ML18159A108

Non-Proprietary, Staff Assessment of Action Item 7 Regarding Inspection Plan for Reactor Internals

ML18159A108
Person / Time
Site:	Crane
Issue date:	07/12/2018
From:	Justin Poole Plant Licensing Branch 1
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
Poole J, NRR/DORL/LPLI, 301-415-2048
References
EPID L-2016-LLL-0002
Download: ML18159A108 (17)

Text

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION July 12, 2018 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION, UNIT 1 - STAFF ASSESSMENT OF ACTION ITEM 7 REGARDING INSPECTION PLAN FOR REACTOR INTERNALS (EPID L-2016-LLL-0002)

Dear Mr. Hanson:

By letter dated September 16, 2016, as supplemented by letter dated April 18, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession Nos.

ML16263A338 and ML18108A287, respectively), Exelon Generation Company, LLC, (the licensee) submitted to the U.S. Nuclear Regulatory Commission (NRC) its evaluation of applicant/licensee action (A/LAI) 7 in accordance with the safety evaluation in MRP-227-A Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (ADAMS Accession No. ML120170453) for Three Mile Island Nuclear Station (TMI), Unit 1.

The NRC staff has reviewed the licensees provided evaluation and determined that the licensee adequately demonstrated that the functionality of the control rod guide tube spacer castings, in-core monitoring instrumentation guide tube spider castings, and vent valve retaining rings at TMI, Unit 1, will be maintained during the period of extended operation. Accordingly, the NRC staff determined that the licensee has adequately resolved A/LAI 7. The NRC staffs assessment is enclosed. contains Proprietary Information. When separated from Enclosure 2, this letter is DECONTROLLED.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION If you have any questions, please contact me at 301-415-2048 or via e-mail at Justin.Poole@nrc.gov.

Sincerely,

/RA/

Justin C. Poole, Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289

Enclosure:

1. Staff Assessment of Action Item 7 (non-proprietary version)

2. Staff Assessment of Action Item 7 (proprietary version) cc w/o Enclosure 2: Listserv

ML18159A030 (OUO-PI): ML18159A108 (Redacted)

• by memo dated OFFICE DORL/LPL1/PM DORL/LPL1/LA DMLR/MVIB/BC*

DORL/LPL1/BC DORL/LPL1/PM NAME JPoole SRohrer DAlley (JPoehle for)

JDanna JPoole DATE 6/13/18 6/12/18 5/29/18 6/18/18 7/12/18

NONPROPRIETARY VERSION STAFF ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION MRP-227-A APPLICANT/LICENSEE ACTION ITEM 7 EVALUATION THREE MILE ISLAND NUCLEAR STATION, UNIT 1 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-289

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION STAFF ASSESSMENT BY THE OFFICE OF NUCLEAR REACTOR REGULATION MRP-227-A APPLICANT/LICENSEE ACTION ITEM 7 EVALUATION THREE MILE ISLAND NUCLEAR STATION, UNIT 1 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-289

1.0 INTRODUCTION

By letter dated September 16, 2016 (Reference 1), as supplemented by letter dated April 18, 2018 (Reference 2), Exelon Generation Company, LLC (licensee), submitted to the U.S. Nuclear Regulatory Commission (NRC) its evaluation of applicant/licensee action (A/LAI) 7 for Three Mile Island Nuclear Station (TMI), Unit 1, in accordance with the safety evaluation (SE) in MRP-227-A Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines (Reference 3). A/LAI 7 requires, in part, applicants or licensees of Babcock and Wilcox (B&W) reactor designs to develop plant-specific analysis to demonstrate that reactor vessel internals (RVI) components made of cast austenitic stainless steel (CASS) and precipitation-hardened (PH) stainless steel will maintain their functionality during the period of extended operation (PEO). By letter dated December 19, 2014 (Reference 4), the NRC staff issued its assessment of all TMI, Unit 1, RVI components and documented the licensees commitment to submit its evaluation of the CASS and PH stainless steel RVI components in accordance with A/LAI 7 of the SE in MRP-227-A. The purpose of the September 16, 2016, submittal, as supplemented, is to fulfill this commitment. Please note that the NRC staff assessment contains licensee proprietary information and is thus marked accordingly with (( )).

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (10 CFR) Part 54 addresses the requirements for managing the effects of aging components during the PEO, and MRP-227-A specifies inspection and evaluation guidelines for adequately managing aging effects in RVI components. The MRP-227-A inspection and evaluation guidelines must be followed if applicants or licensees implement them for their units. RVI include components that are made of CASS and PH stainless steel,

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION which are susceptible to the following aging degradation mechanisms: thermal embrittlement (TE), irradiation embrittlement (IE), or the synergistic effects of TE and IE. For TMI, Unit 1, the three RVI components that are made of CASS or PH stainless steel are the control rod guide tube (CRGT) spacer castings, the in-core monitoring instrumentation (IMI) guide tube spider castings, and the vent valve retaining rings.

