A Safety Concerns and earlof re any concerns with reporting problems or Conscious Cocrs and employees arerasncoen.

Work encouraged to raise such raising concerns.

Environment concerns.

Preventing, A policy for prohibiting This evaluation found no crosscutting Detecting, and harassment and retaliation issues that affected nuclear safety. The Mitigating for raising nuclear safety No existing policy prohibiting harassment N/A Perceptions of concerns exists and is and retaliation for raising nuclear Retaliations consistently enforced.

concerns continues to exist.

Accountability CNP Management defines the line authority and responsibility for nuclear This evaluation found no crosscutting Other safety No issues with nuclear safety accountability.

N/A Continuous CNP ensures that a learning This evaluation found no crosscutting Learning environment exists.

No issues with CNP ensuring a continuous N/A Environment learning environment.

58

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.9 Safety Culture Impact Review NRC Components of Safety Culture Did the evaluation identify any of the Reference corrective Safety Safety following Cross-Cutting Area Components?

actions addressing any Culture:

Culture:

Safety Culture:

Yes/No (Y/N) safety culture Cross-Cre:

Cross-Cross-Cutting Area Document the basis for the safety culture safty c reC Cross-Cutting Area Cutting Area Component Definition Cross-Cutting Area Component as related or Component noted as Cutting Area Component not.yes.

Y/N Basis Other CNP Management uses a Organizational systematic process for Change planning, coordinating, and Management evaluating the safety impacts This evaluation found no crosscutting of decisions related to major is evalu ation al chang N/A changes in organizational No issues with organizational change structures and functions, management.

leadership policies, programs, procedures, and resources.

Other Safety CNP's safety policies and Policies related training establish and This evaluation found no crosscutting reinforce that nuclear safety No issues with CNP's safety policies.

N/A is an overriding priority.

59

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.9 Safety Culture Impact Review INPO 8 Principles of a Strong Nuclear Safety Culture Did the evaluation identify any of the following Reference corrective Safety Culture:

Component Definitions as issues? Yes/No actions addressing any (Y/N) safety culture Cross-Document the basis for the safety culture Cutting Area Component Definition Cross-Component definition as related or not.

Component noted as Y/N Basis yes.

Everyone is personally responsible for nuclear safety.

At CNP, responsibility and authority for nuclear safety are This evaluation found no crosscutting defined and clearly understood. Reporting relationships, N

issues that affected nuclear safety N/A positional authority, staffing, and financial resources support culture.

nuclear safety responsibilities. Corporate policies emphasize the overriding importance of nuclear safety.

Leaders demonstrate commitment to safety.

At CNP, executive and senior managers are the leading advocates of nuclear safety and demonstrate their commitment This evaluation found no crosscutting both in word and action. The nuclear safety message is N

issues that affected nuclear safety N/A communicated frequently and consistently, occasionally as a culture.

stand-alone theme. Leaders throughout the nuclear organization set an example for safety.

Trust permeates the organization.

At CNP, a high level of trust is established in the organization, This evaluation found no crosscutting fostered, in part, through timely and accurate communication.

N issues that affected nuclear safety N/A There is a free flow of information in which issues are raised culture.

and addressed. Employees are informed of steps taken in response to their concerns.

60

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.9 Safety Culture Impact Review INPO 8 Principles of a Strong Nuclear Safety Culture Did the evaluation identify any of the following Reference corrective Safety Culture:

Component Definitions as issues? Yes/No actions addressing any (Y/N) safety culture Cross-Document the basis for the safety culture Cutting Area Component Definition Cross-Component definition as related or not.

Component noted as Y/N Basis yes.

Decision-making reflects safety first.

At CNP, personnel are systematic and rigorous in making decisions that support safe, reliable plant operation. Operators This evaluation found no crosscutting are vested with the authority and understand the expectation.

N issues that affected nuclear safety N/A when faced with unexpected or uncertain conditions, to place culture.

the plant in a safe condition. Senior leaders support and reinforce conservative decisions.

Nuclear technology is recognized as special and unique.

At CNP, the special characteristics of nuclear technology are This evaluation found no crosscutting taken into account in all decisions and actions. Reactivity N

issues that affected nuclear safety N/A control, continuity of core cooling, and integrity of fission culture.

product barriers are valued as essential, distinguishing attributes of the nuclear station work environment.

A questioning attitude is cultivated.

At CNP, individuals demonstrate a questioning attitude by challenging assumptions, investigating anomalies, and considering potential adverse consequences of planned This evaluation found no crosscutting actions. This attitude is shaped by an understanding that N

issues that affected nuclear safety N/A accidents often result from a series of decisions and actions N

uisues that reflect flaws in the shared assumptions, values, and beliefs of the organization. All employees are watchful for conditions and activities that can have an undesirable effect on plant safety.

61

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.9 Safety Culture Impact Review INPO 8 Principles of a Strong Nuclear Safety Culture Did the evaluation identify any of the following Reference corrective Safety Culture:

Component Definitions as issues? Yes/No actions addressing any (Y/N) safety culture Cross-Definition Document the basis for the safety culture Cutting Area Component Cross-Component definition as related or not.

Component noted as Y/N Basis yes.

Organizational learning is embraced.

For this root cause evaluation, extensive At CNP, OE is highly valued, and the capacity to learn from searches of Operating Experience (OE) experience is well developed. Training, self-assessments, were performed. No OE was found that, corrective actions, and benchmarking are used to stimulate N

if acted upon by CNP, would have N/A learning and improve performance.

prevented the issue. No issues were identified by this evaluation regarding self and independent assessments.

Nuclear safety undergoes constant examination.

At CNP, oversight is used to strengthen safety and improve This evaluation found no crosscutting performance. Nuclear safety is kept under constant scrutiny N

issues that affected nuclear safety N/A through a variety of monitoring techniques, some of which culture.

provide an independent "fresh look."

62

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.10 5.10 EQUIPMENT RELIABILITY REVIEW Was the root or Apparent Cause of the equipment failure determined? Yes M No --

r-M Equipment is verified as Run-to-failure and cause determination is not required. (Do not complete the rest of the form)

F-- ER process cause was not determined (External event, Historical, Unknown)(AP-913 PCND)

Comments:

N/A

1. Equipment Reliability Classification Look up the criticality classification of the equipment involved in the Corrective Action Document.

Is the classification of the component correct? Yes ] No M

[

Incorrect classification (AP-913 INCORT)

Li Not Classified (AP-913 NTCLD)

Comments/Corrective Actions:

The LRSS clevis bolts are piece parts, not a piece of equipment. Therefore, they are not classified as critical, non-critical, or run-to-failure. The Corrective Action Document lists the Reactor Vessel, I -OME-1, as Critical, which is correct for the Reactor Vessel.

2. Performance Monitoring Is the System Monitoring Plan and predictive maintenance performed on the equipment adequate?

Yes E] No E] NA M nl Monitored scope inadequate (e.g., levels, temp, pressures, Vibration) (AP-913 PMSLTA)

L-- Monitoring frequency not appropriate (AP-913 PMFNA)

F-l Monitoring execution less than adequate (PMELTA) o Is the monitoring and threshold for action adequate?

o Is there improvement needed in collecting or trending the data?

Comments/Corrective Actions:

Performance monitoring of the LRSS clevis bolts is not performed because they are a structural component. However, inspection of the internals components is governed by the Reactor Vessel Internals Program Plan for Aging Management of Reactor Internals at D.C. Cook Nuclear Plant Unit I (WCAP-17300-NP).

3. Preventive Maintenance (PM)

Is PM program adequate? Yes -] No E NA [;

El PM did not exist (AP-913 PAITDNE)

El PM frequency not appropriate (AP-913 PMFNA)

F-- PM task content not appropriate (or less than adequate) (AP-913 PMTCNA)

F-] PM template/basis less than adequate (AP-913 PMTTBL) 63

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.10 F1-PM execution less than adequate (AP-913 PMELTA)

R PM feedback not implemented (AP-913 PMTFNI)

Comments/Corrective Actions:

A PM does not exist for the LRSS clevis bolts. However, inspection of the internals components is governed by the Reactor Vessel Internals Program Plan for Aging Management of Reactor Internals at D.C. Cook Nuclear Plant Unit 1 (WCAP-17300-NP).

4. Work Practices Are the maintenance practices and behaviors appropriate and acceptable? Yes [-] No L-i NA M E] Work planning, instruction, or preparation less than adequate (AP-913 WPIPLA)

FD PMT not performed or PMT less Than adequate (AP-913 PMT)

R Work activities incorrectly performed (AP-913 WAIP)

Comments/Corrective Actions:

Failures were not found to be the result of maintenance practices or behaviors. All possible causes in this category were refuted.

