ML20151X805

Forwards RAI Re Plant,Units 1 & 2 Integrated Plant Assessment Rept for Reactor Vessel Internals Sys.Response Requested within 30 Days of Receipt of Ltr

ML20151X805
Person / Time
Site:	Calvert Cliffs
Issue date:	09/03/1998
From:	Dave Solorio NRC (Affiliation Not Assigned)
To:	Cruse C BALTIMORE GAS & ELECTRIC CO.
References
TAC-M98835, TAC-M98837, TAC-M99181, NUDOCS 9809170259
Download: ML20151X805 (9)

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September 3,1998 Mr. Chart:s H. Cruss, Vica Pr:sid;nt g Nuclear Energy Division Baltimore Gas & 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, UNIT NOS.1 & 2, INTEGRATED PLANT ASSESSMENT REPORT FOR REACTOR VESSEL INTERNALS SYSTEMS (TAC NOS. M98835, M98837, AND M99181)

Dear Mr. Cruse:

By letter dated May 23,1997, Baltimore Gas and Electric Company (BGE) submitted for review the Reactor Vessel internals System (4.3) integrated plant assessment technical report as attached to the " Request for Review and Approval of System and Commodity Reports for License Renewal." BGE requested that the Nuclear Regulatory Commission (NRC) staff review report 4.3 to determine if the report meet the requirements of 10 CFR 54.21(a), " Contents of application-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 report 4.3 against the requirements of 10 CFR 54.21(a)(1),

10 CFR 54.21(a)(3). By letter dated April 4,1996, the 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 additional information is needed to complete its review.

Please provide a schedule by letter or telephonically for the submittal of your responses within 30 days of the receipt of this letter. Additionally, the staff would be willing to meet with BGE prior to the submittal Sincerely, hMb David L. Solorio, Project Manager License Renewal Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket Nos. 50-317 and 50-318

Enclosure:

Request for AdditionalInformation cc w/ encl: See next page DISTRIBUTION:

See next page DOCUMENT NAME G \ WORKING \SOLORIO\RVI RAl LTR ,

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. . j Mr. Charles H. . Cruse Calvert Cliffs Nuclzar Power Plant

_ Baltimore Gas & Electric Company Unit Nos.1 and 2 l cc:

President Mr. Joseph H. Walter, Chief Engineer l Calvert County Board of Public Service Commission of l _ Commissioners ' Maryland  !

175 Main Street Engineering Division Prince Frederick, MD 20678 6 St. Paul Centre Baltimore, MD 21202-6806 ,

' James P. Bennett, Esquire  !

Counsel Kristen A. Burger, Esquire Baltimore Gas and Electric Company Maryland People's Counsel P.O. Box 1475 6 St. Paul Centre l

Baltimore, MD 21203 Suite 2102  !

l Baltimore, MD 21202-1631  !

Jay E. Silberg, Esquire Shaw, Pittman, Potts, and Trowbridge Patricia T. Birnie, Esquire l 2300 N Street, NW Co-Director Washington, DC 20037 Maryland Safe Energy Coalition i P.O. Box 33111 l Mr. Thomas N. Prichett, Director Baltimore, MD 21218 NRM Calvert Cliffs Nuclear Power Plant Mr. Loren F. Donatell l 1650 Calvert Cliffs Parkway NRC Technical Training Center Lusby, MD 20657 4702 5700 Brainerd Road

, Chattanooga, TN 37411-4017  :

! Resident inspector _

U.S. Nuclear Regulatory Commission David Lewis P.O. Box 287 Shaw, Pittman, Potts, and Trowbridge St. Leonard, MD 20685 2300 N Street, NW Washington, DC 20037 j Mr. Richard I. McLean l

Nuclear Programs Douglas J. Walters

Power Plant Research Program Nuclear Energy institute l Maryland Dept. of Natural Resources 1776 l Street, N.W.

l Tawes State Office Building, B3 Suite 400 l

Annapolis, MD 21401 Washington, DC 20006-3708 DJW@NEl.ORG l Regional Administrator, Region 1  !

U.S. Nuclear Regulatory Commission Barth W. Doroshuk l 475 Allendale Road Baltimore Gas and Electric Company

King of Prussia, PA 19406 Calvert Cliffs Nuclear Power Plant i 1650 Calvert Cliffs Parkway  ;

l NEF ist Floor  !

l Lusby, Maryland 20657 1

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HARD COPY SeulNWIIDs PUBLIC PDLR R/F MEl-Zeftswy DISTRIBUTION: E-MAIL:

FMiraglia (FJM)

JRoe (JWR)

DMatthews (DBM)

CGrimes (ClG)

TEssig (THE) . 4 GLainas (GCL) .

