ML12093A484

NRC Staff Pre-Filed Hearing Exhibit NRC000114D, RAI Set #3 Final Responses - 07/08/10

ML12093A484
Person / Time
Site:	Indian Point
Issue date:	10/29/2010
From:	NRC/OGC
To:	Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML12093A482	List: ML12093A483 ML12093A484 ML12093A485 ML12093A486
References
RAS 22204, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01, MRP-227-A
Download: ML12093A484 (106)

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NRC000114D Resubmitted: April 2, 2012 RAI Set #3 Final Responses - 07/08/10 Figure 3.0-1 A METHOD TO IDENTIFY SSCs AND INTENDED FUNCTIONS WITHIN THE SCOPE OF LICENSE RENEWAL [10 CFR 54.4(a) & (b)]

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RAI Set #3 Final Responses - 07/08/10 Figure 4.1-1 A METHOD TO IDENTIFY SSCs AND INTENDED FUNCTIONS WITHIN THE SCOPE OF LICENSE RENEWAL [10 CFR 54.4(a) & (b)]

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RAI Set #3 Final Responses - 07/08/10 Figure 4.2-1 Identification of Aging Effects Requiring Management 13

RAI Set #3 Final Responses - 07/08/10 FIGURE 4.3-2 Aging Management Program Review 14

RAI Set #3 Final Responses - 07/08/10 RAI 3-3 By letter dated February 20, 2009, the NRC granted a fee waiver under the provisions of 10 CFR 170.1 l(a)(1)(iii) to support a non-fee billable review of TR MRP-227, Rev. 0. The fee waiver was granted by the NRC on the basis that the methodology of the report would be used to update the NRC staff's aging management review (AMR) items for PWR RVI components, as given in the following NRC license renewal guidance documents: (1) the SRP-LR; and (2)

NUREG- 1801, "Generic Aging Lessons Learned Report" (GALL), Volumes 1 and 2.

The NRC's recommended AMR items in Tables of the GALL Report, Revision 1, Volume 1 (henceforth referred to as Table 1 AMR items) are given in the following AMR column format:

Table X. Summary of Aging Management Programs for the .... System Evaluated in Chapter IV of the GALL Report ID Type Component Aging Effect/ Aging Further Related Unique Mechanism Management Evaluation Generic Item Programs Required Item The NRC's recommended AMR items in Tables of the GALL Report, Revision 1, Volume 2 (henceforth referred to as Table 2 AMR items) are given in the following AMR column format:

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B2 (or B3; B4) Reactor Vessel Internals (PWR) - Westinghouse (Combustion Engineering; Babcock and Wilcox)

Item Link Structure Material Environment Aging Aging Further and/or Effect/Mechanism Management Evaluation Component Required Program (AMP)

These AMR item formats have been adopted for use in Tables 3.0-1 and 3.0-2 of NEI 95-10, Revision 6, and endorsed for use in Regulatory Guide 1.188, Revision 1.

Provide Table 1 AMR items for Westinghouse (W), Combustion Engineering (CE), and Babcock

& Wilcox (B&W) RVI component commodity groups that are in conformance with the guidelines for formatting Table 1 AMR items in Figure 3.0-1 of NEI 95-10, Revision 6. The NRC staff also requests that you provide the Table 2 AMR items for W, CE, and B&W RVI component commodity groups that are in conformance with the guidelines for formatting Table 1 AMR items in Figure 3.0-2 of NEI 95-10, Revision 6.

Response: MRP Letter MRP-2009-09 1, Dennis Weakland (MRP) to Tanya Mensah (NRC), dated December 2, 2009, quoted in part below, provided material that is responsive to this RAI:

"The following five documents provide the MRP's initial draft input to assist the NRC staff in updating NUREG 1801, "Generic Aging Lessons Learned Report" (GALL):

1) EPRIDRAFT Input (12-01-09): GALL ChapterXI.M16

2) EPRIDRAFT Input (12-01-09): GALL Table IV.B2 (W)

3) EPRIDRAFT Input (12-01-09): GALL Table IV.B3 (CE)

4) EPRIDRAFTInput (12-01-09): GALL Table IV.B4 (B&W)

5) EPRIDRAFT Input (12-01-09): New Appendix A to MRP-227 15

RAI Set #3 Final Responses - 07/08/10 These documents have been forwarded to the Document Control Desk by the referenced letter (copy attached).

RAI 3-4 Provide a new draft GALL AMP XI.M16, "PWR Vessel Internals Program" in a format that conforms to the recommended program element criteria in SRP-LR BTP RLSB Section A. 1.2.3 and that can be adopted for the contents of an applicant's PWR Vessel Internals Program when the license renewal application is submitted to the NRC for NRC staff approval.

Response: See response to RAI 3-3 above.

RAI 3-5 Provide the basis for why the functionality analysis (FA) in Figure 2-2 was applied to Category C RVI components and not to Category B RVI components.

Response: The process outlined in Figure 2-2 was used to categorize components based on the "likelihood and severity of safety and economic consequences". Priority in the functionality analysis was given to the Category C components because they were found to be "potentially significantly affected". However, the analysis was not limited to the Category C items. A lower priority was given to Category B components because the consequences were found to be "potentially moderately significant". SRP-LB BTP RLSB- 1 specifically states that: "The risk significance of a structure or component could be considered in evaluating the robustness of an aging management program."

The detailed functionality analysis of the barrel former and core shroud incorporated both Category C and Category B components. In other cases, it was determined that existing analysis of Category B and C components was sufficient and additional functionality analysis was not required. For instance, although no detailed mechanical analysis of the lower support columns was undertaken in this study, existing analysis indicates that the stresses in these components are primarily compressive.

Because the initial screening and categorization process identified some level of potential safety and economic consequence for all of the Category B and C components, these components were all considered in the development of the MRP-227 program. The disposition of all Category B and C components is shown in Tables 3-1, 3-2 and 3-3. The final disposition was based on the results of the functionality analysis combined with component accessibility, operating experience, existing evaluations and prior examination results.

RAI 3-6 Clarify whether or not the existing methodology in TR MRP-227, Rev. 0, can be applied to a PWR facility whose reactor core loading pattern operating history is not bounded by the assumptions in the report. If the methodology can be applied, justify why that is the case. If the methodology cannot be applied to these PWRs, identify what actions a licensee with a non-conforming PWR would have to take in order to develop a plant-specific AMP for its RVI components, which is consistent with the intent of TR MRP-227, Rev. 0. Identify whether license renewal applicants should demonstrate that their facility's reactor core loading pattern operating history is bounded by the assumptions in the report as part of the license renewal application (i.e.,

should be a license renewal applicant action item).

Response: Similar concerns were expressed in RAI 2-19 from the 8/24/09 inquiries.

Based on the previous discussion the following conclusions may be drawn:

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RAI Set #3 Final Responses - 07/08/10

1. Section 2.4 of MRP-227 states: "The recommendations are thus applicable to all U.S. PWR operating plants as of May 2007 for the three designs identified."

2. Section 2.4 further states: "Plant modifications made or considered after this date should be reviewed to assess impacts on strategies contained in these guidelines."

3. The strategies in MRP-227 do not assume that the core loading patterns used in the analysis are bounding.

4. The inspection recommendations are robust and there is no reason to anticipate plant modifications that would impact the MRP-227 requirements.

5. To apply MRP-227, the license renewal applicant needs to demonstrate that core loading patterns going forward are reasonably represented by the assumptions of the report.

6. MRP-227 is a living document and the industry will monitor any trends in operating practice that might impact the MRP-227 recommendations.

Original Response to 8/24/09 RAI 2-19:

The core loading patterns used in the MRP-227 reference documents were chosen to represent known operating practice, they are not intended to be used as a reference for plant-specific analysis. The intention of using the representative core loading patterns was not to bracket operation, but to perform an analysis that demonstrates both historic and current fuel management programs. The MRP-227 inspection recommendations based on these calculations are robust and do not require the utility to perform additional analysis of core loading patterns to qualify their applicability.

The condition of the internals at the time of the first required inspections is dominated by the power distribution used to represent the first thirty years of full power operation.

During this period the analysis assumed that the fresh fuel was loaded in the peripheral fuel assemblies. This "out-in" loading pattern produced results in relatively high heat loadings and neutron fluences in the near core structure. In practice all plants in the United States abandoned fuel management based on the "out-in" loading prior to thirty years of operation. There are no current or planned fuel management programs that would result in more deleterious conditions than those assumed in this analysis during the first thirty years of operation. For this reason there is no reason to require any plant to perform an analysis to demonstrate adherence to the assumed core loading pattern prior to performing the first round of inspections. The timing and extent of the first round of MRP-227 examinations is governed by damage that has already been accumulated.

The representative power distributions used for the simulation of years 31 to 60 incorporate the effects of aggressive power uprate programs. Qualification of the core loading pattern is considered in the design analysis for the plant uprate. Although it is not possible to anticipate all possible future options, both current fuel management practice, which maximizes fuel utilization, and concerns about neutron damage in the reactor pressure vessel preclude return to the practice of loading fresh fuel in the periphery locations. It is unlikely that future core loading patterns would invalidate the assumptions of the analysis.

Although the shift from "out-in" core loading patterns to low-leakage patterns resulted in a sharp decrease in the peak temperature in the internals structure, the shift had minimal effect on the location of the peak temperature or the character of the peak damage. There 17

RAI Set #3 Final Responses - 07/08/10 is no reason to expect that changing the loading pattern would change the base inspection recommendations. The MRP-227 recommendations are based on reasonable assumptions about the effects of power uprates. In many cases power uprates can be accomplished without significantly increasing the heat or neutron loading to the internals. Return to the more aggressive core loading patterns could conceivably result in a decrease in the re-inspection interval. However, there is no reason to anticipate any change of this scale.

MRP-227 is intended to be a living document. The MRP will monitor both inspection results and plant operating experience and make appropriate modifications. There is currently no need to require plants to demonstrate adherence to any reference core loading practice.

RAI 3-7 Alloy 600 PWR RVI components and their associated welds manufactured from Alloys 82 and 182 are susceptible to primary water stress corrosion cracking (PWSCC) when exposed to PWR reactor coolant water. In Table 3-1 of TR MRP-227, the following Babcock and Wilcox (B&W) Alloy X-750 PWR RVI components were welded with Alloy 82 material and yet they were classified under "N" category which excludes inspections for these PWR RVI components:

(1) dowel-to-core barrel cylinder welds, (2) dowel-to-upper grid rib section bottom flange welds, (3) dowel locking welds, (4) dowel-to-guide block welds, and (5) dowel-to-distributor flange welds. Even though stress levels in these components may not exceed the threshold levels, the NRC staff considers it to be likely that PWSCC can potentially occur due to the introduction of cold work during fabrication. In light of this observation, provide an explanation for excluding inspection requirements for these B&W PWR RVI components.

Response: The explanation for categorization of these dowel welds is provided in MRP-231-Rev. 1. Excerpts are provided below:

Many preliminary Non-Category A welds in Table 1-2 used nickel-based Alloy 69 (INCO 69) and Alloy 82 (1NCO 82) materials, which are susceptible to PWSCC (or SCC as listed in Table 1-2). However, some of these Alloy 69 and Alloy 82 welds are for locking Alloy X-750 alignment dowels which facilitated the internals assembly process.

These dowels do not have any function after the internals items were joined by bolting.

Hence, those preliminary Category B welds were changed to Category A. The updated Non-Category A welds are included in Table 2-8 with the following welds no longer listed. It should be noted that the reclassified Category A welds below based on the functionality assessment are called "No Additional Measures (N)" in Section 3 and are listed in Table 3-8 as "N".

Plenum Cover Assembly Alloy X-750 dowels-to-plenum cover bottom flange welds These welds are used for locking two Alloy X-750 dowels, which were used to align the plenum cover bottom flange with the plenum cylinder top flange. After the plenum cover bottom flange is bolted to the plenum cylinder top flange with 64 bolts, the two Alloy X-750 dowels and their locking welds no longer have any function.

Upper Grid Assembly Alloy X-750 dowel-to-upper grid rib section bottom flange welds 18

RAI Set #3 Final Responses - 07/08/10 These welds are used for locking two Alloy X-750 dowels, which were used to align the upper grid rib section with the upper grid ring forging. After the upper grid rib section is bolted to the upper grid ring forging with 36 cap screws, the two Alloy X-750 dowels and their locking welds no longer have any function.

Core Barrel Assembly Alloy X-750 core barrel-to-former plate dowels and the locking welds These welds are used for locking the 32 Alloy X-750 dowels, which were used to align the former plates with the core barrel cylinder at the top and bottom former plate level (16 dowels at each level). After the former plates are bolted to the core barrel cylinder with the CB bolts, these Alloy X-750 dowels and their locking welds no longer have any function. These dowels are not considered in the core barrel assembly functionality analysis.

Lower Grid Assembly Alloy X-750 dowel-to-lower grid shell forging welds These welds are used for locking two Alloy X-750 dowels, which were used to align the lower grid shell forging with the core barrel cylinder bottom flange. After the lower grid shell forging is bolted to the core barrel cylinder bottom flange with 108 LCB bolts, the two Alloy X-750 dowels and their locking welds no longer have any function.

Alloy X-750 dowel-to-lower grid rib section welds These welds are used for locking two Alloy X-750 dowels, which were used to align the lower grid rib section with the lower grid shell forging. After the lower grid rib section is bolted to the lower grid shell forging with 36 cap screws, the two Alloy X-750 dowels and their locking welds no longer have any function.

Flow Distributor Assembly Alloy X-750 dowel-to-flow distributor flange welds These welds are used for locking two Alloy X-750 dowels, which were used to align the flow distributor flange with the lower grid shell forging. After the flow distributor flange is bolted to the lower grid shell forging with 96 LCB bolts, the Alloy X-750 dowels and the locking welds no longer have any function.

Table 3-7 in MRP-231 (Rev. 1) summarizes the remaining dowel welds, which remain as either Primary or Expansion items:

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RAI Set #3 Final Responses - 07/08/10 Table RAI 3-7-1 Summary of Nickel-Based Alloy Welds for PWSCC Item Table 2-8 Category Final Category Lower Grid Assembly Alloy X-750 Dowel to Guide Block Welds B P Upper Grid Fuel Assembly Support Pads:

Alloy X-750 Dowel Locking Weld (except DB) B E Lower Grid Fuel Assembly Support Pads:

Alloy X-750 Dowel Locking Weld (Note 1) B E Note 1, Alloy X-750 Dowel Locking Weld is also listed in Section 3.2.6 for Irradiation Embrittlement, and is also categorized as "Expansion".

The lower grid assembly dowel-to-guide block locking welds serve as loose part prevention devices. They are not structural. The Alloy 82 locking welds may be susceptible to cracking as a result of stress corrosion cracking (i.e., PWSCC). Small cracks in the locking weld are acceptable since the locking function can be maintained as long as any part of the weld is present. The fillet welds are therefore categorized as Primary items.

The upper and lower grid assemblies Alloy X-750 dowel-to-fuel assembly support pad welds (either Alloy 82 or Alloy 69 material) are also considered loose part prevention devices and are categorized as Expansion items. Evidence of cracking is expected to occur sporadically and not lead to a short-term serious degradation. However, evidence of cracking should lead to enhanced vigilance for possible loose parts monitoring and consideration for implementing the repair scenarios.

RAI 3-8 In Section 2.4 of TR MRP-227, Rev. 0, the MRP assumes that the design of a PWR plant applying the TR MRP-227, Rev. 0, methodology would not include any design changes beyond those identified in either general industry guidance or recommended by the original vendors. The NRC staff is aware that many of the licensees owning PWR facilities have been granted license amendments to implement measurement uncertainty recapture (MUR) power uprates, stretch power uprates, or extended power uprates for their facilities. However, it is not evident to the NRC staff whether any design changes associated with these type of power uprates would be within the scope of the MRP's term "design changes identified in general industry guidance or recommended by original vendors." Clarify whether design changes that will need to be implemented in order to receive NRC approval of a MUR, stretch, or extended power uprate, or that have been implemented as a result of receiving NRC approval of a power uprate, are within the scope this type of assumption.

Response: The TR MRP-227, Rev. 0 recommendations were based on evaluations relevant to current plant operating experience at the time the report was issued. This operating experience includes measurement uncertainty uprates, stretch power uprates and extended power uprates. Therefore, all three uprate types are considered within the scope of "design changes identified in general industry guidance or recommended by original vendors." As it is not possible to anticipate all possible future plant uprates or 20

RAI Set #3 Final Responses - 07/08/10 modifications, MRP-227 clearly states that "Plant-specific commitments remain the responsibility of the owner."

Average core power must be increased to implement a plant uprate. This increase in power necessarily implies an increase in the average neutron flux in the core. However, for those reactor internals components that are subject to neutron radiation damage, the neutron exposure tends to be determined by power levels in the peripheral fuel assemblies, rather than the core average power. The original "out-in" core loading patterns used in many plants produced relatively flat core power distributions. These flat power distributions lead to higher neutron leakage at the edges of the core. In addition to causing high damage rates in the internals, this leakage results in increased neutron exposure of the reactor pressure vessel and relatively poor fuel utilization. Current core design practice utilizes a "low leakage" loading that tends to reduce power levels in the peripheral assemblies, which in turn reduces the neutron exposure of the reactor internals.

While plant uprates may lead to some increase in neutron exposure of the internals, these are increases of an exposure level already reduced by the core loading, and they are generally moderate compared to the overall increase in power level.

Measurement uncertainty recapture uprates take advantage of improved power monitoring systems to allow the plant to increase power inputs by operating closer to the plant allowables. As these uprates remain within the original design basis of the plant, there is no reason to believe that a measurement uncertainty recapture uprate would fall outside the scope of the MRP-227 recommendations.

Stretch power uprates take advantage of excess margins that are buried in the plant operating limits. In many cases, a plants operating below the true plant capacity can demonstrate safe operation at higher power outputs by removing overly conservative limits. In most cases, the stretch power uprates result in power increases in an individual plant, but do not move the plant outside the envelope of fleet operating experience.

Therefore, there is no reason to believe that a stretch power uprate would fall outside the scope of the MRP-227 recommendations.

Extended power uprates produce the largest increases in plant power. An extended power uprate may rely on both more detailed analysis of plant operation and upgrades to plant equipment. The experience base considered in support of the MRP-227 recommendations included plants with extended power uprates. The core power distributions used in the modeling of irradiation-induced aging of the core baffle and shroud structures included a typical extended power uprate. It is the responsibility of the plant owner to demonstrate that the changes in plant operation are consistent with the general assumptions of MRP-227.

The fimite element-based aging analysis of the core baffle-formers and core shroud completed in support of the MRP-227 guidelines were never intended to provide bounding plant results. The recommendations are robust and not dependent on the details of the analysis. The assumption that the representative plant operated for thirty years with "out-in" core loading patterns has a large impact on the results. Most plants moved away from this aggressive core loading pattern much earlier in plant life. The effects of this conservative assumption about plant operating history are generally larger than any potential effect of plant uprate.

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RAI Set #3 Final Responses - 07/08/10 RAI 3-9 The middle of page 4-1 of Section 4 of TR MRP-227, Rev. 0, provides a list of program element activities for the I&E methodology. SRP-LR BTP RLSB-1 Section A. 1.2.3, Program Elements, provides the recommended program element criteria for AMPs. For each bulleted program criterion that is provided on page 4-1 of TR MRP-227, Rev. 0, the NRC staff requests that a reference or link be provided that matches the programmatic criterion in the bulleted list to its corresponding program element subsection in SRP-LR BTP RLSB-1 Section A. 1.2.3. In addition, the NRC staff requests the criteria in the bulleted list be amended to include an aging management criterion that corresponds to the recommended program element criteria for condition monitoring programs in SRP-LR BTP RLSB-1, Section A. 1.2.3.3, "Parameters Monitored/Inspected."

Response: The inspection and evaluation activities shown as bullets in the middle of page 4-1 of Section 4 of MRP-227 were not intended to match completely with the aging management program elements of SRP-LR BTP RLSB-1, Section A. 1.2.3, Program Elements. These bulleted items were intended only to characterize the "typical" concerns that should be addressed when developing an adequate aging management program.

Instead, the industry has prepared a draft of Aging Management Program XI.M 16 that describes an explicit comparison between the contents of MRP-227 and the Program Elements. This draft should be the point of comparison.

However, many of the typical aging management program concerns on page 4-1 can be compared to, or placed in the context of, SRP-LR aging management program attributes.

For example, "selection of items for aging management" is similar to Scope, although additional detail on "selection of items for aging management" could eventually lead to ParametersMonitored/Inspected. The concern about "selection of the type of examination or other methodologies appropriate for each applicable degradation mechanism" is similar to Detection ofAging Effects, as are "specification of the required level of examination qualification" and "schedule of first and frequency of any subsequent examinations." The concern about "sampling and coverage" could be placed with ParametersMonitored/Inspectedor with Detection ofAging Effects, depending on emphasis. The concern about "expansion of scope if sufficient evidence of degradation is observed" is probably best placed with CorrectiveActions, while "examination acceptance criteria" belongs with Acceptance Criteriaand "methods for evaluating examination results not meeting the examination acceptance criteria" also is best placed with CorrectiveActions. The concern about "updating the program based on industry-wide results" matches well with OperatingExperience, while "contingency measures to repair, replace, or mitigate" also fits with CorrectiveActions. It should be repeated that the list on page 4-1 was not intended to parallel in any way the SRP-LR aging management program attributes.

RAI 3-10 Section 4.1.3, Aging Management Methodology Qualification, of TR MRP-227, Rev.

0, (page 4-2) provides the inspection method qualification discussion and basis for the report. The section implies that methodologies may be qualified in accordance with appropriate qualification requirements, standards, or procedures (e.g., the qualification requirements in the ASME Code),

or may be qualified by only the development of a technical justification to explain the applicability of the selected methodology. Please amend the stated section to be more specific on the qualification methods that would be used to qualify a given inspection technique for implementation for both the case where an RVI component is scoped in for license renewal under one of the safety related intended functions mentioned in either 10 CFR 54.4(a)(1)(i),(ii), or (iii),

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RAI Set #3 Final Responses - 07/08/10 and for the case where a non-safety related RVI component is scoped in for license renewal under the scoping requirement of 10 CFR 54.4(a)(2).

Response: The intent of Section 4.1.3 is to demonstrate that the inspection methodologies specified in MRP-227 for aging management are methodologies that are commonly used in industry and have a substantial experience base.

MRP-227 identifies the aging effects to be detected while its companion document, MRP-228, identifies the experience base for each of the examination methodologies selected, determines the level of standardization and contains the requirements specific to the inspection methodologies involved, as well as requirements for qualification of the NDE systems used to perform those inspections. Whereas failure of component items individually often do not affect the ability of the component/assembly to perform its function, there is no differentiation applied between safety related or non-safety related internals examination requirements in either MRP-227 or MRP-228. All components currently selected for examination in MRP-227 are either safety-related or important to safety. For the purposes of specifying qualification of examinations, reactor internals inspections (PWR or BWR) are divided into either; "remote visual" examinations or

non-remote visual" examinations.

For remote examination procedures the generic requirements for visual examination (EVT-1, VT-1, and VT-3) described in Section 2.3 shall be met, including those addressing the personnel training and experience requirements for individuals performing those examinations.

For examinations other than remote visual, a Technical Justification is required for each examination procedure in accordance with Section 2.1.

No inspections have been specified by MRP-227 that would require a higher level of qualification than a technical justification. The use of ASME Section V Article 14 in conjunction with MRP-228 provides a standardized process whereby components selected for examination by means other than visual must use the technical justification process. Utilities and/or vendors can include more rigor in order to achieve successful implementation of the examination requirements, but not less.

The use of ASME Section V Article 14 in conjunction with MRP-228 provides a standard process for qualification of known methodologies (Technical Justifications) but also provides for scenarios where a component may be examined using an unforeseen new methodology, an untried delivery device, or a more discerning supplemental or alternate examination without a substantial experience base. In such cases the utility or vendor could specify a higher degree of initial qualification rigor if warranted. This is in keeping with the goal that MRP-228 continue as a living document with frequent updates as new or improved transducer delivery tooling or transducer packages are added to the experience base as dictated by industry needs and lessons learned. The technical justification process described in ASME Section V Article 14 as incorporated into MRP-228 allows for these gains or adjustments to be made in an open, standardized manner.

Thus MRP suggests the following re-wording of this section:

An extensive experience base for the aging management methodologies described in this section of the I&E guidelines permits selection of known aging management 23

RAI Set #3 Final Responses - 07/08/10 methodologies. Many inspections specified herein are remote visual examinations whether visual VT-1, EVT- 1 or VT-3. For remote visual no procedure qualifications are required other than that in addition to ASME Code Section XI requirements remote visual examinations must meet the additional generic requirements of MRP-228 for equipment and training of personnel, and in the case of visual EVT- 1 a surface condition assessment, and limitations on camera angle and scan speed. All other methodologies specified herein already have well established procedural qualifications, such as volumetric examination of bolting. Thus the level of procedural qualification for examinations other than remote visual is limited to technical justification. This level of qualification is appropriate. Failures of internals do not result in pressure boundary failures. Internals are either of robust design resulting in flaw tolerance well above the detection level that can be established via technical justification or consist of assemblies for which single (or often multiple) component item failures do not prevent the assembly from performing its function. The Inspection Standard [3] provides detailed guidance for conducting and justifying the selected examination techniques and the technical justifications required for different examination methodologies and component configurations.

RAI 3-11 Section 4.2.5 of TR MRP-227, Rev. 0, discusses the criteria when particular aging effect indications need to be coupled to physical measurement methods. However, the discussion in Section 4.2.5 does not prescribe the physical measurement methods or techniques that would be used to quantify these aging effect indications.

Please amend Section 4.2.5 to specify the physical measurement techniques that will be used to quantify the aging effect indications for which they are credited.

Response: The particular techniques to be utilized for physical measurements are not within the scope of MRP-227, but are covered generically by MRP-228 through the statement that a technical justification is required for any examinations other than visual examinations. Sections 4.3.1 and 4.3.3 of MRP-227 describes the physical measurements needed for the B&W internals core clamping items and the Westinghouse internals hold-down spring, respectively. In addition, Tables 4-1 and 4-3 provide the required examination methods and examination coverage and Tables 5-1 and 5-3 provide the acceptance criteria for the physical measurements. Physical measurement is specified in Table 4-2 for the CE units with core barrel shroud assembled in two vertical sections and a gap between the top and bottom core shroud segments is identified first by VT-1.