5. Design /Operation Is the design of this component appropriate for the application? Yes -] No [

Are the operating procedures and practices appropriate? Yes Z No -] NA[-]

M Original design less than adequate - Component not appropriate for its configuration/application (AP-913 ORIG)

L-Design change less than adequate - Component not appropriate for its configuration/application (AP-913 CHANGE)

R-Equipment was not operated within design (AP-913 OPSNOWD)

Comments/Corrective Actions:

The original design of the LRSS clevis bolts was determined to be less than adequate based on the bolts failing by PWSCC. The original design used an Alloy X-750 bolt in a heat treatment unknowingly susceptible to PWSCC. Additionally, all of the bolts failed in the head to shank transition due to the localized high stresses in this region of the bolt. The replacement bolts utilized an Alloy X-750 bolt with a less susceptible heat treatment and an improved head to shank design transition to reduce stress in this region.

6. Manufacturer/Vendor Quality, Procurement, Shipping, or Storage Are parts availability and quality adequate? Yes -- No --] NA M

-- Vendor quality or workmanship issues (manufacturing defects) (AP-913 VQWI)

F1-Procurement less than adequate (ex. Specification, Equivalence) (AP-913 PLTA)

E] Receipt, Inspection, and Storage less than adequate (ex. Environment, Shelf Life, Control of Scavenged Parts, Storage PM) (AP-913 RISLA)

Comments/Corrective Actions:

LRSS clevis bolts installed at the Cook Nuclear Plant were original equipment manufacturer (OEM) components. Replacement bolts are not typically procured or stored.

64

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.10

7. Previous Corrective Action hiplementation Was corrective action to previous similar problems adequate? Yes E-No [-] NA M

- Previous corrective actions less than adequate or untimely OE use less than adequate Comments/Corrective Actions (if inadequate):

Degradation of LRSS clevis bolts had not been observed at Cook Nuclear Plant or elsewhere in the industry prior to this event.

8. Long Range Plan (Obsolescence/Life Cycle Management)

Is the Long Range Plan adequate? Yes M No L-i NA [

Aging / obsolescence concern, Asset Management/LCM Plans less than adequate (AP-913 AOCAML)

F Previous Business Plan related items not implemented, untimely, or deferred (.4P-913 BPNIUD)

Comments/Corrective Actions:

The Reactor Vessel Aging Management Program has been developed, but has not been completely implemented. This is a new program required for license renewal with a governing document that was completed in February 2011. Procedures must be updated and items must be put in the long range plan for inspections.

9. Other Is configuration management complete and accurate for this CAP product? Yes [:] No Z NA D

Is the equipment referenced in ActionWay correct? Yes N No I-] NA 1-Comments/Corrective Actions:

N/A

10. Manufacture/Vendor Quality Check Is there a concern with the quality of parts, shipping or handling? Yes [-I No Z NA -

Comments/Corrective Actions:

Results of failure analysis indicate the failed bolts met the requirements of the OEM specification.

11. Problem/Issue Management Review Have previous issues not been adequately addressed including but not limited to aging, obsolescence, chronic problem, scheduling, or business planning? Yes [-] No - NA Z
12. Unknown or Different Cause Did the equipment fail due to an unknown cause or other cause than listed in steps 1 through 11 above? Yes -] No M NA--1 65

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.11

11 ADDIT WnAIpT P W

Type Doc.#

Title/Description Review Comments External 46 - OE 99 -

In-Service Failure Of During a refueling outage in 1999 at Nine Mile Point Unit 1, they found Alloy 009452 Alloy X-750 Cap X-750 3/8" cap screws broken during a visual inspection of a core shroud repair Screw On Core tie rod assembly.

Shroud Repair Tie Rod Assembly Causes: IGSCC in conjunction with large, sustained differential thermal expansion stress due to fastening of dissimilar materials with the cap screw.

External 53 - OE19374 Failed RV Upper During the Fall 2004 Farley Unit I reactor vessel upper internals guide tube Internals Guide Tube support (split) pin replacement project, six (6) Alloy X-750 reactor vessel upper Support Pins internals guide tube support pins failed due to Primary Water Stress Corrosion Cracking (PWSCC).

Causes: Primary Water Stress Corrosion Cracking (PWSCC)

External 50 - SOER 84-5 Bolt Degradation or Since 1974, an increasing number of bolt failures in reactor coolant systems have Failure in Nuclear been reported at nuclear power plants. Many examples of Alloy X-750 material Power Plants failure due to stress corrosion cracking.

External 51 - OE1 1453 Reactor Internal Split During a Refueling outage on September 21, 2000, McGuire Unit 2, personnel Pin Failures discovered three cracked and broken guide tube support pins during an inspection. The inspection was the result of industry operating experience of similar support pin failures.

Cause: Pure Water Stress Corrosion Cracking (PWSCC) after extended operation of Alloy X-750 External 54 - NUREG-Generic Aging The Generic Aging Lessons Learned (GALL) Report from the NRC identifies 1801_R2 Lessons Learned the clevis insert bolts as having a possible aging effect/mechanism as loss of (GALL) Report material due to wear. This report supports the use of EPRI MRP-227 for managing the aging effects of clevis insert bolts.

External 47 -

Control Rod Drive Since 1978, several failures of the control rod drive (CRD) Guide Tube NRC_1N82-29 (CRD) Guide Tube Support Pins have occurred. The material is Alloy X-750, which, depending on Support Pin Failures the manufacturer and the fabrication date, has been solution heat-treated and age at Westinghouse hardened at various temperatures and for various times.

66

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.11 Type Doc.#

Title/Description Review Comments PWRS Cause: Stress Corrosion Cracking (SCC)

External 56 - OE25769 Five Control Rod On October 15, 2007 during a refueling outage, Braidwood five (5) support pins Guide Tube Support were found to have failed. All five support pins came out of the Upper Internals Pins found failed in one piece; however, each had a leaf break off in the support pin removal during planned station.

replacement activities Cause: Pressurized Water Stress Corrosion Cracking (PWSCC) of the Alloy X-750 pins.

External 63 - EAR PAR Corrosion in Control On 28-03-06,. Vandellos II had an unplanned shutdown because of indications of 06-047 Rod Guide Tube loose parts detected possibly in the Steam Generator.

Support Pins Cause: Primary Water Stress Corrosion Cracking (PWSCC) and also to the phenomenon of Irradiation-Assisted Stress Corrosion Cracking (IASCC)

External 52 - OE13865 Unplanned Shutdown On May 13, 2002, Wolf Creek had an unplanned Shutdown because of an due to Loose Part in indication that there was a loose part in their Steam Generator. After shutting Steam Generator down from 100% power a control rod guide tube split pin nut and locking device was found in the Steam Generator.

Cause: Primary Water Stress Corrosion Cracking (PWSCC)

External 57 - OE7883 Shutdown because of On May 29, 1996, Vogtle Unit 1 control room personal detect a loose part in the detection of Foreign Steam Generator. This is the same scenario that Wolf Creek (above) went Object in Steam through in May of 2002. The part was also a nut from a control rod guide tube Generator support pin. Another part was found lodged in the tube sheet and others were not located.

Cause: Primary Water Stress Corrosion Cracking (PWSCC)

Internal GT 2012-1808 /

D.C. Cook CRGT The original material of the split pins was Alloy X-750 with a low temperature 2012-1809 split pin GTLRP heat treat. These pins were susceptible to primary water stress corrosion cracking (PWSCC) and failed. Split pins were replaced by Areva in 1985 and 1986 with new Alloy X-750 split pins with improved stresses and a high temperature heat treat. These improved features reduced susceptibility to PWSCC, but did not 67

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.11 Type Doc.#

Title/Description Review Comments eliminate the problem. The comparison of the split pin failure to the clevis bolts was discussed by Westinghouse in WCAP-14577 Rev 1-A, March 2001. While it was known that the Alloy X-750 clevis bolt material was susceptible to PWSCC, the failure was not considered likely due to difference in fluence.,

temperature, and stresses between the two bolts.

Internal AR 2010-10940 D.C. Cook Unit 1 This was the root cause evaluation for the core baffle bolt failures in the Unit 1.

core baffle bolt reactor vessel. The root cause of failed baffle-former bolts was Irradiation failures Assisted Stress Corrosion Cracking (IASCC) in conjunction with thermal and irradiation induced loss of preload in several baffle-former bolts. IASCC was considered in the Support/Refute analysis for this evaluation.

68

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 5.12 DOCUMENTS REVIEWED All Documents are in the directory: \\\\cnp 10 1\\Engineering\\Root Cause\\LRSS Root Cause\\LRSS Exhibits\\

Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Report 1 - WCAP-12527 Connecticut Yankee Thermal Shield Removal 11/14/13 pg225wordCaptureError.pdf Westinghouse-The purpose of this report is to address concerns of removing the Thermal Westinghouse Document WCAP-12527, Revision 0, Connecticut Shield at Connectiut Yankee Nuclear Reactor Yankee Thermal Shield Removal Program Report. April 1990.

External OE 2 - 0E2570 Thermal Shield Support Failure and Repair 11/14/13 Connecticut Yankee.pdf INPO - Thermal Shield Support Failure and INPO Document 0E2570, Thermal Shield Support Failure and Repair Connecticut Yankee Repair. April 1988.

