ML20151V364

Forwards Request for Addl Info Re Licensee 970730 Submittal for Review Reactor Pressure Vessels & Control Element Drive Mechanisms/Electrical Sys (4.2) Integrated Plant Assessment Technical Rept

ML20151V364
Person / Time
Site:	Calvert Cliffs
Issue date:	08/26/1998
From:	Dave Solorio NRC (Affiliation Not Assigned)
To:	Cruse C BALTIMORE GAS & ELECTRIC CO.
References
TAC-M99206, TAC-M99587, TAC-M99588, NUDOCS 9809140276
Download: ML20151V364 (8)

Text

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, August 26, 1998

,. Mr. Charirs H. Crusa, Vice Prssidsnt i'. . Nuclear Energy Division l Baltimore Gas and Electric Company 1650 Calvert Cliffs Parkway

Lusby, MD 20657-47027

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE  ;

CALVERT CLIFFS NUCLEAR POWER PLANT, UNITS NOS.1 & 2, INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR l 5'

PRESSURE VESSELS AND CONTROL ELEMENT DRIVE l MECHANISMS / ELECTRICAL (TAC NOS. M99587, M99588, AND M99206)

Dear Mr. Cruse:

By letter dated July 30,1997, Baltimore Gas and Electric Company (BGE) submitted for review the Reactor Pressure Vessels and Control Element Drive Mechanisms / Electrical System (4.2)

integrated plant assessment technical report as attached to the " Request for Review and l Approval of System and Commodity Reports for License Renewal." BGE requested that the Nuclear Regulatory Commissior) staff review the Reactor Pressure Vessels and Control

]  : Element Drive Mechanisms / Electrical System (4.2) integrated plant assessment technical report I to determine if the report meets the requirements of 10 CFR 54.21(a), " Contents of application-i technical information," and the demonstration required by 10 CFR 54.29(a)(1), " Standards for issuance of a renewed license," to support an application for license renewal if BGE applied in

! the future. By letter dated April 8,1998, BGE formally submitted its license renewal application.

The NRC staff has reviewed the Reactor Pressure Vessels and Control Element Drive q Mechanisms / Electrical System (4.2) integrated plant assessment technical report against the i requirements of 10 CFR 54.21(a)(1),10 CFR 54.21(a)(3). By letter dated April 4,1996, the i staff approved BGE's methodology for meeting the requirements of 10 CFR 54.21(a)(2). Based on a review of the information submitted, the staff has identified in the enclosure, areas where ,

l additionalinformation is needed to complete its review. 1 f

Please provide a schedule by letter or telephonically for the submittal of your responses within

' 30 days of the receipt of this letter. Additiorially, the staff would be willing to meet with BGE prior to the submittal of the responses to provide clarifications of the staff's requests for additionalinformation.

Sincer David L. Solorio, Project Manager g91j0g gg License Renewal Project Directorate P PDR ,

Division of Reactor Program Management )

!u'"" Office of Nuclear Reactor Regulation . 3 Docket Nos. 50-317 and 50-318

Enclosure:

Request for AdditionalInformation . \ i O. I' cc w/ encl: See next page s'e'n"e' t p!'oe"' RBC Hl.E CENTER C9PY DOCUMENT NAME:G:\ WORKING \SOLORIO\RPV_RALLTR OFFICE LA:PDig PDLR/DRM %DRPyknSC PDLR/DRPM:D NAME SLittleV DSolorialb Slee K CGrimes @

DATE 8Mg8 8gsp/98

[ 8/ /98\ 8/2fj98 OFFICIAL RECORD COPY

Mr. Charles H. Cruse Calvert Cliffs Nuclear Power Plant Baltimore Gas & Electric Company Unit Nos.1 and 2 cc:

1 Mr. Joseph H. Walter, Chief Engineer Prerident Public Service Commission of l Calvert County Board of Maryland

)

Commissioners Engineering Division  ;

175 Main Street 6 St. Paul Centre  !

Prince Frederick, MD 20678 Baltimore, MD 21202-6806 l James P. Bennett, Esquire Kristen A. Burger, Esquire Counsel Maryland People's Counsel Baltimore Gas and Electric Company 6 St. Paul Centre P.O. Box 1475 Suite 2102 Baltimore, MD 21203 Baltimore, MD 21202-1631 l l

Jay E. Silberg, Esquire Patricia T. Birnie, Esquire Shaw, Pittman, Potts, and Trowbridge Co-Director ,

2300 N Street, NW Maryland Safe Energy Coalition l Washington, DC 20037 P.O. Box 33111 Baltimore, MD 21218 Mr. Thomas N. Prichett, Director NRM Mr. Loren F. Donatell Calvert Cliffs Nuclear Power Plant NRC Technical Training Center 1650 Calvert Cliffs Parkway 5700 Brainerd Road Lusby, MD 20657-4702 Chattanooga, TN 37411-4017 Resident inspector David Lewis U.S. Nuclear Regulatory Commission Shaw, Pittman, Potts, and Trowbridge P.O. Box 287 2300 N Street, NW -

St. Leonard, MD 20685 Washington, DC 20037 Mr. Richard I. McLean Douglas J. Walters Nuclear Programs Nuclear Energy Institute Power Plant Research Program 1776 l Street, N.W.