Also, refer to the response to RAI 3-12 for the Westinghouse internals hold-down spring.

RAI 3-12 Section 5.2 indicates that acceptance criteria for physical measurement techniques for W-designed RVI components are not included in TR MRP-227, Rev. 0, because the tolerances are available on a plant-specific or design-specific basis. Clarify that the intent of this statement is that licensees of W-designed facilities must obtain acceptance criteria for specified physical measurements based on information in their plant's CLB. Hence, the physical measurements taken during as part of the plant's license renewal AMP must demonstrate that the condition of the affected component remains consistent with the plant's current licensing basis. Identify that this would be an applicant action item for W designed facilities who plan to implement TR MRP-227, Rev. 0.

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RAI Set #3 Final Responses - 07/08/10 Response: Licensees of Westinghouse-designed facilities would need to obtain acceptance criteria for specified physical measurements based on information in their plant's Current Licensing Basis. Section 5.2 applies specifically to type 304 hold-down springs in Westinghouse designed plants. The inspection is intended to demonstrate that the hold-down spring has sufficient remaining compressibility to maintain the hold-down forces on the internals. Required hold-down forces are part of the current licensing basis.

Calculation of the spring height required to maintain the hold down forces and determination of appropriate margin against further deformation prior to the next scheduled inspection would be the plant responsibility. This determination would be part of the pre-inspection engineering program implemented by the utility.

RAI 3-13 Clarify whether the acceptance criterion for eddy current (ET) inspections is based on a "pass - no pass" acceptance criterion (i.e., any ET signals indicating a relevant ET indication would fail the acceptance criterion).

Response: This RAI is very similar to RAI 2-3. Therefore, the response will be similar.

Section 4.2.3 of MRP-227 specifically identifies eddy current surface examination as an electromagnetic testing (ET) method that can be used to supplement visual examination methods, in order to further characterize any detected relevant indications. As indicated in RAI 3-10 response, MRP-228 requires technical justifications for qualification of NDE systems other than visual examinations.

Eddy current examination is called out in Table 4-9 as an existing program carried out for flux thimble tubes in Westinghouse plants, in accordance with plant commitments made in response to NRC I&E Bulletin 88-09. In the latter case, the acceptance criteria for any detected indications are part of the plant commitments. All other applications of eddy current examinations will be supplemental examinations and, as such, this examination technique is not one of the prime examination methods for which specific examination acceptance criteria are required. When eddy current surface examination is used to supplement visual examination, the purpose will not be to again identify the relevant condition, but instead to further characterize the indication by - for example - confirming the crack-like nature of the indication and more accurately sizing its surface-breaking length. In such a case, the acceptance criteria to be applied will not be examination acceptance criteria, but evaluation acceptance criteria. These evaluation acceptance criteria are referred to in the context of supplementary examinations, engineering evaluations, and repair/replacement in Section 6 of MRP-227 (evaluation acceptance criteria have been developed for the MRP by the PWR Owners Group 3, and will be provided to NRC by the MRP).

RAI 3-14 Bolts in some RVI components may be subject to stress relaxation resulting in reduction in preload due to thermal and irradiation effects and, as such, are inspected at every 10 year interval under the ASME Code,Section XI, ISI program. During this interval, reduction in preload in these bolts should not be large enough to cause loss of component functionality prior to the next examination. The evaluation of the need to maintain bolt preload should also consider the impact of loss of preload on vibrational fatigue damage to the bolt and/or the component itself.

3 Westinghouse Non-Proprietary Class 3 Report, "ReactorInternalsAcceptance CriteriaMethodology and DataRequirements, WCAP-I 7096-NP, Revision 2 " December 2009 25

RAI Set #3 Final Responses - 07/08/10 Explain (based on the minimum number (percentage) of bolts/springs in each component that are required to maintain preload) how the proposed 10 year frequency is adequate to maintain functionality of each component under all design basis conditions.

Response: Past operating experience and the results from previous PWR internals bolted assembly examinations have shown that loss of preload is unlikely to change significantly during 10-year ISI intervals. For example, inspections and bolt replacements at European reactors have shown a mixture of preload relaxation and additional loading. In addition, inspection of the baffle-to-former bolts in the internals at least two U.S. units has not identified any failed bolting, either from IASCC or from vibrational fatigue, at a point in operational time when both relaxation and/or reloading of the bolts is expected to have occurred.

This experience base has been confirmed by functionality analyses performed on bolted assemblies in support of the development of the MRP-227, Revision 0 inspection and evaluation guidelines. These analyses clearly show that some bolt locations can be expected to have significant loss of preload, in some cases with loss of preload of 100%;

others in less highly irradiated locations can be expected to have very little loss of preload; and others can be expected to initially have loss of preload, followed by eventual re-loading as the result of void swelling and possibly prying action from connected structure. It should be pointed out that these functionality analyses were limited to long-term, steady-state operation (including neutron irradiation and gamma heating), but did not encompass the entire spectrum of loading conditions. In spite of the limitations, the functionality analyses illustrate the variation in bolting response over time very well. The analyses show loss of preload due to irradiation and overall structural response begins early in plant life. A substantial fraction of the total expected loss of preload occurs in the first 20 years. Subsequent loss of preload is limited and gradual.

26

August 30, 2010 Mr. Neil Wilmshurst Vice President & Chief Nuclear Officer Electric Power Research Institute 1300 West WT Harris Blvd Charlotte, NC 28262

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION NUMBER 4, RE: ELECTRIC POWER RESEARCH INSTITUTE TOPICAL REPORT 1016596, "MATERIALS RELIABILITY PROGRAM (MRP): PRESSURIZED WATER REACTOR INTERNALS INSPECTION AND EVALUATION GUIDELINES (MRP-227 - REV. 0)" (TAC NO. ME0680)

Dear Mr. Wilmshurst:

By letter dated January 12, 2009, Electric Power Research Institute submitted for U.S. Nuclear Regulatory Commission (NRC) staff review Topical Report 1016596, "Materials Reliability Program (MRP): Pressurized Water Reactor Internals Inspection and Evaluation Guidelines."

Upon review of the information provided to date, the NRC staff has determined that additional information is needed to support completion of the review. During discussions with Chuck Welty, Technical Executive, and Anne Demma, Project Manager, we agreed that the NRC staff will receive your response to the enclosed RAI questions by October 29, 2010.

Ifyou have any questions regarding the enclosed RAI questions, please contact me at 301-415-1847.

Sincerely, IRAI Sheldon D. Stuchell, Senior Project Manager Licensing Processes Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project Nos. 669 and 689

Enclosure:

RAI Questions cc w/encl: See next page

August 30, 2010 Mr. Neil Wilmshurst Vice President & Chief Nuclear Officer Electric Power Research Institute 1300 West WT Harris Blvd Charlotte, NC 28262

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION NUMBER 4, RE: ELECTRIC POWER RESEARCH INSTITUTE TOPICAL REPORT 1016596, "MATERIALS RELIABILITY PROGRAM (MRP): PRESSURIZED WATER REACTOR INTERNALS INSPECTION AND EVALUATION GUIDELINES (MRP-227 - REV. 0)" (TAC NO. ME0680)

Dear Mr. Wilmshurst:

By letter dated January 12, 2009, Electric Power Research Institute submitted for U.S. Nuclear Regulatory Commission (NRC) staff review Topical Report 1016596, "Materials Reliability Program (MRP): Pressurized Water Reactor Internals Inspection and Evaluation Guidelines."

Upon review of the information provided to date, the NRC staff has determined that additional information is needed to support completion of the review. During discussions with Chuck Welty, Technical Executive, and Anne Demma, Project Manager, we agreed that the NRC staff will receive your response to the enclosed RAI questions by October 29, 2010.

If you have any questions regarding the enclosed RAI questions, please contact me at 301-415-1847.

Sincerely, IRA!

Sheldon D. Stuchell, Senior Project Manager Licensing Processes Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Project Nos. 669 and 689

Enclosure:

RAI Questions cc w/encl: See next page DISTRIBUTION:

PUBLIC RidsNRRDCICvib PSPB Reading File JJolicoeur (Hardcopy) BParks, NRR RidsNrrDpr GCranston, NRR MHartzman, NRR RidsNrrDprPspb TLupold, RES RTregoning, RES RidsNrrLADBaxley GCheruvenki, NRR EXTERNAL:

RidsAcrsAcnwMailCenter JMedoff, NRR ademma(,epri.com RidsNrrPMTMensah JDozier, NRR ihr(.nei.ora RidsOgcMailCenter MKhanna, NRR vka(~nei orgq BEIliot, NRR RAppajosula,RES ADAMS ACCESSION NO.: ML102310552 NRR-106

Nuclear Energy Institute Project No. 689 Electric Power Research Institute Project No. 669 cc:

Mr. Anthony Pietrangelo, Senior Vice President Mr. John Butler, Director

&Chief Nuclear Officer Operations Support Nuclear Generation Nuclear Energy Institute Nuclear Energy Institute 1776 I Street, NW, Suite 400 1776 I Street, NW, Suite 400 Washington, DC 20006-3708 icb(cnei.orQ Washington, DC 20006-3708 arD(c*nei .ora Mr. James H. Riley, Director Mr. Jack Roe, Director Engineering Security Nuclear Energy Institute Nuclear Energy Institute 1776 1Street, NW 1776 I Street, NW, Suite 400 Washington, DC 20006-3708 Washington, DC 20006-3708 ihr(a.nei.orq iwr(anei.ora Mr. Neil Wilmshurst Mr. Charles B. Brinkman Vice President & Chief Nuclear Officer Washington Operations Electric Power Research Institute ABB-Combustion Engineering, Inc. 1300 West WT Harris Blvd 12300 Twinbrook Parkway, Suite 330 Charlotte, NC 28262 Rockville, MD 20852 nwilmshu(Depri.com brinkmcb(,westinqhouse.com Mr. David J. Modeen Mr. James Gresham, Manager Director, Extemal Affairs Regulatory Compliance and Plant Licensing EPRI Westinghouse Electric Company 1300 W. T. Harris Boulevard P.O. Box 355 Charlotte, NC 28262-8550 Pittsburgh, PA 15230-0355 dmodeen(-epri.com qreshaia(a)westinqhouse.com Dr. Sean Bushart Ms. Barbara Lewis EPRI Assistant Editor 3412 Hillview Avenue Platts, Principal Editorial Office Palo Alto, CA 94304-1338 1200 G St., N.W., Suite 1100 sbushart(a.epri.com Washington, DC 20005 Barbara lewis*,platts.com Mr. Kurt Edsinger EPRI Mr. Alexander Marion, Vice President Nuclear 3412 Hillview Avenue Operations Palo Alto, CA 94304-1338 Nuclear Energy Institute kedsinae(cepri.com 1776 I Street, NW, Suite 400 Washington, DC 20006-3708 amC)nei.orq

Mr. Ken Canavan EPRI 1300 W.T. Harris Boulevard Charlotte, NC 28262-8550 kcanavan(,epri.com Mr. Greg Selby EPRI 1300 W. T. Harris Boulevard Charlotte, NC 28262-8550 pselby(aepri.com Mr. David Steininger EPRI 3412 Hillview Avenue Palo Alto, CA 94304-1338 dsteinin(Depri.com

MRP-227 Request for Additional Information (RAI) #4 REQUEST FOR ADDITIONAL INFORMATION (RAI)

BY THE OFFICE OF NUCLEAR REACTOR REGULATION TOPICAL REPORT (TR) 1016596, "MATERIALS RELIABILITY PROGRAM (MRP):

PRESSURIZED WATER REACTOR INTERNALS INSPECTION AND EVALUATION GUIDELINES" (MRP-227 - REV. 0)

ELECTRIC POWER RESEARCH INSTITUTE (EPRI)

PROJECT NO. 669 In a letter dated January 12, 2009, EPRI submitted a TR MRP-227, Rev. 0, "Pressurized Water Reactor (PWR) Internals Inspection and Evaluation Guidelines," which addresses an aging management program (AMP) for the PWR reactor vessel internal (RVI) components. During the review process, the NRC staff provided three separate Requests for Additional Information letters, which were responded to. The NRC staff continues its review of TR MRP-227, Rev. 0, and supporting information.

The NRC staff has developed the following set of questions based upon its continuing review of topical report MRP-227, "Pressurized Water Reactor Internal Inspection and Evaluation Guidelines," supporting technical reports provided by the Materials Reliability Program (MRP)/Electric Power Research Institute (EPRI) (about which some of the following questions are directed), and MRP/EPRI responses to prior RAls.

The staff has concluded that if the following questions can be answered adequately and completely, the staff can move forward and complete its review and safety evaluation (SE) for MRP-227. However, it should be recognized that inadequate or incomplete answers to the following questions may result in the imposition of limitations and conditions on the use of MRP-227 in the staff's SE. In particular, the staff will strongly consider the resolution of any remaining issues which can be addressed as plant-specific action items by the imposition of limitations and conditions within the SE.

1. Develop a comprehensive roadmap that describes how components that were categorized as non-A (i.e., Category B and C components) by the initial screening analysis were binned into the final recommended inspection categories (i.e., primary, expansion, existing, and no additional measures). The roadmap should include, for a sufficient number of components to demonstrate the general practice, the results of the initial screening analysis, the failure modes, effects, and criticality analysis (FMECA) susceptibility and consequence results and rationale supporting the results, a brief ENCLOSURE 1 Page I of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 summary of the functionality assessment results (if applicable) and a description of the use of these results (i.e., impact) in defining the recommended inspection program, a summary of the recommended inspection program (i.e., inspection type, periodicity, and accessibility requirements), and the supporting basis for this program. As an integral part of this demonstration, ensure that justification/rationale exists for classifying initial B and C components with medium or high failure consequences in any bin other than the primary category. The roadmap should cite references and data sources used to develop information/justification supporting the recommended ranking for each component discussed (e.g., consequence analysis).

The roadmap should identify, for each component discussed, the loading sources that provide the normal operating stresses considered for each component. Loading sources may include, for example, pressure, thermal, deadweight, residual stress (e.g., from fabrication/installation, welding), hydrodynamic, and preload stresses. The roadmap should identify which, if any, of these stresses may produce significant cyclic or transitory stresses under normal operating conditions. Indicate the portion of the normal operating stress due to static loading sources and the portion attributed to significant cyclic or transitory load sources that may contribute to fatigue.

2. Develop a comprehensive roadmap for components that were categorized as A by the initial screening analysis and have medium or high failure consequences. This roadmap should include the consequence results and supporting rationale associated with the recommended categorization. Indicate if and why any of these components were moved into the primary or expansion categories based on either the FMECA or the functionality analysis. The roadmap should cite references and data sources used to develop information/justification supporting the recommended ranking for each component (e.g.,

consequence analysis).

The roadmap should identify, for each of the medium or high failure consequences components, the loading sources that provide the normal operating stresses considered for each component. Loading sources may include, for example, pressure, thermal, deadweight, residual stress (e.g., from fabrication/installation, welding), hydrodynamic, and preload stresses. The roadmap should identify which, if any, of these stresses may produce significant cyclic or transitory stresses under normal operating conditions.

Indicate the portion of the normal operating stress due to static loading sources and the portion attributed to significant cyclic or transitory load sources that may contribute to fatigue.

3. In addition to the information supplied in response to Questions 1 and 2, the FMECA process should be more fully documented to support its use in the development of the Page 2 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 recommended aging management programs. Discuss and describe the following aspects of the FMECA analysis for both the Westinghouse/CE and B&W studies1 :

a) Relate the list of experts who participated in the FMECA with the list of required technical specialties to perform this analysis.

b) Describe the FMECA process used in each study including the following items:

• key assumptions,

• scope and motivation (e.g., what components addressed, use to confirm initial screening results, use to develop recommendations for component classification),

• approach (e.g., ranking and consensus process used to obtain results, consideration of either single degradation mechanisms or combined mechanisms, ranking definitions, development of classification matrices, use of classification matrices to develop component classification recommendations, evaluation of the effects and consequences of component failure for Westinghouse, CE, and B&W designs), and

• analysis and results (e.g., how individual estimates were amalgamated to determine final estimate, how ranking biases among the various experts were addressed and reconciled, results of both susceptibility and consequence analyses, the impact of the FMECA on the final severity of component failure rankings with a focus on instances where the FMECA was used to change initial severity rankings).

In particular, identify and then discuss differences between the Westinghouse/CE and B&W FMECA studies. This discussion should address, for example, differences in the susceptibility/failure likelihood and severity/damage likelihood definitions, consideration of the effects of either single or multiple degradation mechanisms, and recommended inspection categories based on the tabulated matrix values.

Additionally, this discussion should identify any Westinghouse, CE, and B&W components having a similar or equivalent function that received different aging management program classifications based on the FMECA, and, as appropriate, either document why the differences exist or describe how differences in the Westinghouse/CE and B&W FMECA results for similar components were reconciled.

c) One of the FMECA assumptions states "...no consideration was given to manufacturing errors, maintenance errors, installation errors, transport errors, or any other type of random or human errors." Why aren't these failure modes considered for components that have complex manufacturing, maintenance, or installation Note that some of this information was presented during the meeting on June 8, 2010 between the industry and the NRC.

Page 3 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 procedures? What is the justification for not considering these effects for components with medium or severe failure consequences?

d) Section 6.2 of MRP-191 indicates that a "...FMECA does not serve well to identify multiple failures." What is meant by multiple failures (i.e., common cause, indirect, cascading failures, or another type)? How is this deficiency in the FMECA process addressed as part of the aging management strategy?

e) Section 6.2 of MRP-191 also indicates "...operability, reliability, and availability issues were also considered." This statement is unclear. What explicit issues were considered in the FMECA and how were they considered by the experts? For example, was there an explicit process to factor these issues into the FMECA rankings or were these issues used to determine which of several components with similar degradation mechanisms and likelihood and/or failure consequence would be chosen for the primary inspection category?

4. In addition to the information provided in response to Question 1, discuss the impact of the functionality analysis in terms of determining or modifying the final inspection requirements for components. Specifically, provide an overview of how the functionality analysis impacted the recommended component classifications (i.e., primary, expansion, existing programs, and no additional measures) that were initially developed prior to conducting the functionality analysis. For primary and expansion components indicate how the functionality analysis was used to determine the type of inspection and the inspection periodicity. Finally, identify all similar Westinghouse, CE, and B&W components (i.e., those that perform a similar function and have similar failure consequences) that have different final inspection classification and requirements based on the functionality analysis. Provide justification for any differences that exist.

5. In addition to the information provided in response to Questions 1 and 3, discuss in general how susceptibility to multiple degradation mechanisms was considered when developing the final inspection recommendations. The final inspection category recommendations for each component appear to typically be based on the susceptibility and consequences associated with the single most-dominant degradation mechanism.

However, many components are subject to multiple degradation mechanisms and it is not clear how the synergistic effect of multiple degradation mechanisms was considered in the final recommendations. The concern is that a component subjected to multiple degradation mechanisms may be more likely to experience a greater level of total degradation than a component that is subject to a single mechanism (even though the component with the single mechanism may be more susceptible to that mechanism).

Discuss the acceptability of the recommended inspection method for primary components that are susceptible to multiple degradation mechanisms. Demonstrate that the recommended method is capable of identifying degradation due to all significant contributing mechanisms (and not just the single most-dominant mechanism) before component design margins are exceeded.

Page 4 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4

6. Several previous RAts (e.g., RAIs 2-11, 2-18, and 3-8) have questioned whether plant-specific analyses are required to demonstrate that each plant is appropriately represented by MRP-227 such that the proposed aging management programs (AMPs) are applicable. That is, confirmation that the plant's initial design and operating conditions fall within the scope of the MRP-227 evaluation, the plant complies with important assumptions underlying the MRP-227 analysis, and changes in plant design or operating conditions (e.g., resulting from power uprates) have been appropriately considered. Meeting these conditions is necessary to ensure that the plant-specific AMP inspection requirements (i.e., the primary inspection components, inspection type and periodicity) would not be different from the MRP-227 recommendations determined through more generic evaluation.

The responses to these various RAls have indicated that a plant-specific analysis to demonstrate the applicability of MRP-227 guidance is not required because plant-specific differences have been considered by: (a) evaluating operating experience throughout the commercial fleet; (b) using a conservative "out-in" core loading pattern in the functionality analysis; and, (c) assessing several known plant-specific conditions in the FMECA. The responses also justify the representativeness of MRP-227 because:

(a) base load operational profiles (i.e., fixed power levels) are similar among plants, and (b) no design changes have been enacted by plants other than those identified in generic industry guidance or recommended by the original nuclear steam system supply vendors. However, given the variability in design and operational conditions that currently exists in PWR plants, the staff is not convinced the MRP-227 AMP requirements are necessarily appropriate for each plant. For instance, it is not clear that the "out-in" core loading pattern is conservative given that some degradation mechanisms do not initiate until low-leakage core conditions are imposed in the functionality model.

Therefore, the staff requests that guidance be developed that will allow individual licensees to assess the applicability of the MRP-227 method and results. This guidance should particularly focus on demonstrating the applicability of (a) the FMECA and functionality assessments, and (b) the recommended inspection category, inspection method and periodicity for each component. Specifically, this guidance should allow a licensee to determine if plant-specific differences in the RVI design or operating conditions (i.e., power uprate level) result in different component inspection categories (i.e., primary, expansion, existing, and no additional measures) than recommended within MRP-227. Alternatively, additional analysis or justification may be provided to demonstrate that the MRP-227 approach and results are generically applicable such that plant-specific differences in the RVI design or operating conditions do not result in different component inspection categories than recommended within MRP-227.

Page 5 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 In the absence of adequate guidance, the staff will consider the need to implement limitations and conditions on the use of MRP-227, which would address plant-specific action items necessary to address this issue for each facility.

7. Provide guidance on the process that should be followed by licensees if the plant-specific application of the MRP-227 guidelines identifies that inspection or aging management of a primary component (i.e., as defined in MRP-227) is not necessary.

The guidance should address, for example, the plant-specific criteria and process for reclassifying the aging management program for a primary component, disposition of linked expansion components, and identification of (an) alternative plant-specific primary component(s) to be used in lieu of the generic MRP-227 recommendation for that degradation mechanism.

The response to this question should specifically propose text that would be added to the "-A" version of MRP-227 to address this issue.

8. This question discusses accessibility requirements for primary and expansion components. Define the appropriate inspection coverage to ensure the component being inspected does not lose its intended function and the process to be followed if the inspection does not meet the inspection coverage. Provide additional guidance on the component accessibility requirements for each primary and expansion component (i.e.,

those in MRP-227 Tables 4-1 through 4-6, 4-8 and 4-9) such that the results of the inspection can be credited as satisfying the requirements of the aging management program. This guidance should include, at a minimum, the following considerations:

a) For each component, identify the location(s) where degradation is expected.

b) Define the appropriate inspection coverage at this location to ensure that enough of the surrounding material is inspected such that there is assurance that the degradation will be identified before it challenges component or system integrity (i.e.,

the intended function of the component is retained).

c) Describe the procedure that a licensee should follow if inspection accessibility is insufficient to provide the required inspection coverage or if the inspection does not meet other minimum requirements as specified in MRP-227 and MRP-228.

This procedure must address providing an appropriate justification for continued operation with the reduced examination requirements to the NRC for review and approval. The guidance should address the process for adjusting the inspection area and/or coverage interval for both welded and non-welded components as a function of the component being inspected and/or the degradation mechanism being assessed during plant-specific inspections.

With respect to inaccessible components, the MRP should:

Page 6 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 a) Identify any components that are; (1) totally inaccessible (cannot be inspected) and (2) the management of their aging effects is dependent on the inspection results from another primary, expansion, or an existing component.

b) For the components identified in "a)," identify the primary, expansion, or other existing components that are the surrogate for the inaccessible components and explain why the surrogates are the limiting components for the aging effects that need management.

c) For the totally inaccessible components or for the inaccessible portion of primary or expansion components, what are the requirements that the licensee must follow to ensure that the components do not lose their intended function as a result of flaws in the accessible components.

The response to this question should specifically propose text that would be added to the "-A" version of MRP-227 to address this issue.

9. A number of components are identified as being covered by existing programs.

However, there is no summary of existing RVI programs provided in MRP-227 or supporting documentation. Add a summary of existing programs being credited to MRP-227. If an existing program is consistent with a program definition given in the staff's Generic Aging Lessons Learned (GALL) report, it is sufficient to simply identify the related GALL program definition. For existing programs lacking such a convenient reference, a summary should be provided which describes the following program requirements for an acceptable existing program: (a) scope (i.e., components inspected/monitored), (b) the applicable inspection, monitoring, or testing requirements and acceptance criteria, (c) the periodicity of the program, and (d) any other relevant requirements. This summary should identify the degradation mechanism(s) that are intended to be monitored or mitigated by each existing program and provide justification that each program is sufficient to monitor and/or mitigate all the expected degradation mechanisms identified in MRP-227 for the applicable component(s).

The response to this question should specifically propose text that would be added to the "-A" version of MRP-227 to address this issue.

10. MRP-227 guidance is used to develop component-level aging management programs and inspection requirements. Further, the development of these programs and requirements has not considered the effects of transitory design basis events (DBE) on the performance of degraded components or structures. However, as indicated in the response to RAI 2-16, the current industry expectation is that "...when age-related degradation effects are detected during the examinations specified in MRP-227, the suitability of the degraded component for continued service will necessarily take into consideration the full range of design basis event (DBE) effects." Therefore, staff Page 7 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 believes that guidance and requirements should be provided to ensure that licensees perform consistent plant-specific evaluations of the effects of degraded components under both normal and transitory DBEs (i.e., normal, upset, emergency, and faulted loading conditions). These evaluations should provide reasonable assurance that the systems associated with the degraded components will maintain required design margins and that inspection, repair, and replacement requirements are both adequate and timely. The guidance and requirements should, in part:

a) Identify the number or percentage of related primary or expansion components that should be inspected and the allowable number of degraded or non-functioning components for each system to ensure acceptable performance under DBEs.

