ML20195J613
| ML20195J613 | |
| 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 9811250031 | |
| Download: ML20195J613 (18) | |
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o CnAnt.ra H. CRtJSE Baltimore Gas and Electric Cornpany Vice President Calvert Cliffs Nuclear Power Plant
. . , Nuclear Energy 1650 Calvert Cliffs Parkway Lusby, Maryland 2%57 410 495-4455 November 19,1998 I
l U. S. Nuclear Regulatory Commission Washington, DC 20555 ATI'ENTION: Document Control Desk
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, Units 1 & 2, Integrated Plant Assessment Report for the Reactor Pressure Vessels and Control Element Drive Mechanism / Electrical System. and Errata
REFERENCES:
(a) Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated July 30,1997, " Request for Review and Approval of Reactor Pressure Vessels and Control Element Drive Mechanism / Electrical System Report for License Renewal" (b) Letter from Mr. D. L. Solorio (NRC) to Mr. C. H. Cruse (BGE),
August 26,1998," Request for Additional Information for the Review of the Calvert Cliffs Nuclear Power Plant, Unit Nos.1 & 2, Integrated Plant Assessment Report for the Reactor Pressure Vessels and Control Element Drive Mechanism / Electrical" (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 Cliffs Nuclear Power Plant License Renewal Application Submitted by the Baltimore Gas and Electric Company" l
Reference (a) forwarded the Baltimore Gas and Electric Company (BGE) system report for license l- renewal for the Reactor Pressure Vessels and Control Element Drive Mechanism / Electrical System.
Reference (b) forwarded questions from NRC staff on that report. Reference (c) forwarded a numbering system for tracking BGE's response to all of the BGE License Renewal Application requests for f
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 Reference (c).
Attachment (2) provides errata for Section 4.2, Reactor Pressure Vessels and Control Element Drive Mechanism / Electrical System, of the BGE LRA.
! 9811250031.981119
! PDR ADOCK 05000317 NRC Distribution Code A036D
Document Control Desk NovGmber 19,1998 l Page 2 l
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Should you have further questions regarding this matter, we will be pleased to discuss them with you.
l Very truly yours, STATE OF MARYLAND :
- TO WIT:
COUNTY OF CALVERT :
I, Charles H. Cruse, being duly sworn, state that I am Vice President, Nuclear Energy Division, i Baltimore Gas and Electric Company (BGE), and that I am duly authorized to execute and file this i l
response on behalf of BGE. To the best of my knowledge and belief, the statements contained in this j i 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 I information has been reviewed in accordance with company practice and I believe it to be reliable.
/
Subscribed and sworn before me, a Notary Public in and for the State of Maryland and County of Calo4tf ,this /9 dayof /An>wfu ,1998.
WITNESS myIland and Notarial Seal:
Notary Public l My Commission Expires: [ . OOOd GDate' CllC/KRE/ dim l Attachments: (1) Response to Request for Additional Information: integrated Plant Assessment i
Report for the Reactor Pressure Vessels and Control Element Drive l Mechanism / Electrical System
- (2) Errata to Section 4.2, Reactor Pressure Vessels and Control Element Drive Mechanism / Electrical System; License Renewal Application 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. I. McLean, DNR II. J. Miller, NRC J. II. Walter, PSC 1
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4 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISMS / ELECTRICAL SYSTEM I
Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant November 19,1998
j A'ITACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISM / ELECTRICAL SYSTEM NRC Ouestion No. 4.2.1 We noted that page 4.3-5 of Section 4.3, " Reactor VesselInternals,"(of the Baltimore Gas and Electric Company (BGE) License Renewal Application (LRA)] 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 LRA. . However, Figure 4-2 (Revision 18) provided in Chapter 4 of the Calvert Cliffs Nuclear Power Plant (CCNPP) Updated Final Safety Analysis Report (UFSAR) for Units 1 and 2 shows a component attached to the closure head of the reactor pressure vessel (RPV), 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 1 addressed in th: LRA. If not, provide the basis for their exclusion.
BGE Response The lifting lugs were included as an integral part of the RPV closure head plates. The closure head plates were evaluated for aging management as described in Section 4.2.2 of the BGE LRA.
!- NRC Ouestion No. 4.2.2 Figure 4-2 (Revision 18) of the CCNPP UFSAR shows that the closure head insulation is attached to the closure head of RPV. 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(aXIXi)-(iii).
BGE Response Insulation performs none of the intended functions listed in Section 4.2.1.1 on page 4.2-5 and, j therefore, is not within the scope oflicense renewal.
l NRC Onestion No. 4.2J Please clarify whether the components identified in comment (d) of Table 4.2-2 of Section 4.2.1 as a
" Core Stop Lug" is the same component labeled as the core support lug in Figure 4-2 (Revision 18) l provided in Chapter 4 of the CCNPP UFSAR. If these components are not the same, please describe the l 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(aXIXi)-(iii).
l l BGE Response The " core stop lug," referred to in comment (d) of Table 4.2-2, Section 4.2.1, of the BGE LRA is the i same as the " core support lug" in Figure 4-2 provided in Chapter 4 of the CCNPP UFSAR.