Maryland Dept. of Natural Resources Suite 400 Tawes State Office Building, B3 Washington, DC 20006-3708 Annapolis, MD 21401 DJW@NEl.ORG Regional Administratc; Region i Barth W. Doroshuk l U.S. Nuclear Regu!atory Commission Baltimore Gas and Electric Company I' 475 Allendale Road Calvert Cliffs Nuclear Power Plant King of Prussia, PA 19406 1650 Calvert Cliffs Parkway NEF ist Floor Lusby, Maryland 20657 j

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Distribution:

HARD COPY Docket friles' PUBLIC PDLR R/F MEl-Zeftawy OGC DISTRIBUTION: E-MAIL:

FMiraglia (FJM)

JRoe (JWR)

DMatthews (DBM)

CGrimes (ClG)

TEssig (THE)

Glainas (GCL)

JStrosnider (JRS2)

GHolahan (GMH)

SNewberry (SFN)

GBagchi(GXB1) {

RRothman (RLR) 1 JBrammer(HLB) i CGratton (CXG1) l JMoore (JEM) I MZobler (MLZ) I RWeisman (RMW)

SBajwa (SSB1)

ADromerick (AWD)

LDoerflein (LTD)

BBores (RJB)

SDroggitis (SCD)

RArchitzel(REA)

CCraig (CMC 1)

LSpessard (RLS)

RCorreia (RPC)

RLatta (RML1)

DMartin (DAM 3)

WMcDowell(WDM)

DSolorio (DLS2)

SStewart (JSS1) l PDLR Staff I TCollins (TEC)

RCaruso (RXC)

MRazzaque (MMR1)

PPatnaik (PXP)

TSullivan (EJS)

BElliot (BJE)

RWesman (RHW)

JFair (FRF)

Slittle (SLL)

1

• REQUEST FOR ADDITIONAL INFORMATION l GALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE

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MECHANISMS / ELECTRICAL INTEGRATED PLANT ASSESSMENT. SECTION 4.2

) QQCKET NOS.60-317 AND 50-318 i

Section 4.2.1 - Scoolna

1. We noted that page 4.3-5 of Section 4.3 indicated that the reactor vessel head lifting rig

! is discussed with the Fuel Handling Equipment and Other Heavy Load Handling Cranes

of Section 3.2 of the license renewal application (LRA). However, Figure 4-2 (Rev.18)

} provided in Chapter 4 of the Calvert Cliffs Nuclear Power Plant (CCNPP) Updated Final i

Safety Analysis Report (UFSAR) for Units 1 and 2 shows a component attached to the 1 l closure head of reactor pressure vessel, which is called a lifting lug. Are lifting lugs  !

} included within the scope of license renewal? If so, provide a cross reference to where  !

they are addressed in the LRA. If not, provide the basis for their exclusion. )

2. Figure 4-2 (Rev.18) of the CCNPP UFSAR shows that the closure head insulation is

! attached to the closure head of reactor pressure vessel. Please describe the functions of closure head insulation, and indicate if the closure head insulation is required to support one of the functions listed in 10 CFR 54.4(a)(1)(i)-(iii).

3. Please clarify whether the component identified in comment (d) of Table 4.2-2 of Section 4.2.1 as a " Core Stop Lug" is same component labeled as the core support lug in Figure 4-2 (Rev.18) provided in Chapter 4 of the CCNPP UFSAR. If these components are not the same, please describe the functions of core support lug and indicate if the core support lug is required to support one of the functions listed in 10 CFR 54.4(a)(1)(i)-(iii).

4. What changes to the scope or other aspects of the Boric Acid Inspection Program have been made in response to the experiences documented in Section 4.2.2 (pg 4.2-14) of the LRA? l Section 4.2.2 - Agina Management

5. Pursuant to 10 CFR Part 50, Appendix G, provide an analysis of the vessel beltline n'sterial to demonstrate that they will maintain at least 50 ft-Ib Charpy upper-shelf energy (USE) during the period of extended operation, based on the projected neutron fluence and the chemistry of the beltline material. Provide all Charpy USE material data for each beltline material.

