ML14261A145
| ML14261A145 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 09/05/2014 |
| From: | Gideon W R Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RNP-RA/14-0097, TAC ME9633 | |
| Download: ML14261A145 (18) | |
Text
William. R. GideonH. B. Robinson SteamElectric Plant Unit 2DUKE Site Vice Presidein,ENERGYe Duke Energy Progress3581 West Entrance RoaoHartsville, SC 295500:843 857 1701F: 843 857 1319Randy.Gideon@duke-energy.con.10 CFR 54.21Serial: RNP-RA/14-0097SEP 0 5 2014ATTN: Document Control DeskUnited States Nuclear Regulatory CommissionWashington, DC 20555-0001H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2DOCKET NO. 50-261/RENEWED LICENSE NO. DPR-23RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TOTHE PRESSURIZED WATER REACTOR INTERNALS PROGRAM PLAN FOR AGINGMANAGEMENT OF REACTOR INTERNALS (TAC NO. ME9633)Ladies and Gentlemen:The NRC requested that Duke Energy Progress, Inc., respond to two requests for additional information(RAI) regarding the Aging Management Program for the Reactor Vessel Internals at Robinson NuclearPower Plant (RNP). The Duke Energy Progress response to these RAIs (RAI-3-1 and RAI-3-3)(NRCADAMS Accession Numbers.: ML14014A097, ML13240A499 and ML14167A345) is provided in theenclosures to this letter. A copy of Westinghouse Technical Bulletin (TB), Reactor Internals LowerRadial Support Clevis Insert Cap Screw Degradation (TB-14-5), is also provided in the enclosure to thisletter. Based upon communications from the NRC, dated September 3, 2014, the Staff did not requireinclusion of a response to Item 2 of RAI-3- 1, for this submittal.On a clarification call with the NRC on August 28, 2014, related to RAI-3-3, the Staff concurred with theconclusion of the above Westinghouse Technical Bulletin that degradation of the Clevis Insert Bolts posesno nuclear safety or operational concern. The Staff further stated that the following items are deemed toconstitute a complete response to RAI-3-3:1. Summary of the latest inspection results on Clevis Insert Bolts2. Revision of the MRP-227-A inspection requirements to integrate an enhanced visual inspection tobe utilized in the next 10-year ASME Section XI, ISI examinations of the Reactor VesselInternalsIn addition to the existing ASME Section XI, ISI required Visual Test (VT-3), RNP will perform anEnhanced Visual Test (EVT-1), which meets and exceeds VT-3 requirements, in the next 10-year ASMESection XI, ISI examinations (i.e., Fifth 10-year ISI examinations). As it was stated above, althoughdegradation of the Clevis Insert Bolts do not pose a nuclear safety or operational concern, R.NP has begunKlee-United States Nuclear Regulatory CommissionSerial: RNP-RA/14-0097Page 2 of 3performing an impact assessment review arising from the above Technical Bulletin. This effort will giverise to development of an asset management program that would mitigate degradation of the Clevis InsertBolts and associated reactor internals (i.e., Clevis Insert accessible welds, bolts, locking bars, and dowelpins) reflected in the subject Technical Bulletin.There are no regulatory commitments made in this submittal. If you have any questions regarding thissubmittal, please contact Mr. R. Hightower at (843) 857-1329.I declare under penalty of perjury that the foregoing is true and correct.Executed On: SinWilliam. R. GideonSite Vice President United States Nuclear Regulatory CommissionSerial: RNP-RA/14-0097Page 3 of 3WRG/amEnclosure 1:RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TOTHE PRESSURIZED WATER REACTOR INTERNALS PROGRAM PLAN FORAGING MANAGEMENT OF REACTOR INTERNALSH. B. Robinson Unit 2 Summary Report for the Cold Work AssessmentEnclosure 2:RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TOTHE