External OE 3 - Ringhals OE Lower Radial Support wear 2013.pdf 11/14/13 Vattenfall - Ringhals Lower Radial Support Nilsson, P., Internals Support Wear. Presentation by Vattenfall to wear 2013 the Electric Power Research Institute. October 28,2013.

E-mail 4 - wiIsonEmaiI20120924.pdf 11/14/13 Westinghouse - Bryan Wilson's response to E-mail from Bryan Wilson, Westinghouse, to Kevin Kalchik, MPR Review of LRSS Clevis Insert Bolting American Electric Power, RE: MPR Review of LRSS Clevis Insert Analysis Bolting Analysis. September 2012.

Report 5e - DC Cook Clevis Bolts FINAL Report 12-19-13.pdf 11/14/13 B&W - This is an updated report covers laboratory examinations performed by Babcock & Wilcox Technical Services Group Report S-1473-002, Babcock & Wilcox Technical Services Examination of Clevis Bolts Removed from D. C. Cook Nuclear Group (B&W TSG) on failed clevis bolts Plant. December 2013.

removed from D. C. Cook Unit 1, Hot Cell Report. - Final Report 69

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Info Gram 6 - IG-10-1.pdf 11/14/13 Westinghouse - Info Gram from Westinghouse -oIf Cos Reamfom nenl Westinghouse Document IG-10-1, Revision 0, Reactor Internals Westinghouse of Cooks Reactor internals lower radial support clevis insert cap screw Lower Radial Support Clevis Insert Cap Screw Degradation.

degradation March 2010.

degradation Video file 7 -Core Support Lugs.mpg 11/15/13 CNP Document DC Cook UlC23 10 Year ISI, Engineering Programs, DC Cook Outage U1C23 10 Year ISI Disk 1 of 4, WesDyne International-1 DVD, August 2010.

CNP Document DC Cook UlC23 10 Year ISI, Engineering Programs, DC Cook Outage U1C23 10 Year ISI Disk 2 of 4, WesDyne International-1 DVD, August 2010.

Video of ULC23 VT Inspection CNP Document DC Cook U1C23 10 Year ISI, Engineering Programs, DC Cook Outage U1C23 10 Year IS/ Disk 3 of 4, WesDyne International-1 DVD, August 2010.

CNP Document DC Cook U1C23 10 Year ISI, Engineering Programs, DC Cook Outage U1C23 10 Year ISI Disk 4 of 4, WesDyne International-1 DVD, August 2010.

Evaluation 8a - LTR-RCPL-10-41_R1.pdf 11/14/13 Westinghouse Document LTR-RCPL-10-41, Revision 0, Westinghouse - Evaluations Supporting Evaluations Supporting Operability Assessment for American Operability Assessment for American Electric Electric Power Service Corporation Donald C. Cook Nuclear Plant Power Service Corporation (1 Cycle)

Unit 1. March 2010.

70

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Evaluation 8b - LTR-RCPL-10-41_R1.pdf 11/14/13 Westinghouse Document LTR-RCPL-10-41, Revision 1, Westinghouse - Evaluations Supporting Evaluations Supporting Operability Assessment for American Operability Assessment for American Electric Electric Power Service Corporation Donald C. Cook Nuclear Plant Power Service Corporation (2 Cycles)

Unit 1. January 2011.

Analysis 9 -WCAP-17588-P Revision 1.pdf 11/14/13 Westinghouse - Minimum Bolting Pattern Westinghouse Document WCAP-17588-P, Revision 1, D.C. Cook Aasis Unit 1 Lower Radial Support Clevis Insert Acceptable Minimum Bolting Pattern Analysis. March 2013.

Evaluation 10 - LTR-RIDA-10-134[1].pdf 11/14/13 Westinghouse - D.C. Cook Unit 1 comparison Westinghouse Document LTR-RIDA-10-134, Revision 0, D.C. Cook to Unit 2 Engineering Evaluation of the Radial Unit 2 Engineering Evaluation of the Radial Support System Clevis Support System Clevis Insert Bolts and insert Bolts and Operation through Spring 2012. September Operation 2010.

Root Cause 11 - WCAP-15271 Rev. 0, Guide Tube Support Pin Degradation 11/14/13 Investigation Root Cause Investigation.pdf Westinghouse -Guide Tube Support Pin Degradation Root Cause Investigation from Westinghouse Document WCAP-15271, Revision 0, Guide Tube Westinghouse - Maanshan Unit 1 Support Pin Degradation Root Cause Investigation. August 1999.

Technical Report 12 - matsHdbkNukeAppEPRI.pdf 11/14/13 EPRI - Materials Handbook for Nuclear Plant Materials Handbook for Nuclear Plant Pressure Boundary Pressure Boundary Applications Applications (2013). EPRI, Palo Alto, CA: 2013. 3002000122.

Presentation Slides 13a - MAI-Overview.pdf 11/14/13 The Materials Ageing Institute - General van der Lee, J., The Materials Ageing Institute: General Overview.

Overiew The Materials Ageing Institute: Materials Degradation Course for Engineers in the Nuclear Industry, Vail, CO, USA, June 5-8, 2012.

71

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Presentation Slides 13b - Nuclear Plant Environment Todd Allen.pdf 11/14/13 The Materials Ageing Institute -

Nuclear Plant Allen, T., Nuclear Plant Environment. The Materials Ageing Environment Institute: Materials Degradation Course for Engineers in the Nuclear Industry, Vail, CO, USA, June 5-8, 2012.

Presentation Slides 13c - Operational Experience Peter Scott.pdf 11/14/13 The Materials Ageing Institute -

Operational Scott, P., Vaillant, F., Operational Experience -An Overview. The experience Materials Ageing Institute: Materials Degradation Course for Engineers in the Nuclear Industry, Vail, CO, USA, June 5-8, 2012.

Presentation Slides 13d - Materials and Their Use in Plant Components Francois 11/14/13 Cattant.pdf The Materials Ageing Institute - Materials and Their Use in Plant Components Francois Cattant, F., Materials and Their Use in Plant Components. The Cattant Materials Ageing Institute: Materials Degradation Course for Engineers in the Nuclear Industry, Vail, CO, USA, June 5-8, 2012.

Presentation Slides 13e - Fundamentals of Radiation Effects Allen.pdf 11/14/13 Allen, T., Fundamentals of Radiation Effects: What makes The Materials Ageing Institute -

reactors special?. The Materials Ageing Institute: Materials Fundamentals of Radiation Effects Degradation Course for Engineers in the Nuclear Industry, Vail, CO, USA, June 5-8, 2012.

Presentation Slides 13f - PWR RPV Internals Anne Demma.pdf 11/14/13 Demma, A., Electric Power Research Institute: PWR Vessel EPRI - PWR RPV Internals Internals Integrity Issues. The Materials Ageing Institute:

Materials Degradation Course for Engineers in the Nuclear Industry, Vail, CO, USA, June 5-8, 2012.

72

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Presentation Slides 13g - Fundamentals of Plant Chem Keith Fruzzetti.pdf 11/14/13 Fruzzetti, K., Electric Power Research Institute: Fundamentals of Plant Chemistry BWRs and PWRs. The Materials Ageing Institute:

Materials Degradation Course for Engineers in the Nuclear Industry, Vail, CO, USA, June 5-8, 2012.

Report 14 -SCCinitiationStainlesslnconelEPRI.pdf 11/14/13 EPRI - Stress Corrosion Cracking Initiation Stress Corrosion Cracking Initiation Modelfor Stainless Steel and Model for Stainless Steel and Nickel Alloys, Nickel Alloys: Effects of Cold Work. EPRI, Palo Alto, CA: 2009.

Effects of Cold Work 1019032.

Chart 15-GDTChart.pdf 11/14/13 MulitMac - Geometric Dimensioning Chart Geometric Dimensioning Chart. Multimac, Athens, OH.

CAP Product 16 -AR 2010-1804.pdf 11/13/13 CNP Document AR 2010-1804, Rx Vessel Core Support Lug Rx Vessel Core Support Lug Bolting Anomalies Bolting Anomalies. Initiated March 21, 2010.

Correspondence 17 -AEP-11-5.pdf 11/14/13 Westinghouse Document AEP-11-5, Revision 0, American Electric Westinghouse - Correspondence to Phil Power Donald C. Cook Unit I Engineering Services for One (1)

Lozmack from Daniel Beddingfield extending Cycle Extension of Unit 1 Operability Determination. February repair plan for one Cycle.

2011.

Evaluation 18 - EVAL-10-21.pdf 11/14/13 Westinghouse - Revision to D.C. Cook-1 Loose Westinghouse Document EIES-10-67, Revision 0, Revision to D.C.

Parts Operability Determination (LTR-RCPL Cook-1 Loose Parts Operability Determination (LTR-RCPL-1O-41)

41) and 50.59 Screen (EVAL-10-21) and 50359 Screen (EVAL-l0-21). December 2010.

73

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Excel Document 19 -boltinspectionResultsSummary.xlsx 11/14/13 Side by Side picture comparison of Clevis Bolts CNP Document, 19-boltlnspectionReultsSummary.xlsx. Microsoft and Dowel Pins from UIC23 to U1C25 Excel file, April 2013.