Discuss the appropriateness of developing generic versus plant-specific inspection requirements for each system. Alternatively, the guidance should describe how the plant-specific analysis should be performed to determine the inspection sample and allowable number of degraded components for each system, and b) Describe the additional inspection requirements that should be triggered if degraded components are found as part of the primary inspection.

The consideration of item b) should provide guidance for increasing the sample size to inspect other similar components within the system that are subject to the observed degradation mechanism. It should either provide guidance for expanding the inspection to components in other systems that are subject to the same degradation mechanism or justify the adequacy of existing expansion criteria in MRP-227. As an example, this guidance should specify the percentage of baffle-to-former bolts that should be inspected and the percentage that may be degraded before system performance under design basis loading conditions is affected. If degraded baffle-to-former bolts are found, this guidance should next specify the additional baffle-to-former bolts that should be inspected and, for instance, the number of expansion baffle-to-baffle or core barrel-to-former bolts that should be inspected.

11. RAI 2-1 asked for justification of the inspection periodicity recommended in MRP-227 for reactor vessel internal components given that there is little operating experience for basing inspection periodicity and that analysis to evaluate the evolution of degradation in these components has a large degree of uncertainty. The response to that question primarily justifies the adequacy of the recommended inspection intervals based on the functionality analysis, which predicts that degradation will gradually worsen over time and will not suddenly progress. However, the inspection periodicity is not based, as is typically the case, on an evaluation of the maximum level of degradation that is acceptable for components to fulfill their intended design requirements, and the predicted time to reach this level of degradation based on the extent of degradation found during the inspection and evaluation of the rate of degradation with continuing operation. Therefore, the staff requests additional justification for recommended inspection intervals. This justification should address why the current MRP-227 Page 8 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 approach is appropriate for determining inspection periodicity and that determining periodicity based on a component level evaluation to ensure that the required component design margins are retained between inspections is not required.

Alternatively, provide information about the plans of licensees to perform the initial primary inspections required by MRP-227. Staff understands that some licensees are planning to inspect all required primary components during the first refueling outage after entering into the license renewal period. Staff therefore seeks to determine if this approach is being adopted by other licensees that have or will shortly enter the license renewal period. Clarify if this approach is either recommended or required within MRP-227.

12. RAI 2-21 asked about the need to develop more definitive acceptance criteria for inspections to ensure uniform interpretation and implementation from plant to plant. The industry response indicated that more definitive criteria in MRP-227 is not needed because the inspectors will receive component-specific training and that any observable degradation will require further disposition through the corrective action process.

However, staff remains concerned that this approach is not sufficient given the variability associated with inspection conditions and interpretation of inspection results. Therefore, staff requests that more definitive inspection acceptance criteria be developed for the VT-1, EVT-1, UT, and VT-3 inspections for each of the primary and expansion components. These criteria should be a function of the required accuracy and precision of the particular technique and also the application of this technique to each particular component (i.e., accessibility limitations, expected degradation location, expected degradation type).

In the absence of adequate guidance, the staff will consider the need to implement limitations and conditions on the use of MRP-227, which would address plant-specific action items necessary to address this issue for each facility.

13. Provide a description of how international and the US operating experience is (or is planned to be) documented, tracked, and updated so that it will support continued refinement of MRP-227 guidance and inform plant-specific aging management programs.

The staff believes that it would be advisable for documentation regarding the US and international operating experience related to degradation in RVI components be compiled in a single document that could be used to support the process of updating MRP-227.

14. Verify that neither MRP-210, "Materials Reliability Program: Fracture Toughness Evaluation of Highly Irradiated PWR Stainless Steel Internal Components," nor Section 6 of MRP-227 will be used to disposition (i.e., determine need to repair, need to replace, or inspection periodicity) degraded components identified during RVI inspections and that Page 9 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 this guidance will instead be provided by WCAP-1 7096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2, December 2009. Ifthis is not the case, provide a description of the relationship between MRP-21 0, Section 6 of MRP-227, and WCAP-17096 and identify the aspects of the disposition analysis that will be governed by each document.

15. There have been several previous RAIs related to cast austenitic stainless steel (CASS) materials, but several questions/issues remain.

a. The industry is currently supporting using a minimum irradiation embrittlement (IE) threshold of 1 displacement per atom (dpa) to determine susceptibility of CASS components to IE, yet available data seems to support a threshold of 0.3 dpa or less. There is little data between 0.05 and 1 dpa and current data indicates some toughness decrease between 0.3 and 1 dpa. Would a reduction in the screening threshold from 1 to 0.3 or 0.05 dpa result in additional components screened in for IE? Ifso, identify the CASS components that would be screened in for IE susceptibility due to these lower screening thresholds.

Finally, many CASS components are in the A inspection category. Provide the basis/justification for placing these components in the A category.

b. The fracture toughness of CASS can degrade significantly due to thermal embrittlement (TE) and the toughness of both CASS and other stainless steel materials can decrease significantly as neutron fluence increases. When the dose exceeds 5 dpa, available data indicates that fracture toughness can be extremely degraded in many materials. The staff's concern is that the fracture toughness in CASS components may get so low due to TE and/or IE that preexisting fabrication or service-induced flaws that are smaller than the inspection resolution may challenge component integrity under normal loading or under design basis events. Additional guidance to licensees may be needed either in MRP-227 or WCAP-1 7096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2, December 2009, to address this situation. Describe how existing or planned guidance addresses this issue.

Otherwise, justify why such guidance is not needed.

16. MRP-227 identifies several components that require plant-specific aging analysis (e.g.,

fatigue analyses) to determine the appropriate inspection category. However, MRP-227 does not discuss or reference approved methods or acceptance criteria for conducting such analyses. Discuss why guidance is not necessary to ensure consistent application and interpretation of plant-specific aging analyses. Alternatively, if the industry plans to provide such guidance, discuss the plans, approach, and schedule for developing this guidance. This discussion should address how environmental effects should be treated in these analyses.

Page 10 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4

17. MRP-227 Tables 3-1, 3-2, and 3-3 identify component/aging mechanism combinations (where the combination is identified as either "primary," "expansion," or "covered by existing programs" in the tables) that are not identified in Tables 4-1 through 4-6, 4-8 and 4-9. Tables 3-2 and 3-3 also identify components (e.g., In-Core Instrumentation Thimble Tubes in CE internals and control rod guide tube support pins in Westinghouse Internals) that are not identified in Tables 4-1 through 4-6, 4-8 and 4-9 at all. Identify the component/aging mechanism combination in Tables 3-1, 3-2, and 3-3 that are not identified in Tables 4-1 through 4-6, 4-8 and 4-9. Explain how each of these aging mechanisms will be managed and revise MRP-227, if appropriate. Ifthe aging management review (AMR) line items that were previously provided to update the Generic Aging Lessons Learned Report in conjunction with the staffs review of MRP-227 need to be revised, provide recommended changes to the AMR line items.

18. As a follow-up to RAI 2-26, clarify if components that are predicted to locally exceed 5%

swelling by volume are inspected for cracking at those locations. Provide justification why any such components that exceed this criterion are not recommended for inspection.

19. The following components, listed as an example only, were originally identified for potential aging degradation but they were dispositioned under "No Measures" category.

Provide an explanation for not performing any analysis prior to binning them under the "No Measures" category.

COMBUSTION ENGINEERING COMPONENTS MRP-232 Referenc -

Component Aging Effect Reference Core Support Plate Bolts Irradiation Embrittlement Table 2-11 Fuel Alignment Pins (304 Irradiation Embrittlement Table 2-11 stainless steels)

Core Shroud Tie Rods Irradiation Embrittlement Table 2-11 Core Shroud Tie Rods Irradiation Induced Stress Table 2-16 Relaxation Tables2-52-3 and Core Support Plate IASCC, Wear 2-5 Page 11 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 WESTINGHOUSE COMPONENTS MRP-232 Referenc -

Component Aging Effect Reference Lower Core Plate Fuel Irradiation Embrittlement Table 2-12 Alignment Pin Bolts Lower Core Plate Fuel Irradiation Induced Stress Table 2-17 Alignment Pin Bolts Relaxation Bottom Mounted Tables 2-4 and Instrumentation (BMI) IASCC, Irradiation Embrittlement 2-12 Column Bodies IASCC, Thermal Embrittlement, Tables 2-4. 2-10 and Irradiation Embrittlement and 2-12

20. Many licensees have incorporated ANS 51.1, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," which categorizes transient events in a classification scheme by condition, into facility licensing bases. According to the standard, an acceptance criterion for a Condition II event is that by itself, a Condition II incident cannot generate a more serious incident of the Condition III or IVcategory without other incidents occurring independently or result in a consequential loss of function of the reactor coolant system or reactor containment barriers. For example, an anticipated operational occurrence, such as a turbine trip from full power, should not cause a degraded component inside the reactor vessel to fail in such a way that a control element assembly ejection could occur. Further detailed discussion regarding this criterion is available in NRC Regulatory Issue Summary 2005-29, "Anticipated Transients that Could Develop Into More Serious Events."

For those components that the FMECA or functionality analyses provided a basis to reduce or eliminate inspection requirements, address whether consideration of this "non-escalation" criterion affects this basis.

21. Address the effects of failures of uninspected components and components with failure modes that aren't detectable during normal operations (i.e., undetectable failure modes) through the following considerations:

a) Discuss whether the failure of any such component(s) could be an initiating event for a plant transient or other accident.

b) Discuss the effect of failure of any such component(s) on system performance assuming a design basis event (i.e., plant transients, accidents, and seismic events representative of upset, emergency and faulted loading conditions) occurs prior to mitigating the failure. As part of this discussion, describe any analysis that has been Page 12 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 performed, or any plant-specific analysis that is needed, to demonstrate that acceptable system design margins are retained under this scenario.

Finally, discuss whether the final recommendation not to inspect these components is affected by addressing the scenarios described in a) and b) above.

22. MRP-1 90, Section 3 discusses component failure modes that aren't detectable during normal operations (i.e., undetectable failure modes). Provide specific examples of important components susceptible to these failure modes. Describe any special consideration or weighting that components susceptible to these failure modes received in either the FMECA (e.g., through the failure severity rankings) or the final MRP-227 inspection recommendations (e.g., by elevating the component to the primary inspection category) given that the component failure may not be discovered until the next refueling outage (i.e., up to 2 years after failure occurs). Provide specific examples to illustrate the process used to evaluate these components.

23. Identify any components that should be replaced either prior to the period of extended operation or during the period of extended operation because they may not be able to perform their intended function during design basis events (normal, transient, emergency and faulted conditions) based on the results of the FMECA or functionality assessment.

24. Tables 2-18 and 2-19 in MRP-232 and Table 3-8 in MRP-231 indicate that a licensee's aging management program will inspect CE, Westinghouse and B&W RVI components for thermally or irradiation-enhanced stress relaxation. However, various CE, Westinghouse and B&W RVI components that are susceptible to thermally and irradiation-enhanced stress relaxation have been downgraded from Categories B or C to the "No Additional Measures" Category.

Document the basis of the evaluation utilized to downgrade these components to the "No Additional Measures" Category. Demonstrate that both inspected and uninspected components susceptible to thermal or irradiation-enhanced stress relaxation maintain their design function during emergency and faulted events postulated at the end of the period of extended operation. This demonstration should show that the recommended inspection method is adequate for identifying or assessing stress relaxation before design margins become inadequate.

Ifa generic evaluation of the adequacy of such components under design basis loading is not possible, identify plant-specific action items that must be performed by licensees to ensure these components will be able to maintain their design function during design emergency and faulted conditions at the end of the period of extended operation.

In particular, identify the projected loss of preload due to stress relaxation at the end of the period of extended operation for the following bolts.

Page 13 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 a) Combustion Engineering---Core Support Column Bolts; Core Shroud Bolts; Guide Lug Insert Bolts; Barrel-Core Shroud Bolts b) Westinghouse---- Baffle-edge Bolts; Baffle-former Bolts; Lower Support Column Bolts c) Babcock and Wilcox----Baffle-to-Baffle Bolts; Core Barrel-to-Former Bolts; Baffle-to-Former Bolts.

Explain why this loss in preload will not result in the loss of the intended function for these bolts during design basis events that are postulated at the end of the period of extended operation.

25. The effects of radiation on material ductility is a TLAA for B&W vessel internals. Section 4.2.6 of the Three Mile Island Nuclear Station Unit 1 License Renewal Application indicates the following:

The effects of irradiation on the materials properties and deformation limits for the reactor vessel internals was evaluated for the current licensing basis in Topical Report BAW-10008, Revision 1, Appendix E.

This analysis concluded that at the end of the forty years, the internals will have adequate ductility to absorb local strains at the regions of maximum stress intensity, and that irradiation will not adversely affect deformation limits. This analysis is a TLAA that will be managed by the PWR Vessel Internals program for the period of extended operation.

Explain how this issue has been addressed for B&W vessel Internals program. Are the effects of radiation on material ductility a TLAA for CE and Westinghouse vessel internals? If that is not the case, provide an explanation for not performing a TLAA evaluation in CE and Westinghouse vessel internals. Ifit is a TLAA, explain how this issue is addressed in the PWR vessel internals program.

26. RAI 2-20 asked about how the linkage between primary and expansion components was determined and how the expansion criteria (i.e., the results of the primary inspection that triggers an expansion inspection) was developed. While the response to RAI 2-20 is clear and the process is generally understood by staff, there is still a lack of explicit justification for many of the linkages and the explicit expansion criteria. That is, there is not a clear basis why the primary component was selected and why the expansion linkage is both appropriate and comprehensive (i.e., no other components should be linked).

The basis for the criteria used to trigger expansion inspections, and the acceptability of this basis, should be provided for each of the primary and expansion linkages. As an example, expansion criteria for the core barrel and baffle barrel bolts are not triggered unless there is a 5% or higher failure rate in the baffle former bolts. Similarly, a 10% rate Page 14 of 15 August 30, 2010

MRP-227 Request for Additional Information (RAI) #4 of rejection for either the upper core barrel or lower core barrel bolts triggers the expansion items. The basis for these expansion criteria should demonstrate that the failure rate or rate of rejection specified for the baffle former bolts and the upper core barrel or lower core barrel bolts are sufficient to ensure significant degradation is not occurring in the expansion components such that the design margin requirements for expansion components and associated systems are satisfied.

27. MRP-227 and supporting reports do not clearly document how the consideration of degradation mechanisms associated with weld heat-affected zones, weld repair, and variability in welding processes and parameters was addressed in the susceptibility evaluation. Provide an overview of how these issues were evaluated to determine the final AMP recommendations for welded components and provide specific examples to illustrate the impact of these issues on the final inspection requirements.

Page 15 of 15 August 30, 2010

SELECTRIC POWER RESEARCH INSTITUTE MRP Materials Reliability Program MRP 2010-066 (via email)

October 29, 2010 Document Control Desk U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852

Subject:

EPRI MRP Final Response to: REQUEST FOR ADDITIONAL INFORMATION NUMBER 4, RE: ELECTRIC POWER RESEARCH INSTITUTE TOPICAL REPORT 1016596,'MATERIALS RELIABILITY PROGRAM: PRESSURIZED WATER REACTOR INTERNALS INSPECTION AND EVALUATION GUIDELINES (MRP-227-REV. 0)' (TAC NO. ME0680), August 30, 2010 To Whom It May Concern:

Enclosed are two copies of the subject document provided in PDF format on compact disc (CD) storage devices. This material is non-proprietary.

If you have any questions on this item, please contact Anne Demma (ademma(&epri.com) at 650-855-2026.

Sincerely, Terry McAlister SCANA Chairman, Materials Reliability Program Cc: James Lash, First Energy Sheldon Stuchell, NRC (with 8 copies of Subject document)

Victoria Anderson, NEI David Steininger, EPRI Chuck Welty, EPRI Together . . Shaping the Future of Electricity PALO ALTO OFFICE 3420 Hillview Avenue, Palo Alto, CA 94304-1338 USA

• 650.855.2000

• Customer Service 800.313.3774

• www.epri.com

RAI Set 4- Final Responses: 10/29/2010 Final Responses to 4 th Set RAIs On MRP-227, Rev 0 October 29, 2010 1

RAI Set 4 - Final Responses: 10/29/2010 MRP-227 Request for Additional Information (RAI) #4 Titles of MRP Reports Referenced in MRP-227 or Referred to in RAI Responses MRP # Title EPRI #

MRP-128 MaterialsReliability Program:CharacterizationofDecommissionedPWR 1008202 Vessel Internals MaterialSamples - Material Certification,Fluence, and Temperature, 2003 MRP-134 MaterialsReliability Program:Frameworkand Strategiesfor Managing 1008203 Aging Effects in ReactorInternals, 2005 MRP-135 MaterialsReliability Program:Development of Material Constitutive Model 1018291

- Rev. 1 for IrradiatedAustenitic Stainless Steel, 2009 MRP-156 MaterialsReliability Program:PressurizedWater ReactorIssue Management 1012110 Table, PWR-IMT, Consequence ofFailure, 2005 A4RP-157 MaterialsReliability Program:Updated B&W Design Informationfor the 1012132 Issue Management Tables, 2005 MRP-J175 MaterialsReliability Program:PWR Internals MaterialAging Degradation 1012081 Mechanism Screeningand Threshold Values, 2005 MRP-189 MaterialsReliability Program:Screening, Categorization,and Ranking of 1018292

- Rev. 1 B& W-DesignedPWR Internals,2009 MRP-190 MaterialsReliability Program:FailureModes, Effects, and Criticality 1013233 Analysis of B& W-Designed PWR Internals,2006 MRP-191 MaterialsReliabilityProgram:Screening, Categorizationand Ranking of 1013234 Reactor Internalsof Westinghouse and Combustion EngineeringPWR Designs, 2006 MRP-210 MaterialsReliabilityProgram:FractureToughness Evaluation of Highly 1016106 IrradiatedPWR Stainless Steel Internal Components, 2007 MRP-211 MaterialsReliabilityProgram:PWR InternalsAge-Related Material 1015013 Properties,DegradationMechanisms, Models, and Basis Data- State of Knowledge, 2007 MRP-228 MaterialsReliabilityProgram:Inspection Standardfor ReactorInternals, 1016609 2008 2

RAI Set 4 - Final Responses: 10/29/2010 MRP # Title EPRI #

MRP-229 MaterialsReliability Program:FunctionalityAnalysisfor B& W-Designed 1019090

- Rev. 1 RepresentativePWR Internals, 2009 MRP-230 MaterialsReliability Program:FunctionalityAnalysisfor Westinghouse & 1019091

- Rev. 1 CE-DesignedRepresentativePWR Internals, 2009 MRP- MaterialsReliability Program:Aging Management Strategiesfor B&W- 1019092 231-Rev. DesignedPWR Internals, 2009 1

MRP-232 MaterialsReliability Program:Aging Management Strategiesfor 1016593 Westinghouse and Combustion EngineeringPWR Internals, 2008 3

RAI Set 4 - Final Responses: 10/29/2010 RAI 4-1: Develop a comprehensive roadmap that describes how components that were categorized as non-A (i.e., Category B and C components) by the initial screening analysis were binned into the final recommended inspection categories (i.e., primary, expansion, existing, and no additional measures). The roadmap should include, for a sufficient number of components to demonstrate the general practice, the results of the initial screening analysis, the failure modes, effects, and criticality analysis (FMECA) susceptibility and consequence results and rationale supporting the results, a brief summary of the functionality assessment results (if applicable) and a description of the use of these results (i.e., impact) in defining the recommended inspection program, a summary of the recommended inspection program (i.e., inspection type, periodicity, and accessibility requirements), and the supporting basis for this program. As an integral part of this demonstration, ensure that justification/rationale exists for classifying initial B and C components with medium or high failure consequences in any bin other than the primary category. The roadmap should cite references and data sources used to develop information/justification supporting the recommended ranking for each component discussed (e.g., consequence analysis).

The roadmap should also identify, for each component discussed, the loading sources that provide the normal operating stresses considered for each component. Loading sources may include, for example, pressure, thermal, deadweight, residual stress (e.g., from fabrication/installation, welding), hydrodynamic, and preload stresses. The roadmap should identify which, if any, of these stresses may produce significant cyclic or transitory stresses under normal operating conditions. Please indicate the portion of the normal operating stress due to static loading sources and the portion attributed to significant cyclic or transitory load sources that may contribute to fatigue.

Response: A separate roadmap document has been developed to augment the RAI responses and is included as part of the overall response package as Appendix B. It describes the eight-step process from component identification through assignment of inspection recommendations. It also points to the MRP documents which contain the details of the evaluations performed in the development of MRP-227. The roadmap does contain examples of how components were screened in or out, classified and re-evaluated.

The following are examples of the process used for binning components.

Example: Westinghouse Bottom Mounted Instrumentation Column Bodies listed as Expansion Item Original screening results: MRP- 191 Table 5-1 4

RAI Set 4- Final Responses: 10/29/2010

" Mechanisms: SCC, IASCC, Irradiation Embrittlement, Fatigue, Void Swelling Functional

Description:

" MRP-232 Section 4.2.6: BMI column assemblies provide a path for the flux thimbles into the core from the bottom of the vessel and protect the flux thimbles during operation of the reactor.

FMECA

Conclusion:

MRP-191 Table 6-5

• Medium Failure Probability, Low Consequence Analysis of Degradation Mechanisms: MRP-232 Section 4.2.6.1

" Expansion based on cracking in CRGT lower flanges o The primary function of the BMI columns is to allow insertion and withdrawal of the flux thimbles, and as was noted several times, failures within the columns would be indicated by difficulty with the insertion of the flux thimbles during a refueling outage. Thus, detailed inspections are not required, and this component is classified as being an Expansion inspection component, required only when the regular withdrawal and insertion of the flux thimble indicates malfunction.

o Analysis of lower core plate indicated irradiation effects are overestimated.

o BMI system has no structural function.

Example: B&W Core barrel cylinder The core barrel supports the fuel assemblies, lower grid, flow distributor, and incore instrument guide tubes. The primary function of the core barrel cylinders and welds during normal power operation is to provide a flow envelope for the core and, thereby limit core bypass flow.

The core barrel cylinders and welds therefore do not have a direct core support safety function; however, they do have a safety function to control bypass around the core during a loss-of-coolant-accident (LOCA).

Initially screened in as Non-A and ultimately grouped as Expansion

" Screened in as Non-A for SCC, fatigue, and irradiation embrittlement in Step 3 (austenitic stainless steel, Type 304 with welds), all other mechanisms screened out

" FMECA expert panel determined that fatigue as an aging mechanism to have a low susceptibility that is supported by no known operating experience of fatigue, and the design criteria containing a significant amount of margin 5

RAI Set 4- Final Responses: 10/29/2010

" FMECA results identified SCC susceptibility as "B" and safety consequences as "1 ," which preliminarily categorizes this item as "Category A"

" FMECA results identified IE susceptibility as "C" and safety consequences as "1," which preliminarily categorizes this item as "Category B" (see table in Step 5)

" As shown in Section 3.2.3 (MRP-23 1) the core barrel cylinder is considered inaccessible and is not part of the standard 10-year ISI inspection. However, limited access to the former plates, core barrel cylinder, and otherwise inaccessible bolt locking devices is available through the flow bypass holes should a limited examination become necessary

• The baffle plates are the primary item for inspection from IE while the core barrel cylinder is considered to be Expansion item due to its low safety consequences and lower dose Regarding the general treatment of loads; specific loads on individual components were not explicitly considered. However, design loading for normal operating conditions (e.g.

pressure, thermal, dead-weight, residual stress, etc.) as well as design-basis loads were considered on the basis of relative magnitude and impact on operability and functionality using engineering judgment through the FMECA process. Additional insights into the treatment of loads by AREVA for B&W plants and Westinghouse for the CE and Westinghouse plants follow.

B&W:

Screening of PWR RV internals components was performed using the MRP-175 screening criteria in Table 3-2 of MRP-175. The screening process and results for the B&W RV internals are documented in MRP- 189 Rev. 1. Of the eight aging degradation mechanisms, only SCC and IASCC require applied tensile stress value as a screening parameter.

The original qualification of the B&W RV internals for B&W 177-FA was accomplished by both analytical and test methods. Therefore, original stress from the late 1960s and 1970s analyses using ASME Section III as a guidance were found only for a limited number of RV internals components including the stress analyses intended for the B&W 205-FA. In general, the maximum calculated stress values under normal and upset condition were used for screening SCC and IASCC. For non-bolting components, the loading sources for the stress values used included hydraulic pressure loads, dead weight, preload, operating basis earthquake (OBE), and flow-induced vibration (FIV) loads while the thermally-induced load was not included. For bolting components, the loading 6

RAI Set 4- Final Responses: 10/29/2010 sources for the stress values used included hydraulic pressure loads, dead weight, preload, OBE, and FIV and thermally-induced loads.

In addition to the earlier stress analyses described above, detailed stress calculations were performed in the 1980s for a number of RV internals high-strength bolts (i.e., Alloy A-286 and Alloy X-750) to address IGSCC identified at the time. The stress values listed in MRP- 189 for screening these high-strength bolts were from the stress calculations performed in the 1980s. In these cases, the loading sources included flow, dead weight, preload, and FIV and thermal loads from steady-state and transient operations.

Note, the applied stress defined in MRP-175 includes normal operating stress under steady-state condition and residual stress from fabrication and welding. However, residual stress due to welding or fabrication processes was not considered in any of the above stress calculations. Residual stress due to welding and fabrication processes was only addressed by use of the additional screening parameters "Multiple-Pass Weld" and "Cold-Work > 20%" during the screening process.

Westinghouse:

Specific loads on individual components were not explicitly considered on a plant by plant or design grouping basis. Typical loadings for normal operating conditions (e.g. pressure, thermal, dead-weight, residual stress, water hammer, etc.) were considered as the primary drivers for manifestation of degradation. This view was based on currently known operating histories. Design-basis loadings and combinations of loadings for the range of as licensed conditions (e.g. normal, upset, emergency, faulted) were considered on the basis of relative magnitude against typical allowable limits as defined from historical plant evaluations. In all cases the loads employed were within the acceptable bounds for various plant conditions based on a variety of as licensed Code requirements. Relative magnitudes varied widely as a result of the variety of as licensed conditions for the individual units when considering seismic and LOCA conditions as well as the range of normal operating transient effects. Where available, fatigue was considered from both a usage factor and range of stress perspective in considering effects. Potential postulated impacts on operability and functionality using engineering judgment through the FMECA process aided in determining levels of concern regarding operational functionality and safety.