PRESSURIZED WATER REACTOR INTERNALS PROGRAM PLAN FORAGING MANAGEMENT OF REACTOR INTERNALSH. B. Robinson Unit 2 Summary Of Inspection Results for the Clevis Insert BoltsRESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TOTHE PRESSURIZED WATER REACTOR INTERNALS PROGRAM PLAN FORAGING MANAGEMENT OF REACTOR INTERNALSEnclosure 3:Westinghouse Technical Bulletin (TB), Reactor Internals Lower Radial Support Clevis Insert CapScrew Degradation (TB-14-5)cc: Mr. V. M. McCree, NRC, Region IIMs. Martha Barillas, NRC Project Manager, NRRNRC Resident Inspector, HBRSEP Unit No. 2 United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 1 of 15 (including cover)ENCLOSURE 1RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO THEPRESSURIZED WATER REACTOR INTERNALS PROGRAM PLAN FOR AGING MANAGEMENTOF REACTOR INTERNALS (TAC NO. ME9633) DOCKET NO. 50-261H. B. Robinson Unit 2 Summary Report for the Cold Work Assessment United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 2 of 15 (including cover)In Request for Additional Information RAI-3-1 [1 and 5], the U.S. Nuclear RegulatoryCommission (NRC) advised Duke Energy that resolution of Applicant/Licensee Action Item(A/LAI) 1 of MRP-227-A would need to be resolved as part of the staff's review of the H. B.Robinson License Renewal Amendment and Pressurized Water Reactor (PWR) VesselInternals Program."RAI-3-1 (Action Item 1 of the NRC staff's SE for MRP-227-A):Do the RNP RVI have non-weld or bolting austenitic stainless steel components with20% cold work or greater, and if so do the affected components have operating stressesgreater than 30 ksi? If so, perform a plant-specific evaluation to determine the agingmanagement requirements for the affected components."Westinghouse has evaluated the H. B. Robinson reactor internals components according toindustry guideline MRP 2013-025 [2], as well as the MRP-191 [3] industry generic componentlistings and screening criteria (including consideration of cold work as defined in MRP-175 [4],noting the requirements of Section 3.2.3). In addition to consideration of the materialfabrication, forming, and finishing process, a general screening definition of a resulting reductionin wall thickness of 20% was applied as an evaluation limit. It was confirmed that all H. B.Robinson Unit 2 components, as applicable for the design, are included directly in the MRP-191component lists.The evaluation included a review of all plant modifications affecting reactor internals and theplant operating history. The components were procured according to ASTM International orASME material specifications through applicable quality-controlled protocols. H. B. RobinsonUnit 2 components were binned according to the following categories for material fabricationand cold work potential:1. Cold work categories include the following:cast austenitic stainless steel (CASS) (Category 1)hot-formed austenitic stainless steel (Category 2)annealed austenitic stainless steel (Category 3)fasteners austenitic stainless steel (Category 4)cold-formed austenitic stainless steel without subsequent solution annealing(Category 5)2. Cold work potential was based on MRP-227-A generic criteria:No (N) typically applies to cold work Categories 1, 2, and 3.Yes (Y) typically applies to cold work Categories 4 and 5.Where multiple options existed for a component or assembly, the bounding condition, taken asincluding cold work, was selected for the purpose of the assessment. In some instancescascading fabrication would appear to mitigate any potential for cold work; however, since thehistorical record was not detailed, the potential is noted, but a conservative approach wasselected for this assessment.
