ML20195J632

Forwards Responses to 980903 RAI for Review of Ccnpp,Units 1 & 2 Integrated Plant Assessment Rept for Reactor Vessel Internals Sys,Per License Renewal Application

ML20195J632
Person / Time
Site:	Calvert Cliffs
Issue date:	11/19/1998
From:	Cruse C BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9811250039
Download: ML20195J632 (17)

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CHARLES H. CCtJSE Baltimore Gas and Electric Company l Vice President Calvert Cliffs Nuclear Power Plant 1650 Calven Cliffs Parkway

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• Nuclear Energy Lusby, Maryland 20657 410 495-4455 l

November 19,1998 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk j

SUBJECT:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Response to Request for Additional Information for the Review of the Calvert Cliffs Nuclear Power Plant, Uqits 1 & 2, Integrated Plant Assessment Report for the Reactor Vessel Internals System

REFERENCES:

(a) Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated May 23,1997, " Request for Review and Approval of System and Commodity Reports for License Renewal" (b) Letter from Mr. D. L. Solorio (NRC) to Mr. C. H. Cruse (BGE),

September 3,1998," Request for AdditionalInformation for the Review of the Calvert Cliffs Nuclear Power Plant, Unit Nos.1 & 2, Integrated Plant Assessment Report for Reactor Vessel Internals Systems" (c) Letter from Mr. D. L. Solorio (NRC) to Mr. C. H. Cruse (BGE),

September 24,1998, " Renumbering of NRC Requests for Additional Information on Calvert Cli.ffs Nuclear Power Plant Liccuse Renewal Application Submitted by the Baltimore Gas and Electric Company" Reference (a) forwarded Baltimore Gas and Electric Company (BGE) systee and commodity reports for license renewal. Reference (b) forwarded questions from NRC staff on one of those reports, the Integrated Plant Assessment Report on the Reactor Vessel Internals System. Reference (c) forwarded a numbering system for tracking BGE's response to all of the BGE License Renewal Application requests for additional information and the resolution of the responses. Attachment (1) provides our responses to the questions contained in Reference (b). The questions are renumbered in accordance with l Reference (c). i 9811250039 981119 PDR ADOCK 05000317 P PDR NRC Distribution Code A036D

, Document Control Desk -

November 19,1998 Page 2 Should you have further questions regarding this matter, we will be pleased to discuss them with you.

Very truly yours,

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'A STATE OF MARYLAND  :

TO WIT: i COUNTY OF CALVERT  :  !

I, Charles H. Cruse, being duly sworn, state that I am Vice President, Nuclear Energy Division, Baltimore Gas and Electric Company (BGE), and that I am duly authorized to execute and file this response on behalf of BGE. To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal i knowledge, they are based upon information provided by other BGE employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it t be reliable.

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Subscr' ed and worn before me a Notary Public in and for the State of Maryland and County of

,this N day of /7 m lart.,1998.

WITNESS my Hand and Notarial Seal: -

Notary Public I My Commission Expires: '

OM fate CHC/KRE/ dim

Attachment:

(1) Response to Request for Additional Information; Integrated Plant Assessment Report for the Reactor Vessel Internals System I

cc: R. S. Fleishman, Esquire C. I. Grimes, NRC J. E. Silberg, Esquire D. L. Solorio, NRC '

S. S. Bajwa, NRC Resident Inspector, NRC A. W. Dromerick, NRC R.1. mci 2an, DNR H. J. Miller, NRC J. H. Walter, PSC I

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE i

REACTOR VESSEL INTERNALS SYSTEM i

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i Baltimore Gas and Electric Company-Calvert Cliffs Nuclear Power Plant November 19,1998 va , - ,- . . -

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! ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FORTIIE REACTOR VESSEL LNTERNALS SYSTEM NRC Ouestion No. 4.3.1 Figure 3.3-6 (Revision 21) of the Calvert Cliffs Nuclear Power Plant (CCNPP) Updated Final Safety l l Analysis Report (UFSAR) shows the fuel assembly hold down (FAHD) structure. One of the intended j functions of FAHD is to prevent fuel assemblies from being lifled out of position under accident loading l conditions. Please clarify whether the FAHD was subjected to an aging management review (AMR),

particularly the springs in it, which may lose their required force at extended age.

BGE Response Figure 3.3-6, " Fuel Assembly Hold Down," illustrates the relationship between the fuel alignment l plate (which is part of the reactor vessel internals (RVI]) and an individual fuel assembly. Except for the fuel alignment plate, all the components shown on Figure 3.3-6, including the upper end fitting, spring, spider, and upper end fitting posts, are part of the fuel assembly. Since the upper end fitting components of a fuel assembly are discharged with that assembly and since fuel assemblies are replaced after only a few years in the reactor, fuel assemblies (including the upper end fitting components) are considered short lived and are not subjected to an AMR.

NRC Ouestion No. 4.32 Figure 3.3-14 (Revision 21) of the CCNPP UFSAR shows the upper guide structure (UGS) assembly.