Evaluation 20 -CN-RIDA-10-27_R2.pdf 11/14/13 Westinghouse -Structural Evaluation of As-Westinghouse Document CN-RIDA-10-27, Revision 2, Structural Found Degraded Bolting Pattern for D.C. Cook Evaluation of As-Found Degraded Bolting Pattern for D.C. Cook Unit 1 Lower Radial Support Clevis Insert Unit 1 Lower Radial Support Clevis Insert. January 2011.

Evaluation 21 - LTR-RIDA-10-75_RO.pdf 11/14/13 Westinghouse-Evaluation of Potential for Westinghouse Document LTR-RIDA-10-75, Revision 0, Evaluation Loose Parts due to Damaged Core Barrel of Potentialfor Loose Parts due to Damaged Core Barrel Support Support Lug Bolts for D.C. Cook Unit 1 Lug Bolts for D.C. Cook Unit 1. March 2010.

Evaluation 22 - LTR-RIDA-10-76_Rl.pdf 11/14/13 Westinghouse - D.C. Cook Unit 1 Lower Radial Westinghouse Document LTR-RIDA-10-76, Revision 1, D.C. Cook Support Clevis Cap Screw Locking Bar Wear Unit i Lower Radial Support Clevis Cap Screw Locking Bar Wear Evaluation Evaluation. November 2010.

Evaluation 23 - LTR-RIDA-10-78_R2.pdf 11/14/13 Westinghouse -Structural Evaluation Westinghouse Document LTR-RIDA-10-78, Revision 2, Structural Summary of As-Found Degraded Bolting Evaluation Summary of As-Found Degraded Bolting Pattern for Pattern for D.C. Cook Unit I Lower Radial D.C. Cook Unit 1 Lower Radial Support Clevis Insert. January Support Clevis Insert 2011.

Evaluation 24 - LTR-RIDA-10-82_Rl.pdf 11/14/13 Westinghouse - D.C. Cook Clevis Insert Bolts:

Westinghouse Document LTR-RIDA-10-82, Revision 1, D.C. Cook SessiCorosi Cr of Alloy I-tBs Clevis Insert Bolts: Stress Corrosion Cracking of Inconel X-750.

December 2010.

1 74

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Evaluation 25 - LTR-RIDA-10-337.pdf 11/14/13 Westinghouse - Evaluation of Commercial Westinghouse Document LTR-RIDA-10-337, Revision 0, Risks Associated with Continued Operation Evaluation of Commercial Risks Associated with Continued with Degraded Clevis Insert Bolts at D.C. Cook Operation with Degraded Clevis Insert Bolts at D.C. Cook Unit 1.

Unit 1 January 2011.

Evaluation 26 - PE-10-18_Rl.pdf 11/14/13 Westinghouse - D.C. Cook Unit I Lower Radial Westinghouse Document PE-10-18, Revision 1, D.C. Cook Unit 1 Support Clevis Cap Screw Locking Bar Wear Lower Radial Support Clevis Cap Screw Locking Bar Wear Evaluation - Product Engineering Input, Evaluation - Product Engineering Input, Revision 1. November Revision 1 2010.

Drawing 27 - AEP On-site As Built Loop 4 Reactor Internals Drawing 11/18/13 108D467.pdf Westinghouse - On-site As Built Loop 4 Reactor Westinghouse Drawing 108D467, Revision 2, AEPAs-Built 4 Loop Internals Reactor Internals On-Site As-Built. December 1983.

Drawing 28 - 1 in Socket Headed Cap Screw Drawing 206C037.pdf 11/18/13 Westinghouse Drawing 206C037, Revision 2, 1.000 Soc. Hd. Cap Westinghouse - 1 in Socket Headed Cap Screw Screw (Undercut). March 1969.

Drawing 29 - PWR Lower Radial Support Insert Customizing Drawing 11/18/13 541F473.pdf Westinghouse - PWR Lower Radial Support Westinghouse Drawing 541F473, Revision 4, 173-00-000-000 Insert Customizing PWR Lower Radial Support Insert Customizing. April 1972.

75

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Drawing 30 - PWR Lower Radial Support Insert Preliminary Drawing 11/18/13 541F620.pdf Westinghouse - PWR Lower Radial Support Westinghouse Drawing 541F620, Revision 1, 173-00-000-000 Insert Preliminary PWR Lower Radial Support Insert Preliminary Machining. August 1968.

Drawing 31 - Lower Radial Support Clevis Insert Gaging and Assembly 11/18/13 685J790-S1-R4.PDF Westinghouse - Lower Radial Support Clevis Westinghouse Drawing 685J790, Sheet 1, Revision 4, 173 Insert Gaging and Assembly 000-000 PWR Lower Radial Support Clevis Insert Gaging &

Assembly. November 1971.

Drawing 32 -Core Barrel Support Lug Drawing Sheet 2 685J790-S2-R4.PDF 11/18/13 Westinghouse - Core Barrel Support Lug Sheet Westinghouse Drawing 685J790, Sheet 2, Revision 4, 173 2

000-000 PWR Lower Radial Support Clevis Insert Gaging &

Assembly. November 1971.

Drawing 33 - Core Barrel Support Lug Drawing Sheet 3 685J790-S3-R4.PDF 11/18/13 Westinghouse Drawing 685J790, Sheet 3, Revision 4, 173 Westinghouse - Core Barrel Support Lug Sheet 000-000 PWR Lower Radial Support Clevis Insert Gaging &

Assembly. November 1971.

Drawing 34 - Core Barrel Support Lug Drawing Sheet 4 685J790-S4-R4.PDF 11/18/13 Westinghouse Drawing 685J790, Sheet 4, Revision 4, 173 Westinghouse - Core Barrel Support Lug Sheet 000-000 PWR Lower Radial Support Clevis Insert Gaging &

Assembly. November 1971.

76

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Drawing 35 -AEP Reactor (Internals) as Built Drawing Fig 12 and 14 11/18/13 5656D81_6.pdf Westinghouse - Reactor (Internals) as Built Westinghouse Drawing 5656D81, Sheet 7, Revision 2, AEP Figure 12 and 14 Reactor (Internals) As Built Drawings Fig 13 & 14. May 1973.

Drawing 36 -AEP Reactor as Built Drawing Fig 15 5656D81_8.pdf 11/18/13 Westinghouse Drawing 5656D81, Sheet 8, Revision 2, AEP Westinghouse - Reactor as Built Figure 15 Reactor (Internals) As Built Drawings Fig 13 & 14. May 1973.

Drawing 37 -AEP Reactor Vessel as Built Drawing Fig 9 and 10 DC-11/18/13 132486.pdf Westinghouse - Reactor Vessel as Built Figure Westinghouse Drawing 108D002, Sheet 5, Revision 2, AEP 9 and 10 Reactor Vessel As Builts Fig 9 & 10. May 1971.

Drawing 38 - General Arrangement - Elevation Drawing E-233-440.pdf 11/18/13 Combustion Engineering Drawing 233-440, Revision 2, General Westinghouse - General Reactor Arrangement Arrangement-Elevation For Westinghouse Electric Corp. 173" I.D. Reactor Vessel. March 1969.

Drawing 39 - Bottom Head Forming and Welding Drawing E-233-443.pdf 11/18/13 Combustion Engineering Drawing 233-443, Revision 2, Bottom Westinghouse - Bottom Head Forming and Head Forming & Welding For Westinghouse Electric Corp. 173" Welding I.D. Reactor Vessel. December 1969.

Drawing 40 - Core Support Lug Drawing E-233-450_2.tif 11/18/13 Combustion Engineering Drawing 233-450, Revision 2, Westinghouse -Core Support Lug Miscellaneous Attachments for Westinghouse Electric Corp. 173" I.D. Reactor Vessel. September 1968.

77

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Specifications 41 - Inconel_alloy_X-750.pdf 11/18/13 Special Metals - Specs from Special Metals on Special Metals Publication No. SMC-067, Inconel alloy X-750.

Inconel Alloy X750 Special Metals, Huntington, WV. September 2004.

Report 42-EPRI NP-7338-L.pdf 11/19/13 EPRI - Design and Manufacturing Guidelines Materials Reliability Program: PWR Internals Material Aging for High-Strength Components in LWRs-Alloy Degradation Mechanism Screening and Threshold Values (MRP-X-750 175). EPRI, Palo Alto, CA: 2005. 1012081.

Journal Article 43 -journalArticleX750.pdf 11/18/13 Metallurgical and Materials Transactions Mills, J.I., Lebo, M.R., and Kearns, J.J., Hydrogen Embrittlement, Volume 30A - Hydrogen Embrittlement, Grain Graind Boundary Segregation, and Stress Corrosion Cracking of Boundary Segregation, and Stress Corrosion Alloy X-750 in Low-and High-Temperature Water, Metallurgical Cracking of Alloy X-750 in Low and High-and Materials Transactions A, Volume 30A, June 1999, pp. 1579-Temperature Water 1596.

