ML20207C732

Safety Evaluation of Topical Rept BAW-2248, Demonstration of Mgt of Aging Effects for Reactor Vessel Internals. Rept Provides Individual B&W Nuclear Power Plant Utility Owner with Technical Details for for License Application Renewal

ML20207C732
Person / Time
Issue date:	05/26/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20207C718	List: ML20207C712 ML20207C732
References
PROJECT-683 NUDOCS 9906030098
Download: ML20207C732 (23)

Text

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I DRAFT SAFETY EVALUATION (DSE)

BY THE OFFICE OF NUCLEAR REACTOR REGULATION CONCERNING

" DEMONSTRATION OF THE MANAGEMENT OF AGING EFFECTS  !

FOR THE REACTOR VESSEL INTERNALS" BABCOCK & WILCOX OWNERS GROUP REPORT NUMBER BAW-2248 PROJECT NO. 683 i

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r Tebb of Cont:nts 1.0 I NTR ODU CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 L 1.1 Babcock & Wilcox Owners Group Topical Report . . . . . . . . . . . . . . . . . . . . . . . . 1 1.2 Conduct of Staff Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.0

SUMMARY

OF TOPICAL REPORT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 l 2.1. Components and Intended Functions . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . 2 2.1.1 Intended Functions . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.1.2 Com ponents . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.2 Effects of Aging . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2.3 ' Aging Management Programs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.4 Time-Limited Aging Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.0 STAFF EVALUATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... ..... 5 3.1 Intended Functions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.2 Effects of Aging . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 3.2.1 C racki ng . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 3.2.2 Loss of M aterial . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 1 3.2.3 Reduction of Fracture Toughness . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 3.2.4 Loss of Closure integrity for Bolted Closures . . . . . . . . . . . . . . . . . . . . 9 3.2.5 Change of Dimension . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 j 3.2.6 S u m m a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 3.3 Aging Management Programs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . 10 i I

L 3.3.1 C racking . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 l 3.3.2 Loss of Material Due to Wear . . . . . .. .......................12 3.3.3 Reduction of Fracture Toughness . . . . . . . . . . . . . . . . . . . . . . . . . 13 3.3.4 Loss of Closure Integrity for Bolted Closures . . . . . . . . . . . . . . . 15

, 3.3.5 Change of Dimension . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 1

3.4 Time-Limited Aging Analyses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 l i 3.4.1 Fatigue - Cracking (Initiation and Growth) . . . . . . . . . . . . . . . . . . . . . . 17 3.4.2 Ductility - Reduction of Fracture Toughness . . . . . . . . . . . . . . . . . . . 18 i

l 4.0 CONC LU SION S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 4.1. Renewal Applicant Action items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 4.2 Topical Report Open items . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 5.0 R E F E R E N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 APPENDIX A: LIST OF CORRESPONDENCE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22

DRAFT S AFETY EVALUATION l

BY THE OFFICE OF NUQ, EAR REACTOR REGULATION I

CONCERNING

" DEMONSTRATION OF THE MANAGEMENT OF AGING EFFECTS FOR THE REACTOR VESSEL INTERNALS" BABCOCK & WILCOX OWNERS GROUP REPORT NO. BAW-2248 PROJECT NO. 683

1.0 INTRODUCTION

Pursuant to Section 50.51 of Title 10 of the Code of Federal Reaulations (10 CFR 50.51),

licenses to operate nuclear power plants are issued by the U.S. Nuclear Regulatory Commission (NRC) for a fixed period of time not to exceed 40 years; however, these licenses may be renewed by the NRC for a fixed period of time including a period not to exceed 20 years beyond expiration of the current operating license. The Commission's regulations in 10 CFR Part 54, (60 FR 22461) published on May 8,1995, set forth the requirements for the renewal of operating licenses for commercial nuclear power plants (Reference 1).

Applicants for license renewal are required by the license renewal rule to perform an integrated plant assessment (IPA). As specified in 10 CFR 54.21(a)(1), the first step of the IPA requires the applicant to identify and list structures and components that are subject to an aging management review (AMR). In addition,10 CFR 54.21(a)(3) requires the applicant to describe and justify the methods used to meet the requirements of 10 CFR 54.21(a)(1). Further, 10 CFR 54.21(a)(3) requires that, for each structure and component identified in 10 CFR 54.21(a)(1), the applicant demonstrate that the effects of aging will be adequately managed so that the intended function (s) will be maintained consistent with the current licensing basis (CLB) for the period of extended operation. Finally, the applicant must provide an evaluation of time-limited aging analyses (TLAAs) as required by 10 CFR 54.21(c), including a list of TLAAs, as defined in 10 CFR 54.3.

1.1 Babcock & Wilcox Owners Group Topical Report By letter dated July 29,1997, the Babcock & Wilcox Owners Group (B&WOG) Generic License Renewal Program (GLRP) submitted topical report BAW-2248, " Demonstration of the I Management of Aging Effects for the Reactor Vessel intemals" (Reference 2), for staff review i and approval. The purpose of the topical report is to provide a technical evaluation of the effects j of aging of the reactor vesselinternals and demonstrate that the aging effects within the scope i of the report are adequately managed for the period of extended operation associated with  !

license renewal. The topical report provides an individual Babcock & Wilcox (B&W) nuclear power plant utility owner in the GLRP with the technical details necessary for submitting an application for license renewal.

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2 1.2 Conduct of Staff Review The staff reviewed the B&WOG topical report to determine whether it satisfied the requirements set forth in 10 CFR 54.21(a)(3) and (c)(1). The staff issued requests for additional information (RAls) after completing the initial review. The B&WOG responded to the staff's RAls.

Requests for additional information, meeting summaries, and other correspondence are listed in j Appendix A.

2.0

SUMMARY

OF TOPICAL REPORT

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The B&WOG topical repoit, BAW-2248, contains a technical evaluation of aging effects related to B&W reactor vessel intemals components, and was provided to the staff to demonstrate that .

B&WOG member plant owners can adequately manage these effects of aging during the period ,

I of extended operation. This evaluation applies to the following B&WOG GLRP member plants:

. ' Arkansas Nuclear One, Unit 1 (ANO-1)

• Oconee Nuclear Station, Units 1,2, and 3 (ONS-1,-2,-3)

. Three Mile Island, Unit 1 (TMI-1)

The topical report also contains evaluations of TLAAs, as defined in 10 CFR 54.3, for the reactor  ;

vesselinternals. However, the topical report indicates that the TLAA of flaw growth acceptance prescribed in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI inservice inspection (ISI) program (Reference 3) is plant-specific, is not

- within the scope of the report, and will be resolved on a plant-specific basis.

In addition, the report indicates that inservice inspection programs identified in the ASME Code Section XI may need to be supplemented because certain components are not easily accessible using current technology.  ;

2.1. Components and Intended Functions 2.1.1 Intended Function 3 In Section 1.0 of the topical report, the following five intended functions for the Reactor Vessel intemals (RVI), and system components were identified based on the requirements of 10 CFR 54.4(a):

a Provide support and orientation of the reactor core (i.e., the fuel assemblies).