Please describe the functions of the Expansion Compensating Ring, and indicate ifits intended functions would meet the definition ofintended function listed in 10 CFR 54.4(a).

BGE Response The expansion compensating ring, called the hold down ring (HDR) in the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA), has the intended function: " provide structural support for the fuel assemblies, CEAs [ control element assemblies], and ICI [incore instrumentation]

so that they maintain the configuration and flow distribution characteristics assumed in UFSAR Chapter 14 analyses." This intended function meets the definition of " intended function" in 10 CFR 54.4(a). It is identified on LRA page 4.3-5 for the RVI and described as passive on page 43-6. All RVI components that perform this function were subject to AMR and described in the LRA.

NRC Ouestion No. 4.3.3 Section 4.1.3.6 (Revisionl8) of the CCNPP UFSAR indicates that vents were added to the reactor vessel and to the pressurizer head in response to the Three Mile Island Lessons Learned Report, NUREG-0737, Item II.B.l. One of the intended functions of the vents is to ensure core cooling during loss-of-coolant accident. Please indicate if this vent system was subjected to an AMR. If so, provide a cross reference to where the vents are addressed in the LRA. If not, provide the basis for their exclusion.

BGE Resnonse The reactor vessel vent system was subjected to AMR. The nozzles were evaluated as part of the reactor vessel heads, LRA Section 4.2, " Reactor Pressure Vessels and Control Element Drive Mechanisms / Electrical System." The vent system includes valves, piping and tubing. The piping and associated valves were evaluated with the Reactor Cooling System (RCS), LRA Section 4.1. See Figure 4.1-1 which clearly shows the reactor vessel vent system. Tubing and associated valves were evaluated in the Instrument Lines Commodity Evaluation, LRA Section 6.4.

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ATTACHMENT (1) l RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FORTHE REACTOR VESSEL INTERNALS SYSTEM The pressurizer vent system was subjected to AMR. As noted in Section 4.1.3.6 of the CCNPP UFSAR, the pressurizer vent line valves are installed in a line that was added as an additional branch off the pressurizer vapor sample line. Part of this vent system was evaluated with the Nuclear Steam

Supply System Sampling System, LRA Section 5.13. The other part was evaluated with the RCS, l- LRA Section 4.1.

l NRC Ouestion No.~ 4.3.4 Clarify whether all the RVI components listed in Table 4.3-1 r.re within the scope of the American Society of Mechanical Engineers (ASME) Code,Section XI, Subsection IWB Inservice inspection (ISI) l _ Program, as mentioned in Page 4.3-12. In addition, describe the applicable acceptance criteria and l describe the methods used for trending for the visual inspection.

BGE Response Baltimore Gas and Electric Company has credited the ISI Program only for Group 1, Group 2, and l Group 7 components subject to the effects of wear, neutron embrittlement, and high cycle fatigue I' respectively, as summarized in Table 4.3-3. Use of ISI to manage aging of Group I components is

!- discussed on pages 4.3-11 through 4.3-13. Use of ISI to manage aging of Group 2 components is discussed on pages 4.3-14 and 4.3-15. Use of ISI to manage aging of Group 7 components is discussed on page 4.3-29.

NRC Ouestion No. 4.3.5 The aging management programs for Group 5 (Stress Relaxation) described starting on page 4.3-24 indicate that plant-specific analysis will be performed to refine the calculated stress levels on CEA shroud bolts and core shroud (CS) tie rods and bolts for verifying low tensile stress during normal operations, and forjustifying no loss of preload due to stress relaxation. Provide the acceptance criteria

, that will be used for this analysis, and the schedule for completion of the analysis.

BGE Response Design preload requirements are the acceptance criteria. A condition where stress relaxation causes components to have preloads less than the design preload requirement is not acceptable. The current schedule for completing this assessment is h ae 30,2001. This schedule is based on current workload

!. and priorities and is, therefore, subject to chnnge.

NRC Question No. 4.3.6

Page 4.3-24 indicates that an examination of the CEA shroud bolts and CS tie rods and bolts would be

!- conducted as a part of an age-related degradation inspection (ARDI) program if the refiried stress level does not show the low stress expected. Assuming the results did warrant an ARDI for these components, l: ' provide a summary discussion of the ARDIs consistent with the NRC staff's request for additional l information on ARDIs in Reference (1).

i BGE Response

Please see BGE's response to Question 4 in Reference (2).

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A'ITACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FORTHE REACTOR VESSEL INTERNALS SYSTEM NRC Ouestion No. 4.3.7 Section 4.3.1.1 indicates Section 3.3.3 of the UFSAR provides a description of the RVI structures.

Section 3.3.3 does not provide sufficient details of the RVI components identified in Table 4.3-1 from Section 4.3.1. Please provide diagrams that show the location of the device types identified in Table 4.3-1.

BGE Response Calvert Cliffs UFSAR references to the components in question are shown in tabular form below.

The " Figure" column lists the UFSAR figures where the component appears (a listing in italics indicates that the component is not specifically labeled). The "Section" column lists the UFSAR text section that describes the component; this informatien is useful in interpreting the Figures, especially when a component is not specifically labeled.