6. Provide an outline of the Reactor Vessel Material Surveillance Program and discuss how they will be used to monitor neutron irradiation for the reactor pressure vessel (RPV) beltline materials during the period of extended operation. Provide a summary of "CCNPP Comprehensive Reactor Vessel Surveillance Program (CRVSP)* so that the staff can determine that CRVSP is complete and adequate. Are there supplemental or standby capsules available to be used?

Enclosure

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7. How is your assessment of pressurized thermal shock (PTS) affected by the results from the McGuire 1 material surveillance program? Include in your evaluation, the results from the McGuire 1 capsule Y, which is contained in Duke Energy letter to the NRC, dated April 22,1998. '

8. Provide pressure-temperature (P-T) limits for the extended operating term and identify the operating window relative to pump operation for the shutdown cooling system.

During the extended licensed term, will there be any limitations in operation of the shutdown cooling system due to American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Appendix G, P-T operating limits and the minimum l permissible temperature of the reactor vessel?

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9. As identified in Section 4.2.1.1 of the submittal under " Unintentional inclusion of Slag l Stringer in RPV", the fabrication flaw in the Unit 1 reactor vessel weld was stated to be I acceptable in accordance with the applicable ASME Code, Section XI during the

- preservice and subsequent inservice inspections. However, the flaw acceptance criteria of the Code have been based on a 40-year operating life, equivalent to four inspection intervals of 10-year duration each. Therefore, the flaw should be evaluated analytically  ;

for the extended term of operation. Provide an evaluation in accordance with IWB-3600 '

of the ASME Code, Section XI. Identify the locaticn of the flaw within the weld. If the location of the flaw designates it as a surface planar flaw in the inside surface of the reactor vessel in accordance with the ASME Code, Section XI, paragraph IWA-3310, provide an analysis that demonstrates that PTS including small-break-loss-of coolant accident with an extended high pressure injection transient is not a concern consistent with the bases for 10 CFR 50.61. Provide initial and adjusted reference nil-ductility transition temperature (RT-NDT), delta RT-NDT, margin, neutron fluence, and chemistry (Copper and Nickel) of the weld containing the flaw in accordance with Regulatory Guide 1.99, Rev. 2.

10. Provide a description of the "CCNPP Alloy 600 Program" which is implemented as an aging management program for discovery and mitigation of age related degradation, particularly primary water stress corrosion cracking (PWSCC) in inconel, and explain how the program will be implemented during the license renewal term.

11. How will BGE determine the condition of partial penetration welds in the vessel head penetrations and in the bimetallic welds of control element drive mechanisms (CEDM) penetration nonles? In particular, discuss how BGE intends to extend its commitments to Generic Letter 97-01, " Degradation of Control Rod (Element) Drive Mechanism Nontes & Other Vessel Closure Head Penetrations", over the proposed extended term of operation for the CCNPP units. Include in the discussion an assessment and reanalysis of the CCNPP CEDM nonles using the latest crack initiation and growth model that was developed for the Combustion Engineering (CE) Owners Group to assess postulated flaws in CE designed CEDM penetration nonles. With respect to this reanalysis, provide what'the probability will be for cracks to have initiated in the CEDM penetration nonles at the end of the current license and at the end of the proposed extended terms, and state what the anticipated degree of crack growth is for postulated flaws in the CEDM penetration noujes at the end of the current license and at the end of

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3 i the proposed extended operating terms. Identify any volumetric examination of Calvert Cliffs or of other plants CEDM penetration nozzles that will confirm your susceptibility j analysis that cracking will not occur during the license renewal term.

l. 12. Table 4.2 2 of the submittalidentifies stress corrosion cracking of the RPV flow skirt, as

being a plausible age-related degradation mechanism. Discuss how BGE intends to l monitor the flow skirt-to-vessel weld for PWSCC during the extended term of operation.

) To what extent will the flow induced vibration in the flow skirt affect integrity of the subject i weld?-

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13. In Section 4.2.2 of the submittal, you have identified the aging related degradation mechanisms (ARDMs) for various RPV components. Based on these aging

}= mechanisms, how will your inservice inspection (ISI) program be tailored to monitor age

! related degradation due to these mechanisms for these components? Is there any weld j on these components that is not examined due to physical constraints or geometry?

[ Provide your plan to request any relief from the Code-required examination of such l

welds during the renewal term.

j 14. Based on its evaluation of operating experience, the NRC has determined that potential l

aging effect mechanisms in components of PWR vessels are as indicated in the Table 3.1-3 of the Draft Standard Review Plan for License Renewal. Table 3.1-3 identifies I

! components that are considered part of the reactor pressure vessel (RPV) and identifies the associated aging effects for the components. Identify the equivalent components in the Calvert Cliffs RPV and identify the aging effects applicable to these components.