Report 44 - EPRI Design and Manufacturing Guidelines for Alloy 11/19/13 X750.pdf EPRI - Design and Manufacturing Guidelines Design and Manufacturing Guidelines for High-Strength for Alloy X-750 Components in LWRs-Alloy X-750. EPRI, Palo Alto, CA: 1991. NP-7338-L.

Report 45 - NP-6392-SD Microstructure and SCC resistance of X-11/19/13 750_718_286.pdf EPRI - Microstructure and Stress Corrosion Microstructure and Stress Corrosion Resistance of Alloys X-750, Resistance of Alloy X-750, 718, and A-286 in 718, and A-286 in LWR Environments. EPRI, Palo Alto, CA: 1989.

LWR Environments NP-6392-SD.

78

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description External OE 46 - OE 99 - 009452.pdf 11/20/13 INPO Document OE10154, In-Service Failure Of AlloyX-750 Cap INPO - OE from Nine Mile Point Unit 1 Screw On Core Shroud Repair Tie Rod Assembly - rod assembly upper spring assembly. August 1999.

NRC Information 47 - NRCIN82-29.pdf 11/20/13 Notice NRC - OE from NRC Information Notice No. 82-NRC Information Notice IN 82-29, Control Rod Drive (CRD) Guide 29 Tube Support Pin Failures at Westinghouse PWRs. July 1982.

Report 48 - RVIAMP_ROa_FINAL.pdf 11/20/13 AEP - Donald C. Cook Nuclear Plant Reactor CNP Document Reactor Vessel Internals Aging Management Vessel Internals Aging Management Program Program. September 2012.

Presentation Slides 49 -clevisinsertboltsdec20lO.ppt 11/20/13 Lott, R., Clevis Insert Bolt Issue Update. Pressurized Water PWR Owners Group - Clevis Insert Bolt Issue Reactor Owners Group: Materials Subcommittee. Marco Island, Update - Randy Lott FL: December 7-9, 2010.

External OE 50 -SOER 84-5.pdf 11/20/13 INPO - Bolt Degradation or Failure in Nuclear INPO Document SOER 84-5, Bolt Degradation or Failure in Power Plants Nuclear Power Plants. September 1984.

External OE 51 - OE11453.pdf 11/20/13 INPO Document OE11453, Reactor Internal Split Pin Failures.

INPO - OF from McGuire Unit 2 October 2000.

External OE 52 - OE13865.pdf 11/20/13 INPO Document OE13865, Unplanned Shutdown due to Loose INPO - OF from Wolf Creek Part in Steam Generator. August 2004.

1 79

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description External OE 53 - OE19374.pdf 11/20/13 INPO Document OE19374, Failed RV upper Internals Guide Tube INPO - OE from Farley Unit 1 Support Pins. October 2004.

Report 54 - 13627 -1801_R2.pdf 11/20/13 NRC - Generic Aging Lessons Learned (GALL)

NRC Document NUREG-1801, Revision 2, Generic Aging Lessons Report (rev. 2)

Learned (GALL) Report. December 2010.

Report 55 - WCAP-14577R1-A.pdf 11/20/13 Westinghouse Document WCAP-14577, Revision 1-A, License Westinghouse - License Renewal Evaluation:

Renewal Evaluation: Aging Management for Reactor Internals.

Aging Management for Reactor Internals March 2001.

External OE 56 -OE25769.pdf 11/20/13 INPO Document OE25769, Five Control Rod Guide Tube Support INPO -OE from Braidwood Unit 1 Pins Found Failed During Planned Replacement Activities.

November 2007.

External OE 57 -OE7883.pdf 11/20/13 INPO Document OE7883, Shutdown Because of Detection of INPO - OE from Vogtle Unit 1 Foreign Object in Steam Generator. October 1998.

Evaluation 58 - LTR-RIDA-10-134[1].pdf 11/20/13 Westinghouse Document LTR-RIDA-10-134, Revision 0, D.C. Cook Westinghouse - D.C. Cook Unit 2 Engineering Unit 2 Engineering Evaluation of the Radial Support System Clevis Evaluation of the Radial Support System Clevis insert Bolts and Operation through Spring 2012. September Insert Bolts and Operation through Spring 2010.

2012

[Redundant to Exhibit 10]

80

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Time line 59 - Timeline for Unit 1 Reactor Pressure Vessel internals 11/20/13 examinations.docx Roy Hall - Timeline for Unit 1 Reactor Pressure CNP Document, Unit 1 RPV 1O Year Visual Examinations. 59-Vessel internals examinations Timeline for Unit 1 Reactor Pressure Vessel internals examinations.docx. Microsoft Word File, November 2013.

Time line 60 - Timeline for Unit 2 Reactor Pressure Vessel internals 11/21/13 examinations.docx Roy Hall - Timeline for Unit 2 Reactor Pressure CNP Document, Unit 2 RPV 10 Year Visual Examinations. 60 -

Vessel internals examinations Timeline for Unit 2 Reactor Pressure Vessel internals examinations.docx. Microsoft Word File, November 2013.

Information Slides 61 - ClDesigns.pdf 11/21/13 Westinghouse - Geometry Comparisons of Westinghouse Slides, Geometry Comparisons. Provided Clevis Insert Designs November 2013.

Internal OE 62 - OE30993.pdf 11/22/13 INPO - OE of DC Cook Clevis bolting INPO Document OE30993, During 10-Year Reactor Vessel In-degradation Service Inspection Seven Failed Core Barrel Lower Lateral degradation Restraint Bolts were Observed. March 2010.

External OE 63 - EAR PAR 06-047.pdf 11/22/13 EAR PAR 06-047, Corrosion in Control Rod Guide Tube Support INPO - OE of Vandellos 11 Pins. March 2006.

Periodical 64 - nuclearNewsPlantListing20llANS.pdf 11/25/13 Nuclear News, Volume 54, Number 3, 13th Annual Reference Nuclear News - 1 3th Annual Reference Issue Issue. American Nuclear Society, Inc., LaGrange Park, IL: March 2011.

81

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Design Information 65 - DIT-S-00705-17.pdf 11/25/13 Transmittal AEP - Donald C. Cook Nuclear Plant Units 1 & 2 CNP Document DIT-S-00705-17, Unit 1 and Unit 2 Burnup Data.

- Burnup Data October 2013.

Program Document 66 - ISI Program.pdf 11/25/13 AEP - Donald C. Cook Nuclear Plant Units 1 & 2 CNP Document DCC01.G03, Revision 3, Donald C. Cook Nuclear

- ISI Program Plan Fourth Ten-Year Inspection Plant Units 1 & 2 ISI Program Plan Fourth Ten-Year Inspection Interval Interval. January 2013.

Memorandum 67 - UlthermalSheidlndication.pdf 11/25/13 AEP - Resolution of the surface indication CNP Memo From D. A. Patience to R. L. Otte, Unit 1 Thermal found on the Unit 1 thermal shield on June 30 Shield Indication. September 10, 1985.

Report 68 - WCAP-17484-P.pdf 11/26/13 Westinghouse - Reactor Internals Aging Westinghouse Document WCAP-17484-P, Revision 0, Reactor Management MRP-227-A Implementation Internals Aging Management MRP-227-A Implementation Manual for Westinghouse designed Nuclear Manualfor Westinghouse - designed Nuclear Steam Supply Steam Supply Systems Systems. August 2012.

Report 69 - WCAP-14522_RCAUlBarrelBolt.pdf 11/26/13 Westinghouse - Root Cause Determination of Westinghouse Document WCAP-14522, Revision 0, Root Cause Core Barrel - Baffle Former Bolting Failure at Determination of Core Barrel-Baffle Former Bolting Failure at DC Cook Nuclear Plant, Unit 1 Donald C. Cook Nuclear Plant, Unit 1. December 1995.

Working Notes 70 - seatSurfAcceptCrit20130425.doc 11/26/13 AEP - This document provides LRSS Engineering Working Notes, LRSS Bolt Seating Surface requirements of clevis seating surfaces for Acceptance Criteria. April 2013.

1 1 replacement clevis bolts.

82

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Design Change 71 DCP-0125.pdf 11/26/13 AEP - Design Change Package 0125, CNP Document 1-DCP-0125, Revision 0, Reactor Vessel Core Replace barrel-former bolts at locations A-Barrel. June 1997.

4, A-5, and A-6 CAP Product 72 - GT2013-1241.pdf 11/26/13 AEP - GT 2013-1241, Dislodged Thermal CNP Document GT 2013-1241, IER Lvl 3 (13-2) - Dislodged Sleeves Thermal Sleeves. Initiated January 28, 2013.

Video Files 73 -VIDEOTShighlights 11/26/13 AEP - Folder of Videos from the Inspection and CNP Unit 1 Video Highlights from Barrel-Former Bolt Discovery repair of the barrel-former Bolts and Repairs, 1994-1997.