Component UFSAR Figure UFSAR Section CEA Shroud and Bolts (CEASB) 3.3-14 3.3.3.6 CEA Shroud Extension Shaft Guides (ESG) 3.1-1 3.3-14 3.3.2.4 CS 3.1-1 3.3-1 3.3-13 3.3.3.1, 3.3.3.4 l Core Shroud Tie Rod and Bolts 3.3-13 3.3.3.4 Core Support Barrel (CSB) 3.1-1 3.3-1 3.3-12 3.3.3.1, 3.3.3.2 Core Support Barrel Alignment Key (CSBA) 3.3-11 3.3.3.2 Core Support Barrel Snubber and Snubber Bolts

• 3.3-12 3.3.3.2 Core Support Columns 3.1-1 3.3.3.1, 3.3.3.3 Core Support Plate (CSP) 3.1-1 3.3.3.3 l Flow Baffle (Flow Skirt)* 3.1-1 3.3.3.5 Fuel Alignment Pins (Fuel Assembly Locating 3.1-1 3.3.3.3 Pins)

Fuel Alignment Plate / Guide Lug insert (FP) 3.3-14 3.3.3.6 Holddown Ring (HDR)(Expansion Ring) 3.1-1 3.3-14 3.3.3.6 ICI Thimble Support Plate (ITSP) 3.1-1 7.5.4.2 ICI Thimbles

• 3.1-1 7.5.4.2 i Lower Support Structure Beam Assembly (LSSBA) 3.1-1 3.3.3.1, 3.3.3.3

, Upper Gui_de Structure Support Plate (UGSP) 3.3-14 3.3.3.6

• Note: Evaluated in BGE LRA Section 4.2 NRC Ouestion No. 4.3.8 Do the RV1 intended functions include: (a) support for the irradiation surveillance capsules, and (b) shielding for the reactor pressure vessel? If so, summarize what components perform these intended functions and explain whether these components are within the scope oflicense renewal.

BGE Respsmac The intended function for RVI is stated on page 4.3-5 of the LRA. It does not include support for the irradiation surveillance capsules and shielding for the reactor pressure vessel. The surveillance capsules are supported by the reactor vessel. The reactor vessel function of supporting the surveillance capsules is stated on page 4.2-8.

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RESPONSE To REQUEST FOR ADDITIONAL INFORMATION; IN1EGRATED PLANT ASSESSMENT REPORT FOR TIIE REACTOR VESSEL INTERNALS SYSTEM NRC Ouestion No. 4.3J Calvert Cliffs' LRA addresses certain applicable aging effect for specific RVI components. Describe, in summary form, the extent to which the following aging effects were determined to be either non-plausible or non-potential, for the specific components: stress corrosion cracking (SCC) and irradiation assisted stress corrosion cracking (IASCC), corrosion for the UGSP; SCC, IASCC, corrosion, and wear for the CEA shrouds; IASCC and corrosion for the CEASB; SCC, IASCC, corrosion, and wear for the fuel alignment plate; SCC, IASCC, corrosion for the CSB; SCC, IASCC, corrosion, and neutron embrittlement for the core support barrel upper flange; SCC, IASCC, and corrosion for the CS; SCC, IASCC, corrosion, and stress relaxation for the core shroud assembly bolts; SCC, IASCC, and corrosion for the CS tie rods; SCC, IASCC, and corrosion for the fuel alignment plate guide lugs; SCC, IASCC, and corrosion for the CSP; SCC, IASCC, corrosion, and stress relaxation for the fuel alignment pins; SCC, IASCC, ard corrosion of the lower support structure beam assemblies; SCC, IASCC, and corrosion for the core support columns; SCC, IASCC, corrosion, neutron embrittlement, and stress relaxation of the core support column bolts.

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FORTHE REACTOR VESSEL INTERNALS SYSTEM BGE Response NRC Component; tNRCARpMs? 4ComponentCoveredj 4 Plausible LRA$m

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yy gpggg ' : 6F FARDMs? M UGSP SCC /lASCC yes wear CEA Shrouds SCC /lASCC/ corrosion & yes neutron embrittlemer.t, wear thermal aging, high cycle fatigue CEA Shroud Bolts IASCC/ corrosion yes neutron embrittlement, stress relaxation, SCC /lASCC Fuel Alignment Plate SCC /lASCC/ corrosion & yes wear, neutron thermal embrittlement, embrittlement wear CSB SCC /lASCC/ corrosion yes wear, neutron embrittlement CSB upper flange SCC /'ASCC/ corrosion & yes wear, neutron neutron embrittlement embrittlement Core Shroud . Assembly SCC /lASCC/ corrosion & yes neutron embrittlement, low Bolts stress relaxation cycle fatigue, stress relaxation CS Tie Rods SCC /lASCC/ corrosion yes neutron embrittlement, low cycle fatigue, stress relaxation FAP Guide Logs SCC /lASCC/ corrosion yes wear, neutron embrittlement CSP SCC /lASCC/ corrosion yes neutron embrittlement, low cycle fatigue Fuel Alignment Pins SCC /lASCC/ corrosion & yes wear, neutron stress relaxation embrittlement Lower Support Structure SCC /lASCC/ corrosion yes neutron embrittlement Beam Assemblies Core Support Columns SCC /lASCC/ corrosion yes neutron embrittlement, low cycle fatigue, thermal aging Core Support Column SCC /lASCC/ corrosion & There are no Core Support N/A Bolts neutron embrittlement, Column Bolts.