- Provide support, orientation, guidance and protection of the control rod assemblies.

- Provide a passageway for the distribution of the reactor coolant flow to the reactor core.

• Provide a passageway for support, guidance, and protection for the incore instrumentation.

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3 Provide a secondary core support for limiting the downward displacement of the core support structure in the event of a postulated failure of the core barrel. 4 l

1 The portions of the RVI that were identified in the topical report (BAW-2248) as not within the 1 scope of this topical report, are discussed in Section 2.5 of the topical report. These items are:

thermal shield, thermal shield upper restraint assemblies, and upper thermocouple guide tube assemblies. The staff believes that there is an additionalintended function of the RVI as discussed in Section 3.1 of the safety evaluation.

2.1.2 Components As described in the topical report, the RVI scope consists of two major structural sub-assemblies that are located within, but not integrally attached to (i.e., not welded to) the reactor pressure vessel (RPV). These major sub-assemblies are the plenum assembly (PA) and the core support assembly (CSA). For the purpose of defining materials, fasteners, construction, and assembly, the CSA can be further divided into three principal sub-assemblies; the core support shield assembly (CSS), the core barrel assembly (CBA), and the lower intemals assembly (LIA).

The mechanical fasteners (bolting) joining these sub-assemblies and associated items are within the scope of this topical report. The welds within the scope of the reactor vessel intemals report include the major structural welds that form or join the major sub-assembly cylinders and flanges and minor structural welds joining parts such as lifting lugs, support pipes and tubes to the major sub-assemblies. There are no pressure-retaining or pressure boundary welds within the scope of this topical report.

The control rod assemblies (CRA), fuel assemblies (FA), and the incore monitors (IMS) are not considered part of the RVI and are not covered in the topical report. The thermal shield and upper thermocouple guide tube assemblies are RVI items; however, it is concluded in the topical report that they do not perform intended functions as defined in 10 CFR Part 54 and are, therefore, not within the scope of the topical report. The staff disagrees with the conclusion that the thermal shield is not within the scope of Part 54 (see Section 3.1of this safety evaluation).

Portions of the Internal vent valve assemblies are active components that do not require an aging management review under 10 CFR 54.21. The surveillance specimen holder tube assemblies (SSHT) are not part of the RVI for the plants included in the report. As such, the l

. SSHT assemblies are not within the scope of the topical report. Physical and functional l' descriptions of the individualitems within each of the four principle sub-assemblies are presented in Sections 2.1 through 2.4 of the topical report. l l

2.2 Effects of Aoina Section 3.0 of the topical report discusses the aging effects applicable to the reactor vessel intemals described above for the period of extended operation for the participating B&W plants.

The topical report states that the following effects of aging could result in adverse impact or loss of any of the reactor vessel intemals intended functions :

. cracking (initiation and growth) l

. loss of material

.- reduction of fracture toughness

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4 loss of mechanical closure integrity (for bolted connections)

Table 3-2 of the topical report provides a detailed list of the subassemblies in each of the reactor vessel intamal assemblies, and identifies the aging effect applicable to each subassembly, as determined by the B&WOG's evaluations. These evaluations included a review of industry operating experience to identify past incidents of aging effects applicable to the reactor vessel intamals. This review is discussed in Section 3.5 of the topical report.

l The following is a summary of Table 3-2 of the topical report:

Maior RVI Assemblies ADDii eble Aaina Effects

' Plenum Assembly Cracking 3 Loss of material 1 Reduction of fracture toughness Loss of closure integrity Core Support Shield Assembly Cracking Loss of material Reduction of fracture toughness Loss of closure integrity Core Barrel Assembly Cracking Reduction of fracture toughness Loss of closure integrity Lower Intemals Assembly Cracking Loss of material Reduction of fracture toughness Loss of closure integrity 2.3 Aoina Manaaement Proarams Section 4.0 of the topical report discusses the B&WOG bases for demonstrating that the I applicat le aging effects identified in Section 3.0 of the topical report can be managed by existing programs at ANO-1, ONS-1,-2,-3, and TMI-1 during the periods of extended operation of those plants. Table 4-1 in the topical report provides a detailed summary of the existing programs that manage aging effects that are applicable to each subassembly of the four major reactor vessel intemals assemblies identified above. These programs are the following:  ;

1 ASME B&PV_ Code,Section XI, inservice inspection Program l

= ~ Reactor Vessel Intemals Aging Management Program (RVIAMP)

• Plant Technical Specifications for Vent Valve Bodies in Core Support Shield Assemblies in ANO-1, and TMI-1 )

. Pump and Valve in-Service Test Programs for Vent Valve Bodies in Core Support Shield Assemblies of ONS-1,-2,-3 i

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The topical report proposes that the RVIAMP supplement the ASME Section XI ISI program, since the report concludes that the inspection program required by Examination Category B-N-3 of the ASME Section XI program, subsection IWB, may not be adequate to detect aging effects for certain reactor vessel internal components. The RVIAMP addresses the specific aging effects of SCC, lASCC, (neutron) irradiation embrittlement and stress relaxation.

2.4 Time-Limited Aoino Analyses Section 4.5 of the topical report identifies the following TLAAs that are applicable to the reactor vessel internals, and presents the B&WOG's proposed aging management programs for each TLAA:

a Fatigue - Cracking (Initiation and Growth)

. Ductility - Reduction of Fracture Toughness However, the topical report indicates that the TLAA of flaw growth acceptance in accordance with the ASME Section XI ISI program (Reference 3) is plant-specific, is not within the scope of the report, and will be resolved on a plant-specific basis, in addition, the report indicates that inservice inspection programs identified in the ASME Code Section XI may need to be supplemented because certain components are not easily accessible using current technology.

3.0 STAFF EVALUATION The staff reviewed the topical report and additional information submitted by the B&WOG to determine if they demonstrated that the effects of aging of the reactor vessel internal .

components covered by the report will be adequately managed so that there is reasonable assurance that the components will perform their intended functions consistent with the CLB for j the period of extended operation, in accordance with 10 CFR 54.21(a)(3). This is the last step in  !

the IPA described in 10 CFR 54.21(a). l l

Besides the IPA, Part 54 requires an evaluation of TLAAs in accordance with 10 CFR 54.21(c).

The staff reviewed the topical report and additional information submitted by the B&WOG to determine if the TLAAs covered by the report were evaluated for license renewal in accordance with 10 CFR 54.21(c)(1).

1 3.1 Intended Functions The staff reviewed Sections 1.0 and 2.0 of the subject topical report to determine whether there is reasonable assurance that the RVi components and supporting structures subject to AMR have been identified in accordance with the requirerr.ents of 10 CFR 54.21(a)(1). This was accomplished as described below.