The NRC staff provided the detailed regulatory evaluation of the requirements of 10 CFR Part 54 and the inspection and evaluation guidelines of MRP-227-A in the December 19, 2014, staff assessment of the TMI, Unit 1, implementation of MRP-227-A (Reference 4), in which the licensee committed to submit its evaluation of the CASS and PH stainless steel RVI components in accordance with A/LAI 7 of the SE in MRP-227-A.

3.0 TECHNICAL EVALUATION

Section 4.2.7 Plant-Specific Evaluation of CASS Materials of the SE in MRP-227-A, states:

As discussed in Section 3.3.7 of this SE, the applicants/licensees of B&W, CE

[Combustion Engineering], and Westinghouse reactors are required to develop plant-specific analyses to be applied for their facilities to demonstrate that B&W IMI guide tube assembly spiders and CRGT spacer castings, CE lower support columns, and Westinghouse lower support column bodies will maintain their functionality during the period of extended operation or for additional RVI components that may be fabricated from CASS, martensitic stainless steel or precipitation hardened stainless steel materials. These analyses shall also consider the possible loss of fracture toughness in these components due to thermal and irradiation embrittlement, and may also need to consider limitations on accessibility for inspection and the resolution/sensitivity of the inspection techniques. The requirement may not apply to components that were previously evaluated as not requiring aging management during development of MRP-227.

That is, the requirement would apply to components fabricated from susceptible materials for which an individual licensee has determined aging management is required, for example during their review performed in accordance with Applicant/Licensee Action Item 2. The plant-specific analysis shall be consistent with the plants licensing basis and the need to maintain the functionality of the components being evaluated under all licensing basis conditions of operation.

The applicant/licensee shall include the plant-specific analysis as part of their submittal to apply the approved version of MRP-227. This is Applicant/Licensee Action Item 7.

The NRC staffs assessment of the A/LAI 7 evaluation for TMI, Unit 1, focused on the technical information in proprietary report ANP-3479P MRP-227-A Applicant/Licensee Action Item 7 Analysis for Three Mile Island, Unit 1, Revision 0, which the licensee included as Attachment 1 to the September 16, 2016, letter. The licensee included technical evaluations in ANP-3479P, Revision 0, to demonstrate that the CRGT spacer castings, IMI guide tube spider castings, and vent valve retaining rings will maintain their functionality during the PEO, and thus to demonstrate that loss of fracture toughness due to TE, IE, or both TE and IE, with respect to these components will be adequately managed during the PEO.

The NRC staff reviewed the methodology that the licensee used for the technical evaluations in ANP-3479P for the TMI, Unit 1, CRGT spacer castings, IMI guide tube spider castings, and vent valve retaining rings. The basic elements of this approach are identifying appropriate inputs for

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION evaluation, determining the likelihood of failure of the component, and determining the effect of a postulated failure on functionality of the component. The NRC staff finds the methodology the licensee used for the technical evaluations in ANP-3479P acceptable.

The following sections describe the function of each of the three subject components, summarize the licensees technical evaluation of the component, and provide the NRC staffs assessment of the licensees evaluation. By letter dated February 26, 2018 (Reference 5), the NRC staff issued requests for additional information (RAIs) to support its assessment.

3.1 Assessment of the CRGT Spacer Castings 3.1.1 Description and Function of the CRGT Spacer Castings The CRGT spacer castings are part of the brazement subassembly, which includes the vertical control rod guide tubes and vertical control rod guide sectors. There are ((10 parallel CRGT spacer castings)) per brazement subassembly. The function of the CRGT spacer castings is to provide structural support and alignment to the vertical control rod guide tubes and vertical control rod guide sectors. The brazement subassembly guides the control rod assembly over the entire range of the vertical withdrawal or insertion path. The CRGT spacer castings do not have a core support function, but broken CGRT spacer castings could hinder the insertion of the control rods into the core within the normal anticipated time, and thus hindering shutdown capability of the reactor.

The CRGT spacer castings are made from CF3M grade CASS and, per Table 4-1, B&W plants Primary components of MRP-227-A, are susceptible to TE.