Report 74 - MRP-191.pdf 11/26/13 EPRI - Materials Reliability Program - This report describes the process and results of Materials Reliability Program: Screening, Categorization, and categorizing Westinghouse and Combustion Ranking of Reactor Internals Components for Westinghouse and engineering designed pressurized water Combustion Engineering PWR Design (MRP-191). EPRI, Palo Alto, reactor (PWR) internals components according CA: 2006. 1013234.

to age-related degradation and significance.

Report 75 -WCAP-17300-NP Cook AMP Unit 1.pdf 11/26/13 Westinghouse - Reactor Vessel Internals Westinghouse Document WCAP-17300-NP, Revision 0, Reactor Program Plan for Aging Management of Reactor Internals at D.C. Cook Nuclear Plant Vessel Internals Program Plan for Aging Management of Reactor U nit 1 Internals at D.C. Cook Nuclear Plant Unit 1. February 2011.

Quality Assurance 76a 9202929-000, QADP special 1 inch bolts.pdf 11/26/13 Areva - LRSS Clevis Replacement Contingency Data Package Bolt Assembly Standard Threads For AEP / DC AREVA Document 23-9202929, Revision 0, Quality Assurance Cook Unit 1 AEP Contract 1500256 Release Data Package: LRSS Clevis Replacement Contingency Bolt Number 55 AREVA NP Inc. Contract No.

Assembly Standard Threads For AEP/ DC Cook Unit 1. April 2013.

F.501639 83

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Quality Assurance 76b 9195126-001 QADP for 5 remachined bolts.pdf 11/26/13 Areva - LRSS Clevis Replacement Bolt Data Package Assemblies For AEP / DC Cook Unit 1 AEP PO AREVA Document 23-9195126, Revision 1, Quality Assurance 01559035 AREVA NP Inc. Contract No.

Data Package: LRSS Clevis Replacement Bolt Assemblies For AEP /F015 35 DC Cook Unit 1. April 2013.

F.501639 Quality Assurance 76c 9195126-002.pdf 11/26/13 Data PckageAreva

- LRSS Clevis Replacement Bolt Data Package DAssemblies For AEP / DC Cook Unit 1 AEP PO AREVA Document 23-9195126, Revision 2, Quality Assurance 01559035 Rev. 001 & 002 AREVA NP Inc.

Data Package: LRSS Clevis Replacement Bolt Assemblies For AEP /ContRaN.

F01&3N DC Cook Unit 1. April 2013.

Instant Message 77 - Roy E HallAEPIN-20131210-1042.html 12/10/13 Record Instant message conversation between Kevin Instant Message Conversation between Kevin Kalchik, American Kalchik and Roy Hall discussing ASME Code Electric Power, and Roy Hall, American Electric Power. December inspections, code relief, and core barrel pulls 10, 2013.

CAP Product 78 -AR 00124109 Missile Block Drop.pdf 11/26/13 AEP - AR written documenting Missile Block CNP Document AR 00124109, A stop work order has been drop in Unit 2 initiated in response to a. Initiated March 27, 2006 Report 79 - MRP-80.pdf 12/2/13 EPRI - Materials Reliability Program - This report documents the results of a review of Materials Reliability Program: A Review of Thermal Aging PWR materials summarizing the available data Embrittlement in Pressurized Water Reactors (MRP-80), EPRI, and recommendations for follow-on testing to Palo Alto, CA: 2003.1003523.

determine the effects of thermal aging.

Presentation Slides 80 -W Cook Clevis April 2010 MSC.pdf 12/3/13 Lott, R., Radial Support Clevis Bolting Fractures: Industry Westinghouse - Plant Ranking Based on Implications. Pressurized Water Reactor Owners Group:

Materials Subcommittee. Lake Buena Vista, FL: April 20-22, 2010.

84

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Procedure 81 OHP-4050-FHP-044.pdf 12/3/13 AEP - Reactor Vessel Lower Internals Removal CNP Document 12-OHP-4050-FHP-044, Revision 10, Reactor Vessel Lower Internals Removal. January 2013.

Procedure 82 OHP-4050-FHP-045.pdf 12/3/13 AEP - Reactor Vessel Lower Internals CNP Document 12-OHP-4050-FHP-045, Revision 11, Reactor Replacement Vessel Lower Internals Replacement. January 2013.

Report 83 - LTR-RIDA-13-183_6.pdf 12/4/13 Westinghouse - The purpose of this letter is to provide the applicable material specifications Westinghouse Document LTR-RIDA-13-183, Revision 0, Material forite loe adial suppri syemf(LRs Specifications for Lower Radial Support System (LRSS) Clevis cte inser boltsuatoD. C.sCook Unit Inset Bots. ecemer 213.clevis insert bolts at D. C. Cook Unit I Insert Bolts. December 2013.

Report 84 - MRP-134.pdf 12/03/13 EPRI - Materials Reliability Program -This Materials Reliability Program: Framework and Strategies for report describes a framework of associated Managing Aging Effects in PWR Internals (MRP-134). EPRI, Palo PWrite rnals.

Alto, CA: 2005. 1008203.

Report 85a - MRP-156.pdf 12/03/13 EPRI - Materials Reliability Program -This report provides initial input to the Issue Materials Reliability Program: Pressurized Water Reactor Issue Management Table to address the Management Table, PWR-IMT Consequence of Failure (MRP-consequences of failure for the identified 156). EPRI, Palo Alto, CA: 2005. 1012110.

components in a reactor coolant system.

Report 85b - MRP-156supplement.pdf 12/03/13 EPRI - Materials Reliability Program -This Report is to document some potential errors EPRI Supplement, Attention Recipientts of EPRI Report 1012110.

and omissions that were discovered during EPRI, Palo Alto, CA: December 2005.

reviews of MRP-156.

85

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Report 86 - MRP-191.pdf 12/03/13 EPRI - Materials Reliability Program -This report describes the process and results of Materials Reliability Program: Screening, Categorization, and categorizing Westinghouse and Combustion Ranking of Reactor Internals Components for Westinghouse and Engineering designed PWR internals Combustion Engineering PWR Design (MRP-191). EPRI, Palo Alto, components according to age-related CA: 2006. 1013234.

degradation and significance.

Report 87a - MRP-227.pdf 12/03/13 EPRI - Materials Reliability Program - PWR Materials Reliability Program: Pressurized Water Reactor Internals inspection and evaluation Guidelines Internals Inspection and Evaluation Guidelines (MRP-227-Rev. 0).

(rev. 0)

EPRi, Palo Alto, CA: 2008. 1016596.

Report 87b -MRP-227-A.pdf 12/03/13 EPRI - Materials Reliability Program - PWR Materials Reliability Program: Pressurized Water Reactor Internals inspection and evaluation Guidelines Internals Inspection and Evaluation Guidelines (MRP-227-A).

(A)

EPRI, Palo Alto, CA: 2011. 1022863.

Report 88 - MRP-232.pdf 12/03/13 EPRI - Materials Reliability Program -This report provides the technical basis for the Materials Reliability Program: Aging Management Strategies for aging management requirements of Wesitnghouse and Combustion Engineering PWR Internals (MRP-Westinghouse and Combustion Engineering 232). EPRI, Palo Alto, CA: 2008. 1016593.

reactor internals Report 89 - NUREG-1801 Vol 2 Chapter Xl Section M.pdf 12/04/13 NRC Document NUREG-1801, Revision 0, Generic Aging Lessons NRC - Generic Aging Lessons Learned (GALL)

Learned (GALL) Report, Chapter Xh Aging Management programs Report (Chapter X1)

(AMPs). April 2001.

Spreadsheet 90 - coreBurnupData.xlsx 12/09/13 AEP - Shows Effective Degradation Years for CNP Document, 90 - coreBurnupData.xlsx. Microsoft Excel file, Unit 1 and Unit 2 December 2013.

1 1

86

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Report 91 - WCAP-13627.pdf 12/04/13 Westinghouse -This report provides the methods and results of a Westinghouse Westinghouse Document WCAP-13627, Revision 0, Pilot Owners Group pilot study to demonstrate and Application of Risk-Based Inspection Methods to Westinghouse-set precedence for the application of risk-Designed Reactor internals. June 1993.

based technologies in the development of nuclear power plant component inspection programs.

Procedure 92 EHP-5034-SPV-001.pdf 12/04/13 CNP Document 12-EHP-5034-SPV-001, Revision 0, Single point AEP - Single Point Vulnerability Management Vulnerability Management. March 2013.

Report 93 - ap-913 rev 4.pdf 12/04/13 INPO - This document describes an equipment reliability process offered to assist member INPO Document AP-913, Revision 4, Equipment Reliability utilities to maintain high levels of safe and Process Description. October 2013.

reliable plant operation in an efficient manner Presentation Slide 94 - dowelpinlooseningscenario.pdf 12/05/13 Westinghouse - Broken Dowel Pin Tack Weld Westinghouse Slides, Broken Dowel Pin Tack Weld and Pin and Pin Loosening Scenario Loosening Scenario. Provided December 2013.

Procedure 95 - DTG-EQR-002.pdf 12/04/13 CNP Document DTG-EQR-002, Revision 5, Component Scoping, AEP - Component Scoping September 2012.