I As part of the evaluation, the staff determined whether the applicant has properly identified the systems, structures, and components within the scope of license renewal, pursuant to  ;

10 CFR 54.4. The staff reviewed portions of the Updated Final Safety Analysis Reports  ;

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(UFSARs) on the RVI for the applicable operating plants, and compared the information in the i . UFSARs with the information in the report to identify any portions of the RVI that the report did not identify as within the scope of license renewal. The staff then reviewed the structures and components not identified as within the scope of Part 54, and as described below, and requested the B&WOG to provide additional information and/or clarifications for a selected l number of structures and components to verify that they do not have any intended functions as I

delineated in 10 CFR 54.4(a). The staff also reviewed the UFSARs for any safety-related

system functions that were not identified as intended functions in the report to verify that l structures and components having intended functions were not omitted from consideration.

After completing the initial review, by letter dated December 2,1998, the staff issued RAls regarding the RVI, and by letter dated February 18,1999, the B&WOG provided responses to

_ those RAls. ' Section 1.4 of the topical report identifies the intended functions of the RVI. It does not include, " provide shielding for the RPV" as one of the intended functions. As a result of this omission of an RVI intended function, the components that support this intended function, namely, the thermal shield and the thermal shield upper restraint assemblies were omitted from the scope of license renewal, and were not identified as rei Ling AMR. NRC RAI # 1 pointed out this omission, and requested clarsit 1 tion. In response, .e B&WOG agreed that the function

" provide shielding for the RPV" was ne :luded within the scope of the report. Therefore, the B&WOG recommended in its respont ist each license renewal applicant that references l BAW-2248 must determine if the funr%n " provide shielding for the RPV" is an RVI intended function. If not an intended function, ti . !icense renewal applicant should provide justification for that conclusion. Should a license renewal applicant determine that the function " provide shielding for the RPV" is an intended function of the RVI, then the items that support this intended function, such as the thermal shield and the thermal shield upper restraint assemblies, must be identified and reviewed in accordance with 10 CFR 54.21(a)(3). This is Renewal Applicant Action item 3. The B&WOG also indicated in its response that BAW-2248 will be i revised to provide this clarification. Based on the supporting information in the UFSARs, and the 1 B&WOG's response to the staff's request for additional information, the staff has found no j additional omissions in the report and, therefore, concludes that there is reasonable assurance 4 that the report adequately identified those portions of the RVI and its associated (supporting) l structures and components that fall within the scope of license renewal in accordance with 10 CFR 54.4.

As discussed above, the staff has reviewed the information provided in Sections 1.0 and 2.0 of the subject topical report (BAW-2248) and the additional information provided by the B&WOG in response to the staff's RAls. Based on that review, the staff concluded that, except for the  ;

omission as discussed in the above paragraph, there is reasonable assurance that the topical l report has appropriately identified the portions of the RVI and the associated structures and components thereof, that are within the scope of license renewal and are subject to AMR in accordance with the requirements of 10 CFR Part 54.

i 3.2 Effects of Aaina As discussed in Section 2.2 above, the effects of aging evaluated in BAW-2248 included re' duction of fracture toughness, cracking, loss of material, and loss of mechanical integrity (for bolted connections). The B&WOG reviewed these aging effects for their applicability to the RVI

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assemblies within the scope of the report. The B&WOG reviewed RV internals service history of cracking of weld locations (due to mechanical failure, fatigue and other causes), loss of material

' (extemal wall thinning), and loss of closure integrity (wear and erosion). The B&WOG findings about these effects were incorporated into the aging management program.

3.2.1 Crackina Stress Corrosion Crackina (SCC) and Irradiation-Assisted Stress Corrosion Crackina (IASCC)

The topical report identifies cracking as a potential aging effect due to either SCC or IASCC.

SCC results from the synergistic action of a susceptible material subjected to tensile stresses in a corrosive environment, which is specific to the material. The material may be inherently susceptible, or can become sensitized during fabrication. The tensile stresses can be due to the operational loading or residual fabrication stresses. The environmental parameters considered to be critical in SCC are the dissolved oxygen, halide and sulfide contents in the coolant. lASCC is a mechanism in which the presence of the neutron irradiation can make the material more susceptible to SCC.

For SCC, the report uses reactor coolant chemistry control, in particular dissolved oxygen less than 5 ppb and halides less than 150 ppb, as the basis for generally ruling out SCC as potentially significant. The staff believes that RVI components will not be susceptible to SCC with coolant containing these dissolved elements because at these low values the coolant corrosive environment is sufficiently low to preclude SCC. For RVI botting applications, the topical report indicates that SCC is a potential aging effect due to the potential for occluded environment conditions in the crevice area typically associated with bolting. The specific applications cited as potentially susceptible to SCC are: core support shield to core barrel bolts, lower internal assembly to core barrel and thermal shield bolts, core barrel to thermal shield bolts, shell forging to flow distributor bolts, and Alloy 600 locking devices on the modified vent valve assembly.

For IASCC, the report uses a neutron fluence threshold of 1-2 x 102' n/cm2 (E>1 MeV) to determine susceptibility to IASCC. The NRC staff has reservations concerning this threshold fluence approach, and has proposed an aging management program which obviates the need for a threshold fluence consideration. RVI components determined to be susceptible to IASCC are: core barrel assembly base metal and welds, baffle to baffle and former bolts, core support  !

shield to core barrel bolts, lower intemals assembly to core barrel bolts, and upper and lower grid assembly base metals and welds. Of these components, baffle-former and baffle-baffle  ;

bolts are expected to be the first to exhibit indication of IASCC because they are nearest the core and have cracked in PWR plants.

Baffle Former Bolt Crackina The technical evaluation in the BAW-2248 report, Section 3.5 included a review of the historical i performance of the reactor vessel internals (RVI) to identify and assess past incidents of aging i effects applicable to RVI. The assessment included a review of information in the nuclear plant  ;

reliability data system (NPRDS), Licensee Event Reports (LERs) and NRC Generic Letters (GL), i information Notices (lN) and Bulletins. In the BAW-2248 report, B&WOG identified NRC IN 91- i l

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. 05 as providing information regarding cracking in Alloy A-286 bolts used in reactor coolant pumps and the B&W-designed RVI. However, the RVI historical performance review does not include the aging effects applicable to RVI baffle bolting described in the more recently issued }

NRC IN 98-11, " Cracking of Reactor Vessel Intemal Baffle Former Bolts in Foreign Plants,"

issued on March 25,1998.

3.2.2 . Loss of Material Loss of Material Due to Wear in Section 3.3 of the topical repcrt, B&WOG identifies the RVI items that are subject to loss of material due to wear. The items identified include the fuel assembly support pads on the upper and lower grid assemblies, the plenum rib pads, the guide blocks on the lower grid, the top flange on the core support cylinder, and the locking devices on the original vent valve assembly.

Wear occurs as a result of movement at mating surfaces that may result from flow-induced vibration during plant operation and differential thermal expansion and contraction movements during plant heat-up, cool-down and changes in power operating cycles. The resulting relative movement between the interfacing and mating surfaces causes wear. The severity of the wear depends upon the frequency, duration of the motion and the loads imposed on the affected surfaces. The RVI items identified as subject to wear are typical of the RVI construction items found in locations of structural interfaces and mating surfaces that experience relative motion during plant operation. The identified RVi items subject to wear require programmatic aging management.