3.1.2 Licensees Evaluation The licensee provided a detailed evaluation of the TMI, Unit 1, CRGT spacer castings in Section 3.0 CRGT Spacer Castings of ANP-3479P. The evaluation is summarized below.

a) Regarding the possibility of existing flaws: The licensee reviewed available certified material testing reports (CMTRs) for the CRGT spacer castings and determined that the CRGT spacer castings received ((satisfactory penetrant testing (PT) and radiographic testing (RT) evaluations at the time of construction)). ((The licensee stated that information on actual flaw sizes was not retrievable)). Therefore, given that the CRGT spacer castings received ((satisfactory PT and RT evaluations at the time of construction, the licensee concluded that no relevant indications larger than acceptable pre-service indications existed at that time)).

b) Regarding degraded material properties: The licensee stated that ((a large percentage))

of the CRGT spacer castings are susceptible to TE. The licensee investigated the time to reach saturated material property values (such as impact energy, and, correspondingly, fracture toughness) for the CRGT spacer castings and determined that saturation ((was reached prior to the PEO)).

c) Regarding the likelihood of failure: The licensee stated that visual testing (VT)-3 per MRP-227-A guidance of 100 percent of the accessible surfaces at each of the four screw locations of the CRGT spacer castings, performed between 2012 and 2014 at three B&W units, revealed no recordable indications. The licensee also stated that there is no known failures of CASS materials due to embrittlement that have been reported in the industry.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION Finally, the licensee explained that since there is ((no apparent cracking mechanism in the CRGT spacer castings, large defects due to service loading are not expected)). For these reasons, and considering that the material properties of the CRGT spacer castings have ((reached saturation prior the PEO)), the licensee concluded that failure of the CRGT spacer castings during the PEO is unlikely.

d) Regarding the effect of failure on functionality: The licensee evaluated the amount of distortion that will permit the control rods to freely pass through the brazement subassembly and determined that ((a single failure at any of the four screw locations in one CRGT spacer casting would not lead to a restricted guide path for the control rods.

The licensee stated that if two screw locations in the same CRGT spacer castings were to fail, control rod alignment could be affected. However, the licensee explained that two failed screw locations in the same CRGT spacer castings are unlikely because this postulated condition would require a precise location for two existing defects, and the two failures would essentially have to occur at the same time)). Furthermore, the licensee stated that stress analysis of the CRGT spacer castings reinforces why simultaneous failures at two screw locations are unlikely. First, the stress analysis shows that highest stresses are at the four screw locations, which implies that flaws must be ((in these locations to have sufficient stress to result in failure)). Second, the stress analysis showed that ((after one failure occurs at a screw location, the balance of the casting stresses are either significantly reduced or remain essentially the same as before the failure)).

The licensee stated that the brazement subassembly has two redundant features: ((the 10 parallel CRGT spacer castings per brazement subassembly and the brazed connection of CRGT spacer castings to the vertical control rod guide tubes and vertical control rod guide sectors)). The licensee analyzed the brazement subassembly and determined that a single failure at any screw location would be acceptable and not restrict the control rod guide path, and that multiple single failures of CRGT spacer castings in the same brazement subassembly are also acceptable. Additionally, the ((stiffness of the vertical control rod guide tubes and sectors provide some support to a degraded spacer casting)). Also, ((if the failed location of the spacer casting is at any location other than the screw locations, the screws themselves would retain the spacer casting segment in-place even if preload is lost)).

Finally, the licensee stated control rod drop-times are tested at the beginning of each cycle per the TMI, Unit 1, technical specifications (TSs) to ensure they are within the requirements. The licensee investigates unusual drop-times when they occur. The licensee stated that to date, slow trip times have been due to unusual fuel bow or issues with the control rod drive mechanism.

For the reasons in the preceding paragraphs in item (d), the licensee concluded that ((no additional actions are required for the CRGT spacer castings)).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.1.3 NRC Staffs Assessment The NRC staff reviewed the information in Section 3.0 of ANP-3479P for the CRGT spacer castings, as summarized in items (a) through (d) in Section 3.1.2 above, and provides its findings below for each corresponding item.

a) The NRC staff finds it reasonable to assume that the CRGT spacer castings received

((satisfactory PT and RT evaluations at the time of construction)) based on the licensees review of the CMTRs. Additionally, the NRC staff verified in MRP-189, Revision 1 (Reference 6), that the CRGT spacer castings do not screen in for ((irradiation assisted stress corrosion cracking (IASCC), stress corrosion cracking (SCC), and fatigue)),

mechanisms that can extend cracks. Accordingly, since the CRGT spacer castings received ((satisfactory PT and RT evaluations at the time of construction)), the NRC staff finds the licensees assumption that ((no relevant indications larger than acceptable pre-service indications)) to be reasonable ((even though information on actual flaw sizes was not retrievable)).

b) In the April 18, 2018, supplement, in its response to RAI-1, the licensee explained how it concluded that the CRGT spacer castings ((reached saturation prior to the PEO)). The licensee referenced ((NUREG/CR-4513, Revision 1 (Reference 7), and stated that fracture toughness of CF3 grade CASS saturates between 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> to 1,000,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> or more and that saturation would be reached sooner for the CRGT spacer castings since they made of CF3M grade CASS, which contains molybdenum)). The licensee did not specify a value of saturated fracture toughness and stated that the reduction in Charpy impact energy with time correlates with a reduction in fracture toughness.