Explain how the aging effects that are identified as "Significant" or " Unresolved" in the table are addressed for both Calvert Cliffs RPVs.

15. Section 4.2.2 includes a discussion that the ISI walkdown inspections (VT-2) after reactor shutdown and prior to plant startup must ensure that all components that are the subject of lasue Reports, where boric acid leakage has been found, are examined in accordance with the requirements of the program. Does the scope of components covered by the Boric Acid Inspection Program include all of the components for which general corrosion caused by boric acid is plausible, or only those which have been the subject of issue

. Reports?

16. Section 4.3 of the LRA entitled " Reactor Vessel Intemals (RVI) System" indicates that the core support barrel snubber and snubber bolts are addressed in this Section 4.2. The NRC staff did not find these devices described in Section 4.2, therefore, please describe how and where these components are addressed in the LRA.

17. Section 4.2.2 of the LRA states "The threshold for onset of neutron effects for RPV materials is conservatively defined to be a fast neutron fluence that exceeds 1E17n/cm

citing Appendix H of 10 CFR Part 50. The staff believes that Appendix H cites the indicated neutron fluence as a threshold below which a reactor vessel material surveillance program is not required for the vessel. Appendix H thereby creates in effect a " regulatory threshold" for neutron fluence, but clearly not a mechanistic threshold below

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which neutron effects do not occur Please provide your basis for concluding that there are negligible effects from neutron fluence below 1E17n/cnf.

18. Inconel alloy and stainless steel components become susceptible to IASCC at neutron '

fluence greater than 5E20 n/cnf . Since the flow skirt or flow baffle is located between the core and the reactor pressure vessel, the component would be expected to i experience a large neutron fluence. What is the peak fluence for this component and what are the consequences of neutron embrittlement for this component given any potential susceptibility to irradiation assisted stress corrosion cracking (IASCC) or stress corrosion cracMag (SCC)? Are there any other Inconel alloy components (such as the surveillance capsule holders) that receive a sufficiently large neutron fluence (greater than or equal to SE20 n/cnf) that are potentially susceptible to IASCC7 In such cases, -

i what is the peak fluence for these components and what are the consequences of  ;

neutron embrittlement on these components given their potential susceptibility to IASCC (or SCC)?

19. For the components identified with a plausible ARDM, identify any components which are not routinely inspected as a part of the ISI Program or any other program.

20. Section 4.2 indicates that the locations of interest for low cycle fatigue are the RPV main I coolant outlet nozzles and closure head flange studs. The report further indicates that all other RPV components and/or subcomponents are considered to have low susceptibility  !

to low-cycle fatigue. Describe the specific criteria used to determine that the other RPV l components and/or subcomponents have a low susceptibility to low-cycle fatigue.

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21. Section 4.2 indicates that the Fatigue Monitoring Program (FMP) monitors and tracks low-cycle fatigue usage for the selected components of the nuclear steam supply system and the steam generators. Describe the parameters that are monitored by the FMP that are applicable to the RPV Also describe how the monitored parameters are compared to the fatigue analysis of record.

22. Section 4.2 indicates that in order to stay within the design basis, corrective action is initiated well in advance of the cumulative usage factor approaching one or the number i of cycles approaching design allowable. Describe the criteria used to determine when '

corrective actions will be initiated.

23. Section 4.2 indicates that the FMP "will perform an engineering evaluation to determine if the low-cycle fatigue usage for the Control Element Drive Mechanisms (CEDMs)/ Reactor Vessel Level Monitoring System (RVLMS) components are bounded by the existing bounding components." Describe the fatigue criteria used for the design of the 4 CEDM/RVLMS components. Please indicate the reason the FMP is performing an engineering evaluation on these components. l i

24. Section 4.2 indicates that the current usage factors for the critical RPV components are well below one. Provide the usage factors prajected for the critical RPV components at the end of the proposed extended period of operation including a summary discussion of how they were derived.  ;

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25. Section 4.2 of the license renewal application indicates that the licensee in conjunction ,

with the Electric Power Research Institute has initiated an additional study to evaluate  :

the effects of low-cycle fatigue on various fatigue critical plant locations. Provide a 1 description of this study and describe its applicability to the Calvert Cliffs RPV and CEDM/RVLMS components.

I

26. Are there any parts of the systems, structures and components within the RPV or CEDM '

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 j

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 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 effects 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 adequawly managed.