Loss of Material Due to Corrosion The topical report cites three possible mechanisms for loss of material due to corrosion. These mechanisms are: (1) erosion and erosion-corrosion, (2) uniform attack / general corrosion, and, l (3) pitting and crevice corrosion.

Erosion and erosion-corrosion are not considered to be applicable since all of the RVI components are fabricated from stainless steel or nickel-base alloys, and these materials have j

' been found to be resistant to erosion and erosion corrosion.

Uniform attack / general corrosion are not considered to be applicable since all of the RVI components are fabricated from stainless steel or nickel-base alloys, and these materials have been found to be resistant to general corrosion due to protective passivation layers which mitigate the susceptibility of these materials.

Pitting and crevice corrosion are not considered to be applicable due to the low oxygen levels in the reactor coolant as a result of the water chemistry controls.

Based on the RVI materials and reactor coolant environment, the staff concludes that loss of material due to corrosion is not considered to be an applicable aging effect for any of the RVI components.

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9 3.2.3 Badyction of Fracture Touchness The topical report identifies reduction of fracture toughness in RVI components as an applicable aging effect due to either thermal embrittlement or neutron irradiation embrittlement. Thermal embrittlement can occur in cast austenitic stainless steel (CASS) and martensitic stainless steel parts exposed to high temperatures typical of reactor operating conditions. Neutron irradiation embrittlement occurs in all steels due to exposure to high neutron flux conditions typical of many RVi components.- Both of these mechanisms result in increased hardness and tensile strength, along with reduced ductility, impact strength, and fracture toughness of the material.

For RVI components fabricated from CASS and hence subject to thermal embrittlement, concurrent exposure to high neutron fluence levels can result in a synergistic effect wherein the cervice-degraded fracture toughness is reduced from the levels predicted independently for either of the mechanisms. Therefore, components determined to be subject to thermal embrittlement require an additional consideration of the neutron fluence of the component to determine the full range of degradation mechanisms applicable for the component.

Thermal embrittlement was determined to be applicable to intemal vent valve bodies, vent valve retair;inp rings, control rod guide tube (CRGT) and incore guide tube spiders, and the core support shield outlet nozzles at Oconee Nuclear Station Unit 3. All of these components are fabricated from CASS, except for the vent valve retaining rings, which are composed of precipitation-hardened stainless steel.

Determination of RVI components subject to neutron irradiation embrittlement was handled in the topical report using a fluence threshold to screen-out components with a neutron fluence level below 5 x 102 n/cm: (E>1 MeV). Components found to be subject to neutron irradiation embrittlement include the upper grid assembly, core support shield to core barrel bolts, core barrel assembly, baffle-baffle and baffle-former bolts, lower intemal assembly to core barrel bolts, and the lower grid assembly. The NRC staff has reservations conceming this threshold fluence approach: however, the proposed aging management program would obviate the need for a threshold fluence consideration (See Section 3.3 of this safety evaluation).

3.2.4 Loss of Closure Inteority for Bolted Closures in Section 3.4 of the topical report, the B&WOG indicates that bolting stress relaxation is j considered an applicable aging mechanism for those components where maintaining a preload  ;

is important to the structural integrity function (s) of the RVI. These RVI bolts include: the control rod guide tube (CRGT) to upper grid fasteners; the core support shield to core barrel bolts; the core barrel to thermal shield bolts; lower intemals assembly to core barrel bolts; the lower grid  ;

rib-to-shell fasteners; the shell forging to flow distributor bolts; and the lower internals assembly to thermal shield bolts.

3.2.5 Chance of Dimension Section 3.1 of the topical report dismisses change of dimension of the RVI components due to void swelling as a significant aging effect due to the lack of evidence of void swelling under PWR conditions. However, EPRI TR-107521 (Reference 4) cites several sources with conflicting results. One source predicts swelling as great as 14% for PWR baffle-former assemblies over a 40-year plant lifetime, whereas results from another source indicate that swelling would be less i

10 than 3% for the most highly irradiated sections of the intemals at 60 years. The issue of concem to the staff is the impact of change of dimension due to void swelling on the ability of the RVI to perform their intended function. The B&WOG should identify; (1) How much of a change in dimension would be required before the intemals would not be able to meet their intended function; (2) What ongoing programs, if any will evaluate the impact of the void swelling on the intended function of the intemals; and (3) When these programs will provide data to determine whether void swelling could impact the intended function of the intemals.

Should it be determined that change of dimension by void swelling can impede the ability of the RVI to perform their intended functions, then an appropriate aging management program would be required to assure that the need for corrective actions can be properly identified. The determination of the need for an aging management program for changes in dimension is Topical Report Open item 1.

3.2.6 Summary With the exception of cnanges in dimension due to void swelling, the staff agrees with the B&WOG identification of applicable RV component aging effects that are subject to aging management as a condition of license renewal. The staff finds that the B&WOG should address the aging effect of change of dimension due to the aging mechanism of void swelling, as described in Topical Report Open Item 1.

3.3 Aoina Manaaement Proarams -

As described in Section 2.3, the aging management programs discussed by the B&WOG include the RVIAMP, ASME Section XI requirements, plant technical specifications, plant-specific test programs and licensee commitments in response to NRC generic communications.

Applicants for license renewal will be responsible for describing any such commitments and identifying the appropriate regulatory control.

The principal change to aging management programs is the addition of the RVIAMP. The

_ generic UFSAR supplement provided as Appendix A to the topical report BAW-2248 states that an applicant will continue to investigate the potential aging effects identified in BAW-2248, through the RVIAMP, and will establish appropriate monitoring and inspection programs prior to the expiration of the current license. In addition, an applicant would be required to provide annual written status reports to the NRC on the RVIAMP beginning one year after issuance of the renewed license.

3.3.1 Crackina Stress Corrosion Crackina (SCC) and Irradiation-Assisted Stress Corrosion Crackina (IASCC)

Management of SCC and IASCC is achieved through two means. The first means for management of cracking is through the existing inservice inspection (ISI) program, which requires visual VT-3 examination in accordance with Examination Category B-N-3 of Section XI of the ASME Code. However, the topical report indicated that this visual examination may not be adequate to detect cracking for all of the susceptible RVI components due to accessibility concerns, except for the modified vent valve assembly. This concem is addressed through the

- i 11 second means to manage cracking, which is a planned Reactor Vessel Intemals Aging Management Program (RVIAMP). The purposes of this program are to continue the i investigation of the potential aging effects that have been identified in the topical report for the RVI, and to establish appropriate monitoring and inspection programs that will continue to ,

maintain the RVI in a functional state during the period of extended operation. Further, in ,

response to NRC RAI #3, the B&WOG indicated that an industry group, the PWR Materials ~;

Reliability Project (MRP) is addressing neutron embrittlement of RVI components under the auspices of an RPV Internals issue Task Group. The RAl response stated that the results of this MRP activity will be incorporated in the development of the RVIAMP.