The NRC staff reviewed ((Section 3.2.2 of Fracture Toughness J-R Curve of NUREG/CR-4513)) which includes a discussion of the correlation between Charpy impact energy and fracture toughness for CASS. Based on this review, the NRC staff finds acceptable that a reduction in Charpy impact energy with time correlates with a reduction in fracture toughness with time. The staff reviewed the information in ((Figure 19 of NUREG/CR-4513)), which shows Charpy impact energy plotted against aging time for heats of CF3 grade CASS. The staff noted that ((the Charpy impact energy of some CF3 heats do not saturate between 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> to 1,000,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. However, the staff finds it reasonable that the CRGT spacer castings, which are made of molybdenum-bearing CF3M grade CASS, would saturate sooner compared to CF3 grade CASS that does not have molybdenum)). The guidance for TE screening of CASS (Reference 8, Grimes letter), which shows CASS having high molybdenum content being more susceptible to TE, supports this conclusion. Therefore, the NRC staff finds the licensees conclusion that the CRGT spacer castings have ((reached saturation before the PEO))

acceptable.

c) The NRC staff finds that the operating experience (OE) for the three B&W units and the industry-wide OE for CASS materials are favorable with regards to cracking of CRGT spacer castings. As previously mentioned, the NRC staff verified in MRP-189, Revision 1, that the CRGT spacer castings do not screen in for ((IASCC, SCC, and fatigue)).

Therefore, the NRC staff finds that the assumption that ((large defects due to service loads do not exist in the CRGT spacer castings)) is reasonable. The NRC staff noted that TMI, Unit 1, entered the PEO on April 19, 2014. Therefore, considering that the CRGT spacer casting material ((has already reached saturation prior to the PEO (if there were

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION casting failures due to embrittlement, they would likely have occurred already and been detected during inspection))), the favorable OE, ((the improbability of defects due to service loading, and no cracking apparent cracking mechanisms)), the staff finds acceptable the licensees conclusion that failure of the CRGT spacer castings during the PEO is unlikely. The staff also notes that its findings on the stress analysis of the CRGT spacer castings discussed in item d below support the conclusion that failure of the CRGT spacer castings during the PEO is unlikely.

d) The NRC staff reviewed the distortion discussion and determined that ((the brazement subassemblys design has inherent stiffness that would prevent control rod misalignment should a single failure occur)), as discussed in the next paragraph. The staff also reviewed the licensees stress analysis and finds that it corroborates the conclusion that

((failure occuring at two screw locations in the same CRGT spacer castings is unlikely)).

In the April 18, 2018, supplement, in its response to RAI-2, the licensee stated that ((peak stress in the CRGT is less than 60 ksi (kilopound per square inch), is due to the preload on the screws, and would decrease in the event one screw fails)). The NRC staff, therefore, finds the licensees conclusion that ((a single failure at a screw location would not lead to a restricted guide path for the control rods)) to be acceptable.

The NRC staff reviewed and confirmed the two redundant features of the brazement subassembly of which the CRGT spacer castings are a part. The staff reviewed the information in Section 2.3.4 Control Rod Guide Tube Assembly of MRP-189, Revision 1, and determined that ((the brazement subassemblys design has inherent stiffness that would adequately prevent control rod misalignment should a single failure of one CRGT spacer casting or multiple single failures in the same brazement subassembly occur)).

The NRC staff noted that control rod drop-time testing per the TMI, Unit 1, TSs would detect unusual drop-times, whether the cause is related to the degraded CRGT spacer castings or not. This technical specification requirement provides the staff reasonable assurance that the licensee would take the proper action to correct unusual control rod drop-times should they occur.

Based on the preceding paragraphs in item (d), the NRC staff finds that the licensee adequately considered the effect of postulated CRGT spacer castings failures on the components functionality.

Based on the discussion in items (a) through (d) above, the NRC staff determined that the licensee adequately demonstrated that it will maintain the functionality of the CRGT spacer castings during the PEO. Accordingly, the staff determined that the licensee will adequately manage the aging effects of the CRGT spacer castings during the PEO.

3.2 Assessment of the IMI Guide Tube Spider Castings 3.2.1 Description and Function of the IMI Guide Tube Spider Castings The IMI guide tube spider castings resemble a four-leafed butterfly nut, with each leaf or leg welded to one of four walls of a cubbyhole of the lower grid rib section. The center hub of an IMI guide tube spider casting slides over, ((with tight tolerance fit)), the upper tip of the IMI guide tube. The function of the IMI guide tube spider casting is to provide lateral restraint for the IMI guide tube and the function of the spider fillet welds is to hold the IMI guide tube spider casting in place. The IMI guide tube provides a continuous protected guide path for the IMI from their entry

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION into the reactor pressure vessel (RPV), through the RPV instrumentation nozzles, to the entrance into the fuel assembly instrument guide tube. ((The IMI guide tube spider castings (including their welds) have a minimal role in supporting the weight of the IMI guide tube and do not have a core support function)), but broken IMI guide tube spider castings (including their welds) could impair entry of IMI. Impaired entry of IMI could adversely affect monitoring of core parameters critical to ensuring reactor safety.