stress relaxation All ARDMs listed in the question were considered potential. Thermal embrittlement is described as thermal aging in LRA section 4.3. This information is contained in the LRA in Table 4.3-2.

NRC Ouestion No. 43.10 Section 4.3.1.2 of the LRA indicates that a component level scoping and component pre-evaluation were not applied to the RVI before the aging evaluation to determine which components were subject to an AMR. Instead, all components of the ItV! ' vere initially included in the AMR. Section 4.3.1.2 of the LRA further indicates, "some components wert, determined not to be within the scope oflicense renewal since they are not required for the RVI to perform their intended function." Describe which components were considered to be outside the scope oflicense renewal and clarify the criteria that were used to conclude that these components were not required for the RVI "to perform their intended function."

Identify the components that provide a structural integrity function.

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ATTACHMENT (1)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FORTHE REACTOR VESSEL INTERNALS SYSTEM BGE Response Incore instrumentation thimbles were considered to be outside the scope of license renewal, except for the ICI tube flanges that were evaluated in Section 4.2. Please see note (e) on page 4.2-9.

Fuel assemblies were also not evaluated as they are replaced on regular intervals.

The criterion that was used to conclude that ICI thimbles were not required for the RVI "to perform their intended function" was their lack of contribution to the intended function stated on page 4.3-5.

The components that provide a structural integrity function are listed in Table 4.31. Note the intended function on page 4.3-5 and 4.3-6.

NRC Ouestion No. 4.3.11 Section 4.3.2 of the LRA indicates that IASCC is not plausible for Cab est Cliffs RVI because IASCC has not been observed for components with the temperature, oxygen and radiation levels present for the Calvert Cliffs RVI, either in operating plants or in laboratory tests. Identify the operating plant experience and laboratory test data that forms the basis for this conclusion. Identify the RVI components 2

at Calvert Cliffs that are subject to a neutron fluence greater than 5xl E20 n/cm . For these components, identify the temperature, oxygen, radiation levels and stress levels. What inspections or aging management programs will be performed for these components during the extended period of operation to ensure that these components do not exhibit IASCC during the license renewal term?

How does the information in Information Notice 98-11 impact this evaluation? Since bolt cracking has occurred at the junction of bolt head and shank, which is not accessible for visual inspection, how will CEASB and other RVI bolting that is subject to IASCC be examined? What inspections or aging management programs will be performed for these components during the extended period of operation to ensure that these components do not exhibit IASCC during the license renewal term?

BGE Response No IASCC-induced failure of RVI components has been reported in the United States for pressurized water reactors (PWRs). This is the operating experience (or lack thereof) that form the basis for the

- conclusion that IASCC is not plausible for Calvert Cliffs RVI components. All the known cracking of baffle /former bolts have been limited to the six oldest Electricit6 de France (EdF) 900 MW PWR

' units of the "CPO" type design and no failures have been observed in the later EdF "CPY" reactors, ,

even though some have accumulated the same number of hours of operation as the "CPO" units. '

Baltimore Gas and Electric Company does not have baffle /former bolts similar to the ones that have been observed to have failed in French "CPO" plants. Instead of using a high strength, cold worked bolt torqued to high values to attach formers and baffles, the Calvert Cliffs design has annealed plate welded together in a relatively flexible design. No failures have been reported due to IASCC in any operating PWR in annealed and welded construction.

In the' laboratory and in boiling water reactors, IASCC of stainless steels has been produced at 22 2 fluences approaching lx10 and 5x10 , respectively (n/cm2 E>0.1 MeV). Due to the beneficial effects of hydrogen over-pressure in PWRs, the threshold fluence for onset ofIASCC susceptibility is much higher in PWR than for boiling water reactors using normal water chemistry. Service experience at the only PWRs in the world where IASCC has been observed (EdF "CPO" units) 6

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ATTACH. MENT (1) l RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FORTHE REACTOR VESSEL INTERNALS SYSTEM indicate the earliest onset ofIASCC susceptibility at 8x1021 n/cm2. Once again, Calvert Cliffs does not have the components that fractured at the EdF plants.

l- Since IASCC has not been observed for RVI components except for components that Calvert Cliffs j does not have, we conclude we do not have components susceptible to IASCC.

j There is considerable research being performed around the world in an effort to determine whether IASCC might eventually occur in PWR RVI components. Additionally, several inspections of domestic PWR units' baffle bolting will be per%rmed over the next several years. Results of l additional research and ISI may someday indicate that the RVI components are susceptible to IASCC.