The NRC staff proposed to the B&WOG a modified approach to manage cracking of RVI components. In particular, a two-pronged approach was proposed. The two pronged approach is to perform inspection and to perform tests and analysis of irradiated material. The inspection part of the approach is a supplemental (enhanced VT-1) examination of the components believed to be the limiting components for cracking, considering both the susceptibility of the 1 component to the aging mechanism as well as the material properties (in particular the fracture toughness) and the operating stresses on the component. These examinations would be included as part of the 10-year ISI program. This supplemental examination (enhanced VT-1) applies to all RVI components except for bolting. Aging management for the limiting RVI bolting, i baffle former bolting, is described below. I Since the examination addresses the limiting components, plant-specific neutron fluence evaluations are not necessary. Initial consideration by the B&WOG indicated that the limiting .I components with respect to highest neutron fluence were the baffle plates and baffle former bolts. The B&WOG must identify the limiting components and modify the ISI program for j cracking in the topical report accordingly. This is Topical Report Open item 2.

The second part of this approach would determine the need for continuing the supplemental {

examination.- Should data or evaluations from the MRP or any other industry research activity indicate that the supplemental (enhanced VT-1) examinations can be modified or possibly eliminated, each applicant would be required to provide plant-specific justification to demonstrate the basis for the modification or elimination.

Baffle Former Bolt Crackina in BAW-2248, Section 3.1, and Table 3-2, the B&WOG identifies baffle former bolting as a component subject to aging effects. Section 4.1 of the report describes the demonstration of aging management for cracking and identifies baffle former bolts as RVi items that are subject to programmatic aging management, and which require Category B-N-3 examination in accordance with the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, ISI Program, Subsection IWB. Subsection lWB provides requirements for the visual inspection of removable core support structures. Further, the B&WOG indicates that visual inspections may not be adequate to detect bolting cracking that is inaccessible. Therefore, the B&WOG implemented a program to manage the effects of aging due to cracking as described in Section 4.6 of the

. topical report.

In BAW-2248, Section 4.6, the B&WOG outline generalized activities of the implemented program. One of the defined purposes of the aging management program, according to the i

j

1 12 report, is to establish appropriate monitoring, inspection techniques and inspection programs

- that will continue to maintain the RVI functional through the extended life of the plant.

During meetings with the staff subsequent to the submittal of the BAW-2248, the B&WOG described current and ongoing bame bolt activities that included preparation for a possible augmented beme bolt inspection during the next 10-year interval at a B&W lead plant.

By letter dated December 2,1998 (Reference 5), NRC forwarded requests for additional information (RAl) with regard to bame bolt cracking, NRC RAI #12 and RAI #13.

NRC RAI #12 requested the B&WOG to describe the bame bolt inspections that will be conducted prior to the start of the period of extended operation and indicate how these actions providea basis for assuring that the baffle bolt monitoring and inspection techniques that are planned for implementation during the period of extended operation are appropriate.

NRC RAI #13 requested B&WOG to describe the program that will be implemented as outlined in Section 4.6 of BAW-2248 with regard to the aging meagement of reactor internals baffle bolts, and to describe the overall inspection program, including aspects such as intervals, monitoring and inspection techniques.

By letter dated February 18,1999, (Reference 6), the B&WOG provided a response to the RAls.

The response to NRC RAI #4 indicates that the technical elements of the B&WOG RVI aging management program were presented during a meeting with the NRC on April 23,1998. Since that meeting, the industry has initiated a project to address generic materials issues and the scope of the B&WOG RVI aging management program has changed. The PWR MRP was established during the second quarter of 1998 to address and resolve existing and emerging PWR materials issues. An lasues Task Group (ITG) was formed to manage emerging RVI materials issues. The ITG on reactor vessel intemals is currently addressing the issue of

cracking, reduction of fracture toughness and loss of preload related to baffle bolts and associated materials. The data and information acquired from these various ITG activities will be used to determine the necessary steps in managing the effects of aging on bame bolts, including future inspection plans. These plans are expected to be outlined on a plant specific basis, possibly beginning with the inspection at Oconee Unit 1 during its fourth inservice inspection (ISI) interval.

- Each renewal applicant will be responsible for using the tools provided by both the ITG and owners groups to determine the necessary steps (e.g., inspections, operability determinations, and replacements) to manage the applicable beme bolt aging effects. Therefore, the requested information of the B&WOG in RAI #12 and RAI #13, with regard to the management of the effects of aging on beme bolts is converted to a renewal applicant action item based on this transfer of responsibility from the B&WOG to individual applicant. This is Renewal Applicant Action item 4.

3.3.2 Loss of Material Due to Wear

The B&WOG proposes to continue the ASME B&PV Code Section XI program to manage the loss of material of RVI items due to wear that could cause loss of the RVI function (s) during the period of extended operation. The RVI items are managed by Examination Category B-N-3 of

13 inspection (VT-3) for removable RVI structures. These requirements define conditions in lWB-3520.2 which, if detected, must be corrected prior to continued service. These conditions include wear of mating surfaces that may lead to loss of function. Because of our experience with inspections performed in accordance with ASME B&PV Code,Section XI, Subsection lWB, the staff concludes that VT-3 inspections are capable of detecting the loss of material due to wear.

3.3.3 Reduction of Fracture Touchness Thermal Embrittlement The topical report stated that thermal embrittlement of all RVI components is managed by visual inspection (VT-3) in accordance with Examination Category B-N-3 of the ASME Section XI inservice inspection program, Subsection IWB. Aging management for vent valve bodies and retaining rings is also accomplished through vent valve testing and (visual) inspection requirements (at each refueling outage) in accordance with plant technical specifications at ANO-1 and TMI-1, and the Pump and Valve in-Service Test Program at ONS-1, -2, and -3.

NRC staff does not agree that loss of fracture toughness can be managed through VT-3 inspection, and instead has proposed an attemative management methodology. VT-3 inspection may not be capable of detecting fine cracks that could lead to failure of thermally embrittled components.

The vent valve retaining rings would be subject to supplemental (enhanced VT-1) examination.

This examination could be modified or eliminated, provided that the applicant can demonstrate through data (including microstructural considerations) and evaluation that loss of fracture toughness by thermal embrittlement and/or neutron irradiation embrittlement is not significant for the vent valve retaining rings. Such a demonstration could follow the same frame work as that proposed below for CASS RVI components.