The IMI guide tube spider castings are made from CF8 grade CASS and, per Table 4-1 of MRP-227-A, are susceptible to both TE and IE.

3.2.2 Licensees Evaluation The licensee provided a detailed evaluation of the TMI, Unit 1, IMI guide tube spider castings in Section 4.0 IMI Guide Tube Spider Castings of ANP-3479P. The evaluation is summarized below.

a) Regarding the possibility of existing flaws: The licensee reviewed available ((CMTRs for the IMI guide tube spider castings, and determined that the IMI guide tube spider castings received nondestructive examination, which included RT. The licensee stated that information on actual flaw sizes was not available. The licensee also evaluated and screened-out service-induced flaws resulting from IASCC, SCC, and fatigue, for the TMI, Unit 1 guide tube spider castings. Therefore, given that the IMI guide tube spider castings received RT, the licensee concluded that subsurface defects due to the fabrication process in the IMI guide tube spider castings, if they exist)), are limited to ASME (American Society of Mechanical Engineers) Code allowable sizes in castings for pressure boundary components.

b) Regarding degraded material properties: The licensee stated that there is, generally, lack of fracture toughness data, especially in the neutron fluence range relevant to the TMI, Unit 1, IMI guide tube spacer castings. This neutron fluence range is ((1 - 2 displacements per atom (dpa), or specifically, ~1.8 dpa at 54 effective full power years (EFPY))). The licensee explained that at this fluence level, the IMI guide tube spider castings retain enough fracture toughness such that the ((mode of fracture is stable crack extension near or above the yield stress)) based on the fracture toughness categorization defined in NUREG/CR-7027 (Reference 9). Additionally, to address the effect of IE, the licensee referenced an NRC evaluation included in the Grimes letter (Reference 8) and the results of a joint effort on the effect of neutron fluence on fracture toughness by the Boiling Water Reactors Vessel and Internals Program and Materials Reliability Program (BWRVIP/MRP).

Specifically, the licensee determined that ((the reduced fracture toughness of the TMl, Unit 1, IMI guide tube spider castings after 54 EFPY of IE is greater than)) the fracture toughness criterion in the Grimes letter, and, thereby, concluded that the fracture toughness reduction due to IE is not significant. The licensee stated that the IMI guide tube spider castings are not ((susceptible to TE based on low ferrite content)).

c) Regarding the likelihood of failure: The licensee concluded that the IMI guide tube spider castings are unlikely to fail due to IE for three reasons: (1) ((large pre-service flaws or service-induced flaws do not exist)) for the reasons summarized in item (a); (2) the dominant crack driving force is ((self-limiting weld residual stress, which means that as a flaw grows, weld residual stress dissipates as constraint is lost and crack extension ceases)); and (3) the IMI guide tube spider castings ((retain significant fracture toughness after 54 EFPY of operation)) for the reasons summarized in item (b).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION d) Regarding the effect of failure on functionality: The licensees structural analysis demonstrated that stresses due to ((steady state reactor coolant flow)), safe shutdown earthquake (SSE), and LOCA ((were low)) and that ((weld residual stress was the dominant stress)). The structural analysis also shows that ((weld residual stress tend to decrease as additional legs fail)). Additionally, the flow induced vibration analysis shows that ((none of degraded configurations will experience vibration due to flow excitation)). For these reasons, the licensee concluded that ((one failed leg)) in a IMI guide tube spider casting

((will not increase the probability of failure of additional legs)) in the same casting. Having established that ((only one failed leg in one IMI guide tube spider casting is expected to fail)), the licensee further reasoned that ((the remaining legs and the inherent design of the IMI guide tube provide sufficient stiffness and lateral support to maintain the function of the IMI guide tube spider casting)).