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JStrosnider (JRS2)

GHolahan (GMH) )

SNewberry (SFN) '

GBagchi(GXB1)

RRothman (RLR)

JBrammer (HLB) ,

CGratton (CXG1)- l

- JMoore (JEM)

MZobler/RWeisman (MLZ/RMW)

SBajwa/ADromerick (SSB1/AXD)

- LDoerflein (LTD)-

. BBores (RJB)

SDroggitis (SCD)

RArchitzel(REA)  !

' CCraig (CMC 1) i LSpessard (RLS) i RCorreia (RPC)

RLatta (RML1)

EHackett (EMH1) .

AMurphy (AJM1)

TMartin (TOM 2)

DMartin (DAM 3)

GMeyer (GWM)

WMcDowell(WDM)

SStewart (JSS1)

THiltz (TGH)

SDroggitis (SCD)

DSolorio (DLS2)

PDLR Staff TSullivan (EJS)

BElliot (BJE)

AHiser (ALH,1)

RWessman (RHW)

. FGruberlich (F?G)

Shou (SNH)

Slittle (SSL) 4

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REQUEST FOR ADDITIONAL INFORMATION l CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NOS.1 & 2 '

REACTOR VESSEL INTERNALS INTEGRATED PLANT ASSESSMENT. SECTION 4.3  ;

DOCKET NOS. 50-317 AND 50-318 Section 4.3.1 - Sconina

1. Figure 3.3-6 (Rev. 21) of the Calvert Cliffs Nuclear Power Plant (CCNPP) Updated Final Safety Analysis Report (UFSAR) shows the fuel assembly hold down (FAHD) structure.

One of the intended functions of FAHD is to prevent fuel assemblies from being lifted out of position under accident loading conditions. Please clarify whether the FAHD was i subjected to an aging management review (AMR), particularly the springs in it, which I may loose their required force at extended age.

2. Figure 3.3-14 (Rev. 21) of the CCNPP UFSAR shows the upper guide structure (UGS)

Assembly. Please describe the functions of the Expansion Compensating Ring, and indicate if its intended functions would meet the definition of intended function listed in 10 l CFR 54.4(a).  ;

3. Section 4.1.3.6 (Rev.18) 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  !

Leamed Report, NUREG-0737, item ll.B.1. One of the intended functions of the vents is i 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 license renewal application (LRA). If not, provide the basis for their exclusion.

Section 4.3.2 - Aalna Manaaement j

4. Clarify whether all the reactor vessel internal (RVI) components listed in Table 4.3-1 are within the scope of the ASME Code,Section XI, Subsection IWB inservice inspection program, as mentioned in Page 4.3-12. In addition, describe the applicable acceptance criteria and describe the methods used for trending for the visual inspection.

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 control element assembly (CEA) shroud bolts and core shroud tie rods and tc!ts for verifying low tensile stress during normal operations, and for justifying 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.

6. Page 4.3-24 indicates that an examination of the CEA shroud bolts and core shroud tie rods and bolts would be conducted as a part of an age related degradation inspection (ARDI) program if the refined stress level does not show the low stress expected.

Assuming the results did warrant an ARDI for these components, provide a summary Enclosure

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oiscussion of the ARDis consistent with the NRC staff's request for additionalinformation i on ARDIs in letter dated August 28,1998, " Request for Additional Information For the Review of the Calvert Cliffs Nuclear Power Plant, Unit Nos.1 & 2, Integrated Plant Assessment Report." i i

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 i 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.

8. Do the RVI intended functions include: (a) support for the irradiation surveillance capsules, and (b) shielding for the reactor pressure vessel (RPV)? If so, summarize ,

what components perform these intended functions and explain whether these components are within the scope of license renewal.

9. CCNPP license renewal application addresses certain applicable aging effect for specific 1

reactor vessel internals 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 upper guide structure support plate; SCC, lASCC, corrosion, and wear for the control element assembly (CEA) shrouds; IASCC and corrosion for the CEA shroud bolts; SCC, IASCC, corrosion, and wear for the fuel alignment plate; SCC, lASCC, corrosion for the core support barrel; SCC, lASCC, corrosion, and neutron embrittlement for the core support barrel upper flange; SCC, IASCC, and corrosion for the core shroud, SCC, lASCC, corrosion, and stress relaxation for the core shroud assembly bolts; SCC, IASCC, and corrosion for the core shroud tie rods; SCC, lASCC, and corrosion for the fuel alignment plate guide lugs; SCC, lASCC, and corrosion for the core support plate; SCC, IASCC, corrosion, and stress relaxation for the fuel alignment pins; SCC, IASCC, and 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.