The RVI components fabricated from CASS are potentially subject to a synergistic loss of fracture toughness due to the combination of thermal and neutron irradiation embrittlement. This enhanced loss of fracture toughness is not accounted for by current CASS screening criteria either in BAW-2243A nor in revisions to EPRI TR 106092 (Reference 7). To account for this synergistic loss of fracture toughness, a modified approach for CASS RVI components is proposed. This modified approach consists of either a supplemental (enhanced VT-1) examination of the affected components as part of the applicant's 10-year ISI program during the license renewal term, or a component-specific evaluation to determine the susceptibility to loss of fracture toughness. The proposed evaluation will first examine the neutron fluence of I the component. If the neutron fluence is greater than 1 x 10" n/cm2(E > 1 MeV), a mechanical loading assessment would be conducted for the component. This assessment will determine the maximum tensile loading on the component during ASME Code Level A, B, C and D ,

conditions. If the loading is compressive or low enough to preclude fracture of the component, l then the component would not require supplemental inspection. Failure to meet this criterion i would require continued use of the supplemental (enhanced VT-1) inspection. If the neutron l fluence is less than 1 x 10" n/cm 2(E > 1 MeV), an assessment would be made to determine if the affected component (s) are bounded by the screening criteria in EPRI TR-106092 (Reference 7), modified as described below. In order to demonstrate that the screening criteria in EPRI TR- l 106092 (Reference 7) are applicable to RVI components, a flaw tolerance evaluation specific to ,

I

i 14' the reactor vesselinternals would b3 performed. If the screening criteria are not satisfied, then a supplemental (enhanced VT-1) inspection will be performed on the component.

The CASS components should be evaluated to the criteria in EPRI TP 106092 with the following i additional criteria:

Statically cast componevits with a molybdenum content meeting the requirements of SA-351 Grades CF3 and CF8 and with a delta ferrite content less than 10 percent will not need supplemental examination.

Ferrite levels will be calculated using Hull's equivalent factors or a method producing an equivalent level of accuracy (16 percent deviation between measured and calculated values).

Cast stainless components containing Niobium are subject to supplemental examination.

Flaws in CASS with ferrite levels less than 25 percent and no niobium may be evaluated using ASME Code IWB-3640 procedures.

Flaws in CASS with ferrite levels exceeding 25 percent or containing niobium must be evaluated using ASME Code IWB-3640 procedures. If Flaws are discovered in such components, fracture toughness data must be provided on a cas(-by-case basis.

Components that have delta ferrite levels below the screening criteria have adequate fracture toughness and do not require supplementalinspection. Components that have delta ferrite levels exceeding the screening criteria may not have adequate fracture toughness, as a result of 1 thermal embrittlement, and do require supplemental inspection. l This proposed program was discussed with the B&WOG. It needs to be incorporated into the topical report. This is Topical Report Open item 3.

The B&WOG methodology for estimating neutron fluence is contained in topical report BAW-2241. This methodology was approved by the staff; however, it is only applicable for calculating neutron fluence in the radial direction between the core edge to the reactor vessel cavity. Hence, the methodology can not be applied to RVI components above and below the core.

To determine whether CASS components are above or below the threshold value of 1 x 10" n/cm , as discussed in Section 3.3.3 of the safety evaluation, the B&WOG must provide estimates of the neutron fluence of each CASS component at the expiration of the license renewal term, identify the method of determining the neutron fluence and provide justification for applicability of the method to components above or below the core. This is Topical Report Open item 4.

Neutron Embrittlement For non-CASS RVI components, the topical report stated that management of neutron embrittlement would be addressed through a planned RVIAMP. The purposes of this program

15 are to continue the investigation of the potential aging effects that have been identified in the topical report for the RVI, and to establish appropriate monitoring and inspection programs that will continue to maintain the RV1 functional through the period of extended operationt. Further, in response to an RAI, the B&WOG indicated that an industry group, the PWR MRP, is addressing neutron embrittlement of RVI components under the auspices of an RPV Intemals issue Task Group. The RAI response stated that the results of this MRP activity will be incorporated in the development of the RVIAMP.

The NRC staff proposed to the B&WOG a modified approach to manage neutron embrittlement of RVI components. In particular, a two-pronged approach was proposed. The two pronged approach is to perform inspections and to perform iests and analysis of irradiated material. The inspection part of the approach is supplemental (enhanced VT-1) examination of the components believed to be the limiting components for neutron irradiation embrittlement, considering both the susceptibility of the component to the aging mechanism as well as the  !

material properties (in particular the fracture toughness) and the operating stresses on the 1 component. These examinations would be included as part of the 10-year ISI program. Since the examination addresses the limiting components, plant-specific neutron fluence evaluations j are not necessary. Initial consideration by the B&WOG indicated that the limiting components I with respect to highest neutron fluence were the baffle plates and baffle former bolts. The i B&WOG must identify the limiting components and incorporate this program into the topical l report. This is Topical Report Open Item 5. Note that the supplemental (enhanced VT-1) examination does not apply to bolting due to accessibility limitation. Section 3.3.1 of the safety l evaluation describes aging management for baffle former bolting.

The se cond part of this approach would be to determine the need for continuing the ,

supplemental examination. Should data or evaluations from the MRP or any other industry  !

research activity indicate that the supplemental (enhanced VT-1) examinations can be modified or possibly eliminated, each applicant would be required to provide plant-specific justification to demonstrate the basis for the modification or elimination of examinations.

3.3.4 Loss of Closure Intearity for Bolted Closures '

In Section 3.4 of the topical report, the B&WOG indicates that bolting stress relaxation b considered an applicable aging mechanism for those components where maintain;ng a preload  !

is impodarit to the structural integrity function (s) of the RVI. These RVI bolts include: the control rod guide tube (CRGT) to upper grid fasteners; the core support shield to core barrel bolts; the core barrel to thermal shield bolts; lower intemals assembly to core barrel bolts; the lower grid rib-to-shell fasteners; the shell forging to flow distributor bolts; and the lower intemals assembly to thermal shield bolts In Section 4.4 of the topical report, the B&WOG indicates that the required programmatic mmagement for these bolts is the VT-3 visual examination required by Examination Category B&3 of the ASME B&PV Code,Section XI ISI Program, Subsection IWB. The VT-3 visual examination is specificciij designed to determine the general mechanical and structural conditions, including structural distortion and displacements, loose or missbg parts, and wear of mating surfaces that may lead to the loss of integrity at bolted connections. lWB-3142 provides options for correcting the relevant condition (s), such as: (1) acceptance by supplemental surface and/or volumetric examint Ln, in order to further characterize the condition,

1. ,

• 3 w

16 1

~'(2) acceptance by corrective measures (i.e., re-establishing the preload) or repairs;

- or (3) acceptance by replacement of the item. However, the B&WOG indicates that it recognizes that visual examination may not be adequate to detect the loss of mechanical closure integrity of the RVI, and the GLRP has implemented a program to manage these aging )

effects as discussed in Section 4.6 of the topical report. ]

In Section 4.6 of the topical report, the B&WOG indicates a comprehensive aging management program will be developed to supplement the programmatic management for stress relaxation of RVI botting contained in Section 4.4. The supplemented AMP will be implemented such that the

]

RVI can perform their component intended function (s) for the period of extended operation. The )

B&WOG indicates the purpose of the program is to: (1) continue the investigation of the aging effects; and (2) establish the appropriate monitoring and inspection programs to continue to maintain the RVI functions. The B&WOG indicates the elements of the program that will be implemented to address stress relaxation of RVI bolting are: (1) to determine critical locations; and (2) establish appropriate monitoring and inspection techniques.

in the topical report, B&WOG indicates that the commitments to implement this AMP and to notify the NRC staff regarding the status of the program activities on a regular basis will be part of the updated FSAR supplement and will be included in any plant-specific license renewal application.