3.2.3 NRC Staffs Assessment The NRC staff reviewed the information in Section 4.0 of ANP-3479P for the IMI guide tube spider castings, as summarized in items (a) through (d) in Section 3.2.2 above, and provides its findings below for each corresponding item.

a) The NRC staff finds it reasonable to assume that the IMI guide tube spider castings received ((nondestructive examination that included RT based on the licensees review of the CMTRs)). The NRC staff verified in MRP-189, Revision 1, that the IMI guide tube spider castings do not screen in for ((IASCC, SCC, and fatigue)). Since the IMI guide tube spider castings received ((RT)) and cracking mechanisms are not present, NRC staff finds it reasonable that ((only subsurface fabrication defects)) deemed acceptable by the ASME Code exist in the IMI guide tube spider castings even though information on

((actual flaw sizes was not available)).

b) The NRC staff established its position on loss of fracture toughness for CF3 and CF8 grade CASS (with less than 20 percent ferrite) due to IE and TE in its SE of BWRVIP-234 (Reference 10). The licensee stated that the IMI guide tube spider castings (made of CF8 grade CASS) have ((low ferrite content)), but did not specify a value. The pressurized water reactors owners group (PWROG) issued report PWROG-15032-NP (Reference 11) to evaluate loss of fracture toughness for CF8 grade CASS due to TE. In its assessment (Reference 12), of PWROG-15032-NP, the NRC staff stated that static-cast CF8 grade CASS can be shown to have a high probability of ferrite content below 20 percent. Therefore, the NRC staff determined that ((low ferrite content)) likely means

((less than 20 percent ferrite)) and that, accordingly, the SE of BWRVIP-234 is applicable for evaluating the effects of IE and TE on the spider castings. The bottom curve in Figure A1 of the SE of BWRVIP-234 shows that at a fluence of ((1.8 dpa)), the fracture toughness is a little below 200 kJ/m2 (kilojoule/square meter and megajoule/square meter). The value of 200 kJ/m2 is the fracture toughness acceptance criterion established in Section 3.3.8 of the SE of BWRVIP-234. The NRC staff noted that this fracture toughness criterion is less than 255 kJ/m2 from the Grimes letter that the licensee referenced, but also that the Grimes letter criterion was established for pressure boundary components. As the staff noted in in Section 3.3.8 of the SE of BWRVIP-234, RVI components do not need the same level of toughness as pressure boundary components. Therefore, the NRC staff determined that the reduction of fracture toughness of the IMI guide tube spider castings due to IE and TE would not be significant.

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION c) The NRC staff finds that assuming ((large pre-service flaws or service-induced flaws do not exist)) in the IMI guide tube spider castings is reasonable for the reasons discussed in item (a). The staff finds that due to the ((self-limiting nature of weld residual stresses, the assumption that as a postulated flaw grows, the weld residual stress decreases such flaw extension ceases)), is reasonable. The NRC staff cannot verify that the IMI guide tube spider castings ((will retain significant fracture toughness due to IE after 54 EFPY of operation)) for the reasons discussed in item (b).

d) The NRC staff finds it reasonable that ((weld residual stress to be the dominant driving force for a postulated crack in the IMI guide spider castings)) because weld residual stresses are well-known to cause cracking in component joints. The staff finds it appropriate that the licensee considered design basis loads in the structural analysis in addition to ((weld residual stress)). In the April 18, 2018, supplement, in its response to RAI-3, the licensee stated that ((the maximum stress levels decreased as additional legs failed (from a configuration with all legs intact or one failed leg) and provided values for these stress levels determined from finite element analysis)). For these reasons, the NRC staff determined that ((the likelihood of the other three legs failing due to the failure of one leg is unlikely)).

The NRC staff reviewed the description of ((the two additional support connections of the IMI guide tube in Section 2.6.2 Flow Distributor Assembly of MRP-189, Revision 1. One support connection near the bottom of the tube is a welded connection at the 2-inch thick flow distributor head plate. The second connection, further up the tube, is an interference fit with the 2-inch thick IMI guide tube support plate. The tubes are secured on top of the IMI guide tube support plate with nuts and washers)). Although the licensee did not calculate a ((specific displacement of the hub with one failed leg)) as the licensee stated in the April 18, 2018, supplement, the NRC staff finds that ((the two support connections described above and the three remaining legs would provide sufficient stiffness of the IMI guide tube to prevent misalignment of the monitoring instrumentation should one spider casting leg fail)).

The NRC staff also determined that the licensees flow induced vibration analysis that

((ruled out high-cycle fatigue and flow excitation of degraded spider casting configurations)) further supports that ((one failed leg)) is not a concern. Accordingly, the NRC staff determined that the licensee provided reasonable assurance that it will maintain functionality of the IMI guide tube spider castings during the PEO.

Based on the discussion in items (a) through (d) above, the NRC staff determined that the licensee adequately demonstrated that it will maintain the functionality of the IMI guide tube spider castings during the PEO. Accordingly, the NRC staff determined that the licensee will adequately manage the aging effects of the IMI guide tube spider castings during the PEO.