! At the present time, however, there is no data that iPJicate any Calvert Cliffs RVI components will 4

ever be susceptible to IASCC. Baltimore Gas and Eiectric Company will follow this work. l During the current licensing period and durirg the renewal period, BGE performs and will perform ASME Section XI examinations of RVI components. Cracking of the highest fluence components (CS plates and welds) would be identifiable by visual examination. Note that these inspections are l required by Code and current licensing basis. Baltimore Gas and Electric Company has no aging management program for any RVI component due to IASCC since we have concluded it is not plausible. Additionally, BGE is not providing the other requested parameters since we have concluded IASCC is not plausible.

Components that are likely to exceed 5x102 : n/cm2 fluence include LSSBA, CSB, CSP, Fuel Alignment Pins, Core Support Columns, CS Tie Rods and Bolts, CEASB, FP, and CS.

l l Information Notice 98-11 refers to cracking of a component that is not installed at CCNPP.

l Baltimore Gas and Electric Company will follow developments in research and in plants that have l similar components to those described in Information Notice 98-11, to determine whether events described in 98-11 could occur at Calvert Cliffs. At this point in time, there is no evidence that the events described in 98-11 are relevant to Calvert Cliffs.

NRC Ouestion No. 43.12 Section 4.3.2 of the LRA indicates, " procedures will be enhanced if modified to specifically identify each component of the RVI which relies on this program for aging management for license renewal."

Which RVI components have had or will have their procedures modified as a result of the review of aging management for license renewal? Briefly summarize the reasons for the modifications.

! BGE Response  ;

l As stated in Table 4.3-3 of the LRA (page 4.3-30), the existing implementing procedures for ISI of j l ASME Code Section XI components will be modified for the device types in Groups 1,2, and 7. The affected' components for each of these groups are identified in Table 4.3-2 of the LRA (page 4.3-8 and 9). l l

} As stated on pages 4.3-12,15, and 29 of the LRA, BGE determined that the existing ISI procedures I would be enhanced if modified to specifically identify each component of the RVI that relies on ISI

{ for aging management for license renewal. This was considered an optimization since it would l provide specific identification of the applicable ARDMs associated with each affected RVI i component.

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FORTHE REACTOR VESSEL INTERNALS SYSTEM NRC Ouestion No. 4113 Section 4.3.2 of the LRA indicates that of the three U.S. suppliers of light water reactors the most fatigue-susceptible RVI components have been identified for PWR plants. What is the most-fatigue susceptible RVI component? Explain how this was determined? If the usage factor for these components exceeds 0.5 (criteria specified in the LRA), what additional actions will be initiated.

Additionally, indicate to what degree would the scope of components being evaluated be expanded as a result ofexceeding the usage factor of 0.5 for the components normally evaluated.

I BGE Response Table 4.3-2 on pages 4.3 8 and 43-9 identifies components for which fatigue is plausible. Baltimore Gas and Electric Company has demonstrated management of the effects of fatigue. No attempt was made to identify the "most fatigue susceptible RVI component."

Determination of fatigue plausibility and discussion ofits effects and the management of the effects is included in two groups of Section 4.3. Group 3, beginning on page 4.3-15, discusses low cycle fatigue. Group 7, beginning on page 4.3-27 discusses high cycle fatigue.

The usage factor question is pertinent to Group 3 only. If the usage factor for these components exceeds 0.5, then actions will be taken r.s described on page 4.3-18. l Baltimore Gas and Electric Company has determined components for which fatigue is plausible as discussed above. If evidence of fatigue or a concern about fatigue are noted for any other l components, it would be addressed by the site Corrective Actions Program.

NRC Ouestion No. 4114 Section 4.3.2 of the LRA indicates, " Thermal aging is potentially significant for: (1) centrifugally-cast parts with delta ferrite content above 20%; (2) statically-cast parts with molybdenum content meeting CF3 and CF8 limits and with a delta ferrite content above 20%; and (3) statically-cast parts with molybdenum content exceeding CF3 and CF8 limits with delta ferrite content above 14%." Provide the basis for the conclusion that thermal aging is not significant below these levels. How is the amount of delta ferrite in cast stainless steel RVI components determined? What are the uncertainties in these test methods? How are the uncertainties incorporated into the estimate of the delta ferrite?

If the delta ferrite values exceed the limits in the LRA, Section 4.3.2 indicates that an examination will be performed. Provide a fracture mechanics analysis to demonstrate the critical flaw size at the end of the license renewal term for these limits. Identify the inspection procedures and the capability of the inspection to de tect flaws smaller in size than the critical flaw size.

BGE Respog The conclusiion that thermal aging is not significant below the limits detailed in Section 4.3.2 of the 4 LRA is base:1 on the Electric Power Research Institute screening criteria presented in Reference (3).

The limits described in Section 4.3.2 of the LRA constitute a screening criteria for thermal embrittlement of cast austenitic stainless steel (CASS). The screening criteria are based on an extensive body of fracture toughness test data for thermally aged CASS components representative of those in PWR applications.