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 RVI were initially included in the AMR. Section 4.3.1.2 of the LRA further indicates,"some components were determined not to be within the scope of license renewal since they are not required for the RVI to perform their intended function." Describe which components were considered to be outside the scope of license renewal and clarify the criteria that were us ed 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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11. Secton 4.3.2 of the LRA indicates that IASCC is not plausible for Calvert Cliffs RV1 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

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l laboratory tests. Identify the operating plant experience and laboratory test data that forms the basis for this conclusion. Identify the RVI components si Calvert Cliffs that are subject to a neutron fluence greater than 5x1 E20 n/cm2. For these components, identify the temperature, oxygen, radiation levels and stress levels. What inspections or aging management programs (AMP) 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, " CRACKING OF REACTOR VESSEL INTERNAL BAFFLE FORMER BOLTS IN FOREIGN PLANT" impact this ovaluation? Since bolt cracking has occurred at the junction of bolt head and shank, which is not accessible for visual inspection, how will core shroud and bolts (CEASB) and other RVI bolting that is subject to lASCC be examined? What inspections or aging management programs (AMP) 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?

12. Section 4.3.2 of the LRA indicates, " procedures will be enhanced if modified to specifically identify each component of the RV1 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.

13. Section 4.3.2 of the LRA indicates that of the three U.S. suppliers of light water reactor the most fatigue-susceptible RV1 components have been identified for pressurized water reactor (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 of exceeding the usage factor of 0.5 for the components normally evaluated.

14. Section 4.3.2 of the LRA indicates, " Thermal aging is potentially significant for. (1) centrifugally-cast parts with detta ferrite content above 20%; (2) statically-cast parts with molybdenum content meeting CF3 and CF8 limits and with a detta 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 be determined? What are the uncertainties in these test methods? How are the uncertainties incorporated into the estimate of the delta ferrite?

If the detta 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 l the critical flaw size at the end of the license renewal term for these limits. Identify the i inspection procedures and the capability of the inspection to detect flaws smaller in size than the critical flaw size.

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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?

16. Table 4.3-2 indicates erosion, erosion / corrosion, general corrosion / uniform attack, hydrogen damage and pitting / crevice corrosion are not plausible. Explain the bases for these conclusions.

17. Section 4.3.2 indicates stress corrosion cracking /lGSCC/intergranular attack are potential age related degradation mechanism (s) (ARDM(s)) for RV1 components fabricated from 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 reactor vesselintemal components that were fabricated using this material or other nickel base alloys and describe the aging management program that will be used during the extended period of operation to ensure these components are not susceptible to primary water stress corroslor , racking.

18. Table 4.3 indicates that many components (CEASB, CS, CSTR, CSB, CSC, CSP, FAPFP, 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 degradation (i.e., smaller crack sizes) that are lower than those if the part was not in an embrittled condition. Identify for each component that is susceptible to neutron embrittlement, the peak neutron fluence at the end of the extended period of operation, and the materials used to fabricate the specific component. For the limiting component (censidering 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 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.

19. Section 4.3.2 states that "No instances of degradation of RVi for PWRs have been recorded which have definitely been attributed to neutron embrittlement," and "Calvert Cliffs has not discovered any thermal-aging related damage for the RIV. Also there have not been RVI damage events at other PWRs that were identified as thermal aging failur'e." Based on the staff's experience the degradation in material 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.

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20. Section 4.3.1.1 of the LRA indicates that the aging evaluation of RV1 *creets" 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 I control? Describe the " credits" assumed for each ARDM and affected component and justify the credits assumed.

21. Section 4.3 indicates that changes in the design of the hold down rings (HDRs) installed  !

at Calvert Cliffs Units 1 and 2 were made as a result of wear experienced in a similar component at uther reactor plant and the discovery for the need to provide for additional fuel assembly growth. Table 4.3-1 identified the HDRs 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 during a refueling outage, and the RVI 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 deflecting the inner and outer edges of the ring spring in a direction because flattening of the ring. In deflecting the HDR, the outer edge of the top surface l and inner edge of the bottom surface of the ring contact and load the Pressure Vessel '

Closure and the Upper Guide Structure (UGS) flange, respectively. Provide a description '

of the accessibility to the bottom surface of the HDR that contacts the UGS flange, the l UGS flange and the undersurface of the vessel closure for visualinspection. Your j description should account for the accuracy required in the use of visual indications of l detectable wear to reliably determine changes in the HDR load developing capability. ,

t in addition, any such wear, if it occurs, may gradually reduce the HDR clamping force  !

and induce core 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.

22. 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 HDR has not and will not be reduced to impair its intended function during the period of extended operation.

23. Describe the visual examinations of the CEASB that have been previously performed or that will be performed to maintain the structural integrity of the RVI consistent with the current licensing basis during the period of extended operation. Describe the portions 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.

24. Are there any parts of the systems, structures and components within the RV1 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 of

t inaccessible 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 adequately managed.

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