3.3 Assessment the Vent Valve Retaining Rings 3.3.1 Description and Function of the Vent Valve Retaining Rings The vent valve retaining rings are part of the vent valve assembly within the core support shield assembly. There are eight vent valve assemblies in the core support shield assembly. Each vent valve assembly includes a top retaining ring and a bottom retaining ring. The function of the retaining rings is to retain the vent valve body in the vent valve nozzle. In the event of a pipe rupture in the RPV inlet pipe, the vent valve opens to permit steam generated in the core to flow

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION directly to the break. This permits the core to be flooded and adequately cooled when emergency core coolant is supplied to the RPV. A secondary function of the vent valve is to

((prevent unacceptable bypass flow during normal operation)). Failure of a retaining ring or portion of a retaining ring results in loss of support function for the vent valve body ((and the eventual complete release of the vent valve body)) and ((could affect the secondary bypass flow function, but such a failure is not expected to prevent the function of relieving pressure in the interior of the core support assembly during a cold leg large break loss-of-coolant accident (LOCA))).

The vent valve retaining rings are made from Type 15-5 PH stainless steel and, per Table 4-1 of MRP-227-A, are susceptible to TE.

3.3.2 Licensees Evaluation The licensee provided a detailed evaluation of the TMI, Unit 1, vent valve retaining rings in Section 5.0 Vent Valve Retaining Rings of ANP-3479P. The evaluation is summarized below.

a) Regarding the possibility of existing flaws: The licensee reviewed available CMTRs for the vent valve retaining rings, and determined that the TMI, Unit 1, vent valve rings ((received satisfactory ultrasonic testing (UT) and PT evaluations)). The licensee stated that information on ((actual flaw sizes was not retrievable)). Therefore, given that the vent valve rings received ((satisfactory UT and PT evaluations, the licensee concluded that the vent valve rings are free of defects that exceeded the UT and PT acceptance criteria)).

b) Regarding degraded material properties: ((The licensee stated that it did not locate any material property information for Type 15-5 PH stainless steel in the same tempered condition used in the vent valve retaining rings and in the same thermally aged condition.

However, based on aged material properties for other PH steels and for Type 15-5 PH stainless steel in a different tempered condition, the licensee concluded that saturation has been reached prior to the PEO for the vent valve retaining rings)). Additionally, the licensee stated that a reasonable lower bound fracture toughness for the vent valve retaining rings is ((52 MPam (47.3 ksiin)).

c) Regarding the likelihood of failure: The licensee stated that given the vent valve retaining rings ((have already reached saturation prior to the PEO)), the ((stresses during normal operation in the vent valve retaining rings are insufficient to cause failure)). The licensee stated that having no known cracking or failures of vent valve retaining rings confirms this conclusion. Because of this, the conclusion in item (a) regarding the ((improbability of flaws exceeding the UT and PT acceptance criteria)), and the expected fracture toughness of the vent valve retaining rings to be ((larger than lower bound fracture toughness assumed in item (b))), the licensee concluded that failure of the vent valve retaining rings is not expected during the PEO.

d) Regarding the effect of failure on functionality: The licensee stated that if vent valve retaining rings break, ((the vent valve body would be pushed inward into the annulus between the plenum assembly and the core shield support)), but it would not ((block RCS flow for pressure relief)) during a large cold leg break LOCA. The licensee also stated that with broken vent valve retaining rings, ((nuclear instrumentation and asymmetry in the hot leg temperature measurements would detect substantial bypass flow)).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION 3.3.3 NRC Staffs Assessment The NRC staff reviewed the information in Section 5.0 of ANP-3479P for the vent valve retaining rings, as summarized in items (a) through (d) in Section 3.3.2 above, and provides its findings below for each corresponding item.

a) The NRC staff finds it reasonable to assume that the vent valve rings received

((satisfactory UT and PT evaluations)) based on the licensees review of the CMTRs.

Additionally, the NRC staff verified in MRP-189, Revision 1, that the vent valve retaining rings do not screen in for ((IASCC, SCC, and fatigue)), mechanisms that can extend cracks. Accordingly, the NRC staff finds the licensees assumption that the vent valve retaining rings are ((free of defects that exceeded the UT and PT acceptance criteria)) to be reasonable even though information on ((actual flaw sizes was not retrievable)).

b) The licensee stated that ((based on saturated material properties of other PH steels and of Type 15-5 PH stainless steel in a different tempered condition, the vent valve retaining rings reached saturation prior to the PEO)). The NRC staff requested clarification on how the licensee reached this conclusion. In the April 18, 2018, supplement, in its response to RAI-4, the licensee clarified that, ((based on the fracture toughness saturation due to TE of 17 EFPY for Type 17-4 PH and 25 EFPY for 15-6 PH stainless steels, and due to the composition and temperature effects on TE, the saturation time of Type 15-5 PH stainless steel is expected between 17 EFPY and 25 EFPY, referencing a publication on thermal aging of martensitic steels)). The licensee also clarified that ((the fracture toughness value of 52 MPam (47.3 ksiin) is for saturated Type 17-4 PH stainless steel and referenced toughness properties for martensitic stainless steels)). The NRC staff finds it reasonable to cite saturation information for ((other PH stainless steels)), considering ((no saturation data is available for Type 15-5 PH stainless steel in the same tempering condition as the vent valve rings)). Accordingly, the NRC staff finds acceptable the licensees conclusion that the vent valve retaining rings have ((reached saturation before the PEO)).