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j In general, the amount of delta ferrite can be determined either by direct measurement using a i calibrated magnetic instrument (Severn Gauge) or estimated from the material chemical composition using an empirical formula. Estimation using the formula is the only feasible method for RVI components due to high dose rates and difficult accessibility since remote test equipment for direct measurement of delta ferrite does not exist. The chemical compositions of RVI CASS components l will be obtained from certified material test reports for the individual components. The delta ferrite percentage will then be estimated and compared to the screening criteria. A plot of measured versus calculated delta ferrite (Reference 4) shows that virtually all the measured values fall within two standard deviations of the mean deviation, which is approximately plus or minus six weight percent l of the calculated delta ferrite. For the purposes of conservatively determining the fracture toughness l loss of the component, an appropriate margin of some weight percentage delta ferrite will be added to the calculated delta ferrite. The margin will be based on the most current available data comparing calculated to measured delta ferrite.

The loss of fracture toughness can be estimated based on published data or empirical formulas.

Fracture mechanics analyses specific to RVI CASS components have not been performed, but will be performed prior to the end of the current licensing period for RVI CASS components predicted to fall below the design basis fracture toughness requirements specific to the component. If inspection is deemed necessary based on the results of the fracture mechanics analyses, the most probable inspection technique is remote visual inspection.

NRC Ouestion No. 4.3.15 Section 4.3.2 of the LRA indicates: "A stress analysis will be performed specifically to evaluate the potential for SSC of CEA shroud bolts." Provide the criteria that will be used in this evaluation. Provide the data that will be used to establish the criteria that A-286 CEA shroud bolts are not subject to SCC during the extended period of operation. What type of examination, extent of examination and acceptance criteria are applicable for A 286 CEA shroud bolts under the ARDI program?

BGE Response As noted in Question 5 and on pages 4.3-23 and 4.3-24, plant specific analyses will be performed to refine the calculated stress levels on these bolts. The SCC discussion on page 4.3-25 indicates, "A generally accepted value for threshold stress is the yield stress of the material . . ." A value less than yield will be used as the criterion. Please see BGE's response to Question 4 in Reference (2) for the requested ARDI discussion.

NRC Ouestion No. 4.3.16 Table 4.3-2 indicates erosion, erosien/ corrosion, general corrosion / uniform attack, hydrogen damage and pitting / crevice corrosion are not plausible. Explain the bases for these conclusions.

BGE Respnic Baltimore Gas and Electric Company has determined that erosion, erosion / corrosion, general corrosion / uniform attack, hydrogen damage, and pitting / crevice corrosion are not plausible for the l

RVI. The following paragraphs describe the reasons the ARDMs are not plausible for the RVI.

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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FORTHE REACTOR VESSEL INTERNALS SYSTEM 4

i General Corrosion / Uniform Attack l

The use of resistant materials and a pure water system make general corrosion not plausible for all RVI device types. The austenitic stainless steels, alloy steels, and nickel-based alloys used in the RVI components are quite resistant to general corrosion in a benign operating environment.

PittingfCrevice Corrosion The amount of oxygen and chlorides in the surrounding reactor coolant are not sufficient to cause pitting / crevice corrosion. Pitting is not a concern in the RVI environment unless the reactor has a history oflong outages without proper chemistry control.

Erosion / Erosion Corrosion The high operating pressure, relatively low fluid velocity, and low level of particulates in the Reactor Coolant System ensure that erosion / erosion corrosion are not plausible for nickel-based alloys, high alloy steels, and stainless steel RVI components in the PWR environment. The operating pressure, 2250 psia, precludes cavitation erosion, and the purity and particulate control of the reactor coolant climinate the possibility of particulate-based erosion.

Hydrogen Damage The amount of available hydrogen in the surrounding reactor coolant is not sufficient to cause hydrogen damage for austenitic stainless steels, low alloy steels, and nickel-based alloy materials of the RVI device types. Above 400 F hydrogen ditTuses rapidly in steel and is eliminated by off gassing. Hydrogen damage is not plausible since the temperature of the Reactor Coolant System is greater than 400 F and the hydrogen can easily diffuse. I NRC Ouestion No. 4.3.17

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l Section 4.3.2 indicates SCC /IGSCC/intergranular attack are potential ARDM(s) for RVI components fabricated from [ Aerospace AfaterialSpeci/lcation] AMS 5735 iron base superalloy A-286; but does not identify any Inconel 600 components. Primary water stress corrosion cracking in PWR environment has occurred in Inconel 600 components. Identify the RVI components that were fabricated using this ,

material or other nickel base alloys and describe the aging management program that will be used during l the extended period of operation to ensure these components are not susceptible to primary water stress corrosion cracking.

BGE Response l There are no RVI components requiring AMR that are fabricated from Inconel 600 or other nickel

base alloys.