RAI 4-2: Develop a comprehensive roadmap for components that were categorized as A by the initial screening analysis and have medium or high failure consequences. This roadmap should include the consequence results and supporting rationale associated with the recommended 7

RAI Set 4 - Final Responses: 10/29/2010 categorization. Indicate if and why any of these components were moved into the primary or expansion categories based on either the FMECA or functionality analysis. The roadmap should also cite references and data sources used to develop information/justification supporting the recommended ranking for each component (e.g., consequence analysis).

The roadmap should also identify, for each of the medium or high failure consequences component, the loading sources that provide the normal operating stresses considered for each component. Loading sources may include, for example, pressure, thermal, deadweight, residual stress (e.g., from fabrication/installation, welding), hydrodynamic, and preload stresses. The roadmap should identify which, if any, of these stresses may produce significant cyclic or transitory stresses under normal operating conditions. Please indicate the portion of the normal operating stress due to static loading sources and the portion attributed to significant cyclic or transitory load sources that may contribute to fatigue.

Response: A separate roadmap document has been developed to augment the RAI responses and is included as part of the overall response package as Appendix B. It describes the eight-step process from component identification through assignment of inspection recommendations. It also points to the MRP documents which contain the details of the evaluations performed in the development of MRP-227. The roadmap does contain examples of how components were screened in or out, classified and re-evaluated.

Using the Issue Management Tables (MRP-156 and MRP-157) the failure consequences are defined as:

A. Precludes the ability to reach safe shutdown B. Causes a design basis accident C. Causes significant onsite and/or offsite exposure D. Jeopardizes personnel safety E. Breaches reactor coolant pressure boundary F. Breaches fuel cladding G. Causes a significant economic impact For those components placed in Category A, all but one component (Lower Support Casting or Forging) the only failure consequence identified was "G, Causes a significant economic impact". As noted in Table 2 of the roadmap the Lower Support Casting or Forging also had a consequence "A, Precludes the ability to reach safe shutdown" assigned to it. As is noted in the roadmap section for the FMECA process:

"...the only component with potential safety-related consequence of failure identified in the IMT was the lower core support forging. (The cast stainless steel version of this component was screened-in due to thermal embrittlement 8

RAI Set 4- Final Responses: 10/29/2010 concerns.) Loss of support due to catastrophic failure of this structure could preclude safe shut down of the reactor. However, the FMECA panel could not identify any potential cause or mode of catastrophic failure that would require aging management of this large forging. Therefore the lower support forging was not reinstated for additional evaluation.

Also, as part of the FMECA all components initial screened as "A" components were reevaluated by the FMECA panels. In each case the "A" classification was confirmed.

Regarding the general treatment of loads; specific loads on individual components were not explicitly considered. However, design loading for normal operating conditions (e.g.

pressure, thermal, dead-weight, residual stress, etc.) as well as design-basis loads were considered on the basis of relative magnitude and impact on operability and functionality using engineering judgment through the FMECA process. Additional insights into the treatment of loads by AREVA for B&W plants and Westinghouse for the CE and Westinghouse plants follow.

B&W:

Screening of PWR RV internals components was performed using the MRP-175 screening criteria in Table 3-2 of MRP-175. The screening process and results for the B&W RV internals are documented in MRP- 189 Rev. 1. Of the eight aging degradation mechanisms, only SCC and IASCC require applied tensile stress value as a screening parameter.

The original qualification of the B&W RV internals for B&W 177-FA was accomplished by both analytical and test methods. Therefore, original stress from the late 1960s and 1970s analyses using ASME Section III as a guidance were found only for a limited number of RV internals components including the stress analyses intended for the B&W 205-FA. In general, the maximum calculated stress values under normal and upset condition were used for screening SCC and IASCC. For non-bolting components, the loading sources for the stress values used included hydraulic pressure loads, dead weight, preload, operating basis earthquake (OBE), and flow-induced vibration (FIV) loads while the thermally-induced load was not included. For bolting components, the loading sources for the stress values used included hydraulic pressure loads, dead weight, preload, OBE, and FIV and thermally-induced loads.

In addition to the earlier stress analyses described above, detailed stress calculations were performed in the 1980s for a number of RV internals high-strength bolts (i.e., Alloy A-286 and Alloy X-750) to address IGSCC identified at the time. The stress values listed in 9

RAI Set 4 - Final Responses: 10/29/2010 MRP- 189 for screening these high-strength bolts were from the stress calculations performed in the 1980s. In these cases, the loading sources included flow, dead weight, preload, and FIV and thermal loads from steady-state and transient operations.

Note, the applied stress defined in MRP-175 includes normal operating stress under steady-state condition and residual stress from fabrication and welding. However, residual stress due to welding or fabrication processes was not considered in any of the above stress calculations. Residual stress due to welding and fabrication processes was only addressed by use of the additional screening parameters "Multiple-Pass Weld" and "Cold-Work > 20%" during the screening process.

Westinghouse:

Specific loads on individual components were not explicitly considered on a plant by plant or design grouping basis. Typical loadings for normal operating conditions (e.g.

pressure, thermal, dead-weight, residual stress, water hammer, etc.) were considered as the primary drivers for manifestation of degradation. This view was based on currently known operating histories. Design-basis loadings and combinations of loadings for the range of as licensed conditions (e.g. normal, upset, emergency, faulted) were considered on the basis of relative magnitude against typical allowable limits as defined from historical plant evaluations. In all cases the loads employed were within the acceptable bounds for various plant conditions based on a variety of as licensed Code requirements.

Relative magnitudes varied widely as a result of the variety of as licensed conditions for the individual units when considering seismic and LOCA conditions as well as the range of normal operating transient effects. Where available, fatigue was considered from both a usage factor and range of stress perspective in considering effects. Potential postulated impacts on operability and functionality using engineering judgment through the FMECA process aided in determining levels of concern regarding operational functionality and safety.

RAI 4-3: In addition to the information supplied in response to Questions 1 and 2, the FMECA process should be more fully documented to support its use in the development of the recommended aging management programs. Discuss and describe the following aspects of the FMECA analysis for both the Westinghouse/CE and B&W studies':

a. Relate the list of experts who participated in the FMECA with the list of required technical specialties to perform this analysis.

1 Note that some of this information was presented during the meeting on June 8, 2010 between the industry and the NRC.

10

RAI Set 4- Final Responses: 10/29/2010

b. Describe the FMECA process used in each study including the following items:

" key assumptions,

• scope and motivation (e.g., what components addressed, use to confirm initial screening results, use to develop recommendations for component classification),

" approach (e.g., ranking and consensus process used to obtain results, consideration of either single degradation mechanisms or combined mechanisms, ranking definitions, development of classification matrices, use of classification matrices to development component classification recommendations, evaluation of the effects and consequences of component failure for Westinghouse, CE, and B&W designs), and

" analysis and results (e.g., how individual estimates were amalgamated to determine final estimate, how ranking biases among the various experts were addressed and reconciled, results of both susceptibility and consequence analyses, the impact of the FMECA on the final severity of component failure rankings with a focus on instances where the FMECA was used to change initial severity rankings).

In particular, identify and then discuss differences between the Westinghouse/CE and B&W FMECA studies. This discussion should address, for example, differences in the susceptibility/failure likelihood and severity/damage likelihood definitions, consideration of the effects of either single or multiple degradation mechanisms, recommended inspection categories based on the tabulated matrix values,. Additionally, this discussion should, identify any Westinghouse, CE, and B&W components having a similar or equivalent function that received different aging management program classifications based on the FMECA, and, as appropriate, either document why the differences exist or describe how differences in the Westinghouse/CE and B&W FMECA results for similar components were reconciled.

c. One of the FMECA assumptions states that "...no consideration was given to manufacturing errors, maintenance errors, installation errors, transport errors, or any other type of random or human errors." Why aren't these failure modes considered for components that have complex manufacturing, maintenance, or installation procedures?

What is the justification for not considering these effects for components with medium or severe failure consequences?

d. Section 6.2 or MRP-191 indicates that a "...FMECA does not serve well to identify multiple failures." What is meant by multiple failures (i.e., common cause, indirect, cascading failures, or another type)? How is this deficiency in the FMECA process addressed as part of the aging management strategy?

11

RAI Set 4- Final Responses: 10/29/2010

e. Section 6.2 or MRP-191 also indicates that "...operability, reliability, and availability issues were also considered." This statement is unclear. What explicit issues were considered in the FMECA and how were they considered by the experts? For example, was there an explicit process to factor these issues into the FMECA rankings or were these issues used to determine which of several components with similar degradation mechanisms and likelihood and/or failure consequence would be chosen for the primary inspection category?

Response: The response to the various parts of this RAI will take several forms. In some cases, the response to a particular part will point to a section in either MRP- 190 or MRP-191 that directly addresses the issue. In other cases, the response has already been provided by the industry in a presentation to the NRC staff during the June 8, 2010, meeting between the industry and the NRC staff, and the response will so indicate.

Finally, in some cases, the response is available in the separate roadmap document that has been developed to augment the RAI responses. That separate roadmap document is included as a part of the overall response package.

RAI 4-3(a) asks for the list of experts who participated in the FMECA process and their technical areas of expertise. The list of experts for the FMECA documented in MRP- 190 and their areas of expertise is shown on the Acknowledgements on Page ix of MRP-190.

MRP-191 does not provide the list of experts and their areas of expertise, but the areas of expertise were provided in the presentations given to the staff on June 8, 2010. For completeness, the Combustion Engineering/Westinghouse FMECA areas of expertise are shown below:

" Component design, testing and repair

" Structural modeling and analysis

" Thermal-hydraulics and systems analysis

• Neutron fluence and radiation analysis

• Materials degradation and failure experience

" Component inspection experience

" Risk assessment

" Inspection requirements

" System function and operating experience

" Licensing and regulatory interaction The list of experts and areas of expertise acknowledged in MRP-190 explicitly called out eighteen people -- designers, materials experts, radiation physics and neutronics/safety experts, thermal hydraulics systems engineers, stress and structural analysts, non-12

RAI Set 4- Final Responses: 10/29/2010 destructive examination experts, safety and accident analysts, license renewal (aging management) specialists, and project managers. While not specifically cited, these individuals in most cases were also experienced in component fabrication, installation, and operation. For completeness, the B&W FMECA areas of expertise are shown below:

• Component Design

• Materials

• Finite Elements/Stress/Dynamics Analysis

• Structural Analysis

• Non-Destructive Evaluation

• Fracture Mechanics

• License Renewal

• Safety/Accident Analysis

• Thermal-Hydraulics

• Neutronics/Radiation Physics RAI 4-3(b) essentially asks for a description of the FMECA process used by each vendor, with any substantive differences in process identified. For MRP- 190, most of this type of information is described in Section 3.1, but additional information on the process is given in other portions of Section 3 and the key assumptions are provided in Section 4. While the details of the FMECA deliberations were not recorded, the FMECA process steps can be followed reasonably well by starting with the flow chart shown below, and then using the table format in Appendix A to follow the logical exercise of that flow chart.

13

RAI Set 4- Final Responses: 10/29/2010 The tables in Appendix A of MRP- 190 contain eleven columns, starting with the component name, the FMECA identifier, the degradation mechanism(s), the failure mode(s), and the local and global failure effects. The next three columns reflect the deliberations of the experts after developing the failure mode(s) and failure effect(s) columns, giving the consensus ranking of susceptibility and the safety and economic consequence severity. The transition from the former columns to the latter columns represents the aggregate deliberations of the expert panel. The final two columns - the determination of detectability and the column with comments include insights representing the collective informed judgment of the experts, and are invaluable in understanding how the FMECA process results evolved.

For MRP- 191, the description of the Combustion Engineering/Westinghouse FMECA process is largely provided in Section 6.2, and the flow chart of that process is shown below.

14

RAI Set 4 - Final Responses: 10/29/2010 MI

• CRITERIA # DEGRADATION *WCAP..13627

" DEGRADATION a CONSEQUENCE OF FAILMUR

• SCREENING RESULTS a RANKING PANEL

• REVIEWNALIDATE INPUTS

" IDENTIFY INFO NEEDS

" CONSENSUS PROBABILITY AND CONSEQUENCES ULw/MediumfHigh)

The FMECA expert panel that followed this flow chart process was a multi-disciplinary team assembled to consider failure modes, effects and consequences. Some individuals from the screening panel were also included in the expert panel. However, reactor internals design and analysis was only one of the many disciplines represented on the FMECA panel. The entire FMECA panel was briefed on the screening process as part of their training and data generated by the screening panel was also made available to the FMECA panel. The FMECA panel was asked to review and confirm the results of the screening process.

To supplement the screening process, panels of experienced stress analysts and designers were assembled to augment the available information with conservative estimates of the required values. Separate panels were created for the Westinghouse and CE reactor designs. Each panel was comprised of individuals familiar with the original design basis for the internals and the range of design variants. The screening panels were instructed to generate generic answers to six basic questions:

1. Could the operating stress be >30 ksi?

2. Where is the component located relative to the core?

3. Is there potential for wear?

15

RAI Set 4 - Final Responses: 10/29/2010

4. Could the Cumulative Fatigue Usage Factor (CUF) be >0.25 @ 40 years?

5. Does it contain a structural weld?

6. Is the component bolted or is it a spring?

The questions were structured to conservatively assure compliance with the requirements of the MRP- 175 screening process. Whenever there was a potential for a component to exceed the stated limit, the panel answered the question in the affirmative. The role of the screening panel was limited to generating input data for the screening process.

RAI 4-3(b) also asked about any substantive differences in the two FMECA processes.

The major difference between the two approaches is that the screening and categorization was documented in a separate report (MRP-189) for the B&W plants, while that portion of the work was combined into a single report (along with the FMECA results) in MRP-191 for both CE and Westinghouse plants. The style of reporting the FMECA results is quite different, as well. The summary tables in MRP-190, Appendix A (Tables A-I through A-4) give a wealth of information that was derived from the FMECA process, and might be considered to contain much greater detail than is provided in Tables 6-5 and 6-6 of MRP- 191. The latter tables summarize the results of the FMECA process for CE and Westinghouse internals without the detail of expert elicitation that can be found in Tables A-I through A-4 of MRP-190. Additional detail on the FMECA process deliberations of MRP-191 was provided in the presentations given on June 8, 2010, to the NRC staff, and further discussion is provided in the accompanying roadmap document.

However, at the summary level required for decision-making and incorporation into MVRP-227, any differences between the two approaches are not significant.

Should additional information on the expert panel deliberations that led to the FMECA results cited in Tables 6-5 and 6-6 of MRP- 191, it will be necessary to meet with some of the cognizant individuals involved in the Combustion Engineering/Westinghouse FMECA deliberations under conditions that protect the proprietary nature of the discussions that eventually led to the cited results.

RAI 4-3(c) asks about a statement in Section 6.2 of MRP-191 that "...no consideration was given to manufacturing errors, maintenance errors, installation errors, transport errors, or any other type of random or human errors." This statement is consistent with the purpose of MRP-227, which is the management of age-related degradation effects, with such effects as installation errors handled by other programmatic efforts, such as the plant corrective action program. This statement is also consistent with the guidance provided in Branch Technical Position RSLB- 1 (see pages A. 1-1 through A. 1-9 of NUREG- 1800, Revision 2).

RAI 4-3(d) asks about a statement in Section 6.2 of MRP-191 that "...FMECA does not serve well to identify multiple failures." This statement means that the FMECA process 16

RAI Set 4 - Final Responses: 10/29/2010 provided a limiting level of protection at the initiating event stage and thus did not address hypothetical multiple failures.

RAI 4-3(e) asks about a statement in Section 6.2 of MRP-191 that "...operability, reliability, and availability issues were also considered." The purpose of this statement was to point out that the expert elicitation team for the FMECA process included experts on internals component operability, reliability, and availability. So the answer is yes, the FMECA process includes explicit consideration of these issues by the experts, and those expert opinions were taken into consideration in the development of the inspection recommendations.

RAI 4-4: In addition to the information provided in response to Question 1, discuss the impact of the functionality analysis in terms of determining or modifying the final inspection requirements for components. Specifically, provide an overview of how the functionality analysis impacted the recommended component classifications (i.e., primary, expansion, existing programs, no additional measures) that were initially developed prior to conducting the functionality analysis. Also, for primary and expansion components indicate how the functionality analysis was used to determine the type of inspection and the inspection periodicity.

Finally, identify all similar Westinghouse, CE, and B&W components (i.e., those that perform a similar function and have similar failure consequences) that have different final inspection classification and requirements based on the functionality analysis. Provide justification for any differences that exist.

Response: A separate roadmap document has been developed to augment the RAI responses and is included as part of the overall response package as Appendix B. It describes the eight-step process from component identification through assignment of inspection recommendations. It also points to the MRP documents which contain the details of the evaluations performed in the development of MRP-227. The roadmap does contain examples of how components were screened in or out, classified and re-evaluated.

For the specific overview requested in this RAI, refer to steps 5, 6 and 7 of the roadmap and the MRP document text references identified in those steps.

Components, whether they perform similar functions or not, or have similar failure consequences were evaluated on their own based on multiple criteria. A decision was then made on where it should be binned for inspection. The following examples shown below depict similar components with different classifications..

1. Upper core support barrel flange weld (W & CE) vs. CSS upper flange weld (B&W) 17

RAI Set 4 - Final Responses: 10/29/2010 The upper core support barrel flange weld in W and CE units is "Primary" due to SCC. The equivalent weld location in B&W units is the core support shield (CSS) upper flange weld, which is Category "A" and does not require augmented aging management.

The CSS upper flange weld in B&W units was screened out of all aging degradation mechanisms including SCC as documented in MRP-189 Rev. 1, Tables 3-2 and 3-3 with the FMECA identifier "S.2". This weld is a double U-groove weld made using an automatic submerged arc welding process with Type 308L weld metal. The MRP-175 SCC screening criteria for austenitic stainless welds are 30 ksi and <5% ferrite.

Due to the low applied stress on the CSS upper flange and specified minimum 5%

ferrite for austenitic stainless welds by B&W, this weld location was screened out of SCC for the B&W units. In addition, the CSS cylinder received post-weld stress relief treatment after the top and bottom CSS flanges were welded in the shop during the fabrication.

The upper core support barrel flange weld in W and CE units is identified in MRP-232, Table 2-18 (W) and Table 2-19 (CE) with a initial category "B" or "C", and the final group "Primary".

2. CRGT guide cards (W) vs. CRGT guide tubes and sectors (B&W)

The CRGT guide plates (cards) in W units are "Primary" due to wear. The equivalent components in B&W units are the CRGT guide tubes and guide sectors, which is in the "No Additional Measure" category.

The CRGT guide tubes (C-tubes) and guide sectors (split-tubes) in B&W units were initially categorized as "Not-A" for Wear, and were placed in Category "B" after the FMECA as documented in MRP- 190 and MRP- 189 Rev. 1 Table 4-1. Afterwards, wear of the CRGT guide tubes and sectors was further evaluated by reviewing past wear investigations to the control rods within the guide path as documented in MRP-231 Section 2.3. It was concluded that there was no evidence of wear on the control rod, and thus there should not be any wear on the CRGT guide tubes and guide sectors. Therefore, the CRGT guide tubes and sectors were downgraded to the "No Additional Measure" category.

The designs of the CRGT assembly are quite different for B&W and W units. The guidance for the control rods in B&W units is continuous with full-length guide tubes and sectors. In W units, approximately 1/4 of the CRGT assembly length guidance near the bottom is provided using continuous guide tubes (called C-tubes and sheaths in MRP-191 and MRP-232). The guidance for the remaining CRGT assembly length is discontinuous using guide plates (cards). For W-units, the C tubes and sheaths were initially categorized as "C" in MRP-190 and "Primary" in MRP-232 Table 2-6 for 18

RAI Set 4- Final Responses: 10/29/2010 wear. However, they were finally placed in the "No Additional Measures" category in MRP-232 Table 2-19. Note 2 to MRP-232 Table 2-19 explained that the C-tubes and sheaths were placed in the "No Additional Measures" because decisions on remediation of wear and degradation in the CRGT assembly will be based only on the conditions detected in the "Primary" the guide tubes (cards).

RAI 4-5: In addition to the information provided in response to Questions 1 and 3, discuss in general how susceptibility to multiple degradation mechanisms was considered when developing the final inspection recommendations. The final inspection category recommendations for each component appear to typically be based on the susceptibility and consequences associated with the single most-dominant degradation mechanism. However, many components are subject to multiple degradation mechanisms and it is not clear how the synergistic effect of multiple degradation mechanisms was considered in the final recommendations. The concern is that a component subjected to multiple degradation mechanisms may be more likely to experience a greater level of total degradation than a component that is subject to a single mechanism (even though the component with the single mechanism may be more susceptible to that mechanism).

Also discuss the acceptability of the recommended inspection method for primary components that are susceptible to multiple degradation mechanisms. Demonstrate that the recommended method is capable of identifying degradation due to all significant contributing mechanisms (and not just the single most-dominant mechanism) before component design margins are exceeded.

Response: Susceptibility to multiple degradation mechanisms was considered during the initial screening process, as documented in MRP- 189 for B&W internals components and in MRP-191 for both Combustion Engineering and Westinghouse internals components.

All internals components were evaluated against the screening criteria for each of the eight identified degradation mechanisms separately. However, in some cases, combinations of degradation mechanisms or effects were known to interact, or had the potential for interaction. The most common of these was the case of irradiation stress relaxation/creep combined with wear (loss of material due to wear) and/or fatigue (cracking due to fatigue) for some bolting. Examples of this are shown in Table 5-1 of MRP- 191, where Wear (I) and Fatigue (I) are identified for upper column assembly bolting. The parenthetical (I) is intended to illustrate the effect of irradiation that can lead to potential loss of preload from irradiation-induced stress relaxation/creep, which in turn has the potential to cause loss of material due to wear and/or cracking due to fatigue.

While not directly combined, many components were found to be potentially affected by moderate or significant degradation from more than a single mechanism. Examples of multiple degradation mechanisms with moderate or significant effects can be found in the tables in Section 3 of MRP-227.

The results of the screening process were revisited during the expert elicitation FMECA process, which required a confirmation by the experts that the screening results, including 19

RAI Set 4 - Final Responses: 10/29/2010 those cases of combined effects, were not contradicted by experience. In addition, the FMECA process included consideration by the experts of all of the degradation mechanisms for which moderate or significant effects were suspected, and these experts were capable of evaluating the potential for combinations of degradation mechanisms to cause more harmful effects than might be caused by individual degradation mechanisms.

Therefore, for many components subject to more than one degradation mechanism with moderate or significant effects, the final inspection category recommendation reflected the need for the inspection to detect an effect common to more than one degradation mechanism (e.g., cracking caused by IASCC and fatigue). And, in a few cases, the final inspection category recommendation reflected the need for an inspection capable of detecting more than one effect during the same examination (e.g., distortion caused by void swelling; gross cracking and material separation caused by IASCC). For this particular case, a visual (VT-3) examination encompasses the relevant conditions that describe both distortion caused by void swelling and material separation caused by gross cracking.

In summary, potential susceptibility to the effects from multiple degradation mechanisms was considered by: (1) identifying such combinations during the initial screening based on known interactions (e.g., irradiation-induced stress relaxation of bolt pre-load combined with either wear or fatigue); (2) FMECA expert elicitation of combined effects that resulted in greater consequences; and (3) recommending examinations capable of detecting relevant conditions caused by more than one degradation mechanism or effect.

RAI 4-6: Several previous RAIs (e.g., RAls 11 and 18 (Set #2), and RAI 3-8) have questioned whether plant-specific analyses are required to demonstrate that each plant is appropriately represented by MRP-227 such that the proposed aging management programs (AMPs) are applicable. That is, confirmation that the plant's initial design and operating conditions fall within the scope of the MRP-227 evaluation, the plant complies with important assumptions underlying the MRP-227 analysis, and changes in plant design or operating conditions (e.g.,

resulting from power uprates) have been appropriately considered. Meeting these conditions is necessary to ensure that the plant-specific AMP inspection requirements (i.e., the primary inspection components, inspection type and periodicity) would not be different from the MRP-227 recommendations determined through more generic evaluation.

The responses to these various RAIs have indicated that a plant-specific analysis to demonstrate the applicability of MRP-227 guidance is not required because plant-specific differences have been considered by: (a) evaluating operating experience throughout the commercial fleet; (b) using a conservative "out-in" core loading pattern in the functionality analysis; and, (c) assessing several known plant-specific conditions in the FMECA. The responses also justify the representativeness of MRP-227 because (a) base load operational profiles (i.e., fixed power 20

RAI Set 4- Final Responses: 10/29/2010 levels) are similar among plants, and (b) no design changes have been enacted by plants other than those identified in generic industry guidance or recommended by the original nuclear steam system supply vendors. However, given the variability in design and operational conditions that currently exists in PWR plants, the staff is not convinced that that the MRP-227 AMP requirements are necessarily appropriate for each plant. For instance, it is not clear that the "out-in" core loading pattern is conservative given that some degradation mechanisms do not initiate until low-leakage core conditions are imposed in the functionality model.

Therefore, the staff requests that guidance be developed that will allow individual licensees to assess the applicability of the MRP-227 method and results. This guidance should particularly focus on demonstrating the applicability of (a) the FMECA and functionality assessments, and (b) the recommended inspection category, inspection method and periodicity for each component. Specifically, this guidance should allow a licensee to determine if plant-specific differences in the RVI design or operating conditions (i.e., power uprate level) result in different component inspection categories (i.e., primary, expansion, existing, and no additional measures) than recommended within MRP-227. Alternatively, additional analysis or justification may be provided to demonstrate that the MRP-227 approach and results are generically applicable such that plant-specific differences in the RVI design or operating conditions do not result in different component inspection categories then recommended within MRP-227.

In that absence of adequate guidance, the staff will consider the need to implement limitations and conditions on the use of MRP-227 which would address plant-specific action items necessary to address this issue for each facility.