c) The NRC staff finds that because the vent valve retaining rings are: (1) ((free of defects that exceeded the UT and PT acceptance criteria)); (2) ((have already reached saturation prior to the PEO)); and (3) there has been no cracking or failures of vent valve retaining rings B&W-designed PWRs, the failure of the vent valve retaining rings is unlikely during the PEO.

d) The NRC staff finds the information the licensee provided regarding ((detectability of bypass flow)) and the capability of the vent valve to ((relieve pressure during LOCA with failed retaining rings)) provides reasonable assurance of functionality of the vent valve retaining rings during the PEO.

Based on the discussion in items (a) through (d) above, the NRC staff determined that the licensee adequately demonstrated that it will maintain the functionality of the vent valve retaining rings during the PEO. Accordingly, the NRC staff determined that the licensee will adequately manage the aging effects of the vent valve retaining rings during the PEO.

4.0 CONCLUSION

The NRC staff has reviewed the licensees TMI, Unit 1, evaluation of A/LAI 7 of the SE in the MRP-227-A. Based on the discussions in Section 3.0 of this assessment, the staff determined

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION that the licensee adequately demonstrated that the functionality of the CRGT spacer castings, IMI guide tube spider castings, and vent valve retaining rings at TMI, Unit 1, will be maintained during the PEO. Accordingly, the NRC staff determined that the licensee has adequately resolved A/LAI 7.

5.0 REFERENCES

1. Letter from Edward W. Callan (Exelon) to NRC, Submittal of Inspection Plan for Reactor Internals, A/LAI #7 Component Evaluation, September 16, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16263A338).

2. Letter from David P. Helker (Exelon) to NRC, Response to Request for Additional Information Three Mile Island Nuclear Station, Unit 1 - Request for Additional Information Regarding Inspection Plan for Reactor Internals Action Item 7 (EPID L-2016-LLL-0002), April 18, 2018 (ADAMS Accession No. ML18108A287).

3. Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines (MRP-227-A). EPRI, Palo Alto, CA: 2011. 1022863 (ADAMS Accession Nos.

ML120170453).

4. Letter from NRC to Michael J. Pacilio (Exelon), Three Mile Island Nuclear Station, Unit 1 -

Staff Assessment of the Reactor Vessel Internals Inspection Plan (TAC No. MF1459),

December 19, 2014 (ADAMS Accession No. ML14297A411).

5. Letter from Justin C. Poole (NRC) to Bryan C. Hanson (Exelon), Three Mile Island Nuclear Station, Unit 1 - Request for Additional Information Regarding Inspection Plan for Reactor Internals Action Item 7 (EPID L-2016-LLL-0002), February 26, 2018 (ADAMS Accession No. ML18043B142).

6. Materials Reliability Program: Screening, Categorization, and Ranking of B&W-Designed PWR Internals Component Items (MRP-189-Rev. 1). EPRI, Palo Alto, CA: 2009. 1018292 (PROPRIETARY).

7. NUREG/CR-4513, Revision 1, Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems, August 1994 (ADAMS Accession No. ML052360554).

8. Letter from NRC to Douglas J. Walters (Nuclear Energy Institute), License Renewal Issue No. 98-0030, Thermal Aging Embrittlement of Cast Austenitic Stainless Steel Components, May 19, 2000 (ADAMS Accession No. ML003717179).

9. NUREG/CR-7027, Degradation of LWR Core Internal Materials due to Neutron Irradiation, December 2010 (ADAMS Accession No. ML102790482).

10. Letter from NRC to Tim Hanley (Chairman, BWR Vessel and Internals Project), Final Safety Evaluation of the BWRVIP-234: Thermal Aging and Neutron Embrittlement Evaluation of Cast Austenitic Stainless Steel for BWR Internals (TAC NO. ME5060), June 22, 2016 (ADAMS Accession No. ML16096A002).

OFFICIAL USE ONLY - PROPRIETARY INFORMATION OFFICIAL USE ONLY - PROPRIETARY INFORMATION

11. Pressurized Water Reactor Owners Group Report PWROG-15032-NP, Revision 0, PA-MSC-1288, Statistical Assessment of PWR RV Internals CASS Materials, November 2015 (ADAMS Accession Nos. ML16068A245 and ML16068A246).

12. Office of Nuclear Reactor Regulation Staff Assessment of The Pressurized Water Reactor Owners Group Report PWROG-15032-NP, Revision 0, PA-MSC-1288, Statistical Assessment of PWR RV Internals CASS Materials, September 2016 (ADAMS Accession No. ML16250A001).