NRC Ouestion No. 4.3.18 Table 4.3 indicates that many components (CEASB, CS, CSTR, CSB, CSC, CSP, FAP, FP, and LSSBA) are susceptible to neutron embrittlement, which generally results in loss of fracture toughness in the material composing the component. This loss of fracture toughness is a reduction in resistance to crack growth, which could mean that parts that are macroscopically degraded (through wear or some sort of cracking mechanism such as SCC or fatigue) may fail (fracture) at load levels and/or degradetion (i.e., smaller crack sizes) that are lower than those if the part was not in an embrittled condition. Identify l 10 l

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l for each component that is susceptible to neutron embrittlement, the peak neutron fluence at the end of l the extended period of operation, and the materials used to fabricate the specific component. For the

! limiting component (considering the neutron fluence, material fracture toughness and operating stresses in determining the limiting component), provide a fracture mechanics analysis to determine the critical  ;

flaw size during normal operation and emergency and faulted conditions. Provide data to justify the ,

l fracture toughness assumed in the analysis. Identify the inspection procedure and the capability of the inspection to detect flaws smaller in size than that of the critical flaw.

t BGE Response l There are no existing design requirements for fracture toughness determination on RVI materials.

The use of flaw-tolerant stainless steel materials for vessel internals obviated the need for consideration of toughness in design requirements. It has been hypothesized that neutron irradiation of vessel internals can cause the stainless steel to transform from a flaw-tolerant material to a material

' that could possibly undergo cleavage fracture. This hypothesis is supported by very few data, taken l

under conditions that do not closely duplicate a PWR environment. It is not possible to verify the correctness of the hypothesis, or evaluate the effect of neutrons on toughness, using existing data.

Lacking sufficient information to predict whether neutron embrittlement will affect the stainless steel internals, BGE will use ASME Section XI examinations to discover whether embrittlement is occurring. This is consistent with current regulations.

i Baltimore Gas and Electric Company views this as an emerging technical issue, as discussed in part 6.3.3.5 of Section 2.0 of the BGE LRA, Integrated Plant Assessment Methodology. Baltimore Gas and Electric Company will follow and assist NRC and industry efforts over the next several years

l. to collect additional data on embrittlement of stainless steels. If it is determined that significant embrittlement is probable, then BGE will take appropriate steps to address toughness requirements for the vessel internals, when requirements are published. Appropriate steps would typically include performance of fracture mechanics analyses to determine component critical flaw sizes. If the critical flaw size was determined to be smaller than section thickness, examination techniques capable of finding flaws smaller than the critical size would be developed and applied.

NRC Ouestion No. 4.3.19 Section 4.3.2 states that "No instances of degradation of RVI for PWRs have been secorded which have definitely been attributed to neutron embrittlement," and "Calvert Cliffs has not discovered any thermal-l aging related damage for the RVI. Also there have not been RVI damage events at other PWRs that were identified as thermal aging failure." Based on the staff's experience the degradation in material u properties attributable to these two ARDMs can only be " observed" through evaluation of the results of

. destructive material property testing of degraded components. Therefore, elaborate on the basis for these conclusions.

BGE Response The reference number and section for "No instances of degradation of RVI for PWRs have been L recorded which have definitely been attributed to neutron embrittlement," appears immediately aller l the statement on page 4.3-14. The reference numbers and sections for "Calvert Cliffs has not discovered any thermal-aging related damage for the RVI. Also there have not been RVI damage L

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! events at other PWRs that were identified as thermal aging failure . . ." appears immediately after the l l statement on page 4.3-20. The reference list is on page 4.3-32.

t NRC Ouestion No. 43.20

' Section 4.3.1.1 of the LRA indicates that the aging evaluation of RVI " credits" the primary water

- chemistry control as an aging management program to manage aging of RVI components. Which components and ARDMs are affected by primary water chemistry control? Describe the " credits" assumed for each ARDM and affected component and justify the credits assumed.

i l BGE Response i 1-Al! 107i components are normally immersed in reactor coolant. The corrosior AMDMs in Table 4.3-2

• are affected by primary water chemistry control. As is stated in 4.3.1. son page 4.3-4), the j demonstration of the primary chemistry control program as an aging management program is not j repeated in Section 4.3 of the LRA.

NRC Ouestion No. 43.21 Section 4.3 indicates that changes in the design of the HDRs installed at Calvert Cliffs Units I and 2 j were made as a result of wear experienced in a similar component at another reactor plant and the discovery for the need to provide for additional fuel assembly growth. Table 4.3-1 identified the HDRs l as a device type . subject to AMR. Table 4.3-2 identifies the HDRs as device types subject to wear as an L ARDM. Further, the LRA indicates that wear can be discovered when the reactor vessel is opened . i l during a refueling outage, and the RV1 are subject to a visual examination of accessible surfaces.  !

L The HDR is a near flat ring spring of a rectangular cross section. The hold down force is developed by -

l- deflecting the inner and outer edges of the ring spring in a direction to cause flattening of the ring. In l deflecting the HDR, the outer edge of the top surface and inner edge of the bottom surface of the ring ,

L contact and load the pressure vessel closure and the UGS flange, respectively. Provide a description of l the accessibility to the bottom surface of the HDR that contacts the UGS flange, the UGS flange and the

. undersurface of the vessel closure for visual inspection. Your description should account for the accuracy required in the use of visual indications of detectable wear to reliably determine changes in the HDR load developing capability.