Since the B&WOG has proposed a comprehensive program to manage stress relaxation of RVI components, the staff considers the B&WOG proposal acceptable subject to applicant's fulfilment of the commitments included in the report. This is Renewal Applicant Action item 1.

3.3.5 Chanae of Dimension - 1 i

Should the evaluation described in Topical Report Open item 1 demonstrate the need for an )

AMP, then the topical report would require revision to identify the elements of the AMP. This is addressed as part (c) of Topical Report Open item 1.

I 3.4 Time-Limited Aoina Analyses ' l l

I Time-limited aging analyses are defined in 10 CFR 54.3 as those licensee calculations and analyses that:

(1) involve systems, structures, and components within the scope of !icense renewal, as stated in 10 CFR 54.4(a);

l (2) consider the effects of aging; (3) involve time-limited assumptions defined by the current operating term, for example, I 40 years; (4) were determined to be relevant in making a safety determination;

r. . -)

l l

.17 l l (5) involve conclusions or provide the bases for conclusions related to the capability of the I

system, structure or component to perform its intended functions, as stated in 10 CFR 54.4(b); and '

(6) are contained or incorporated by reference in the current licensing basis. 4 Paragraph 54.21(c)(1) requires the applicant to demonstrate that:

-(i) the analyses remain valid for the period of extended operation; (ii) the analyses have been projected to the end of the period of extended operation; or .

(iii) the effects of aging on the intended functions (s) will be adequately managed for the period of extended operation.

In Section 4.5 of the topical report, the B&WOG identified the TLAA applicable to the RVI items within the scope of the report based on reviewing plant-specific docketed correspondence files, .

plant-specific FSARs, BAW topical reports, and the ASME B&PV Code,Section XI ISI requirements. The topical report identifies four TLAAs appplicable to the RVI:

a flow-induced vitration calculations and measurements to verify during hot function testing that the flow induced vibration stresses are below the endurance limit a fatigue cumulative usage factor calculations, which rely on transient cycle count assumptions used for the design of the RVI calculation to demonstrate the deformation limits in the ASME Code are not violated as a result of neutron irradiation of RVI materials

-= flaw growth calculations performed in accordance with the ASME B&PV Code,Section XI ISI requirements  ;

I The first two of these TLAAs are lumped together in the topical report under the heading

" Fatigue - Cracking (initiation and Growth)," and the third is referred to in the topical report as j

" Ductility - Reduction of Fracture Toughness." The flaw growth acceptance TLAA is identified in  !

the topical report as requiring a plant-specific evaluation, and as such is not evaluated within the  ;

topical report. This is Renewal Applicant Action item 5. J All of the TLAA analyses documents were identified in the topical report as required by 10 CFR 54.21(c)(1).

3.4.1 Fatiaue - Crackina (Initiation and Growth)

The flow-induced vibration endurance limit assumptions were based on 1052 cycles for 40 years.

^

The analysis was extended for the license renewal period by conservatively increasing the number of cycles to 10'8, and then determining the endurance limit using the latest ASME fatigue curves. The component stress values were found to be less than the endurance limit, I

i 18 rendering the evaluation acceptable, in conformance with the requirements of 10 CFR 54.21(c)(1).

In the topical report, B&WOG indicates that the design cyclic loadings and thermal conditions used for the analyses are defined in the component design specifications and that flow-induced vibration input used was obtained from hot functional testing data contained in the listed analyses documents. The ability to withstand cyclic loading without fatigue failure was evaluated using a cumulative usage factor methodology. For each utility the number of transients accrued to date was conservatively extrapolated, and in all cases it was found that the number of design cycles would not be exceeded in the period of extended operation. The B&WOG reported that each of the participating utilities monitors occurrences of design transients and is thus managing the potential for cracking resulting from fatigue. The topical report indicates that the plants must continue to monitor and track occurrences of design transients.

l 3.4.2 Ductility - Reduction of Fracture Touchness Section 4.5.2 of BAW-2248 describes a TLAA related to the acceptability of the reactor vessel l internals under loss-of-coolant-accident (LOCA) and seismic loading. The topical report states that Appendix E to BAW-10008, Part 1, Rev.1, conciudes "that at the end of 40 years, the l internals will have adequate ductility to absorb local strain at the regions of maximum stress  !

intensity, and that irradiation will not adversely affect deformation limits." The topical report j indicates that this TLAA will be resolved on a plant-specific basis per 10 CFR 54.21 (c)(1)(iii) l based on the results and conclusion of the planned RVIAMP. As described in the topical report, the planned RVIAMP program will provide the data necessary to resolve this TLAA.

Therefore, this item should be addressed as a renewal applicant action item on a plant-specific basis pending the results of the RVIAMP. This is Renewal Applicant Action item 6.

4.0 CONCLUSION

S The staff has reviewed the subject B&WOG topical report (Reference 2) and additional information submitted by the B&WOG On the basis of its review, the staff concludes that the B&WOG topical report provides an acceptable demonstration that aging effects on reactor vessel intemal components within the scope of this topical report will be adequately managed for the GLRP member plants, with the exception of the noted renewal applicant action items and topical report open items, so that there is reasonable assurance that the reactor vessel intemal components will perform their intended functions in accordance with the CLB during the period of extended operation. The staff also concludes that, upon completion of the renewal applicant action items set forth in Section 4.1 below, the B&WOG topical report provides an acceptable evaluation of time-limited aging analyses for the reactor vesselinternals for the GLRP member plants for the period of extended operation.

Any B&WOG GLRP member plant may reference this topical report in a license renewal application to satisfy the requirements of (1) 10 CFR 54.21(a)(3) for demonstrating that the effects of aging on the reactor vesselintemal components within the scope of this topical report will be adequately managed and (2) 10 CFR 54.21(c)(1) for demonstrating that appropriate findings be made regarding evaluation of time-limited aging analyses for the reactor vessel intemals for the period of extended operation. The staff also concludes that, upon completion of l

)

r ..

19 the renewal applicant action items set forth in Section 4.1 below, referencing this topical report in ,

' a license renewal application and summarizing in an FSAR supplement the aging management programs and the TLAA evaluations set forth in the topical report will provide the staff with sufficient information to make the necessary findings required by 10 CFR 54.29(a)(1) and (a)(2) for components within the scope of this topical report.

4.1 Renewal Anolicant Action items The following are license renewal applicant action items to be addressed in the plant-specific license renewal application when incorporating the B&WOG topical report in a renewal application:

(1)- The license renewal applicant is to verify that the critical parameters for the plant are l

. bounded by the topical report. Further, the renewal applicant is to commit to programs l described as necessary in the topical report to manage the effects of aging during the period of extended operation on the functionality of the reactor vessel components.