United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 3 of 15 (including cover)The evaluation, consistent with industry guidelines, concluded that the reactor internalsCategories 1, 2, and 3 (non-bolting) components at H. B. Robinson Unit 2 contain no cold workgreater than 20% as a result of material specification and controlled fabrication construction.Category 4 components were already assumed to have the potential for cold work in theMRP-191 generic assessments. Material fabrication specifications used for H. B. Robinson Unit2 would suggest that processes were limiting and precluded the introduction of significant coldwork in some of the Categories 4 and 5 components. Category 4 components with greater than20% cold work have been previously dispositioned in MRP-191 [3]. This disposition isaccounted for in the aging management program; therefore, H. B. Robinson Unit 2 requires noadditional stress analyses for Category 4 components. Following MRP 2013-025 guidelines [2],Category 5 components with greater than 20% cold work are to be analyzed for stresses. H. B.Robinson Unit 2 contains no Category 5 components with greater than 20% cold work;therefore, no stress analyses were performed. The detailed evaluation for the A/LAI for H. B.Robinson Unit 2 cold work assessments concluded both that the plant-specific materialfabrication and design were consistent with the MRP-191 basis and that the MRP-227-Asampling inspection aging management requirements, as related to cold work, are directlyapplicable to H. B. Robinson Unit 2. Therefore, a plant-specific evaluation to determine theaging management for 20% or greater cold work components with operating stresses greaterthan 30 ksi is not needed for H. B. Robinson Unit 2.References1. Email from Siva Lingam (NRC) to Richard Hightower (PGN), "Robinson Unit 2 PWRVessel Internal Program Plan for Aging Management -Requests for AdditionalInformation (RAIs) (TAC No. ME9633) -Correction to RAI-3-3," September 5, 2013.(NRC ADAMS Accession No.: ML13249A000)2. Materials Reliability Program Letter, MRP 2013-025, "MRP-227-A Applicability TemplateGuideline," October 14, 2013.3. Materials Reliability Program: Screening, Categorization, and Ranking of ReactorInternals Components for Westinghouse and Combustion Engineering PWR Design(MRP-191). EPRI, Palo Alto, CA: 2006. 1013234.4. Materials Reliability Program: PWR Internals Material Aging Degradation MechanismScreening and Threshold Values (MRP-175). EPRI, Palo Alto, CA: 2005. 1012081.5. Duke Energy Letter, "Response to NRC Request for Additional Information Related tothe Pressurized Water Reactor Internals Program Plan for Aging Management ofReactor Internals (TAC No. ME9633)," January 9, 2014. (NRC ADAMS Accession No.:ML14014A097)
United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 4 of 15 (including cover)ENCLOSURE 2RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO THEPRESSURIZED WATER REACTOR INTERNALS PROGRAM PLAN FOR AGING MANAGEMENTOF REACTOR INTERNALS (TAC NO. ME9633) DOCKET NO. 50-261H. B. Robinson Unit 2 Summary Of Inspection Results for the Clevis Insert Bolts United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 5 of 15 (including cover)AHB Robinson -RO-27~PruuassEnmvR~doreaile arL a Head at the u4A)Claris Insert Keyways and LowerRadial Supportl Lugs loo" foradebris, er deform-ai andabramilcndilone (0.90, 1181&270 degrees). (" Sid320)Ciash kwuau K I ayii NaA M- 12.02-13 SECT-XI Non-RoilarlN/A r-3CoEN Lengefaied M MIII 095 n3dricebmnsucih as nano, scuirrig.IDA) scratchme. gags porn typicallower Internals assorwnmldgi-anesterly operations wiere noted.Prowenoroly naiad indications wereobeorved-nopacung.. NoRelevant inlndcaionnaTotal Coerarge MW%AE InnLamer ClvisIneairdsadial SuppoutsVT.3 70 V~ 18 Daenns * -2f2O- e iurnd. Narmel(k' PNBd 09:U3227 wdah W'ing apaithons PrevMulyna change.2012-0 ~ i22. 16 otherthwa prolvionity adwtliedKoyweyL h4PExamined By: Victor Morton (VM)Examined By Deras Eri (DNB)Level IlI Review: Andrew Clay (AWCLevel IlI Review Dan Lengenfeld (DL)Level: III Date* 2012021Level: II Date: 20201Date: 202-02-16Date 2012-01m I I United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 6 of 15 (including cover)ENCLOSURE 3RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION RELATED TO THEPRESSURIZED WATER REACTOR INTERNALS PROGRAM PLAN FOR AGING MANAGEMENTOF REACTOR INTERNALS (TAC NO. ME9633) DOCKET NO. 50-261Westinghouse Technical Bulletin (TB), Reactor Internals Lower Radial Support Clevis Insert Cap ScrewDegradation (TB- 14-5)
United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RN14-0097Page 7 of 15 (including cover) Westinghouse Non-Proprietary Class 3)WestinghouseI9ech-nicalgduye tinAn advisory of a recent technical development pertaining to the installation or operation of Westinghouse-supplied nuclear plantequipment. Recipients should evaluate the information and recommendation, and initiate action where appropriate.1000 Westinghouse Drive, Cranberry Township, PA 16066© 2014 Westinghouse Electric Company LLC. All Rights Reserved.Subject: Number:Reactor Internals Lower Radial Support Clevis Insert Cap Screw Degradation TB-14-5System(s): Date:Reactor Coolant System, Reactor Internals, Lower Radial Support System 08/25/2014Affects Safety-Related Equipment S.O.:Yes El No a NAThis Technical Bulletin supersedes InfoGram IG-10-1 (Reference 1).BACKGROUNDWestinghouse IG-1 0-1 was issued to communicate an operating experience (OE) related to clevis insertcap screw (bolt) degradation. Since the issuance of IG-10-1, additional