Response

The starting point for the response to this RAI is from Section 2.4 (Guidelines Applicability) of MRP-227, which have been cited in previous RAI responses - The last two paragraphs state that:

"These assumptions are a conservativerepresentationof U.S. PWR operatingplants, all of which implemented low leakage core loadingpatterns early in operatinglife. The recommendationsare thus applicable to all U.S. PWR operatingplants as of May 2007 for the three designs identified. These guidelines are also consideredapplicable to plants that have replacedcomponents or component assemblies; however, alternatives can be technicallyjustified.

Plant modifications made or consideredafter this date should be reviewed to assess impacts on strategies containedin these guidelines."

21

RAI Set 4 - Final Responses: 10/29/2010 These two paragraphs are based on the review and assessment by vendors that: (1) even though power uprates were not specifically addressed in the representative plant component functionality analyses, all plant uprates and other plant modifications up until May 2007 were considered to be within the envelope of the representative plant analysis results; and (2) no inspection recommendation cited in MRP-231 and MRP-232 would have been altered by a change in the functionality assumption of an earlier conversion over to a low-leakage core loading pattern. The first of these vendor findings is covered by the last paragraph, which clearly states the action required by a plant that has sought a power uprate or has undergone a significant plant modification as of May 2007. No further guidance is needed on the power uprate or major plant modification issue. The second of the vendor findings is not intended to argue that degradation mechanisms only initiate during high-leakage core loading operation, or that degradation effects cannot worsen during low-leakage core loading operation. The finding is simply that the vendors have reviewed the functionality analysis results and have determined that the recommended inspection requirements would not be altered by a change in functionality analysis assumption to an earlier conversion from high-leakage to low-leakage core loading. This core loading assumption only has relevance for those components which were aged and assessed using the detailed irradiation analysis finite element analysis (FEA) model. The aging analyses were conducted to understand the complex interactions between active degradation mechanisms in highly irradiated components.

These detailed modeling efforts were applied to the B&W and Westinghouse baffle-former-barrel structures, and a welded CE core shroud assembly. The intent of the irradiation aging analysis was to identify trends and limits in the component behavior.

The analysis was used to identify factors that could potentially cause component failure.

Representative plant designs with relatively severe irradiation conditions were selected for the irradiated aging analysis. These conditions were chosen to test the capability of the structure and identify points of potential concern. While the results of the FEA work provided insights into where degradation would most be expected, neither the vendors nor other members of the core writing team pinpointed the recommended examination scope solely based on the results of irradiation aging analysis. Instead, the irradiation aging analysis results were combined with engineering judgment and experience to provide examination recommendations. The only limited scope recommendations confirmed by the FEA results were for the CE welded designs where the most highly stressed and irradiated weld seams are specified. Thus, while another damage mechanism could play a more important role in the overall aging of the components when a more realistic core loading history is employed, in no case would the recommendations to detect that degradation change because:

• the anticipated effect and the overall degradation hierarchy would not change and

• no component or component assemblies have inspection requirements directed at local effects predicted in the FEA results to the extent that a shift in degradation 22

RAI Set 4- Final Responses: 10/29/2010 mechanism predominance would necessitate a change in location recommendations.

An excellent example of this is provided by baffle-to-former bolts in B&W and Westinghouse plant designs, where the effects of irradiation-induced stress relaxation of bolt pre-load has been observed to reduce the potential susceptibility to IASCC for the baffle-to-former bolt with the highest radiation exposure, while a baffle-to-former bolt with somewhat lower radiation exposure (somewhat further from the core) would be more susceptible. This shift of susceptibility to baffle-to-former bolts further from the core does not lead to a condition where core barrel-to-former bolts are more susceptible to IASCC than baffle-to-former bolts and, since the examination recommendation is for examination of 100% of the accessible baffle-to-former bolts, no change in the examination recommendation is warranted. Therefore, the additional level of detail provided by the functionality analysis does identify complex structural interactions, particularly in bolted assemblies, but did not lead to recommendations for changing the scope of examinations.

While the MRP agrees with the staff that a wide variety of designs are addressed by the representative plants particularly the Westinghouse and CE designs, the plant designs selected do correspond more closely with plants with earlier implementation dates for the MRP-227 requirements.

In addition, as plants begin the implementation process for Issue Program (IP) guidance, such as implementation of MRP-227 guidance, the responsibility for reviewing and determining the applicability of the explicit assumptions given in Section 2.4 are well understood, as outlined in NEI 03-08, including either the need or the wish to use valid alternatives through the deviation process. Thus it would be a plant-specific action to confirm the guidance in MRP-227 remains applicable within the boundaries indicated in Section 2.4. The general framework for both the determination of applicability and the process for justifying deviations is described in industry documents and is further discussed in the response to RAI 4-7.

RAI 4-7: Provide guidance on the process that should be followed by licensees if the plant-specific application of the MRP-227 guidelines identifies that inspection or aging management of a primary component (i.e., as defined in MRP-227) is not necessary. The guidance should address, for example, the plant-specific criteria and process for reclassifying the aging management program for a primary component, disposition of linked expansion components, and identification of (an) alternative plant-specific primary component(s) to be used in lieu of the generic MRP-227 recommendation for that degradation mechanism.

23

RAI Set 4 - Final Responses: 10/29/2010 The response to this question should specifically propose text that would be added to the "-A" version of MRP-227 to address this issue.

Response: The plant-specific guidance for implementation of the aging management program elements documented in Tables 4-1 through 4-9 of MRP-227 have been determined to be "Needed" under the Materials Initiative (NEI-03-08). Since MRP-227 guidance has been issued generically for the fleet it is anticipated that utilities may not be able to wholly meet the guidance or may choose to take alternative actions that better correspond with their long term aging management plans, e.g. replacement. Any MRP utility member or contractor personnel to an MRP utility may submit an inquiry to the MRP Advisory Panel of the guidelines for interpretation of part of these guidelines' requirements. In the case of MRP-227, the Advisory Panel core members who serve as the advisory panel on inquiries for MRP-227 were chosen to be the same as the utility members from the guidelines core group who wrote MRP-227-Rev. 0. According to the MRP administrative procedures, the responsibilities of the Advisory Panel are to 1) provide responses to inquiries on the meaning of current MRP Guideline Documents and how they should best be interpreted in light of NEI 03-08 and its other referenced documents; it is not the purpose of the Advisory Panel process to develop new guideline requirements, but to interpret, when necessary, existing requirements documented in the guidelines; and 2) offer recommendations to the guidelines committees regarding revisions to specific guidelines reflecting issues that emerge from the Advisory Panel process. The typical Advisory Panel consideration of an inquiry consists of a telephone conference call or meeting. Advisory Panel decisions are determined by vote, based on a simple majority of the members. Typically, the utility member who filed the inquiry is invited to listen to the Advisory Panel's discussions and respond to questions as needed.

The example of the Inquiry Format (submission via email) is provided below.

Advisory Panel-MRP-### - Topic of Inquiry.

Subject:

Point of contact familiar with the reason and basis for inquiry. Name/Phone:

Plant/Utility: Plant Name Unit #/Corporate Name Section: Section(s) of applicable Guideline

Background:

Plant conditions or other circumstances relevant to the inquiry.

Inquiry: Please prepare all statements in a condensed and precise question format.

Where appropriate compose inquiry in such a way that "yes" or "no" (perhaps with provisions) would be an appropriate reply.

Proposed Reply: State what it is believed is the intent of the guidelines.

An example of an inquiry to MRP-227-Rev. 0 is provided below:

24

RAI Set 4- Final Responses: 10/29/2010 Inquiry: For plants that do not have a flange of any kind at the location specified in Figure 4-21 of MRP-227, between the continuously guided region and the rest of the lower guide tube, is the intent that we be exempt from this examination category?

Proposed Reply: No, the intent is to examine a sample of locations with high residual and operating stresses to determine that SCC and fatigue are being adequately managed for the internals as a whole. If there is no similar location in a given reactor, then a deviation report with technical bases should be submitted to the MRP.

(Note that the reply to this inquiry was approved by the MRP.)

When a licensee determines that a needed requirement, is not needed, no longer applicable or cannot be accomplished, requirements are already provided in the Implementation Protocol of NEI 03-08. This guidance stipulates that, when a "Mandatory" or "Needed" work product element will not be fully implemented or will not be implemented in a manner consistent with its intent, a technical justification for deviation shall be developed and retained with the utility's program documentation or owner-controlled tracking systems. In addition, deviations from "Mandatory" and "Needed" work product elements shall be entered into corrective action programs (CAP).

The technical justification shall provide the basis for determining that the proposed deviation meets the same objective, or level of conservatism exhibited by the original work product, and shall clearly state how long the deviation will be in effect. In the context of MRP-227, if the guidance contained in Table 4-1 through 4-9 can not, need not, or will not be implemented the technical justification for the deviation should clearly state what requirement can not, need not, or will not be met and why; what alternative action is being taken to satisfy the objective or intent of the guidance; and, why the alternative action is acceptable. Examples of alternatives that may be justifiable are:

elevation of an Expansion component to Primary; substitution of an equivalent or more rigorous examination than is required by the tables; or destructive testing in lieu of non-destructive examination, such as the case where one or more of the primary components is being replaced. Since the Expansion components are also "needed" requirements, any deviation that would not fully implement a Primary component examination or not implement it in a manner consistent with its intent, would be expected to include in the justification a disposition of associated Expansion components.

Justification for deviations from work products or elements shall be reviewed and approved in accordance with the applicable plant procedures with the additional responsibility for deviation from a 'Needed' element that an internal independent review is performed and that concurrence is obtained from the responsible utility executive. For a deviation from a Mandatory element, an external independent review is performed in addition to the internal independent review and the concurrence from the responsible 25

RAI Set 4- Final Responses: 10/29/2010 utility executive. Further, as stipulated in the Implementation Protocol of NEI 03-08 a utility is required to notify the Issue Program (e.g., the MRP) and the NRC.

Although not requested in the RAI, two important steps in the NEI process regarding deviations rest upon the Issue Program (e.g. MRP, BWRVIP, SGMP) that developed the guidance. When a deviation is received, the Issue Program is responsible for reviewing the deviation for technical adequacy, providing feedback to the utility, and assessing the deviation for potential generic applicability and/or need to modify the requirements.

Modification to guidance documents can be accomplished via revision or, if warranted, interim guidance can be issued.

This guidance, plus the information contained in MRP-227, its supporting documentation, and in the MRP administrative procedures, is considered sufficient to address the issue raised in the RAI. However, to provide clarity, a paragraph will be added with similar wording to the response above to Section 7.1 of MRP-227 providing a direct reference to NEI 03-08, Implementation Protocol and the deviation justification process (see Appendix A to these RAI responses).

RAI 4-8: This question discusses accessibility requirements for primary and expansion components. Define the appropriate inspection coverage to ensure the component being inspected does not lose its intended function and the process to be followed if the inspection does not meet the inspection coverage. Provide additional guidance on the component accessibility requirements for each primary and expansion component (i.e., those in MRP-227 Tables 4-1 through 4-6, 4-8 and 4-9) such that the results of the inspection can be credited as satisfying the requirements of the aging management program. This guidance should include, at a minimum, the following considerations:

a) For each component, identify the location(s) where degradation is expected.

b) Define the appropriate inspection coverage at this location to ensure that enough of the surrounding material is inspected such that there is assurance that the degradation will be identified before it challenges component or system integrity (i.e., the intended function of the component is retained).

c) Describe the procedure that a licensee should follow if inspection accessibility is insufficient to provide the required inspection coverage or if the inspection does not meet other minimum requirements as specified in MRP-227 and MRP-228.

This procedure must address providing an appropriate justification for continued operation with the reduced examination requirements to the NRC for review and approval. The guidance should address the process for adjusting the inspection area and/or coverage interval for both 26

RAI Set 4- Final Responses: 10/29/2010 welded and non-welded components as a function of the component being inspected and/or the degradation mechanism being assessed during plant-specific inspections.

With respect to inaccessible components, the MRP should:

a) Identify any components that are; (1) totally inaccessible (can't be inspected) and (2) the management of their aging effects is dependent on the inspection results from another primary, expansion, or an existing component.

b) For the components identified in "a)," identify the primary, expansion, or other existing components that are the surrogate for the inaccessible components and explain why the surrogates are the limiting components for the aging effects that need management.

c) For the totally inaccessible components or for the inaccessible portion of primary or expansion components, what are the requirements that the licensee must follow to ensure that the components do not lose their intended function as a result of flaws in the accessible components.

The response to this question should specifically propose text that would be added to the "-A" version of MRP-227 to address this issue.

Response: This RAI requests additional information on the description of inspection coverage provided in Tables 4-1 through 4-6 of MRP-227. The RAI also requests additional information on inspection coverage for Tables 4-8 and 4-9 of MRP-227; however, inspection coverage for Existing Program components listed in Tables 4-8 and 4-9 of MRP-227 is covered by the Existing Program requirements - such as the inspection coverage requirements for PWR core support structures listed in Tables 4-8 and 4-9 that are subject to ASME Code Section XI Examination Category B-N-3 visual inspections. Therefore, the response to this RAI will address only the inspection coverage requirements contained in Column 6 of Tables 4-1 through 4-6 of MRP-227.

The RAI requests very specific information to be addressed in Column 6, with the first item as "identify the location(s) where degradation is expected," with the second item as "define the appropriate inspection coverage at this location to ensure that enough of the surrounding material is inspected .... ".

Because the RAI deals with the multiple facets of accessibility and coverage, the response is broken into multiple parts.

In response to the question concerning assurance that the sufficient surrounding material is inspected, the response is that all welds inspected for SCC or IASCC must be examined along with the weld heat affected zone (HAZ). This requirement is defined in 27

RAI Set 4- Final Responses: 10/29/2010 MRP-228, which states in 2.3.6.4(a): "For welds, the area of interest is generally considered the entire width of the weld and 3/4 inch of the adjacent base material on each side of the weld." This is consistent with the BWRVIP-03 "area of interest" standards for IGSCC in stainless steel welds in BWR internals.

The RAI also requests more information concerning other accessibility issues. There are several inspection strategies employed in MRP-227 to deal with accessibility and other issues that help ensure an adequate inspection sample. For the purpose of this RAI response, these are broken into the following categories:

1. Selecting 100% of accessible surfaces of a continuous structural weld,

2. Selecting 100% of accessible bolting in a bolted assembly,

3. Selecting 100% of accessible surfaces of a set of components or component items,

4. Selecting a focused examination where analysis clearly indicates which portion of the component, assembly or structure is most susceptible,

5. Selecting a sampling approach where a only portion of the component or component item is accessible without disassembly,

6. Selecting the Primary over the Expansion component where access to the Expansion component is severely limited,

7. Physical measurements for core clamping functionality.

The first situation involves those components where the examination requirement is to examine 100% of accessible surfaces of a continuous structural weld employing visual EVT- I techniques. This requirement applies to the core barrel upper flange welds for both CE and Westinghouse designs. These welds are relatively accessible with no known significant inspection restrictions. In addition, the welds are not particularly susceptible relative to comparable BWR core shroud welds. Thus, this examination is expected to be confirmation of lack of significant degradation. However, since it is important to obtain a large sample and to ensure unanticipated accessibility restrictions are properly addressed, a minimum coverage requirement is proposed to be added to Tables 4-2 and 4-3 for these components and to be included in the '-A' version of MRP-227 (See Appendix A Point 3 of these RAI responses). For completeness, this proposed minimum requirement is also included in this RAI response:

"A minimum of 75% of the total weld length (examined + unexamined) including coverage consistent with the Expansion criteria in Table 5-2 (or Table 5-3), must be examined from either the inner or the outer diameter for inspection credit".

As indicated above, this minimum coverage is justified by the lack of negative examination results from previous B-N-3 examinations, the anticipated low susceptibility 28

RAI Set 4- Final Responses: 10/29/2010 of the core barrel welds, and an aggressive sampling approach that is considered adequate to demonstrate the continued absence of degradation.

The second situation involves the requirement to examine 100% of the accessible bolting in a bolted assembly. This applies to:

" Upper core barrel bolts and locking devices,

" Lower core barrel bolts and locking devices,

" Baffle-to-former and internal baffle-to-baffle bolt locking devices, for the B&W designs,

• Core shroud bolts for CE designs, and

" Baffle former bolts, and

" Baffle edge bolts for Westinghouse designs.

For this situation, the applicable bolted assemblies were included in the irradiation modeling efforts that were a large part of the various functionality analyses. The results of these analyses provided trends and insights into the complex behavior of the structures, from which targeted examination recommendations could have been made. These results when combined with existing minimum bolting assessments and relative lack of negative experience could have justified a limited examination scope. However, it was recognized that, in order to account for uncertainties and obtain a large initial sample, a requirement of 100% of the accessible bolts was logical. This strategy serves three goals, 1) it is not limited to minimum compliance but seeks the best efforts of the examiner, 2) it diminishes the uncertainty in predictability of assembly, manufacturing uniformity, and

3) it reduces the potential for competing degradation effects to progress undetected.

Camera access to the bolt head and locking devices are, with minimal exceptions, without limitation. Thus, where visual VT-3 examination is required for bolt locking devices or general assessment of an undamaged condition, there is virtually no access limitation anticipated. The potential significant limitation on accessibility for this situation only occurs when volumetric examination (UT) is specified and a bolt head design or as-built condition limits access or effectiveness for the UT transducer. Since this is a potential severe limitation, considering the industry's limited examination experience across the entire variety of design variations, the MRP is proposing a minimum coverage requirement to be added to Tables 4-1, 4-2 and 4-3 and to be included in the '-A' version of MRP-227 to further assure that potential limitations on access and examination coverage will be adequately addressed.(See Appendix A Point 3 of these RAI responses).

For completeness, this proposed minimum requirement is also included in this RAI response:

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RAI Set 4 - Final Responses: 10/29/2010 "A minimum of 75% of the total population (examined + unexamined) including coverage consistent with the Expansion criteria in Table 5-1 (or Table 5-2, or Table 5-3) must be examined for inspection credit."

This minimum coverage requirement means that expansion of coverage when degradation is detected at that specific location is not an issue; the examination of 100%

of the accessible surfaces or items will provide the necessary evidence to show whether the degradation is localized or widespread directly, without the need for coverage expansion. Also, the ability to expand from 100% of accessible surfaces or items for a Primary component to 100% of accessible surfaces or items for an Expansion component provides additional evidence about the limits of detected degradation. Throughout the process of determining the final recommendations for examination coverage, access to the Primary components has not been an industry concern and thus the coverage requirements were deemed adequate for the aforementioned reasons. However, given the potential for more severe than anticipated limitations due to the industry's limited examination experience across the entire variety of design variations, the MRP is proposing the above minimum coverage requirements to Table 4-1, 4-2 and 4-3 so that further assurance that potential limitations on access and examination coverage will be adequately addressed.

For both of the above situations, if a minimum coverage is not satisfied, the intent of MRP-227 is not met and, as discussed in RAI response 4-7, a deviation must be prepared and the staff and the MRP notified of the inability to meet the "Needed" requirement for coverage. This requirement already applies where specific recommendations are made for other Primary components and no further changes to MRP-227 need be made as discussed for the remaining situations discussed below.

The third situation involves those components where 100% of accessible surfaces of a set of components or component items is required. This applies to:

• CSS cast outlet nozzles,

• CSS vent valve top retaining ring,

• CSS vent valve bottom retaining ring,

• Lower grid assembly dowel-to-guide block welds,

• IMI guide tube spiders, and

• IMI guide tube spider-to-lower grid rib section welds and is applicable only to B&W Primary components, The strategy employed for these component items is to specify that the accessible surfaces of all of the components in the population be examined to the extent that it would not require disassembly. All of the components are accessible, although some portions of the visible surfaces will be obstructed. In all of these cases, the required VT-3 visual examination is looking for gross degradation, such as separation of material, 30

RAI Set 4 - Final Responses: 10/29/2010 broken or missing locking welds/devices, etc. Thus, no minimum coverage requirement is deemed necessary.

The fourth situation occurs where the scope of the component or the component assembly selected surface focuses on where degradation effects would be expected. The following components were selected based on insights from the materials modeling/aging analysis performed on the core barrel regions:

" CE (welded two sections) - Visual examination of shroud plate-to-former plate welds at the re-entrant core shroud mid-plane to detect cracking (IASCC).

" CE (welded full height) - Visual examination of shroud plate core-side welds at the re-entrant core shroud mid-plane to detect cracking (IASCC).

• CE (welded two sections) - Visual examination of upper and lower section flange interfaces to detect distortion.

" CE (bolted) & Westinghouse - Assembly level visual examination to detect distortion.

For these components, the areas selected consist of the locations where the highest predicted void swelling would manifest itself as gaps and/or displacements at joint interface locations or where the highest irradiation/stress combination would most likely produce cracking due to lASCC. Thus, no minimum coverage requirement is deemed necessary.

The fifth situation occurs where the recommendations involve a sampling strategy employed specifically due to access limitations. One of the basic assumptions is that disassembly of reactor internals would be avoided unless warranted. This sampling examination strategy has been specified for:

" B&W - visual examination around baffle plate bolts holes ("100% of the accessible surface within 1 inch around each flow and bolt hole"). This recommendation provides for examination of an adequate amount of surface area with the bolt holes selected as the most likely point of initiation due to the bolt hole and locking weld acting as a stress riser.

" CE - visual examination of deep beam-to-beam welds axially from top surface to four inches below. This results in a sample of the top four inches of 100% of the deep beam intersections which equates to approximately 10-15% of the available weld length. However, the sample size selected provides reasonable confidence in detecting the extent of potential degradation effects as it encompasses the most highly irradiated portion of the component.

" CE - visual examination of outer peripheral guide tubes of the control element assembly instrument guide tubes. This recommended scope can be accomplished without by unnecessary disassembly with an adequate sample size selected to provide a reasonable level of confidence in detecting the extent of potential 31

RAI Set 4 - Final Responses: 10/29/2010 degradation effects. The sample size is expected to include roughly 15-20% of the population.

" Westinghouse - visual examination of all CRGT guide cards within a 20% CRGT sample. This recommendation avoids unnecessary disassembly with an adequate sample size selected to provide reasonable confidence in detecting the extent of potential degradation effects.

" Westinghouse - visual examination of the outer CRGT lower flange welds. This recommendation avoids unnecessary disassembly with an adequate sample size selected to provide a reasonable level of confidence in detecting the extent of potential degradation effects. The sample size is expected to include roughly 15-20% of the population. All of the guide tube weld locations are equal in their relative susceptibility.

The sixth situation involves choosing as Primary the accessible component over an inaccessible Expansion component. These three cases always elevate a component of equal or greater susceptibility that is accessible over the inaccessible Expansion components. In no instance is a more susceptible component placed in the Expansion pool due to inaccessibility. This, coupled with the conservative threshold for requiring the evaluation of the inaccessible Expansion components, provides reasonable assurance that the appropriate evaluation and/or corrective actions will be taken well before significant degradation occurs.

The seventh situation involves physical measurements to demonstrate acceptable core clamping functionality for the B&W and the Westinghouse designs. These components are accessible to perform these measurements.

In addition, plant-specific fatigue evaluations are to be employed for 3 CE components that may have some access limitations. The results of this evaluation will determine what, if any examinations are recommended. See RAI 4-16 response.

Rules concerning the acceptance and disposition of flaws in a component where limited examination coverage is or can be obtained are contained in WCAP-17096-NP. This document has been submitted for review and approval.

Finally it must be mentioned that one of the Primary component items has recently been determined to be inaccessible. As mentioned in our October 14, 2010, meeting with the staff, AREVA has been working closely with the B&W owners in preparation for their implementation of MRP-227. This effort has included record searches and more meticulous accessibility studies. The result has been that the CSS vent valve disc shaft (or hinge pins) cannot be seen without disassembly of the valve. The portions believed accessible are in fact covered by other journal bushing rings. This will result in a change to Table 4-1 of MRP-227 to evaluate the condition or replacement. As discussed in the B&W supporting documents, these disc shafts/hinge pins are subject to loss of fracture 32

RAI Set 4 - Final Responses: 10/29/2010 toughness due to thermal embrittlement. As indicated in the Table 4-1 footnotes, these valves are exercised each outage to ensure they are functional, i.e. that they will lift within the design limits. Additionally, failure during operation that results in valve displacement would be detected by operators as abnormal by-pass flow. These factors offset the concern with inaccessibility, but the MRP recognizes that a formal evaluation --

either generic for the B&W fleet or on a plant-specific basis -- should be completed or a replacement strategy recommended. Thus, in the interim, Table 4-1 will be changed to indicate that the CSS vent valve disc shafts (or hinge pins) are inaccessible and that an evaluation or a replacement campaign is required (See Appendix A Point 4 Change a of these RAI responses).

RAI 4-9: A number of components are identified as being covered by existing programs.

However, there is no summary of existing RVI programs provided in MRP-227 or supporting documentation. Add a summary of existing programs being credited to MRP-227. If an existing program is consistent with a program definition given in the staff's Generic Aging Lessons Learned (GALL) report, it is sufficient to simply identify the related GALL program definition.

For existing programs lacking such a convenient reference, a summary should be provided which describes the following program requirements for an acceptable existing program: (a) scope (i.e.,

components inspected/monitored), (b) the applicable inspection, monitoring, or testing requirements and acceptance criteria, (c) the periodicity of the program, and (d) any other relevant requirements. This summary should identify the degradation mechanism(s) that are intended to be monitored or mitigated by each existing program and provide justification that each program is sufficient to monitor and/or mitigate all the expected degradation mechanisms identified in MRP-227 for the applicable component(s).

The response to this question should specifically propose text that would be added to the "-A" version of MRP-227 to address this issue.

Response: First, as stated in MRP-227 (Section 4.4), no existing generic industry programs were considered sufficient for monitoring the aging effects addressed by these guidelines for B&W plants. Therefore, no components for B&W plants were placed into the Existing Programs group.

Second, Tables 4-8 and 4-9 of MRP-227 identify and provide a reference to all credited Existing Program components for Combustion Engineering and Westinghouse plants, respectively. All of the credited Existing Programs in both tables are covered by the visual (VT-3) examination requirements of Examination Category B-N-3 of the ASME Code Section XI, except for the flux thimble tubes in Westinghouse plants. For those components covered by ASME Section XI, GALL report Section XI.M1 provides an adequate summary of program requirements, and the ASME Section XI inspections specified in MRP-227 are consistent with GALL XI.M I program requirements. The 33

RAI Set 4 - Final Responses: 10/29/2010 Table 4-9 reference for flux thimble tubes in Westinghouse plants is IEB 88-09, which is based on individual licensee commitments. The MRP-227 recommendations are consistent with the requirements of the GALL XI.M37 Flux Thimble Tube Inspection program, and site-specific commitments are included within this program.