Applicants for license renewal will be responsible for describing any such commitments and identifying the appropriate regulatory control. Any deviations from the aging management programs within this topical report described as necessary to manage the effects of aging during the period of extended operation and to maintain the functionality of the reactor vessel internal components or other information presented in the report, such as materials of construction, will have to be identified by the renewal applicant and evaluated on a plant spacific basis in accordance with 10 CFR 54.21(a)(3) and (c)(1). j (2) A summary description of the programs and evaluation of TLAAs is to be provided in the license renewal FSAR supplement in accordance with 10 CFR 54.21(d).

(3) License renewal applicants must identify whether the intended function of the RVI is to provide shielding for the RPV. If not an intended function, the license renewal applicant should provide justification for that conclusion. Should a license renewal applicant determine that the RVI's intended function is to provide shielding for the RPV, then the items that support this intended function, such as, the thermal shield and the thermal shield upper restraint assemblies, must be identified and reviewed in accordance with 10 CFR 54.21(a)(3).

(4) According to B&WOG, one of its objectives in BAW-2248 states, "It is intended that NRC l ' review and approval of this report will allow that no further review of the matters described herein will be needed when the report is incorporated by reference in a plant specific renewal license application." The license renewal applicant must address matters not described in the report, such as the baffle former bolt cracking issues addressed in Section 3.3.1 of this SE pertaining to References 5 and 6, with regard to the industry ITG project, initiated after April 23,1998, to address generic RVI materials issues. The BWOG indicates this industry effort resulted in subsequent changes in the BWOG RVI aging management program. The ITG is currently addressing the issues of cracking of baff!e bolts. The BWOG indicates that the changes in the aging management program now requires the applicants to be responsible for using the industry ITG project developed information to determine the necessary steps (e.g., inspection, operability determinations, and replacements) in managing the effects of aging on baffle bolts.

I

. 1 I

i 20 v

(5) If flaws have been detected in the reactor vessel intemals, a TLAA plant-specific evaluation must be performed to determine whether the flaw growth is acceptable in accordance with the ASME B&PV Code,Section XI, inservice inspection requirements at the expiration of the renewed license.

(6) Plant-specific analysis is required to demonstrate that, under loss-of-coolant-accident (LOCA) and seismic loading and with irradiation accumulated at the expiration of the renewal license, the intemals have adequate ductility to absorb local strain at the regions of maximum stress intensity and will meet deformation limits. The data to demonstrate that the intemals will meet the deformation limits at the expiration of the renewed license will be developed from the RVAMP.

4.2 Topical Reoort Open items (1) The B&WOG should identify:

(a) How much of a change in dimension would be required before the internals would not be able to meet their intended function, (b) What ongoing programs, if any will evaluate the impact of the void swelling on the intended function of the internals; and (c) When these programs will provide data to determine whether void swelling could impact the intended function of the internals.

(2) The B&WOG must modify their aging management program as described in Section 3.3.1 of the safety evaluation for managing the effects of cracking (SCC and lASCC). One acceptable option is the program described in Section 3.3.1. In addition, ,

the B&WOG must identify the limiting components (excluding RVI bolting).

(3) The B&WOG must modify their aging management program as described in Section 3.3.3 of the safety evaluation, under" Thermal Embrittlement" for managing the effects of thermal embrittlement, and possibly synergistic effects with neutron embrittlement, on the fracture toughness of cast austenitic stainless steel (CASS) RVI components. One acceptable option is the program described in Section 3.3.3 under

" Thermal Embrittlement." .

(4) To determine whether CASS components are above or below the threshold neutron fluence value of 1 x 10"n/cm ,2 as discussed in Section 3.3.3, the B&WOG must provide estimates of the neutron fluence of each CASS component at the expiration of the license renewal term, identify the method of determining the neutron fluence and provide justification for applicablity of the method to components above or below the core.

(5) The B&WOG must modify their aging management program as described in Section 3.3.3 under " Neutron Embrittlement" for managing the effects of neutron embrittlement on the fracture toughness of RVi components. One acceptable option is the program described in Section 3.3.3 of the safety evaluation, under " Neutron Embrittlement." in addition. The B&WOG must identify the limiting components

( excluding RVi bolting).

21 l

5.0 REFERENCES

l

1. 10 CFR Part 54, " Requirements for Renewal of Operating Licenses for Nuclear Power Plant.s," Federal Reaister, Vol. 60, No. 88, pp. 22461-22495, May 8,1995.

2. BAW-2248, " Demonstration of the Management of Aging Effects for the Reactor Vessel Internals," Babcock & Wilcox Owners Group, July 1997.

3. Boiler and Pressure Vessel Code,Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components," American Society of Mechanical Engineers,1989.

4. EPRI Technical Report TR-107521, " Generic License Renewal Technical Issues Summary,"

Electric Power Research Institute, April 1998.

5. Letter from Raj K. Anand, NRC, to David J. Firth, December 2,1998, " Request for Additional Information Regarding the Babcock & Wilcox Owners Group Generic License Renewal Program Topical Report Entitled Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, BAW-2248, July 1997."

6. Letter from William R. Gray to David B. Mathews, NRC, dated February 18,1999, "B&WOG Generic License Renewal Program Topical Report BAW-2248, " Demonstration of the Management of Aging Effects for the Reactor Vessel Internals" (RAls 1 through 14 from

- December 4,1998)."

7. EPRI Technical Report TR-106092, " Evaluation of Thermal Aging Embrittlement for Cast Austenitic Stainless Steel Components in LWR Reactor Coolant Systems," Electric Power Research Institute, September 1997. i 4

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l APPENDIX A l

l LIST OF CORRESPONDENCE t 1. Letter from David J. Firth (B&WOG) to Marylee Slosson (NRC), July 29,1997, transmitting B&WOG Generic License Renewal Program Topical Report BAW-2248,

" Demonstration of the Management of Aging Effects for the Reactor Vessel Internals,"

July,1997,

2. Letter from Raj K. Anand (NRC) to David J. Firth (B&WOG), December 2,1998,

" Request for Additional information Regarding the Babcock & Wilcox Owners Group Generic Licensee Renewal Program Topical Report Entitled Demonstration of the Management of Aging Effects for the Reactor Vessel Internals, BAW-2248, July 1997."

3 Letter from William R. Gray (B&WOG) to David B. Mathews (NRC) February 18,1999, "B&WOG Generic License Renewal Program Topical Report BAW-2248, " Demonstration

]

of the Management of Aging Effects for the Reactor Vessel internals" (RAls 1 through 14 !

from December 41998)."

h 4 NRC Meeting Summary dated May 6,1998, entitled " Summary of Meeting on April 23,  !

1998, between the U.S. NRC staff and B&WOG representatives to discuss the status of

{

the B&WOG Generic License Renewal Program." j l

j l

1 l

l l

L _.