information has become availablerelated to the root cause of the OE that was previously communicated. This Technical Bulletin (TB)provides a summary of the OE as well as root cause findings and the applicability of these findings onWestinghouse and Combustion Engineering (CE) pressurized water reactor designs. This TB alsoreviews the safety implications of the OE and root cause analysis results as well as inspectionrecommendations for licensees to consider including as part of their aging management program toaddress this OE.During a 10-year in-service inspection (ISI) in March 2010, an anomaly was observed in the lower radialsupport clevis inserts. The 10-year ISI program includes a remotely operated visual inspection of thevessel after the core barrel is removed. This visual inspection examines the condition of the six clevisinserts, including general integrity of bolted and welded connections. Figure 1 provides an overview ofthe radial support clevis insert configuration. Each clevis insert at the plant inspected features 8 capscrews and 2 dowel pins which provide for retention of the insert. At a total of 7 (out of 48) boltlocations, there was visual evidence that the bolt head had detached from the shank. On one of the clevisdowel pins, the lock welds on the pin had broken and the pin was slightly rotated and displaced deeperinto the insert. Reviews of video images from the previous 10-year ISI vessel inspection showed noindications of wear, fractures, or other anomalies with the clevis insert cap screws or dowel pins at anylocation.Additional information, if required, may be obtained from Bryan Wilson, (412) 374-3281Author: Reviewer Manager:William J. Smoody John T. Crane James A. GreshamRegulatory Compliance Regulatory Compliance Regulatory ComplianceVerifier: Verifier:Bryan M. Wilson Michael A. Burke, PhDReactor Internals Design and Analysis I Primary Systems Design and RepairNeither Westinghouse Electric Company nor its employees make any warranty or representation with respect to the accuracy, completeness or usefulness of theinformation contained in this report or assume any responsibility for liability or damage which may result from the use of such information.Electronically approved records are authenticated in the electronic document management system United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RAJ14-0097Page 8 of 15 (including cover)TB-14-5Page 2 of 9BACKGROUND (Continued)During the spring 2013 refueling outage, 29 bolts were removed and replaced. The primary concern wasthat the clevis insert bolt would continue to degrade eventually leading to a condition where the clevisinsert would be able to completely dislodge once the lower internals were removed. When the internalsare assembled into the reactor vessel the clevis insert is effectively trapped by the interfacing lowerinternals components. Once the lower internals are removed, and if there is significant degradation of thedesigned fits and hardware, there is a potential for the insert to be dislodged from the mating vessel lug.Once dislodged from the vessel lug, reinstallation was viewed as being of high commercial consequencedue to the customization needed to reestablish the required design gaps for this component. The numberof replacement bolts was determined based on structural evaluations of the bolts assuming that none ofthe original bolts were functional. Evaluations (discussed later) showed that no bolts were required tomaintain the intended safety function. The evaluations, to determine the number of required replacementbolts, were primarily focused on ensuring that the bolts would remain functional for the remaining life ofthe plant.Dowel Pin Cap ScrewLocation Locations/ ___________ ____View AoAFigure 1 -Example of Lower Radial Support Clevis Insert, Cap Screw and Dowel Pin Locations United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 9 of 15 (including cover)TB-14-5Page 3 of 9ROOT CAUSE ANALYSIS RESULTS SUMMARYThe licensee's root cause analysis identifies the cause of the clevis insert bolt failures to be primary waterstress corrosion cracking (PWSCC) due to the use of Alloy X-750 with a susceptible heat treatment.Additional details of particular interest are as follows:-Of the 29 bolts removed, 13 were originally considered intact (later found to be cracked) and 16were fractured.-Of the 16 bolts found to be fractured, only 8 showed signs of wear between the head and the lockbar during visual inspections prior to removal.-All 29 bolts examined contained cracking in the head-to-shank transition.-Fractographic scanning electron microscopy (SEM) analysis and cross section metallographicexaminations confirmed the fracture mode was essentially 100% intergranular on all of the bolts.