Not included in Table 4-8 are the references for Existing Programs for ICI thimble tubes and thermal shield positioning pins in CE plants. Again, the guidance for these components is limited to plant-specific recommendations that preclude the preparation of a generic summary of program elements for all plants. This is consistent with line item IV.B3.RP-357 in GALL Rev. 2 which calls for a plant specific aging management program for ICI thimble tubes. Plant owners will review their plant-specific commitments when developing individual plant aging management programs for these components. A similar situation applies to guide tube support pins (split pins) in Westinghouse plants. Plant-specific recommendations involve owner/operator replacement decisions and not an inspection program. As a result, the decision was made to exclude these components from Table 4-8 and 4-9 for the CE and Westinghouse designs respectively.

The staff may wish to continue to include the plant-specific disposition of the ICI thimble tubes, and thermal shield positioning pins for CE plants and the guide tube support pins as plant-specific actions for license renewal submittals.

RAI 4-10: MRP-227 guidance is used to develop component-level aging management programs and inspection requirements. Further, the development of these programs and requirements has not considered the effects of transitory design basis events (DBE) on the performance of degraded components or structures. However, as indicated in the response to RAI 2-16, the current industry expectation is that "...when age-related degradation effects are detected during the examinations specified in MRP-227, the suitability of the degraded component for continued service will necessarily take into consideration the full range of design basis event (DBE) effects." Therefore, staff believes that guidance and requirements should be provided to ensure that licensees perform consistent plant-specific evaluations of the effects of degraded components under both normal and transitory DBEs (i.e., normal, upset, emergency, and faulted loading conditions). These evaluations should provide reasonable assurance that the systems associated with the degraded components will maintain required design margins and that inspection, repair, and replacement requirements are both adequate and timely. The guidance and requirements should, in part:

a) Identify the number or percentage of related primary or expansion components that should be inspected and the allowable number of degraded or non-functioning components for each system to ensure acceptable performance under DBEs. Discuss the appropriateness of 34

RAI Set 4- Final Responses: 10/29/2010 developing generic versus plant-specific inspection requirements for each system.

Alternatively, the guidance should describe how the plant-specific analysis should be performed to determine the inspection sample and allowable number of degraded components for each system, and b) Describe the additional inspection requirements that should be triggered if degraded components are found as part of the primary inspection.

The consideration of item b) should provide guidance for increasing the sample size to inspect other similar components within the system that are subject to the observed degradation mechanism. It should either provide guidance for expanding the inspection to components in other systems that are subject to the same degradation mechanism or justify the adequacy of existing expansion criteria in MRP-227. As an example, this guidance should specify the percentage of baffle-to-former bolts that should be inspected and the percentage that may be degraded before system performance under design basis loading conditions is affected. If degraded baffle-to-former bolts are found, this guidance should next specify the additional baffle-to-former bolts that should be inspected and, for instance, the number of expansion baffle-to-baffle or core barrel-to-former bolts that should be inspected.

Response: This RAI expresses the staff s preference that "guidance and requirements should be provided to ensure that licensees perform consistent plant-specific evaluations of the effects of degraded components under both normal and transitory DBEs (i.e.,

normal, upset, emergency, and faulted loading conditions)." The MRP recognized the industry's need to have consistent methodologies for evaluating the results of inspection findings from implementation of MRP-227. As a result MRP requested that the PWROG work with the industry to provide guidance for component/component assembly acceptance and evaluation. WCAP-17096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2, December 2009, was developed and submitted to NRC for review with the intent that it provide "consistent, industry-wide analytical methodologies and data requirements for developing:

1. Acceptance Criteria for the Primary and Expansion Components identified in the Materials Reliability Program (MRP) Reactor Internals Inspection and Evaluation (I&E) Guidelines (MRP-227 Rev. 0)

2. Evaluation Procedures for utilities to assess potential safety and functional impacts of degradation in components with observed relevant conditions."

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RAI Set 4- Final Responses: 10/29/2010 MRP-227 does provide guidance on expanding the inspection sample size when degradation is detected in excess of the Expansion criteria in Section 5. That increased sample size is the main purpose behind the sampling strategy represented by the Primary components and the potential enlargement to include Expansion components. Both the Primary and Expansion component inspection coverage in MRP-227 specifies the percentage (100%) of accessible bolts that should be inspected.

WCAP- 17096 has been submitted to the staff for review and evaluation, and detailed information requests regarding evaluation of examination results can be addressed during its review. In order to ensure that consistent methodologies are employed by utilities in the evaluation of degradation the MRP will add a 'Needed' requirement to the approved version of MRP-227 (see Appendix A to these RAI responses).

It is the MRP's expectation that flaw evaluations performed in accordance with NRC-approved methodologies do not require transmittal to or approval by the NRC. However, under no circumstances would any requirements to report flaws or flaws evaluations contained in ASME Section XI be superseded. Flaw evaluations that do not meet NRC-approved guidance (e.g. assumptions, methods, etc.) are to be submitted for approval.

MRP-227 does provide guidance on expanding the inspection sample size when degradation is detected in excess of the Expansion criteria in Section 5. That increased sample size is the main purpose behind the sampling strategy represented by the Primary components and the potential enlargement to include Expansion components. Both the Primary and Expansion component inspection coverage in MRP-227 specifies the percentage (100%) of accessible bolts that should be inspected.

RAI 4-11: RAI 2-1 asked for justification of the inspection periodicity recommended in MRP-227 for reactor vessel internal components given that there is little operating experience for basing inspection periodicity and that analysis to evaluate the evolution of degradation in these components has a large degree of uncertainty. The response to that question primarily justifies the adequacy of the recommended inspection intervals based on the functionality analysis which predicts that degradation will gradually worsen over time and will not suddenly progress.

However, the inspection periodicity is not based, as is typically the case, on an evaluation of the maximum level of degradation that is acceptable for components to fulfill their intended design requirements, and the predicted time to reach this level of degradation based on the extent of degradation found during the inspection and evaluation of the rate of degradation with continuing operation. Therefore, the staff requests additional justification for recommended inspection intervals. This justification should address why the current MRP-227 approach is appropriate for determining inspection periodicity and that determining periodicity based on a 36

RAI Set 4- Final Responses: 10/29/2010 component level evaluation to ensure that the required component design margins are retained between inspections is not required.

Alternatively, provide information about the plans of licensees to perform the initial primary inspections required by MRP-227. Staff understands that some licensees are planning to inspect all required primary components during the first refueling outage after entering into the license renewal period. Staff therefore seeks to determine if this approach is being adopted by other licensees that have or will shortly enter the license renewal period. Clarify if this approach is either recommended or required within MRP-227.

Response: The response to RAI 2-1 (Set #2) described the results of the component degradations in the functionality analyses as very gradual, with no sudden increases during either the 30 years of simulated high-leakage core loading or the subsequent 30 years of low-leakage core loading. From this work, it was determined that: (1) inspections immediately prior to or immediately following the beginning of the extended period of operation would have similar likelihoods of detecting and characterizing any degradation present in the reactor internals Primary components, and (2) no basis for departure from the standard ten-year inspection intervals to more frequent inspection intervals was warranted. The latter determination was supported by both operating experience and the results of ASME Section XI Examination Category B-N-3 inspections conducted during the first 40 years of operation. In addition, since MRP-227 is a living document, which will be updated to reflect both positive and potentially negative information from inspection results obtained by a series of plants entering the period of extended operation over the next several years, the currently-specified inspection periodicity will be reviewed and altered, as needed, to incorporate that information. As currently anticipated, some five or six plants will be conducting all or portions of the inspections specified in MRP-227 in the next four years. Additionally, as can be seen graphically below, the US PWR fleet will have units entering their period of extended operation over intervals sufficient to ensure examination results are regularly being collected.

37

RAI Set 4- Final Responses: 10/29/2010 I Demand for MRP-227 SER PWRs Entering Period of Extended Operation Numberol Plants Entering PEO 9t 10 6

5 -.

4 tI U If 2 I 2M 7o W 20ni

• 7M J 7AMi 9 20[I4 =5li 2M6 Nn 3*dIMO;~t 2915 7WO2 M"*' 7"m" 20n* 20A1 2M 20 20VF# 20 U Even more to the point, the inspection results could result in the detection and sizing of defects that might require engineering evaluation for disposition. In such cases, the engineering evaluations will provide clearer evidence than is now available on the level of degradation that is acceptable, and subsequent inspections will provide information on the rate of increase in that degradation. This type of information will provide additional grounds for sustaining or modifying the current inspection periodicity. Considering the operating history of PWR internals in the United States, the results of ASME Code prescribed examinations on internals over the first 40 years of operation, and the conservative results from the functionality analyses, the current inspection periodicities are considered consistent with the intent of the 10CFR54 requirements. Further, on the same basis, they are considered adequate to identify potential degradation of the non-pressure boundary components associated with the reactor internals with the potential for either increasing or decreasing that periodicity as warranted by initial examination results.

Several plants have provided their plans to perform the initial primary inspections required by MRP-227 to the NRC. Some licensees are planning to inspect all required primary components during the first refueling outage after entering into the license renewal period. This approach is not a direct recommendation within MRP-227-Rev. 0, as the inspection Tables (Tables 4) in MRP-227 provide some flexibility so that a PWR 38

RAI Set 4- Final Responses: 10/29/2010 Unit does not necessarily have to inspect all of its MRP-227 Primary components during the same outage.

The following inspection plans were provided by owners depicting their current plans to perform the initial primary inspections required by MRP-227 to the MRP are shown below.

A Westinghouse 3-loop has provided the following current plan for information.

Inspection 2010 2011 2012 2013 2014 Baffle bolts UT (1) W 3 (1) W 3 loop loop B&W Core Barrel bolts UT Primary EVT-1 (2) W 3 loop Primary VT-3 or (2) W 3 loop measurements Primary VT-3 (1) W 3 loop (1) W 3 loop requiring CB removed 3 B&W Units have provided the following current plan for information. These units currently plan to perform the "Primary Components" inspection from MRP-227, Rev. 0 Table 4-1 during the fall 2012, fall 2013 and spring 2014 respectively, with the exception of two examinations already performed and possible changes from a deviation in progress. The two examinations that have been performed are the onetime physical measurement of the interference fit between the plenum cover weldment rib pads and RV flange, and the UT of the upper core barrel bolts. The timing of the examinations is given in the table below:

Physical Measurement Unit-1 Fall 2006 No change From Base Line during initial installation Unit-2 Fall 2008 No change From Base Line during initial installation Unit-3 Fall 2007 No change From Base Line during initial installation UT of Upper Core Barrel Bolts Unit-1 Spring 2008 100% coverage, no rejections Unit-2 Fall 2008 100% coverage, no rejections Unit-3 Fall 2007 100% coverage, Two bolts rejected, lack of back wall, same as previous inspections The deviation moves two "Primary" components (Vent Valve Disks and ONS-3 Cast Outlet Nozzles) to "No Additional Measures" due to ferrite content being below the screening criteria, and one "Expansion" component (CRGT Spacer Castings) being moved to "Primary" as a result of the other two castings being moved to "No Additional Measures".

39

RAI Set 4 - Final Responses: 10/29/2010 RAI 4-12: RAI-21 (Set #2) asked about the need to develop more definitive acceptance criteria for inspections to ensure uniform interpretation and implementation from plant to plant. The industry response indicated that more definitive criteria in MRP-227 is not needed because the inspectors will receive component-specific training and that any observable degradation will require further disposition through the corrective action process. However, staff remains concerned that this approach is not sufficient given the variability associated with inspection conditions and interpretation of inspection results. Therefore, staff requests that more definitive inspection acceptance criteria be developed for the VT-1, EVT- 1, UT, and VT-3 inspections for each of the primary and expansion components. These criteria should be a function of the required accuracy and precision of the particular technique and also the application of this technique to each particular component (i.e., accessibility limitations, expected degradation location, expected degradation type).

In the absence of adequate guidance, the staff will consider the need to implement limitations and conditions on the use of MRP-227 which would address plant-specific action items necessary to address this issue for each facility.

Response: In the response to RAI 2-21 (Set#2) the MRP described the inspection acceptance criteria contained in Section 5 of MRP-227 for VT-3, VT-1, EVT-1, and UT examinations, which are relatively simple; any detected relevant condition must be reported and placed in the plant corrective action program. Such simple inspection acceptance criteria will be interpreted and implemented consistently at different plants.

MRP-228 has been developed and issued to ensure exactly that.

The MRP agrees that establishing more definitive acceptance criteria leads to more uniform and consistent plant-to-plant implementation especially for signal-related examinations.

The MRP believes that adequate direction has been provided for the ultrasonic examination of bolting by requiring that the technical justification process from MRP-228 (based on ASME Section V Article 14) be followed. A substantial discussion of this process is included in Section 2 of MRP-228. This discussion addresses: (1) whether a formal Technical Justification is required or not (some examination methods are sufficiently standardized and codified that no addition technical justification is needed);

(2) the required accuracy and precision that must be demonstrated by the Technical Justification, including the potential for mock-ups that simulate the range of geometries and types of degradation needed; and (3) the personnel training that is planned to ensure some degree of uniformity in implementation and interpretation.

Definitive (numerical) acceptance criteria is provided for the physical measurement specified for B&W plants (height differential from top of plenum ribs pads to RPV 40

RAI Set 4- Final Responses: 10/29/2010 seating surface) due to uniformity of design characteristics. The physical measurement for Westinghouse plants is the spring height for those plants with Type 304 stainless steel hold down springs. While there is similarity in the basic design for the Westinghouse plants, the numerical acceptance criteria is a plant-specific input due to the variations in designs loads and is also plant-specific as-built dimensions. There are no direct physical measurements specified as Primary or Expansion for CE components.

The MRP believes that uniformity of inspection implementation can be accomplished without technical justifications for visual examinations based on the uniformity of existing industry requirements and Section XI of the ASME Boiler and Pressure Vessel Code (Section XI) examination experience. The MRP requirements for visual examinations have incorporated Section XI and BWR Vessel and Internals Project (BWRVIP) EVT-1 requirements with augmented guidance based on expected or potential degradation. The visual examination rules of Section XI are in IWA-2210 (2010 Edition). These rules are implemented routinely across the entire fleet virtually every outage with oversight by NRC and Authorized Nuclear Inservice Inspectors. Consistent implementation has been established. The BWRVIP developed the EVT-1 examination by utilizing the Section XI visual rules and augmented them where needed based on the type flaws to be expected. The essential elements of the BWRVIP developed EVT-1 will be used by MRP.

Over the last 16 years enhancements have been made to the EVT- 1 examinations based on lessons learned that were shared by all member utilities via the BWRVIP Inspection Focus Group. Additional insights were gained from INPO review visits where peers identified best practices. These lessons are reflected in current BWRVIP EVT- 1 guidance and have been utilized by the MRP. Also, there is current research and testing ongoing regarding visual examination methods. MRP is integrated into the research and will gain from insights that result. MRP will also share lessons learned via its Inspection ITG and through INPO best practices. Where appropriate, the visual examination guidance will be revised to reflect industry experience. Finally, the current, known inspection vendors have performed visual examinations per Section XI and the EVT-1 requirements for many years further ensuring consistent implementation of visual examinations when employed in implementation of MRP-227.

Providing explicit criteria for relevant conditions from visual examinations has the disadvantage of potentially biasing examiners and owners so they inadvertently disregard unanticipated conditions. Considering the scarcity negative OE for reactor internals to date and the fact that many of the Primary visual examinations will be the first instances where a component is inspected, it is important to get a good baseline. To do this and to anticipate the potential for these initial examinations to find the "unexpected", MRP required that ay relevant condition be reported. Explicit relevant conditions are specified when evaluating the need to expand from the Primary to Expansion 41

RAI Set 4 - Final Responses: 10/29/2010 components. Relevant conditions not specified in the Primary to Expansion linkage would not require expansion to comply with MRP-227; however all relevant conditions require reporting and disposition via a plant's corrective action program. Corrective action programs typically require that consideration of extent of condition be provided as part of dispositions. Aligning MRP-227examinations with well proven, well known and used ASME methods of examination and terminology is anticipated to provide more information reporting and subsequent industry scrutiny.

RAI 4-13: Provide a description of how international and US operating experience is (or is planned to be) documented, tracked, and updated so that it will support continued refinement of MRP-227 guidance and inform plant-specific aging management programs.

The staff believes that it would be advisable for documentation regarding US and international operating experience related to degradation in RVI components be compiled in a single document that could be used to support the process of updating MRP-227.

Response: The MRP proposes to use a new Appendix A of MRP-227, which currently contains a summary of US and relevant international operating experience related to degradation of RVI components to date (this was provided to the NRC by MRP Letter 2001-091 dated December 2, 2009; see Appendix A to these RAI responses). This summary was compiled during the preparation of the supporting documentation for MRP-227, and was generated with the intent to combine and place this information in a single location that would be periodically updated with additional operating experience as it becomes available. Section 7.6 of MRP-227 identifies a "Good Practice" recommendation that will help to assure the collection of this operating experience. The MRP recognizes the importance of documenting the operating experience to support the maturation of processes, procedures and requirements in the family of documents centered around MRP-227 that will implement the inspection and evaluation requirements. Thus the MRP proposes that the "Good Practice" recommendation be modified in Section 7.6 to elevate to the "Needed" requirement to provide inspection results. That recommendation would be changed to state that (see Appendix A to these RAI responses):

"Each commercial U.S. PWR unit shallprovide a summary report ofall inspectionsand monitoring, items requiringevaluation, and new repairsto the MRP ProgramManagerwithin 120 days of the completion of an outage during which PWR internalsare examined."

This recommendation is similar to those provided under other NEI 03-08 Issue Programs (IP), e.g. periodic BWRVIP Inspection Summaries, MRP Inspection Survey. While a 42

RAI Set 4- Final Responses: 10/29/2010 final format and method of sharing has not been determined the information obtained from these summary reports will be available to update the operating experience record on a periodic basis. The current format for summarizing and updating that operating experience may continue to be the current format contained in Appendix A of MRP-227, or the format may change to parallel one of the other IPs, or that format may change to adapt to circumstances. It is also the expectation that these summary reports will be provided in similar fashion to the staff either as part of a stand-alone document or combined with other reporting of inspection results (i.e., MRP-219).

Because foreign utilities are not as obligated as part of the NEI 03-08 initiative, obtaining international experience is not as certain. Nevertheless, foreign utilities actively participate in the MRP and the PWROG-MSC and share inspection experience in meeting forums, as well as cooperative efforts to benchmark each others approaches to the managing the aging of reactor internals.

RAI 4-14: Verify that neither MRP-2 10, "Materials Reliability Program: Fracture Toughness Evaluation of Highly Irradiated PWR Stainless Steel Internal Components," nor Section 6 of MRP-227 will be used to disposition (i.e., determine need to repair, need to replace, or inspection periodicity) degraded components identified during RVI inspections and that this guidance will instead be provided by WCAP- 17096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2, December 2009. If this is not the case, please provide a description of the relationship between MRP-2 10, Section 6 of MRP-227, and WCAP- 17096 and identify the aspects of the disposition analysis that will be governed by each document.

Response: Demonstration of the acceptability for continued service of PWR internals components is required when a relevant condition exceeding the examination acceptance criteria of Section 5 of MRP-227 is detected.

MRP-210, "Materials Reliability Program: Fracture Toughness Evaluation of Highly Irradiated PWR Stainless Steel Internal Components," is one of the key and informative reference documents that was used in the preparation of MRP-227, and Section 6 of MRP-227 is considered to be valuable information that could be useful to PWR licensees.

However, neither MRP-2 10 nor Section 6 of MRP-227 contain any "Mandatory" or "Needed" requirements for engineering evaluations of degraded components identified during the examinations and inspections specified in Table 4-1 through 4-9 of MRP-227.

MRP-2 10 was intended to demonstrate the relative flaw tolerance for potential flaws in several typical reactor internals components including irradiated material considerations but was not intended to establish a particular component's critical flaw size nor establish generic requirements.

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RAI Set 4- Final Responses: 10/29/2010 On the other hand, WCAP-17096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2, December 2009, provides generic methodologies tailored more specifically to the component or assembly and the aging effect(s) that can be used by licensees to generate engineering evaluation acceptance criteria. This is the document that will be used as the framework to develop those more specific evaluations either generically where similarities support generic efforts or for plant-specific applications.

RAI 4-15: There have been several previous RAIs related to cast austenitic stainless steel (CASS) materials, but several questions/issues remain.

a. The industry is currently supporting using a minimum irradiation embrittlement (IE) threshold of 1 displacement per atom (dpa) to determine susceptibility of CASS components to 1E, yet available data seems to support a threshold of 0.3 dpa or less as there is little data between 0.05 and 1 dpa and current data indicates some toughness decrease between 0.3 and 1 dpa. Would a reduction in the screening threshold from 1 to 0.3 or 0.05 dpa result in additional components screened in for IE? If so, identify the CASS components that would likely be screened in for IE susceptibility due to these lower screening thresholds. Finally, many CASS components are in the A inspection category. Please provide the basis/justification for placing these components in the A category.

b. The fracture toughness of CASS can degrade significantly due to thermal embrittlement (TE) and the toughness of both CASS and other stainless steel materials can decrease significantly as neutron fluence increases. When the dose exceeds 5 dpa, available data indicates that fracture toughness can be extremely degraded in many materials. The staff's concern is that the fracture toughness in CASS components may get so low due to TE and/or IE and in other components due to IE that preexisting fabrication or service-induced flaws that are smaller than the inspection resolution may challenge component integrity under normal loading or under design basis events. Additional guidance to licensees may be needed either in MRP-227 or WCAP-17096, "Reactor Internals Acceptance Criteria Methodology and Data Requirements," Revision 2, December 2009, to address this situation. Please describe how existing or planned guidance addresses this issue. Otherwise justify why such guidance is not needed.

Response: The MRP has recently completed an assessment of thermal aging and irradiation embrittlement for the CASS materials in PWR internals (MRP-276). In response to item RAI 4-15a, the following information has been excerpted from this report:

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RAI Set 4- Final Responses: 10/29/2010 There are three CASS items in the B&W design PWR internals that are expected to exceed a fluence of lxlO17 n/cm 2, E > 1.0 MeV at the end of a 60-year lifetime.

These three items are:

" IMI guide tube assembly spiders

" CRGT assembly spacer castings

" Plenum cylinder reinforcement castings (DB only)

The IMI guide tube assembly spiders are classified as a primary item in MRP-227. A visual (VT-3) examination of the IMI guide tube spiders is to be performed. The IMI guide tube spiders are being examined to detect spider arms that do not align with the lower fuel assembly support pad center bolt. The recommended methodology for acceptance given in WCAP-17096 is to perform an analysis to show that one or more missing spider arms or a completely missing spider will not result in loss of function of the IMI guide tube. Thus, no fracture mechanics evaluations are needed. Since there are 52 IMI guide tubes in each B&W unit, a redundancy argument may also be adequate.

The CRGT assembly spacer castings are classified as an expansion item in MRP-227.

If necessary, a visual (VT-3) examination of the CRGT spacer castings is to be performed. The spacer castings have limited accessibility from the top or bottom of the CRGT through a center free-path (once the plenum assembly is removed from the vessel). Examination at the quarter points where the threaded connections are present is possible. These lanes are not blocked by the rod guide tubes. The examination would look for cracking of the spacer surface or evidence that the spacer is not approximately centered. The threaded fasteners are welded to the OD of the pipe column so it is possible that a degraded threaded location would not be detected.

Since there are 69 CRDMs in each B&W unit, the recommended methodology for acceptance given in WCAP-17096 is to perform a reactivity analysis to determine the number of CRDMs that are required for shut down of the reactor. Thus, no fracture mechanics evaluations are needed.

There are two plenum cylinder reinforcement castings (at DB only), which have not been included in any MRP or PWROG evaluations as of the preparation of this document. As noted in Table 4-1 of MRP-276, only the lower edge of these castings are expected to exceed a fluence level of lxlO17 n/cm2 , E > 1.0 MeV at the end of a 60-year lifetime. Assuming the evaluations and conclusions would be the same as the wrought reinforcement plates at the other B&W units, this item would be classified as "No Additional Measures." It is also possible that this item could be dispositioned by reviewing the materials records and determining the ferrite content to be below the 45

RAI Set 4 - Final Responses: 10/29/2010 MRP-175 screening criteria, which would also classify it as "No Additional Measures."

Therefore, it is concluded that the CASS items in the B&W design PWR internals are redundant and/or potentially able to be analyzed for functionality in the anticipated degraded conditions. Replacement of the degraded item or component is also a potential option. Thus, no fracture toughness properties would be required for fracture mechanics analyses.

MRP-191 indicated there are only four components fabricated from CASS in the CE design PWR internals:

" Core support columns

" CEA shrouds

" CEA shroud bases

• Modified CEA shroud extension shaft guides The aging management strategy employed for thermal embrittlement in the CE reactor internals requires inspection for cracking at appropriate locations. As the thermal embrittlement susceptibility of a CASS component is determined by the ferrite content and service temperature, it is not possible to use cracking as a leading indicator of this degradation mechanism. Therefore the classification of the CASS components as primary or expansion is determined by the relative susceptibility to cracking. In the case of the CE design, there are no CASS components that are ranked as a primary component for any of the cracking mechanisms (fatigue, SCC or IASCC). The core support columns were dispositioned as an Expansion inspection component in MRP-227 and evaluation of IE and TE must be considered in the evaluation of cracking in this component.

The three remaining internals components (CEA shrouds, CEA shroud bases, and modified CEA shroud extension shaft guides) were not originally screened in based on the MRP-175 threshold fluence. These components would require additional evaluation if the lower threshold for thermal embrittlement were employed.