Procedure 96 - EHI-5054-RPV.pdf 12/04/13 CNP Document EHI-504-RPV, Revision 1, Reactor Vessel Integrity, AEP - Reactor Vessel Integrity May 2013.

87

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Report 97 -LRP-EAMP-01.pdf 12/04/13 CNP Document LRP-EAMP-01, Revision 3, License Renewal Areva - License Renewal Project Evaluation of Project: Evaluation of Aging Management Programs. November Aging Management Programs (rev. 3) 2005.

Report 98 - LRP-MAMR-01.pdf 12/04/13 AEP - License Renewal Aging Management CNP Document LRP-MAMR-01, Revision 3, Aging Management Review of the RCS (rev. 3)

Review of the Reactor Coolant System. September 2005.

Report 99 -WCAP-16198-P-Alloy 600 Program Cook.pdf 12/05/13 Westinghouse - PWSCC Susceptibility Westinghouse Document WCAP-16198-P, Revision 1, PWSCC Assessment of the Alloy 600 and Alloy 82/182 Susceptibility Assessment of the Alloy 600 and Alloy 82/182 Components in D.C. Cook Units l and 2 Components in D. C. Cook Units 1 and 2. July 2004.

Report 100 -WCAP-15271 Rev. 0, Guide Tube Support Pin Degradation 12/05/13 Root Cause Investigation.pdf Westinghouse - Guide Tube Support Pin Westinghouse Document WCAP-15271, Revision 0, Guide Tube Degradation Root Cause Investigation Support Pin Degradation Root Cause Investigation. August 1999.

Procedure 101 - EHI-5070-ALLOY600.pdf 12/05/13 AEP - Alloy 600 Material Management CNP Document EHI-5070-ALLOY600, Revision 4, Alloy 600 Program Material Management Program. January 2012.

Letter 102 - ML13325A973.pdf 12/05/13 Florida Power & Light Company letter L-2013-287 to the USNRC, Florida Power & Light - License Renewal dated October 30, 2013, License Renewal (LR) Reactor Vessel Reactor Vessel Internals Inspection Program Internals (RVI) Inspection Program Response to Request for Response to RAls Additional Information (RAI), Agencywide Documents and Accession Management System (ADAMS) Accession No. ML13325A973.

88

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Letter 103 - ML13270A250.pdf 12/05/13 Entergy Nuclear Northeast letter NL-13-122 to the USNRC, dated September 27, 2013, Reply to Request for Additional Information Entergy - IPEC License Renewal Regarding the License Renewal Application Indian Point Nuclear Generating Units Nos. 2 & 3, Agencywide Documents and Accession Management System (ADAMS) Accession No. ML13270A250.

Report 104 -WCAP-16777-N P.pdf 12/05/13 Westinghouse Document WCAP-16777, Revision 0, Interim Westinghouse - Management Strategies for Report on Aging Management Strategies for Westinghouse and Westinghouse and Combustion Engineering Combustion Engineering Reactor Vessel Internals Components, Reactor Vessel Internals Components April 2007.

Accepted Stores 105 - Flow Restrictor ASP Package.pdf 12/05/13 ProcedureCNP Document ASP-27585, Revision 0, Westinghouse - Covers AEP - Unit 1 and 2 Flow Restrictor ASP Package Restrictor Flow Assy. April 2003.

-Design Information 106 - Flow Restrictor DIT.pdf 12/05/13 Transmittal AEP - D.C. Cook Unit 2 Part Length Control Rod CNP Document DIT-B-03240-00, D.C. Cook unit 2 Part Length Guide Tube Flow Restrictors Control Rod Guide Tube Flow Restrictors. October 2007.

Report 107 -WCAP-11000 U2 failed CRGT pin op rpt.pdf 12/05/13 Westinghouse Document WCAP-11000, Revision 0, D. C. Cook Westinghouse - Estimated Operability with Unit 2 Estimated Operability with Failed Control Rod Gide Tube Failed Control Rod Guide Tube Support Pins Support Pins. January 1985.

Procedure 108 - Ul Spring 2010 Lower Internals Removal Procedure.pdf 12/06/13 AEP-Procedure used during U1C23 removal of CNP Document 12-OHP-4050-FHP-044, Revision 8, Reactor the Internals under WO 55293911-07 Vessel Lower Internals Removal. March 2010.

89

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Data Graph 109 - Unit 2 Tcold.pdf 12/09/13 CNP Data Retrieved From R*Time Points: PPC2.DATA.TO406A RCS LP 1 COLD LEG TEMPER (SNAP), PPC2. DATA.T0426A RCS LP 2 AEP - Graph of Cold Leg Temps of Unit 2 COLD LEG TEMPER (SNAP), PPC2.DATA.T0446A RCS LP 3 COLD January 2003 to December 2013 LEG TEMPER (SNAP), PPC2.DATA.T0466A RCS LP 4 COLD LEG TEMPER (SNAP). 1/1/2003 to 12/9/2013, Snap Data Period 24 hrs. Retrieved December 9, 2013.

Data Graph 110 - Unit 1 Tcold.pdf 12/09/13 CNP Data Retrieved from R*Time Points: PPC1.DATA.T0406A RCS LOOP 1 COLD LEG TEMP (SNAP), PPC1.DATA.T0426A RCS LOOP 2 AEP - Graph of Cold Leg Temps of Unit 1 COLD LEG TEMP (SNAP), PPCI.DATA.T0446A RCS LOOP 3 COLD January 2003 to December 2013 LEG TEMP (SNAP), PPC1.DATA.T0466A RCS LOOP 4 COLD LEG TEMP (SNAP). 1/1/2003 to 12/9/2013, Snap Data Period 24 hrs.

Retrieved December 9, 2013.

Data Graph 111 - Unit 1 Tcold Turbine Outage.pdf 12/09/13 CNP Data Retrieved from R*Time Points: PPC1.DATA.T0406A RCS LOOP 1 COLD LEG TEMP (SNAP), PPC1.DATA.T0426A RCS LOOP 2 AEP - Graph of Cold Leg Temps of Unit 1 COLD LEG TEMP (SNAP), PPC1.DATA.T0446A RCS LOOP 3 COLD During Turbine Shutdown LEG TEMP (SNAP), PPC1.DATA.T0466A RCS LOOP 4 COLD LEG TEMP (SNAP). 9/20/2008 to 11/22/2009, Snap Data Period 24 hrs. Retrieved December 9, 2013.

-Design Information 112 - DIT-B-03415-OO.pdf 12/10/13 Transmittal AEP -Justification for Continued Operation for CNP Document DIT-B-03415-00, Justification for Continued U2C19 Regarding LRSS Operation for U2C19 Regarding LRSS. August 2010.

1_1 90

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Report 113 - MRP-205.pdf 12/10/13 EPRI - Materials Reliability Program -This PWR IMT report identifies, describes, and prioritizes Materials Reliability Program: Pressurized Water Reactor issue 80 open R&D gaps where additional research Management Tables - Revision 3 (MRP-205). EPRI, Palo Alto, CA:

is needed to resolve issues related to 2013. 3002000634.

degradation of PWR NSS components.

Report 114 - LTR-RIDA-13-188.pdf 12/11/13 Westinghouse - The purpose of this letter is to Westinghouse Document LTR-RIDA-13-188, Revision 0, D.C. Cook provide the reactor internals assembly units 1 and 2 Reactor Internals Assembly Specification - 616A225 specification for D.C. Cook Unit 1 and Unit 2.

Rev. 4. December 2013.

E-Mail 115 - e-mail on clevis lugs.pdf 12/11/13 E-mail from Kevin Neubert, Westinghouse, to Kevin Kalchik, Westinghouse - Responses to Cook RCA Action American Electric Power, RE: Responses to Cook RCA Items List ActionltemList2O131127.xlsx. December 11, 2013.

E-Mail 116 - thermal shield flexure material.pdf 12/11/13 E-mail from Kevin Neubert, Westinghouse, to Kevin Kalchik, Westinghouse - Cook Unit 1 and Unit 2 American Electric Power, Cook Units 1 and 2 - Thermal Shield Thermal Shield Flexure Flexure. December 11, 2013.

Purchase Requisition 117 - 1988 U2 RPV Exam.pdf 12/12/13 AEP - Purchase Requisition for the mechanized ultrasonic examinations and remote visual CNP Document 01681-040-8X, 151 Reactor Vessel Examinations.

examination of Unit 2 Reactor Vessel for the March 1988.

1988 ISI inspection Licensee Event Report 118 - chloride LER 1.pdf 12/11/13 CNP Licensee Event Report 82-071/03L-0 Transmittal Letter with AEP - Licensee Event Report from 1982 Report Attached From W.G. Smith, American Electric Power, to J.G. Keppler, USNRC. September 7, 1982.

91

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Licensee Event Report 119 - chloride LER.pdf 12/11/13 CNP Licensee Event Report 81-009/03L-0 Transmittal Letter with AEP - Licensee Event Report from 1981 Report Attached From D.V. Shaller, American Electric Power, to J.G. Keppler, USNRC. April 29, 1981.