-There was no evidence that the bolts failed due to fatigue cracking or mechanical overload.While the conclusion of PWSCC as the failure cause is consistent with the supposition made inReference 1, the new information related to the extent of cracking suggests that clevis insert bolt crackingis potentially a latent issue which has not yet manifested itself as visual indications at other plants. Thisnew information also confirms that visual inspection is less than adequate in detecting clevis insert boltfailures. However, as discussed herein, the lower radial support structure will maintain its intended safetyfunction despite bolt failures.PLANT APPLICABILITYThis issue potentially impacts all Westinghouse- and CE-designed plants. All Westinghouse and CEplant designs have been confirmed to contain a similar susceptible heat treatment of Alloy X-750 to thatused in the subject cap screws. Furthermore, preload stresses in the Alloy X-750 clevis insert bolts (orsnubber lug bolts in the case of the CE design) have generally been found to be consistent across the fleet.While there are differences among the designs used across the Westinghouse and CE fleets, thesedifferences are not considered to have a significant impact on susceptibility.SAFETY IMPLICATIONSIn support of an operability assessment, Westinghouse performed engineering evaluations of the as-foundcondition, including an evaluation of the potential for loose parts. The loose parts evaluation concludedthat the separated cap screw heads will remain captured in the clevis insert counterbores and will notimpact operation. However, lock bars at the degraded cap screw locations were observed to haveexperienced wear-related degradation; therefore, the potential for loose parts from the lock bars to affectother locations in the reactor vessel was evaluated. Westinghouse concluded that no degradation ofmechanical components is expected as a result of the potential presence of loose parts from the lock barsin the primary system.Westinghouse also performed a detailed structural evaluation of the as-found condition of the clevisinserts. The analysis demonstrated that failure of the bolts would not result in a loss of safety function.Furthermore, it was demonstrated that a significant number of redundancies prevent the loss of theintended safety function.Similar evaluations to those performed for the subject clevis insert design have been performed for theother three basic Westinghouse clevis insert designs. These evaluations also considered the results of theroot cause analysis by assuming all bolts on the same insert were non-functional. The conclusionsresulting from these evaluations were consistent with the conclusions reached for the evaluationsperformed for the plant, which are:
United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 10 of 15 (including cover)TB-14-5Page 4 of 9-Failure of the bolts would not result in a loss of safety function-Significant number of redundancies prevent the loss of the intended safety function-Concerns related to bolt failures are generally non-safety in natureNo formal evaluations have been conducted for the CE snubber lug designs, which are in a similarlocation in the reactor as the Westinghouse lower radial support, have the same intended safety function,and also use Alloy X-750 attachment bolts. However, based on a comparison of the design features,installation, and loading, it is concluded that the snubber lug will be able to perform its intended safetyfunction in the event of bolt failures. Furthermore, the snubber lug bolt heads would remain trapped andwould not become loose parts.MAINTENANCE IMPLICATIONSWhile clevis insert bolt failures do not present a safety concern, continued operation with degraded clevisinsert hardware increases the likelihood and consequence of maintenance risks due to the complexity andcost of repair, and the required level of contingency planning. The primary risks are related to significantloosening and disengagement of the clevis insert, which could result in difficulty or inability to removethe core barrel or need to replace the insert to reestablish customized design gaps. Either of these wouldhave a significant outage impact. As previously mentioned, maintenance considerations were one of theprimary considerations which led to replacement of bolts. Other potential implications of operation withfailed clevis insert bolts include the following:-Bolt remnants wearing on clevis insert bolt counterbores, seating surfaces, or mating threadscould lead to additional scope for reconstituting these surfaces prior to bolt replacement.-Bolt head remnants wearing into adjacent surfaces of the radial key could lead to wedging of thebolt head during barrel removal.-Worn lock bars releasing from the insert and becoming loose parts has been evaluated and shownto not be a safety concern. However, as the length of operation with failed clevis insert boltsincreases, the probability of lock bars releasing from the insert increases. Furthermore, thepotential for these lock bars to tumble, break, or continue to wear into smaller fragments withinthe internals also increases. If they become small enough in size, they can potentially enter thefuel and lead to cladding wear, potentially including wear-through (leaking fuel) or foreign objectsearch and retrieval (FOSAR) activities.-Increased time for bolt replacement if a significant number of bolts are broken. It generallyrequires much less effort to remove intact bolts versus separated bolts.Maintenance risks should be considered in evaluating the inspection methods used for managing the agingof clevis insert bolts.INSPECTION RECOMMENDATIONSThe U.S. nuclear industry-through the Materials Reliability Program (MRP)-has developed Inspectionand Evaluation (I&E) guidelines (MRP-227-A; Reference 2) for the management of reactor internals age-related degradation issues in the U.S. pressurized water reactor (PWR) fleet. MRP-227-A focuses onthose internals components identified as susceptible to aging effects and provides information to supporteffective aging management, while simultaneously maintaining safety and reliability. The inspection andevaluation recommendations in MRP-227-A were issued under the Nuclear Energy Institute (NEI)NEI 03-08 (Reference 3) protocol for industry wide implementation.