The CEA shroud bases accumulate enough neutron fluence to be categorized as part of Region 2 per MRP- 191. Components in fluence Region 2 are expected to receive between lxl102 and 7x10 20 n/cm 2, E > 1.0 MeV (0.15 to 1.1 dpa) by the end of the 60-year license period. Thus, the CEA shroud bases are expected to exceed the NRC screening level for synergistic embrittlement. However, CASS versions of the bases are present in only one plant. All other CE plants with this component have bases 46

RAI Set 4- Final Responses: 10/29/2010 fabricated from wrought Type 304 stainless steel. Consideration of the shroud bases in a generic manner is not appropriate, since CASS shroud bases are present in only one plant.

The modified CEA shroud extension shaft guides are located at the top of the CEA shroud assembly approximately in line with the mating surface between the reactor vessel and the reactor vessel head. No specific fluence or stress analyses have been completed for this component, but in this region of the reactor, a very low accumulated fluence (probably lower than the 0.05 dpa screening value) would be expected. Also, this component does not serve as a core support structure, and the loads are expected to be quite low. The knowledge gap for this component is the lack of analyses that address effects of fluence or stress. However, results of any analyses are expected to show low fluence and low stress. MRP-191 classified this component as a Category A component because of the low likelihood for failure.

The CEA shrouds constitute part of a core support structure in the CE design and may receive enough fluence to exceed the 0.05 dpa level. An initial survey of the CE plants indicated that a significant portion of the CE fleet has CEA shrouds fabricated from cast stainless steel. The FMECA panel moved this component to Category A based on a low likelihood of failure. It should also be noted that the CEA shrouds are generally centrifugal castings containing less than the MRP-175 thermal embrittlement screening criterion of 20% delta ferrite. This, in addition to the expected low likelihood of failure, supports the placement of the CEA shrouds in Category A.

There are relatively few CASS components in the Westinghouse design PWR internals. CASS was primarily used in non-structural or redundant components.

Therefore, there are relatively few requirements for fracture toughness determinations. Six of the eight CASS components were already identified in the screening process for potential irradiation embrittlement concerns due to relatively high peak neutron fluences. However, there were no requirements for flaw tolerance analyses in the evaluation procedures for these components; therefore, there were no requirements for fracture toughness data. Thus, there are no data gaps identified for these six screened-in components.

There are two remaining CASS components that were not identified for irradiation embrittlement:

" Intermediate flange in the control rod guide tube assemblies

" Lower support castings 47

RAI Set 4 - Final Responses: 10/29/2010 The intermediate flanges on the control rod guide tube assemblies are not expected to exceed the 0.05 dpa screening limit because these flanges are well removed from the core. Cracking in the intermediate flanges of the control rod guide tube assembly is less probable than cracking in the lower flanges, where both fluences and bending stresses are expected to be higher. At most, the intermediate flange would be an expansion item for the lower flange. Should a crack be observed in an intermediate flange in this component, a full flaw tolerance analysis would not likely be required.

Rather than performing crack growth predictions, a relatively small component like the relevant portions of the intermediate flanges is typically considered failed once a crack is detected. Therefore, there is no apparent need for fracture toughness data for the intermediate flange material.

A small fraction of Westinghouse plants have lower support castings rather than lower support forgings. Although the lower support casting is well-removed from the core, there is a remote possibility that portions of this component may experience fluences greater than 0.05 dpa. In the unlikely event that cracking is observed on the surface of the lower support casting, a flaw tolerance analysis might be undertaken.

Since it is a large component, it may be possible to show by analysis that the stresses at the location of the crack are too low to drive crack propagation. If a flaw tolerance analysis is conducted, an estimate of the fracture toughness in the lower support casting would be required. It may be possible to assume a lower bound toughness equivalent to highly irradiated austenitic steel and demonstrate that significant margin against failure remains.

There is a potential gap for fracture toughness data to evaluate flaws in CASS lower support castings. However, there is a reasonable potential for demonstrating structural integrity with suitably conservative assumptions. There are no other data gaps for CASS components for the Westinghouse design.

In response to item RAI 4-15b:

There are no CASS items in the B&W design internals that are anticipated to exceed 5 dpa (3.3x10 2 ' n/cm 2 , E > 1.0 MeV) at the end of license renewal.

Due to the conservative nature of the screening process, peak fluences in the Westinghouse BMI cruciforms and lower support column bodies as well as the CE core support column bodies were assumed to be greater than 3.3x 1021 n/cm 2, E> 1.0MeV).

All three of these components were placed in Category B and embrittlement effects were considered in the aging management recommendations. The Westinghouse BMI 48

RAI Set 4- Final Responses: 10/29/2010 Cruciforms were eventually placed in the "no additional measures" category because there were no identified cracking mechanisms and the cruciform has minimal structural significance. The support columns in both the Westinghouse and CE designs were identified as Expansion components to be inspected for potential cracking. Although it is recognized that these columns are potentially subject to both irradiation and thermal embrittlement, WCAP-17096 anticipates that the acceptance criteria for these columns will not take structural credit for any cracked support column. The proposed WCAP-17096 methodology requires an analysis to demonstrate that the remaining unflawed columns will provide the required structural support. This approach is similar to the minimum acceptable bolting patterns applied to core baffle and core shroud bolts.

Because there is no flaw tolerance requirement in the proposed evaluation methodology, there is no need to estimate the fracture toughness of the embrittled component. The margins applied to the WCAP-17096 evaluation methodology are intended to account for undetected and newly initiated flaws.

RAI 4-16: MRP-227 identifies several components that require plant-specific aging analysis (e.g., fatigue analyses) to determine the appropriate inspection category. However, MRP-227 does not discuss or reference approved methods or acceptance criteria for conducting such analyses. Discuss why guidance is not necessary to ensure consistent application and interpretation of plant-specific aging analyses. Alternatively, if the industry plans to provide such guidance, discuss the plans, approach, and schedule for developing this guidance. This discussion should address how environmental effects should be treated in these analyses.

Response: Three Primary components for Combustion Engineering (CE) plants in Table 4-2, which were identified for potential fatigue related degradation require inspection when adequate fatigue life cannot be demonstrated by TLAA. This reference recognizes the potential for a CE plant to have a current licensing basis that includes core support structure fatigue design calculations. In such a case, the requirements for aging management are already defined by § 54.21(c) of 10 CFR Part 54. Any subsequent inspection requirements for these components TLAA would arise from the TLAA process.

The entry in the Examination Coverage column of Table 4-2 for these three components refers to a plant-specific fatigue analysis. In reviewing this RAI, it has become apparent that the wording of this entry could possibly be misconstrued. The intent of this entry was to require the utility to identify the potential location and extent of fatigue cracking.

There was no intent to require a utility that was not subject to the TLAA requirements to complete a full fatigue evaluation prior to the inspection. Therefore we would propose to replace the words "plant-specific fatigue analysis" with the words "evaluation to determine the potential location and extent of fatigue cracking."

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RAI Set 4- Final Responses: 10/29/2010 There is already adequate guidance on fatigue analysis available from both regulatory sources, such as Chapter X ofNUREG-1801, or from industry sources, such as NEI 95-

10. While this guidance normally addresses pressure boundary components, the same methodology and procedures, is readily applied to core support structures with a fatigue design basis. We do not believe that there is any need to duplicate this guidance in MRP-227. This same guidance can be used by plant owners that have yet to apply for license renewal. Most, if not all, of these plants are more recent vintage plants with well-defined ASME Class 1 fatigue design bases.

RAI 4-17: MRP-227 Tables 3-1, 3-2, and 3-3 identify component/aging mechanism combinations (where the combination is identified as either "primary," "expansion," or "covered by existing programs" in the tables) that are not identified in Tables 4-1 through 4-6, 4-8 and 4-9.

Tables 3-2 and 3-3 also identify components (e.g., In-Core Instrumentation Thimble Tubes in CE internals and control rod guide tube support pins in Westinghouse Internals) that are not identified in Tables 4-1 through 4-6, 4-8 and 4-9 at all. Identify the component/aging mechanism combination in Tables 3-1, 3-2, and 3-3 that are not identified in Tables 4-1 through 4-6, 4-8 and 4-9. Explain how each of these aging mechanisms will be managed and revise MRP-227, if appropriate. If the aging management review (AMR) line items that were previously provided to update the Generic Aging Lessons Learned Report in conjunction with the staff's review of MRP-277 need to be revised, provide recommended changes to the AMR line items.

Response: A separate roadmap document has been developed to augment this and several other RAI responses, and is included as part of the overall response package. In addition to the description of the eight-step process that was used to develop MRP-227, this document also points to the details of the MRP-227 supporting documents, including details on components in Tables 3-1, 3-2, and 3-3 that were originally screened in as non-Category A for at least one of the eight age-related degradation mechanisms that were found through subsequent evaluations not to be either a Primary nor an Expansion component. As a result of the further evaluations, such components would not be identified in Tables 4-1 to 4-6 as either a Primary or an Expansion component, but could be identified in Tables 4-8 or 4-9 as a component covered by Existing programs. The following examples given in RAI 4-17 (in-core instrumentation thimble tubes in CE plants and control rod guide tube support pins in Westinghouse plants, etc.) are typical of this evaluation process.

In-Core Instrumentation Thimble Tubes in CE Plants. The Zircaloy-4 in-core instrumentation thimble tubes in CE plants were originally screened in for wear in MRP-191, primarily as the result of operating experience at San Onofre Units 2 and 3 and Waterford Unit 3. However, the further evaluation documented in Section 4.1.7 of MRP-50

RAI Set 4- Final Responses: 10/29/2010 232 pointed out that modifications to the fuel alignment plate to alter the flow conditions in the vicinity of the entry point of the thimble tubes into the plate resolved the issue.

Therefore, wear caused by flow-induced vibration is not expected to challenge the integrity of these components in the future. However, although irradiation-induced growth of zirconium alloys was not explicitly identified in MRP-175 as an age-related degradation mechanism to be evaluated as part of the screening process, it has been observed that irradiation-induced growth in the axial direction of the thimble tubes has reduced the clearance between the thimble nose and the bottom of the fuel assembly. As a result, some plants have observed that the thimbles were being compressed when the upper internals structure was replaced after fuel reload. Ten plants have taken some action to address the issue, and six of these plants have replaced the thimble tube assemblies with modified designs that are shorter in length to accommodate expected irradiation-induced growth in the future. Two additional plants have replacement designs in fabrication and have made preparations to install the replacement thimbles in an upcoming outage. The remaining two plants have not yet begun preparations for a full replacement of the thimble tubes, but have instead taken the intermediate step of raising the thimble support plate to accommodate additional axial growth. All of the affected plants will likely have replaced their thimble tubes prior to license extension.

Westinghouse has provided plant-specific evaluations of the projected growth of the original thimble tubes, as well as recommendations for timing the replacement of these thimble tubes based on calculated clearances at the bottom of the fuel assembly to the affected CE-designed plants.

Because of the actions already carried out and the anticipated actions in the future, the management of this issue has been placed in the Existing Programs category.

Control Rod Guide Tube Support Pins in Westinghouse Plants. The control rod guide tube support pins in Westinghouse plants that were originally manufactured from Alloy X-750 material were screened in for primary water stress corrosion cracking (PWSCC), as well as wear and fatigue. The further evaluation of these components is documented in Section 4.2.5.2 of MRP-232. That further evaluation pointed out that failure of the guide tube support pins does not challenge safe plant operation, nor does such failure compromise control rod functionality. However, failure of guide tube support pins can result in a significant loose parts issue for the plant. In order to address this issue, some utilities have opted to replace guide tube support pins with a Type 316 cold-worked stainless steel support pin, others have opted to perform ultrasonic inspections to determine degradation, while some have preferred to take no action at this time. All domestic Westinghouse plants should or will have existing programs to replace or monitor guide tube support pins. As a result of past, current, and anticipated future actions, the management of this issue has also been placed in the Existing Programs category.

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RAI Set 4- Final Responses: 10/29/2010 CE Guide Lug Insert Bolts. Table 3-2 lists wear, fatigue and stress relaxation as "Existing Programs" for the CE Guide Lug Insert Bolts. Based on the following excerpt from MRP-232, stress relaxation and fatigue could have been moved to "No Additional Measures". However based on the observation that wear is identified as the expected manifestation for all three degradation mechanisms and the ASME Section XI exam already inspects this location for wear, it was determined that the wear inspection provided appropriate aging management for these three mechanisms.

Text from MRP-232 Section 4.1.5:

In severalplant designs, the guide lug insert bolts are additionallyretainedby the tight clearance between thefuel alignmentplate keyway and the contactface of the guide lug insert.If all of the guide lug insert bolts were to become loose, the most visible sign of degradationwould be "exacerbatedwear'"in the vicinity of thefuel alignmentplate keyway. Abnormal wear in this region would not be a clear indicationof guide lug insert bolt stress relaxation,and a specific evaluation of the wear and guide lug insert integrity would be requiredto identify if stress relaxationor insert bolt cracking were an issue.

Guide lug inserts and/orthe matingface in the fuel alignmentplate keyway were hard-faced duringfabricationto resist wear degradationin these components, so loss of structuralalignmentas a result of such wear is not plausible.

The likelihoodof stress relaxationis very low, such that wear andfatigue cracking of these bolts is not likely, so this component is consideredto requireNo Additional Measures. However, since current VT-3 inspection during the JO-year ISI is capable of evaluating the condition of wear on this surface, this component alsofalls under the classificationof Existing. Noise monitoring should also continue to be used as a supplementary method to determine ifexcessive vibrationwithin the vicinity of thefuel alignmentplate is occurring duringoperation.

CE Fuel Alignment Pins. Table 3-2 lists IASCC, Wear, Fatigue, irradiation embrittlement and stress relaxation as "Existing Programs" for the CE Fuel Alignment Pins. These concerns are directly relevant to the designs with full height shroud panels.

The evaluation in MRP-232 recommended an inspection for cracking as the appropriate means of detecting IASCC, fatigue and stress relaxation (cracking of tabs). Aging management for irradiation embrittlement is typically incorporated as an evaluation requirement for the cracking mechanisms.

Wear is a potential concern in all fuel alignment pin designs. It has been listed separately as an existing program for the non-full height shroud panel plants.

Text from MRP-232 Section 4.1.4:

Fuel alignmentpins are consideredas components with an Existing inspection, because thefuel alignmentpins are inspected as part of the normal JO-year in-service inspection.

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RAI Set 4 - Final Responses: 10/29/2010 The concerns with IASCC, wear, and irradiation-inducedstress relaxationfrom the screeninganalysis have been dispositionedand loss offunctionality has been determined to be a very low probability event. The adverse effects of stress relaxationshould be eliminated by the welded tabs, as long as the tabs remain intact.A visual inspection of the tabs may be requiredto assure their integrity and to justify reliance on these tabs to disposition the effects of stress-relaxation. Visual inspection of the welded tabs would be recommended as a one-time inspection to be completed during the 10-year in-service inspection (ISI)prior to enteringthe periodofplant life extension. Missing tabs, missing fuel alignmentpins, or abnormalwear of thefuel alignmentpins would be easy to evaluate using a VT-3 visual inspection.

Westinghouse Upper Core Plate Alignment Pins. Table 3-3 lists SCC and wear as "Existing Programs" for the Westinghouse Upper Core Plate Alignment Pins. The first visible manifestation of the degradation effect for these two mechanisms was determined to be wear. The aging management recommendation for both SCC and wear is to inspect for wear. A failure due to SCC might also be reported as a result of a VT-3 exam. The aging management program is based on inspecting for wear as the early indicator of an aging concern.

Text from MRP-232 Section 4.2.9.2:

Upper core plate alignmentpins arefabricatedfrom austeniticstainless steel and are welded intoplace. Based on the structuralwelds that hold these pins in position and the potentialfor localizedresidualstressesfrom welding to exceed the screeningthreshold for stress, they were conservatively screenedin for SCC. However cracking of these components is consideredunlikely and thefirst visible manifestation of damage is expected to be wear.

Westinghouse Core Barrel Flange. Table 3-3 lists SCC as an "Expansion Program" and wear as an "Existing Program" for the Westinghouse Core Barrel Flange. Wear of the core barrel flange is listed in the Existing Programs Table for Westinghouse.

However the designation of the core barrel flange as an expansion requirement for SCC may appear circular as the Primary item is the Core Barrel Flange Weld. The intent was to expand to the remaining core barrel welds. This is clear in the MRP-227 section 4 Tables. Table 3-3 is potentially confusing in this regard.

Text from MRP-232 Section 4.2.2.7:

The potentialfor large residualstresses in the unirradiatedcore barrelwelds make them a potentialleadcomponentfor SCC. Under normaloperatingconditions, the upper flange weld is expected to experience the highest stress. Given the criticalstructuralrole of the core barrel,periodic inspectionfor cracking of the high stress weld is recommended.

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RAI Set 4 - Final Responses: 10/29/2010 The proposed inspection methods are appropriate for degradation when cracking is the primary effect. The cracking-related mechanisms would include SCC, IASCC and fatigue. The VT-3 examination can also be used to detect visible signs of wear. Gross deformation due to swelling may also be detectable in a visual exam, but severe effects of swelling may occur long before the deformation is observable. However, there is no non-destructive inspection technique capable of detecting thermal or irradiation embrittlement. At this time there is no practical way to monitor stress relaxation by measuring loads in reactor internal bolting. Although MRP-227 has identified irradiation embrittlement, thermal embrittlement, void swelling and irradiation induced stress relaxation as primary or expansion degradation mechanisms for multiple components in Tables 3-2 and 3-4, there are no effective inspections techniques for these mechanisms.

Although there are no inspection requirements for these components the proposed Rev.0 would include these mechanisms listed under the effect "Aging Management" in Table 4-2, 4-3, 4-5, 4-6, 4-8 and 4-9.

The aging management strategies for void swelling and stress relaxation must rely on detection of the secondary consequences of these mechanisms. The irradiation aging analysis conducted on the baffle-former structure provides the basis for determining these consequences. The aging analysis does suggest relative displacement along seams in the baffle structure that may be directly observable. The only other observable consequence of void swelling in the baffle-former-barrel assembly is IASCC failure of baffle-former bolts and baffle-edge bolts caused by swelling in the former plates. The timing of the failure is affected by compensating loss of load due to stress relaxation. Therefore, inspections of the bolting systems for IASCC failure provide an indicator of these related degradation mechanisms.

The aging management strategies for thermal embrittlement and irradiation embrittlement rely largely on trend curves compiled from laboratory data. Embrittlement can lead to loss of toughness that reduces the flaw tolerance of the materials. This loss of toughness can have a drastic effect on the acceptable flaw size in the component. Section 6.2.2 of MRP-227 provides guidance on fracture mechanics analysis of irradiated components.

Because the irradiated components and thermally embrittled components have a reduced flaw tolerance, it is particularly important that any active cracking mechanism in these components be actively managed. In the inspection strategy outlined in Tables 38-49, every component with an identified embrittlement concern has a corresponding requirement for inspection related to one or more potential cracking mechanism.

The second question in this RAI concerns component/aging mechanism combinations that are in Tables 3-1 through 3-3 but not identified in Tables 4-1 through 4-9.

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RAI Set 4- Final Responses: 10/29/2010 For the B&W design, Table 4-1 lists "overload" for the core barrel assembly locking devices and locking welds of the baffle-to-former bolts and internal baffle-to-baffle bolts; this was inadvertently not captured in Table 3-1. MRP-227 Table 3-1 will be revised to include reference to "Note 1" for these components and the note will be revised to include the locking devices for these components. This is the only example of a component/aging mechanism combination in Table 3-1 not identified in Tables 4-1 or 4-4 for the B&W design components.

The CE and Westinghouse Tables, Tables 4-2, 4-3, 4-5, 4-6, 4-8 and 4-9, will be updated in the '-A' version of MRP-227 to include the appropriate component/aging mechanisms combinations (see Appendix A to these RAI responses).

RAI 4-18: As a follow-up to RAI-26 (second set of RAIs), please clarify if components that are predicted to locally exceed 5% swelling by volume are inspected for cracking at those locations.

Provide justification why any such components that exceed this criterion are not recommended for inspection.

Response: For the B&W design, there are no locations predicted to locally exceed 5%

swelling by volume. Fro the CE and Westinghouse designs, the MRP-227 I&E Guidelines monitor the distortion caused by the integrated effects of void swelling over the entire volume of a PWR internals assembly. The inspections that implement this requirement are shown in Tables 4-2 and 4-3 of MRP-227 for Combustion Engineering and Westinghouse plants, respectively. The inspections focus on the regions of high swelling and locations where the accumulated effects of swelling over the assembly are most likely to be visible. The particular distortions that led to the recommendations were the result of the functionality analyses reported in MRP-229 and MRP-230.

These functionality analyses showed that the volumes of material for which relatively high "effective irradiation strain growth" was calculated were very small and localized, to regions of combined high fluence, high gamma heating, and limited cooling. (Note: the volumetric swelling is equal to 3x the irradiation growth strain.) In the Westinghouse design, the high swelling regions were limited to high fluence seams where there are joints between the baffle and former plates. Typical Westinghouse high fluence regions are indicated in the accompanying figure. MRP-227 requires VT-3 examinations of the baffle-former assembly and the baffle-edge bolts at these locations. Even though the FEA aging evaluation included these levels of swelling, the analysis did not indicate stresses high enough to initiate IASCC in these components. The high swelling locations are remote from significant structural load paths and cracking at these locations is not expected to have a significant structural impact. Therefore it was determined that the 55

RAI Set 4 - Final Responses: 10/29/2010 VT-3 examinations were adequate to manage cracking due to age related degradation at these locations.

A similar situation occurs near the mating surfaces of the upper and lower core shroud former plates in the CE design (MRP-227 Figure 4-14). The MRP-227 guidelines require inspections of the joint for observable distortion and the adjacent welds for cracking.

Because the welds are considered a structural element, the requirement is for an EVT- 1 inspection at the joint.

The MRP continues to sponsor projects to investigate the expected significance of void swelling. If detrimental effects due to localized void swelling are observed in operating reactors, the MRP and industry will then develop specific guidance to characterize its effects, recommend appropriate changes to inspections, and determine its future acceptability.

Areas in the Westinghouse Baffle-Former Assembly with >5% Swelling at 60 Years

• 0 0 0 0 Linear Strain

-f

- M .016667 1 .019971 M .022908 m .026212

- .029148

• 10*ý - .032452 1.035389 1 .038693

. 04163

.044934

.047871

.051175 m .054112 1 .057416

- .060353

- .063657 Former Level 4 Note: 0.01667 Linear Strain is equivalent to 5% Volumetric Swelling 56

RAI Set 4- Final Responses: 10/29/2010 RAI 4-19: The following components, listed as an example only, were originally identified for potential aging degradation but they were dispositioned under "No Measures" category. The staff requests that MRP provide an explanation for not performing any analysis prior to binning them under "No Measures" category.

COMBUSTION ENGINEERING COMPONENTS MRP-232 Referenc -

Component Aging Effect Reference Core Support Plate Bolts Irradiation Embrittlement Table 2-11 Fuel Alignment Pins (304 Irradiation Embrittlement Table 2-11 stainless steels)

Core Shroud Tie Rods Irradiation Embrittlement Table 2-11 Core Shroud Tie Rods Irradiation Induced Stress Table 2-16 Relaxation Tables 2-5 2-3 and Core Support Plate IASCC, Wear 2-5 i

WESTINGHOUSE COMPONENTS MRP-232 -

Aging Effect Referenc Component Reference Lower Core Plate Fuel Irradiation Embrittlement Table 2-12 Alignment Pin Bolts Lower Core Plate Fuel Irradiation Induced Stress Table 2-17 Alignment Pin Bolts Relaxation Bottom Mounted Tables 2-4 and Instrumentation (BMI) IASCC, Irradiation Embrittlement 2-4a Column Bodies IASCC, Thermal Embrittlement, Tables 2-4. 2-10 and Irradiation Embrittlement and 2-12 Response: A separate roadmap document has been developed to augment this and several other RAI responses, and is included as part of the overall response package. In addition to the description of the eight-step process that was used to develop MRP-227, 57

RAI Set 4- Final Responses: 10/29/2010 this document also points to the details of the MRP-227 supporting documents, including details on components in Tables 3-1, 3-2, and 3-3 that were originally screened in as non-Category A for at least one of the eight age-related degradation mechanisms that were found through subsequent evaluations to require No Additional Measures.

As a specific example of the way in which this roadmap document was prepared, one of the items in the RAI 4-19 table - core shroud tie rods in CE plants subject to irradiation embrittlement and irradiation-induced stress relaxation - will be followed through the entire categorization process. Core shroud tie rods in CE plants were originally screened in (see Table 5-2 in MRP-191) for wear, fatigue, irradiation embrittlement, and irradiation-induced stress relaxation. For this reason, the original categorization based only on screening with respect to potential significance of the eight age-related degradation mechanisms was non-Category A. Based on the FMECA evaluation (see Table 6-6 in MRP-19 1), the FMECA group to which core shroud tie rods was Group 1, the lowest of the three groups, based on likelihood of failure and consequences of that failure. Then, Table 7-3 in MRP-191 shows that the preliminary placement of Category B, which means that the core shroud tie rods are not considered as a candidate for a Primary component, but could be considered as a candidate for an Expansion component.

Then, in Table 4-4 of MRP-232, core shroud tie rods are shown as non-Category A for the same four age-related mechanisms as before (wear, fatigue, irradiation embrittlement, and irradiation-induced stress relaxation), but is also shown as requiring No Additional Measures. The discussion and justification of this recommendation takes place in the last paragraph of Section 4.1.1 of MRP-232, where the results of finite element analysis and associated engineering assessment show that the combined effects of irradiation embrittlement and irradiation-induced stress relaxation do not indicate the need for any additional examinations beyond those required for core support structures, and that the remaining effects of wear and fatigue also do not reach the threshold for further consideration.

Similar types of documentation are available for the other items in the RAI table as well as the B&W design, and that reasoning is provided by the accompanying roadmap.

RAI 4-20: Many licensees have incorporated ANS 51.1, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants," which categorizes transient events in a classification scheme by condition, into facility licensing bases. According to the standard, an acceptance criterion for a Condition II event is that by itself, a Condition II incident cannot generate a more serious incident of the Condition III or IV category without other incidents occurring independently or result in a consequential loss of function of the reactor coolant system or reactor containment barriers. For example, an anticipated operational occurrence, such as a turbine trip from full power, should not cause a degraded component inside the reactor 58

RAI Set 4- Final Responses: 10/29/2010 vessel to fail in such a way that a control element assembly ejection could occur. Further detailed discussion regarding this criterion is available in NRC Regulatory Issue Summary 2005-29, "Anticipated Transients that Could Develop Into More Serious Events."