In addition, any such wear, if it occurs, may gradually reduce the HDR clamping force and induce core l barrel motion under flow excitations. Verify the existence of a program for monitoring and trending the possible core barrel motion, using data from excore neutron detectors.

BGE Response i Baltimore Gas and Electric Company can access the bottom surface of the HDR and the UGS flange if required. The undersurface of the vessel closure head is accessible when it is in the head laydown area.

I Baltimore Gas and Electric Company does not rely on monitoring of core barrel motion to manage j l wear for the HDR. As shown in Table 4.3-2 on page 4.3-8, wear of the HDR is addressed by the l l Group 1 device type. ,As shown in Table 4.3-3 on page 4.3-30, wear of Group 1 device types is

managed by ISI of ASME Code Section XI Components. .

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NRC Ouestion No. 43.22 l Provide the basis for considering the HDR as a device type subject to stress relaxation. Describe any inspections performed, or that will be performed with regard to changes in as-built dimensions or deflection measurements that demonstrate that the hold down load provided by the llDR has not and will not be reduced to impair its intended function during the period of extended operation.

BGE Response Baltimore Gas and Electric Company does not consider stress relaxation plausible for the HDR as )

shown in Table 4.3-2. A discussion of aging management for the effects of wear of the HDR is contained on pages 4.3-10 through 4.3-13.

j NRC Ouestion No. 43.23 Describe the visual examinations of the CEASB that have been previously performed or that will be l performed to maintain the structural integrity of the RVI consistent with the current licensing basis during the period of extended operation. Describe the postions of the CEASB that are accessible for visual examination and discuss how the observations can be used to reliably demonstrate and provide adequate assurance that neutron embrittlement will be managed during the period of extended operation.

BGE Response The visual examinations that have been performed to date have been remote visual exams using l

underwater television cameras operated from the top or side of the refueling pool. These examinations have looked for: 1) structural distortion or displacement of parts; 2) loose, missing, cracked, or fractured parts, bolting or fasteners; 3) foreign materials or accumulation of corrosion products; 4) corrosion or erosion; 5) wear of mating surfaces; and 6) structural degradation of attachments. These exams have covered the portions of the UGS that are readily accessible when the UGS is removed from the reactor vessel. The portions of the CEASB that have been readily accessible for these exams are those items located around the periphery of the UGS or visible from the top or bottom of the UGS; these items include the instrument tubes and supports, the bases, base tubes, shrouds, lower shrouds, external spanner nuts, socket head cap screw, and locking strip.

Indications of parts, bolting, or fasteners that become loose, missing, cracked, or fractured would lead to corrective measures including cause determination. These corrective measures could include i disassembly of UGS components to make additional items accessible for examination.

The discussion ofISI as the aging management program for neutron embrittlement is contained on

pages 4.3 14 and 4.3-15 of the BGE LRA.

NRC Ouestion No. 43.24 l 'Are there any parts of the systems, structures and components within the RVI system that are inaccessible for inspection? If so, describe what aging management program will be relied upon to maintain the integrity of the inaccessible areas. If the aging management program for the inaccessible areas is an evaluation of the acceptability ofinaccessible areas based on conditions found in surrounding

, accessible areas, please provide information to show that conditions would exist in accessible areas that would indicate the presence of, or result in degradation to, such inaccessible areas. If different aging

effects or aging management techniques are needed for the inaccessible areas, please provide a summary to address the following elements for the inaccessible areas: (a) Preventive actions that will mitigate or 13 1 ,_ _, - . _ __ - ,

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, i prevent aging degradation; (b) Parameters monitored or inspected relative to degradation of specific structure and component intended functions; (c) Detection of aging effects before loss of structure and component intended functions; (d) Monitoring, trending, inspection, testing frequency, and sample size to ensure timely detection of aging efTects and corrective actions; (e) Acceptance criteria to ensure structure and component intended functions; and (f) Operating experience that provides objective -

evidence to demonstrate that the effects of aging will be adequately managed.

BGE Response Baltimore Gas and Electric Company can access all RVI components if required.

References

1. Letter from Mr. D. L. Solorio (NRC) to Mr. C. H. Cruse (BGE), dated August 28, 1998,

" Request for Additional Information Fcu the Review of the Calvert Cliffs Nuclear Power Plant, Unit Nos.1 & 2, Integrated Plant Ass & ment Report"

2. Letter from Mr. C. H. Cruse s sGE) to NRC Document Control Desk, dated, November 12,1998, Response to " Request for Additional Information for the Review of the Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Integrated Plant Assessment, Generic Areas"

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

4. NUREG/CR-4513 (ANL-93/22), Revision 1, " Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems," O. K. Chopra, , Argonne National

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I Laboratory, Argonne,IL, August 1994 a j I

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