United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 11 of 15 (including cover)TB-14-5Page 5 of 9Westinghouse-Designed Plants:In MRP-227-A, the clevis insert bolts for the Westinghouse-designed plants are classified as an ExistingPrograms component managed for loss of material (wear). To manage this mechanism, MRP-227-Arecommends the use of the existing ASME Section XI visual inspection (VT3) of all accessible surfaces.While this broadly covers the inspection scope and remains the recommended approach to agingmanagement of the clevis insert bolts, there are some clarifications needed on scope of inspection toensure that aging management inspection programs are focused on monitoring conditions that are directlyrelated to the safety function of the component.As Note 2 of Table 4-5 of MRP-227-A indicates, the bolts were screened-in because of stress relaxationand associated cracking. However, the failures of the bolts alone do not result in a loss of the intendedsafety function of the lower radial supports. The loss of function of the lower radial supports is linkedmore closely to increased motion of the lower end of the core barrel caused by wear on the clevis insert orvessel attachment interface. Therefore, although it is one of many barriers to loosening of the insert,management of the bolt failures does not directly address management of the functional performance ofthe component. To ensure that the functional performance is being managed, it is recommended to ensurethat the MRP-227-A inspections (ASME Section XI, VT3) for the clevis inserts include the followingscope:-Radial key/clevis insert interfacing surfaces; look for aggressive or abnormal wear as comparedto previous inspection, if available. The radial key does not make contact with the full length ofthe clevis insert; if wear is significant it would be visible as a step located toward the bottom endof the clevis insert as shown in Figure 2.-Interface between the clevis insert and vessel lug (see Figure 3); look for signs of looseness ordislocation. Faces of the insert and vessel lug are generally flush; dislocations may be visible bythe insert protruding toward the vessel centerline as compared to the vessel lug.Figure 2 -Example of Wear at the Radial Key/Clevis Insert Interfacing Surface United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 12 of 15 (including cover)TB-14-5Page 6 of 9Figure 3 -Example of Dislocation of the Clevis Insert from the Vessel LugWhile managing the functional performance is the primary goal of the recommended inspections, this TBrecommends that the scope of the inspection also include the following as a means of reducingmaintenance risk associated with clevis insert bolt degradation (Figure 4):-Look for wear between the bolt head and lock bar and/or bolt head dislocation.-Look for broken tack welds and dislocation of the dowel pin.Figure 4 -Examples of Bolt and Dowel Pin Related DegradationWhile these methods have been shown to be less than adequate in detecting bolt failure, the indicationsdescribed have been shown in one case to manifest themselves as a result of bolt failure and prior to anynotable loosening of the clevis insert or increased wear between the radial key and clevis insert.