For those components that the FMECA or functionality analyses provided a basis to reduce or eliminate inspection requirements, address whether consideration of this "non-escalation" criterion affects this basis.

Response: The non-escalation criterion in ANS 51.1 is intended to assure that a relatively frequent Condition II transient cannot generate a more serious (and less frequent) Condition III or Condition IV event without other incidents occurring independently. At issue is whether or not the FMECA evaluations and the functionality analyses took into account the full range of potential consequences of a relatively frequent operational transient causing failure of an aging-degraded component, including the possibility of triggering a much less frequent but more serious event. The answer is that the FMECA evaluations certainly considered such possibilities, with the intent to determine the most conservative range of consequences of aging-degraded component failure.

To be very precise about this point, the first two steps performed in the MRP-227 development process are described with the issue of consequences emphasized. For example, the first step (screening) evaluated potential component susceptibility to the eight age-related degradation mechanisms and their effects, with no consideration of consequences. The results of this step were documented in MRP-189 and MRP-191.

However, in the second step, the FMECA process used expert elicitation to determine the most conservative set of consequences from the failure of a degraded component, regardless of whether that failure was initiated by a Condition II transient or by some other event. The results of this step were documented in MRP-190 and MRP-191.

In this regard, the experts had access to previous work by the industry where the consequences of loose parts caused by component failures were systematically evaluated, including both safety and economic consequences. The expert elicitation process used this information, along with the expert opinions about the consequences of many other postulated failure modes for aged components, as the basis for the assignment of risk consequences. These risk consequences included both the potential for that postulated component failure to escalate into a more serious event, or the observation that the postulated component failure resulted in modest to negligible risk (non-escalation). For a component that was found to be potentially susceptible to one or more of the eight age-related degradation effects, but for which the consequences of postulated failure were found to be negligible, it could be inferred that the non-escalation of consequences was a 59

RAI Set 4 - Final Responses: 10/29/2010 contributing factor to reduced or eliminated inspection requirements. For a component that was found to be potentially susceptible to one or more of the eight age-related degradation effects, and for which the consequences of postulated failure were found to be significant, the escalation of consequences could be inferred to be a contributing factor for the assignment of Primary, Expansion, or Existing Program inspection requirements.

In a few cases, the evaluation of potential escalation was formalized explicitly, as illustrated by the following excerpt from Section 3.3 of MRP-190:

"Duringthe meeting, it was recognized that if the failure mode was not detectable, a second consequence question needed to be posed"would the degradationmechanism result in a more severe consequence (if undetected) when a design basis event occurred (e.g., seismic or LOCA)? The consequence column metric is the most conservative consequence (between normal operationand considerationof a design basis event, when needed). "

Following this logic, Section 3.4.2 of MRP-190 describes a number of these "cascading failures," as shown in Tables 5-2 through 5-6 of MRP-190. Similar treatment of cascading consequences was carried out in the preparation of MRP- 191; however, the process is not explicitly described.

The functionality analyses results described in MRP-229 and MRP-230 were used, in part, to support the recommendations in MRP-231 and MRP-232 that were eventually incorporated into the requirements of MRP-227. These functionality analyses were primarily carried out to evaluate the degree of aging related degradation in the identified components. The focus of the functionality analysis was to determine the effects of long term exposure to normal operating conditions (including normal operating transients) on component condition. There was no consideration of off-normal transients in the functionality analyses. Even so, the functionality analyses provided evidence for non-susceptibility in a few cases - i.e., clear evidence that 60 years of conservative operation resulted in insignificant degradation. Such findings led to the determination that no additional measures were required to monitor aging effects in these components. Stated another way, the insignificant degradation did not affect the ability of the component to perform its intended function, including reasonable assurance that safety-related functions would be maintained.

Therefore, degradation of the reactor internals leading to a failure that could initiate a system transient at any level is considered in MRP-227 as a potential loss of functionality requiring aging management. Analysis of the consequences of the transient or more serious events that could subsequently develop is beyond the scope of the document 60

RAI Set 4- Final Responses: 10/29/2010 RAI 4-21: Address the effects of failures of uninspected components and components with failure modes that aren't detectable during normal operations (i.e., undetectable failure modes) through the following considerations:

a) Discuss whether the failure of any such component(s) could be an initiating event for a plant transient or other accident.

b) Discuss the effect of failure of any such component(s) on system performance assuming a design basis event (i.e., plant transients, accidents, and seismic events representative of upset, emergency and faulted loading conditions) occurs prior to mitigating the failure. As part of this discussion, describe any analysis that has been performed, or any plant-specific analysis that is needed, to demonstrate that acceptable system design margins are retained under this scenario.

Finally, discuss whether the final recommendation not to inspect these components is affected by addressing the scenarios described in a) and b) above.

Response: The RAI concerns PWR internals components for which no inspection requirements are prescribed in MRP-227. There are two sets of such components. The first and largest set of such components are the components that fall into the category of No Additional Measures and which are not core support structures that continue to require periodic visual examination based upon ASME Code Section XI requirements.

The second set is a relatively small set of components that are essentially inaccessible without extraordinary measures, but which are sufficiently susceptible to one or more aging degradation effects that they were placed in the Expansion category. We will refer to the first set as No Additional Measures/Non-Code (NAM/NC) components. We will refer to the second set as Inaccessible Expansion (INEX) components.

With respect to Part a) of the RAI, all components - including NAM/NC and INEX components were evaluated during the FMECA expert elicitation process to determine the consequences of their postulated failure, and the evaluation considered the potential for that failure to be an initiating event for a plant transient or other accident.

Discussions of these evaluations are documented in MRP-190 (Section 3 and the tables in Appendix A) and MRP- 191 (Section X). Special attention should be paid to the discussion of "cascading events" in MRP-190. In particular, the FMECA exercises reported in MRP-190 and MRP-191 both took advantage of previous industry efforts to evaluate the consequences of "loose parts' caused by PWR internals component failure, regardless of the cause of such failures.

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RAI Set 4 - Final Responses: 10/29/2010 With respect to Part b) of the RAI, the expert elicitation process used in the FMECA evaluations documented in MRP-190 (Sections 3.2, 3.3, and 3.4) and MRP-191 (Section Y) did include the consequences of design-basis events that could occur following postulated failure of a component, including NAM/NC and INEX components. The consequences derived from the expert elicitation process were qualitative but conservative, and were based upon the participating expert's knowledge and experience with the systems being considered. In some cases, generic or plant-specific analyses that attempted to simulate the consequences of the postulated failures were also available for consideration.

With respect to the final part of the RAI, the results of the FMECA evaluations for the NAM/NC components showed that even the most conservative estimates did not cause unacceptable safety or economic consequences. Therefore, even with the inclusion of the scenarios described in Part a) and Part b) of the RAI, the final recommendation in MRP-227 was unaffected. This was not the case for the INEX components. For this set of components, the final recommendation - as shown by example in Table 4-4 on Pages 4-27 and 4-28 (Expansion Components for B&W plants) - inspection was not an option (short of disassembly) because of inaccessibility. Therefore, the final recommendation was to permit either engineering evaluation or component replacement as options. It should be noted that, for the choice of engineering evaluation, plant-specific analysis considering the full range of design-basis loadings would be required.

RAI 4-22: MRP-190, Section 3 discusses component failure modes that aren't detectable during normal operations (i.e., undetectable failure modes). Provide specific examples of important components that are susceptible to these failure modes. Describe any special consideration or weighting that components susceptible to these failure modes received in either the FMECA (e.g., through the failure severity rankings) or the final MRP-227 inspection recommendations (e.g., by elevating the component to the primary inspection category) given that the component failure may not be discovered until the next refueling outage (i.e., up to 2 years after failure occurs). Provide specific examples to illustrate the process used to evaluate these components.

Response: Section 3.3 of MRP-190 describes a step in the FMECA expert elicitation process where the experts were asked to consider the potential for an undetected component failure mode and the modification of the resulting consequences if a design-basis event, such as an earthquake or a LOCA were to occur prior to mitigation:

"Duringthe meeting, it was recognized that if thefailure mode was not detectable, a second consequence question needed to be posed: would the degradationmechanism result in a more severe consequence (if undetected) when a design basis event occurred (e.g., seismic or LOCA)? The consequence column metric is the most conservative 62

RAI Set 4- Final Responses: 10/29/2010 consequence (between normaloperation and considerationof a design basis event, when needed)."

The results of that modification to the FMECA expert elicitation process are found in Table A-I of MRP-190, with particular attention to Column 10 (Heading: Detectable) and Column 11 (Heading: Comments). An excellent pair of examples is provided by the Plenum Cover Assembly bottom flange and the Plenum Cover Assembly support flange.

Both of these components received a "No" in Column 10. However, in the former case, the "No" includes a comment that ultrasonic examination (UT) during a periodic ten-year inspection would provide an adequate basis for managing the degradation, whereas the latter presented no operational issues. Another excellent example is the Plenum Cover Assembly top flange, which falls into the same category as the Plenum Cover Assembly bottom flange.

The information contained in Table A-I of MRP- 190 carried over to MRP-23 1, which provided the actual inspection recommendations for inclusion in MRP-227; however, the fundamental evaluation results and discussion are contained in the columns of Table A-1 of MRP-190.

The RAI refers specifically to MRP-190, which applies to B&W designs. MRP-191, which applies to CE and Westinghouse designs, does not describe a similar approach for addressing the potential for a component failure mode that would be undetected during normal operations. However, the FMECA process steps described in MRP-191 contain the essential elements that would have led to results similar to those reported in MRP-190. For example, on page 6-1 of MRP-191, one of the six basic questions to be addressed by the FMECA expert panel was: "How might the failure be detected?" The range of expertise on the panel would have been able to identify those component failure modes that would be undetected during normal plant operation, and would have been able to adjust the worst-case consequences accordingly. These worst-case consequences would then be reflected in the severity rankings which, in turn, would have been reflected in the recommendations for inspection in MRP-232, if warranted.

RAI 4-23: Identify any components that should be replaced either prior to the period of extended operation or during the period of extended operation because they may not be able to perform their intended function during design basis events (normal, transient, emergency and faulted conditions) based on the results of the FMECA or functionality assessment.

Response: No internals components for Babcock & Wilcox, Combustion Engineering, or Westinghouse PWR plants were identified as requiring replacement due to inability to perform their intended function during design basis events prior to the period of extended 63

RAI Set 4 - Final Responses: 10/29/2010 operation or during the period of extended operation, as the result of either FMECA or functionality assessment.

RAI 4-24: Tables 2-18 and 2-19 in MRP-232 and Table 3-8 in MRP-231 indicate that a licensee's aging management program will inspect CE, Westinghouse and B&W RVI components for thermally or irradiation-enhanced stress relaxation. However, various CE, Westinghouse and B&W RVI components that are susceptible to thermally and irradiation-enhanced stress relaxation have been downgraded from Categories B or C to the "No Additional Measures" Category.

Document the basis of the evaluation that was utilized to downgrade these components to the "No Additional Measures" Category. Demonstrate that both inspected and uninspected components susceptible to thermal or irradiation-enhanced stress relaxation maintain their design function during emergency and faulted events postulated at the end of the period of extended operation. This demonstration should show that the recommended inspection method is adequate for identifying or assessing stress relaxation before design margins become inadequate.

If a generic evaluation of the adequacy of such components under design basis loading is not possible, identify plant-specific action items that must be performed by licensees to ensure these components will be able to maintain their design function during design emergency and faulted conditions at the end of the period of extended operation.

In particular, identify the projected loss of preload due to stress relaxation at the end of the period of extended operation for the following bolts.

" Combustion Engineering---Core Support Column Bolts; Core Shroud Bolts; Guide Lug Insert Bolts; Barrel-Core Shroud Bolts

" Westinghouse ---- Baffle-edge Bolts; Baffle-former Bolts; Lower Support Column Bolts

" Babcock and Wilcox ----- Baffle-to-Baffle Bolts; Core Barrel-to-Former Bolts; Baffle-to-Former Bolts.

Explain why this loss in preload will not result in the loss of the intended function for these bolts during design basis events that are postulated at the end of the period of extended operation.

Response: The technical basis for downgrading to No Additional Measures some PWR internals components that were originally screened in as either Category B (moderately susceptible) or Category C (significantly susceptible) to thermally-induced or irradiation-64

RAI Set 4- Final Responses: 10/29/2010 enhanced stress relaxation will be addressed in a separate "roadmap" response. That response will cover the first three paragraphs of this RAI.

The last part of the RAI requests a quantitative "projected loss of preload due to stress relaxation at the end of the period of extended operation" for a list of specified bolts in Combustion Engineering, Westinghouse, and B&W plants, with the intent to "ensure that these components will be able to maintain their design function during design emergency and faulted conditions at the end of the period of extended operation." No such calculations were carried out on either a generic or design-specific basis during the development of MRP-227 or any of its supporting documents. Therefore, it is not possible to respond to the specific request for additional information.

However, the development of MRP-227 and its supporting documents did address several aspects pertinent to the underlying intent of the RAI, which will be discussed in the following paragraphs.

Thermally-InducedStress Relaxation. This topic was addressed thoroughly in the text of MRP-175 and in Appendix B of MRP-191, and that discussion will be briefly summarized here. In the absence of a significant radiation environment, the relaxation of preload for PWR internals is of the order of 10% to 20%. This amount of preload loss is of the same order as the variability of preload in bolted assemblies that are subjected to very careful bolt torque patterns, depending upon the number of torque passes and the flexibility of the underlying structure. The loss of preload from thermally-induced stress relaxation occurs over a period of a few thousand hours, after which further exposure with time essentially shows saturation of the effect, with very little additional loss. Both the loss of preload from thermally-induced stress relaxation and the variability of preload from the torque pattern are accounted for by increasing the specified preload slightly, so that the residual preload is sufficient to maintain function.

Irradiation-Enhanced Stress Relaxation. The functionality analyses of a representative core barrel assembly in a B&W plant, a representative core shroud assembly in a Combustion Engineering plant, and a representative baffle-former-barrel assembly in a Westinghouse plant evaluated bolt preload changes caused by the combination of irradiation-enhanced stress relaxation (or creep) and void swelling. The analyses were based upon 30 years of normal operation in a high-leakage core environment, followed by 30 years of normal operation in a low-leakage core. No design-basis emergency conditions or design-basis faulted conditions were included in the analyses. As a result, the calculated reductions (and increases) in bolt preload cannot be considered to be projected losses of preload due to stress relaxation with the intent to ensure the capability to maintain design function. In addition, the estimates of preload change are highly 65

RAI Set 4- Final Responses: 10/29/2010 variable, depending on the location of a particular bolt relative to the core (for radiation exposure) and the location of a particular bolt relative to the prying action caused by integrated void swelling effects and associated global deformation.

MRP-229 and MRP-230 provide the detailed discussion of the functionality analyses of the three representative assemblies listed above, while MRP-231 and MRP-232 summarize the essential results. Suffice it to say that, during the first 30-year period of high-leakage core loading, loss of bolt preload due to irradiation-enhanced stress relaxation tends to dominate any potential increase in bolt preload from prying caused by void swelling, with complete loss of preload in some cases and very large losses in preload (say up to 70% loss of preload) not uncommon. After the second 30-year period, when the combination of effects is more mixed, the results are even more variable, since the prying action caused by the integrated effects of void swelling counteract the effects of irradiation-enhanced relaxation for many bolts. A typical example is provided by Figures 3-54 to 3-63 in MRP-230, which show the calculated baffle-former bolt preload for the representative Westinghouse plant at six year intervals, beginning with the 6 th year of high-leakage core loading. The reduction in preload is larger for the bolts nearer the active core region (say rows 3 through 6) than in rows 1 and 8. At year 30, when the transition from a high-leakage core to a low-leakage core takes place, the variability is at its extreme, with a few baffle-former bolts showing the dark blue color signifying essentially complete loss of preload, while others near or slightly above the initial preload. Examining the plots for the period of low-leakage core loading all the way out to 60 years of simulated operation shows that roughly half of the bolts has lost all or very nearly all of their preload, while the top row of baffle-former bolts continues to maintain the initial preload.

Similar results were observed for the representative B&W plant core barrel assembly functionality calculations reported in MRP-229 and evaluated in MRP-231 (see Section 3.2.2.3). The portion of the MRP-229 Section 4 paragraph summarizing the results is repeated here for convenience:

"Irradiation-inducedstress relaxationis most significantfor the baffle-to-baffle bolts at internalbaffle corners, where relaxation of over 90% (in some case complete loss of bolt load) is typical. The baffle-to-former bolts atformer elevations 2 to 7 experience large amounts of stress relaxationof up to 90%. Maximum relaxationof the external baffle-to-baffle and the core barrel-to-formerbolts is about 50% and 40%, respectively. In general,however, these bolts experience much smaller magnitudes of stress relaxation."

Two other topics are relevant to this discussion. First, bolt locking devices are a significant deterrent to complete bolting failure resulting from complete loss of preload, and the supporting documentation for MRP-227 describes the benefits of bolt locking devices in some detail. Second, examining bolts to determine potential loss of preload is an ongoing research activity that has yet to lead to practical and effective application.

66

RAI Set 4- Final Responses: 10/29/2010 For example, in situ measurement of the length of a bolt by ultrasonic testing to determine its current length relative to its initial preloaded length is a potential technique that has yet to be demonstrated as practical and workable. The accuracy required for the measurement and/or the need to compare to a very accurate baseline measurement represent obstacles that have yet to be overcome. Even the possibility of detecting complete loss of preload through some type of impact-echo system has yet to shown to be cost-effective.

In summary, quantitative calculation of the loss of preload through the end of the renewal period term, along with analytical demonstration that the calculated loss of preload does not prevent bolted assemblies from performing their intended function, is not within the scope of activities carried out in the MRP-227 development program. Instead, the aging management program elements for PWR internals components that are subject to potentially significant irradiation-enhanced stress relaxation or creep are based on detection and management of subsequent aging effects that ensue from the loss of preload, which include wear, fatigue, and IASCC.

RAI 4-25: The effect of radiation on material ductility is a TLAA for B&W vessel internals.

Section 4.2.6 of the Three Mile Island Nuclear Station Unit 1 License Renewal Application indicates the following:

The effects of irradiation on the materials properties and deformation limits for the reactor vessel internals was evaluated for the current licensing basis in Topical report BAW-10008, Revision 1, Appendix E.

This analysis concluded that at the end of the forty years, the internals will have adequate ductility to absorb local strains at the regions of maximum stress intensity, and that irradiation will not adversely affect deformation limits. This analysis is a TLAA that will be managed by the PWR Vessel Internals program for the period of extended operation.

The staff requests that the MRP explain how this issue has been addressed for B&W vessel Internals program. Are the effects of radiation on material ductility a TLAA for CE and Westinghouse vessel internals? If that is not the case, provide an explanation for not performing a TLAA evaluation in CE and Westinghouse vessel internals. If it is a TLAA explain how this issue is addressed in PWR vessel internals program.

Response: This question was originally identified as a Time Limited Aging Analysis (TLAA) in Section 5.1.5 of BAW-2248A and in the associated staff Final Safety Evaluation Report (FSER). The TLAA was defined as renewal applicant action item number 12 in the FSER, but as noted in Section 5.1.5 of BAW-2248A, this TLAA for 67

RAI Set 4 - Final Responses: 10/29/2010 B&W RV internals is to be resolved on a plant-specific basis per 10CFR 54.21 (c)(1)(iii),

based on the results and conclusions of the RV internals aging management program (RVIAMP) discussed in Section 4.6 of BAW-2248A.

As described in the attached roadmap (which is a response for RAI 4-1 and various other RAI responses that have been grouped within it), the B&W Owners Group (B&WOG) disbanded and the ongoing RVIAMP efforts were superseded and are currently being completed through the MRP efforts, which ultimately will establish the appropriate monitoring and inspection programs to be performed (i.e., the MRP-227 requirements).

As stated above in RAI 4-25, the TMI- 1 LRA indicates that the PWR Vessel Internals program will manage this TLAA and this was concluded by the staff to be adequate.

However, AREVA, through the MRP, has addressed this TLAA on a generic basis for the B&W RV internals with issuance of a non-proprietary report (AREVA 47-9048125-002, provided to the NRC by MRP Letter 2010-064, dated October 26, 2010), which is available for each of the B&W unit licensees. This document is provided as an attachment to the RAI responses. A proprietary version will be submitted to the NRC by AREVA NP Inc. Corporate Regulatory Affairs with an accompanying affidavit in accordance with 10 CFR2.390(b). Both of these documents are being provided to the Staff as information in support of the MRP-227 review, which clearly show that for the irradiation levels at the end of 60-year lifetime for this component, there will be adequate ductility at operating temperature to absorb local strains at the regions of maximum stress intensity, and that irradiation will not adversely affect deformation limits.

No TLAA related to material ductility has been identified for either CE or Westinghouse reactor internals. However, it should be noted that the effects of loss of fracture toughness and reduced material ductility are managed for both CE and Westinghouse reactor internals through aging management program elements defined in the Section 4 tables of MRP-227, and that these effects will be taken into consideration for any engineering evaluations of detected conditions that exceed the examination acceptance criteria contained in Section 5 of MRP-227.

RAI 4-26: RAI-20 (Set #2) asked about how the linkage between primary and expansion components was determined and how the expansion criteria (i.e., the results of the primary inspection that triggers an expansion inspection) was developed. While the response to RAI-20 is clear and the process is generally understood by staff, there is still a lack of explicit justification for many of the linkages and the explicit expansion criteria. That is, there is not a 68

RAI Set 4 - Final Responses: 10/29/2010 clear basis why the primary component was selected and why the expansion linkage is both appropriate and comprehensive (i.e., no other components should be linked).

The basis for the criteria used to trigger expansion inspections, and the acceptability of this basis, should also be provided for each of the primary and expansion linkages. As an example, expansion criteria for the core barrel and baffle barrel bolts are not triggered unless there is a 5%

or higher failure rate in the baffle former bolts. Similarly, a 10% rate of rejection for either the upper core barrel or lower core barrel bolts triggers the expansion items. The basis for these expansion criteria should demonstrate that the failure rate or rate of rejection specified for the baffle former bolts and the upper core barrel or lower core barrel bolts are sufficient to ensure that significant degradation is not occurring in the expansion components such that the design margin requirements for expansion components and associated systems are satisfied.

Response: The third item (Item 3) in the response to RAI-20 (Set #2) provided a considerable amount of detail on the basis for the linkages between Primary and Expansion components, including a reference to the expert panel elicitations from which the recommendations for those linkages were derived -- "Letter to Reactor Internals Focus Group from MRP,

Subject:

Minutes of the Expert PanelMeetings on Expansion Criteriafor Reactor Internals I&E Guidelines,MRP 2008-036 (via email), June 12, 2008)". This RAI acknowledges that that response was clear and the process generally understood by staff, so the process will not be described further. The RAI is asking for additional technical support for the expansion criteria themselves other than the results of the expert panel elicitation; i.e., what was the basis for the expert panel to debate and to eventually agree on a specific expansion criterion, such as the 5% failure rate in baffle-former bolts or the 10% failure rate for upper core barrel bolts?

An attempt will be made to respond to the RAI by example, rather than by going through lengthy and perhaps speculative discussions for each of the Primary and Expansion links.

Two examples will be provided - one of which deals with expansion criteria for a bolted assembly and the other of which deals with a welded components.

First, the expert elicitation process by which the linkages and criteria were developed took place at a point in time where a draft C of MRP-227 was available (out of eventually a draft J), and early drafts of MRP-231 and MRP-232 were also available (the supporting documents with specific inspection recommendations for MRP-227). Second, preliminary results were also available for functionality analyses of such assemblies as the B&W core support shield assembly, the Combustion Engineering core shroud assembly, and the Westinghouse baffle-former assembly. Third, many of the experts involved in the expert elicitation process were familiar with or had been involved with safety evaluations of internals bolting patterns carried out in the 1990s.

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RAI Set 4 - Final Responses: 10/29/2010 With this back ground in place, consider the particular combination of a Primary component - Core Shroud Assembly core shroud bolts in bolted core shroud Combustion Engineering plants - and the linked Expansion component - Core Shroud Assembly barrel-shroud bolts in bolted core shroud Combustion Engineering plants. The degradation effect is cracking from either fatigue or IASCC, and the environment comparison provides the justification for selecting one set of bolts as Primary and the other set of bolts as Expansion.

The experts discussed a number of potential expansion criteria, starting with a recommendation that, for similar bolting configurations, functionality could be maintained by 50% or fewer of the bolts, with certain restrictions on functional bolt distribution. Some experts argued that observing a much lower number of failures in the core shroud bolts, such as 10%, would be appropriate to indicate that expansion to include barrel-shroud bolts. The argument for the 10% figure was supported by the observation that the 100% of accessible bolts examination requirement for the core shroud bolts would also cover a large number of core shroud bolts with much lower fluence. The discussion among the experts then led to options for considering the number of failed core shroud bolts detected for different core shroud plates with different fluence levels, and whether failures for high-exposure core shroud bolts should carry more influence than failures for lower-exposure core shroud bolts. The consensus that was eventually reached was based on: (1) essentially all of the core shroud bolts are accessible, which means that both high-exposure bolts and low-exposure bolts will be part of the sample space; (2) a relatively large number of core shroud bolts failures can be tolerated in the high-exposure region without compromising assembly function; and (3) functionality is optimally assured by limiting the number of failures in the low-exposure region.

This eventual consensus of the experts led to the recommendation that confirmed failure of a number greater than 5% of core shroud bolts on the four lowest shroud plates at the largest distance from the core, which are the low-exposure shroud bolts, would trigger the examination of the Expansion linkage components. This was felt to be a very conservative recommendation, since the low-exposure core shroud bolts still have roughly five times the exposure of the barrel. The conservatism of the recommendation provided adequate assurance that the specified expansion criterion for the core shroud bolts would ensure that significant degradation would not occur in the barrel-shroud bolts prior to the expansion examinations.

The second example is the combination of a Primary component - the upper core barrel flange weld in Westinghouse plants - and the linked Expansion components - the 70