Procedure Figure 120 - cooldown hydrogen control band.pdf 12/12/13 CNP Document 12-THP-6020-CHM-110 - Figure 1, Revision 36, AEP THP-6020-CHM-110 Figure 1 RCs Cool RCS Chemistry - Shutdown and Refueling: RCS Cooldown Down Dissolved Hydrogen Hydrogen Control Band. October, 2013.

Report 121 - MRP-274.pdf 12/12/13 EPRI - This report evaluates the potential for aging degradation of components inside the Materials Reliability Program: Assessment of Westinghouse and reactor vessel that are fabricated from Combustion Engineering Nickel-Based Alloy Orphan Locations austenitic nickel-based alloys and their (MRP-274). EPRI, Palo Alto, CA: 2010. 1021025.

associated weld metals and that have not been addressed in detail in other documents and programs.

Report 122 - MRP-48.pdf 12/12/13 EPRI - This report contains the plant rankings using the time-at-temperature model and PWR Materials Reliability Program Response to NRC Bulletin provides comments regarding applicable 2001-01 (MRP-48), EPRI, Palo Alto, CA: 2001. 1006284.

regulatory requirements.

Inspection Plan 123 - DC Cook Unit-2 RPV Visual Inspection Plan (2009) (Rev-3) 12/13/13 (3).pdf AREVA-RPV Visual Examination ISI Inspection AREVA Document, DC Cook Unit-2, Outage 2009 U2C-18 RPV Plan for U2C18 Visual Examinations. AREVA Contract No. A001717. March 2009.

92

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Report 124 - Underwater Remote Visual Examination of the Reactor 12/13/13 Vessel and Internals Unit 1 Cycle 23 RFO.pdf CNP Document DC Cook UIC23 10 Year ISI, Engineering Programs, DC Cook Outage U1C23 10 Year ISI Disk 1 of 4, WesDyne International-1 DVD, August 2010.

CNP Document DC Cook U1C23 10 Year ISI, Engineering Underwater remote visual examination of the Programs, DC Cook Outage U1C23 10 Year ISI Disk 2 of 4, Reactor Vessel and Internals of U1C23 WesDyne International-1 DVD, August 2010.

Summary Report CNP Document DC Cook UIC23 10 Year ISI, Engineering Programs, DC Cook Outage UIC23 10 Year ISI Disk 3 of 4, WesDyne International-1 DVD, August 2010.

CNP Document DC Cook UIC23 10 Year ISI, Engineering Programs, DC Cook Outage U1C23 10 Year ISI Disk 4 of 4, WesDyne International-1 DVD, August 2010.

Procedure 125 THP-6020-CHM-110 (rev. 8).pdf 12/13/13 AEP THP-6020-CHM-110 rev. 8 CNP Document 12-THP-6020-CHM-110, Revision 8, RCS Chemistry-Shutdown/Refueling. January 2002.

Spreadsheet Data and 126 -Turbine Outage Hydrogen.xls 12/13/13 Graph AEP -Graph of Hydrogen during the Ul CPN Data Retrieved from WinCDMS Database Point: Unit 1-RCS-Turbine Event H2. 9/1/2008 to 12/31/2010. Retrieved December 13, 2013.

Report 127 - MPR-118.pdf 12/13/13 EPRI - The goals of this report were to review Materials Reliability Program: Suitability of Emerging new technologies and to determine if they Mateial ReiablityProram Sutabiityof meringwere viable for PWR applications for mitigating Technologies for Mitigation of PWSCC (MRP-118). EPRI, Palo wSCCr I Alto, CA: 2004. 1009500.

93

Unit 1 Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description CAP Product 128 - GT 2013-18674-1 CARB Approval of Mid Brief.pdf 12/13/13 AEP - Mid CARB Brief on Reactor Vessel Core CNP Document GT 2013-18674-1, CARB 828 Meeting Minutes Support Lug Bolting Anomalies Tracking. Initiated December 6, 2013.

Meeting Agenda 129 - CARB 828 Agenda with LRSS Mid Brief Material.pdf 12/13/13 CNP Document, Management Screening Committee (MSC)

AEP - Mid GARB Brief on Reactor Vessel Core Corrective Action Review Board (CARB) #828 Agenda. December Support Lug Bolting Anomalies 6, 2013.

Report 130 - DZO White PaperZinc 2010.doc 12/16/13 AEP -Technical Evaluation of Zinc Addition For Miller, D.W., Management White Paper Technical Evaluation of Cook, Units 1 & 2 Zinc Addition For Cook, Units I & 2. December 28, 2010.

Report 131a - LTR-RIDA-13-189.pdf 12/18/13 Westinghouse - D.C. Cook Unit 1 Lower Radial Westinghouse Document LTR-RIDA-13-189, Revision 0, D.C. Cook Support System Quality Control Systems Unit 1 Lower Radial Support System Quality Control System Requirements, Trip Reports and Field Requirements, Trip Reports and Field Deficiencies. December Deficiencies 2013.

Specification 131b - QCS-1_R3.pdf_1387291934_LTR-RIDA-13-189.pdf 12/18/13 Westinghouse - This specification establishes requirements for manufacturer's systems for Westinghouse Document QCS-1, Revision 3, Manufacturer's control of quality during manufacture, Quality Control System Requirements. May 1967.

including inspection plans.

Trip Reports 131c -

12/18/13 AEPFieldDeficientiesandTripReports.pdf_1387293684_LTR-Westinghouse - Surveillance covering the RIDA-13-189.pdf installation of the clevis inserts and Westinghouse Document LTR-RIDA-13-189, Attachments 2 reinstallation of the lower internals assembly through 5, Revision 0, D.C. Cook Unit 1 Lower Radial Support for the final measurement of the clevis fit and System Quality Control Systems Requirements, Trip Reports and nozzle gaps.

Field Deficiencies. December 2013.

94

Unit I Rx Vessel Core Support Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.12 Documents Reviewed Type Exhibit - Document Name Date Reviewed Title/Description Report 132 - MRP-280.pdf 12/18/13 EPRI-The objective of this report was to determine effective methods involving Materials Reliability Program: Mitigation of Stress Corrosion changes in primary water chemistry for the Crack Growth in Nickel-Based Alloys in Primary Water by mitigation of PWSCC in thick-wall components Hydrogen Optimization and Zinc Addition (MRP-280). EPRI, Palo constructed of Alloy 600 and its weld metals in Alto, CA: 2010. 1021013.

PWR coolant systems.

Regulatory Guide 133 - RG1.133_MLOO3740137.pdf 2/24/14 USNRC Regulatory Guide 1.133, Revision 1, Loose-Part Detection Loose parts monitoring Reg Guide Program for the primary System of Light-Water-Cooled Reactors.

May 1981. ML003740137.

Report 134 - RPT-0025-1304 MPR Oversight for B&W Hot Cell 2/25/14 Analysis and Testing of Failed LRSS Clevis Bolts.pdf MPR Document RPT-0025-1304-1, Revision 0, Summary MPR Letter report documenting MPR oversight of Technical Oversight of B& W Hot Cell Laboratory Analysis and B&W hot cell testing and report Testing of Failed LRSS Clevis Bolts from D.C. Cook Unit 1.

February 2014.

Vendor Technical 135 - VTD-FANP-0001.pdf 2/25/14 Document CNP Document VTD-FANP-0001, Revision 1, Framatome ANP, CNP loose parts monitoring instruction manual Inc. Instruction manualfor Loose Parts Monitoring System -

(LPMS-V) [PUB. #01-5021870-00]. August 2008.

95

Unit I Rx Vessel Core Supvoort Lug Bolting Anomalies (AR 2010-1804-10) ATTACHMENT 5.13 5.13 CARB COMMENTS CARB Comments:

Page 4, first paragraph under Basis, remove last sentence starting with "It is not..."

DONE Page 7, section 1.6.3, remove paragraph starting with "PMP-7030_CAP-005..." and the associated action.

DONE Page 7, section 1.6.3, define the due dates on the actions.

DONE Page 34, add discussion concerning the means to detect loose parts.

DONE, added RG 1.133 and VTD-FANP-0001 to references to support discussion NRC Resident Administrative Comments:

Clarify reason for bolt replacement when there is not safety or operability concern.

  • DONE, Added discussion starting on Page 25.

Annotate proprietary information because the NRC may be subject to public request for information, this will protect proprietary information.

  • DONE, no proprietary information found in the LRSS RCE
  • Hot Cell Report will be marked proprietary before submitting to the NRC Include more discussion in section 3.3 indicating that CNP was aware of X-750 issues and followed industry guidance.
  • DONE, indicated that CNP follows ASME Code, EPRI-MRP guidance, and PWROG-MSC guidance Define "Literature" in supporting information column for 1.4 Corrosion Failure in the Potential Failure mode Evidence Support/Refute Matrix.
  • DONE, referenced EPRI Materials Handbook Provide more detail in the Reference section such that a reader outside CNP may be able to find and retrieve references.
  • DONE, references section updated with more explicit detail Administrative CARB Comments:

No administrative comments received from any CARB member 96

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