United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 13 of 15 (including cover)TB-14-5Page 7 of 9Some optional methods that could be used to improve the reliability of crack detection and enhance themanagement of maintenance risk include activities such as ultrasonic examination of bolts and/or boltreplacement.CE-Designed Plants:For the CE-designed plants, the components similar to the clevis insert bolts are the core stabilizing lugbolts. Neither these bolts, nor the core stabilizing lug shims (CE components similar to the Westinghouseclevis inserts) were screened in as part of MRP- 191 (Reference 4). As a result, no inspectionrecommendations are provided for these components in MRP-227-A. However, considering these boltsare Alloy X-750, installed with similar preload stress, and at similar operating temperatures as theWestinghouse clevis insert bolts, these bolts are also potentially susceptible to PWSCC. Since thefunction of core stabilizing lugs is identical to that of the Westinghouse lower radial supports, similarinspections for looseness of the insert should be conducted to ensure function is maintained. Specificrecommendations for inspection scope are as follows:-Core support barrel snubber lug/core stabilizing lug shim interfacing surfaces (see Figure 5); lookfor aggressive or abnormal wear as compared to previous inspection, if available. The coresupport barrel snubber lug key does not make contact with the full depth of the core stabilizinglug shim. If wear is significant it would be visible as a step located toward the vessel side of theclevis insert as shown in Figure 5.-Interface between the core stabilizing lug shim and the core stabilizing lug (see Figure 6); lookfor signs of looseness or dislocation. Looseness may be visible by gapping between the corestabilizing lug shim and the core stabilizing lug when looking radially outward toward the lug.Figure 5 -Example of Wear at the Core Support Barrel Snubber Lug/Core Stabilizing Lug ShimInterfacing Surface United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 14 of 15 (including cover)TB-14-5Page 8 of 9Figure 6 -Example of Dislocation of the Shim from the Core Stabilizing LugWhile managing the functional performance is the primary goal of the recommended inspections, it isrecommended that the scope of the inspection also include the following as a means of reducingmaintenance risk associated with clevis insert bolt degradation (Figure 7):-Bolt heads; look for wear between the bolt head and the shim and/or bolt head dislocation.-Bolt threaded end; look for signs of the bolt threads backing. Bolt threads should generally berecessed from the face of the shim; look for differences in the amount of recess between the twobolts threaded into the shim being viewed.-Locking pins; look for dislocation, wear, or failure of the locking pin.Bolt Threaded EndBolt HeadLocking PinFigure 7 -Identification of Hardware United States Nuclear Regulatory CommissionEnclosures to Serial: RNP-RA/14-0097Page 15 of 15 (including cover)TB-14-5Page 9 of 9As previously discussed, these methods may be less than adequate in detecting bolt failure, however theymay manifest themselves as a result of bolt failure and prior to any notable loosening of the corestabilizing lug shim or increased wear between the core support barrel snubber lug and core stabilizinglug shim.Similar to the recommendations for the Westinghouse-designed plants, optional methods such asultrasonic examination of bolts or bolt replacement could be used to improve the reliability of crackdetection and enhance the management of economic risk.CONCLUSIONOther than the single occurrence discussed, Westinghouse is unaware of any other occurrences of clevisinsert bolt degradation. Based on the engineering evaluations performed to date and consideringinformation made available through the investigations conducted on removed failed bolts, there are nosafety or operability concerns to be communicated to the industry. The recommendations for inspectionsof clevis insert bolt-related degradation to enhance focus on the conditions that may result from clevisinsert bolt failures, is based on available OE. Westinghouse will continue to monitor the results of clevisinsert bolt inspections conducted throughout the industry and will communicate to the industry anyrelevant OE which changes the recommendations provided herein.REFERENCES1. Westinghouse InfoGram, IG-10-1, Revision 0, "Reactor Internals Lower Radial Support ClevisInsert Cap Screw Degradation," March 31, 2010.2. "Materials Reliability Program: Pressurized Water Reactor Internals Inspection and EvaluationGuidelines (MRP-227-A)," EPRI, Palo Alto, CA: 2011. 1022863.3. Nuclear Energy Institute, NEI 03-08, Rev. 2, "Guideline for the Management of MaterialsIssues," January 20104. "Materials Reliability Program: Screening, Categorization, and Ranking of Reactor InternalsComponents for Westinghouse and Combustion Engineering PWR Design (MRP-191)," EPRI,Palo Alto, CA: 2006. 1013234.This document is available at https://partner.westinghousenuclear.com/project/12/SP1510/default.aspx. This site is a free service for Westinghouse Electric Company LLC(Westinghouse) customers and other electric power industry-related organizations. Access will be provided based on Westinghouse judgment of appropriate businessaffiliation. Westinghouse reserves the right, at its sole discretion, to grant or deny access to this site. Requests for access should be made to giampora@westinghouse.com.