NRC Ouestion No. 4.2.4 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 (page 4.2-14) of the LRA7 l
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S ATTACHMENT (1) i RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION;
} INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECIIANISM/ ELECTRICAL SYSTEM i
BGE Resnonae l
Baltimore Gas and Electric Company believes that Section 4.2.2 (page 4.2-14) of the LRA already describes the changes made to the Boric Acid Inspection Program as a result of operational experiences.
"The Boric Acid Inspection Program has evolved to account for operational experience. Reactor Coolant System leakage has been discovered that has affected the focus of subsequent inspections. For example: (1) A Unit 2 pressurizer heater sleeve was discovered to have a leak, and as a result, CCNPP instituted a more comprehensive inspection of the pressurizer heater sleeves in Unit 1; and (2) A seal vent line in containment developed a leak which dripped onto
! and caused surface corrosion of a RCS elbow resulting in increased attention with the program to l such leaks."
" Additionally, the program has evolved with regard to the qualification level of personnel for evaluating boric acid ler.ks. The program dictates a minimum qualification level of non-destructive examination Level II Inspector for the evaluation of boric acid leaks. Any person conducting a walkdown or inspection may discover boric acid leakage. Such leakage would then be documented in an Issue Report (IR) by the individual discovering the leak, and routed to the l ISI group for closer inspection and evaluation by a Level II Inspector. This approach provides for wide boric acid leakage inspection coverage, but ensures boric acid leakage and its effects are evaluated by qualified individuals."
NRC Ouestion No. 4.2.5 Pursuant to 10 CFR Part 50, Appendix G, provide an analysis of the vessel beltline material to demonstrate that they will maintain at least 50 fl-lb 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.
BGE Response Attachment (1) of Reference (1), response to Question 2a, provides the initial USE for all CCNPP RPV plate materials and four of the six seam weld materials. This response states: "All beltline plates contain .15% copper or less. For the bounding combination (74 ft-lbs. and 0.15% copper) the USE is not predicted to fall below 50 ft.-lbs. prior to reaching a fluence value of approximately 3.65 x 10" n/cm 2at the 1/4t thickness position. A fluence value of 3.65 x 10" n/cm 2at the 1/4t thickness l position corresponds to a vessel wall fluence of 6.12 x 10" n/cm2 , which will not be reached in either vessel prior to the expiration of the current license." Both units' are still not predicted to exceed the fluence values provided in Reference (1). As shown in Attachment 1, Table 1, of Reference (2), the predicted end-of-life (EOL) fluence at the vessel wall for 48 effective full power years (EFPY)
. (60 years) for Unit 1 is 4.95 x 10" n/cm 2. The EOL fluence at the vessel wall for 48 EFPY for Unit 2 2
l is 5.77 x 10" n/cm .
Reference (1) also states: "The minimum measured USE for any Calvert Clifts beltline weld is 110 ft.-lbs. The maximum copper content in any beltline weld is 0.23%. For the bounding combination (110 ft.-lbs. and 0.23% copper), the USE is not predicted to fall below 50 ft.-lbs. prior to reaching a fluence value of approximately 6 x 10" n/cm2 , which will not be reached at the 1/4t thickness position in either vessel before the expiration of the current license." Additionally, l
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neither vessel will exceed 6 x 10 n/cm at the 1/4t thickness before the end of any extended operation up to 48 EFPY. This applies to Unit I Weld Seams 2-203-A,B,C and 9-203 and Unit 2 Weld Seams 3-203-A,B,C and 9-203. Baltimore Gas and Electric Company addressed Unit 1 Weld Seam 3-203-A,B,C and Unit 2 Weld Seam 2-203-A,B,C ia repara*e Generic Letter 92-01 responses to NRC.
For the Unit 1 Weld Seam 3-203-A,B,C, Reference (3) piavides the ir.itial USE value of 109 fl.-lbs.
Our predicted 48 EFPY EOL USE for this weld seam is 63 ft.-lbs. 'lb prediction uses a 1/4t fluence 2
of 3.0 x 10 n/cm and a copper content of 0.17 w/o, which corresponds to a Regulatory Guide 1.99, l
" Radiation Embrittlement of Reactor Vessel Materials," predicted decrease of 42%.
For the Unit 2 Weld Seam 2-203-A,B,C, Reference (4) provides the initial USE value of 83.5 ft.-lbs.
Our predicted 48 EFPY EOL USE for this weld seam is 50 ft.-lbs. The prediction uses a 1/4t fluence 2
of 3.4 x 10 n/cm and a copper content of 0.16 w/o, which corresponds to a Regulatory Guide 1.99 predicted decrease of 40%.
No RPV beltline material will fall below the 50 ft.-lbs. USE required by 10 CFR Part 50, Appendix G, before 48 EFPY.
NRC Ouestion No. 4.2.6 Provide an outline of the Reactor Vessel Material Surveillance Program and discuss how they will be used to monitor neutron irradiation for the RPV beltline materials during the period of extended operation. Provide a summary of "CCNPP Comprehensive Reastor 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?
BGE Response Page 4.2-25 of the LRA provides a description our material surveillance program. Additionally, Reference (1) provides a detailed description of our surveillance program and why it meets the requirements of 10 CFR Part 50, Appendix H. Calvert Cliffs currently does not have a 60-year capsule withdrawal schedule. However, CCNPP will submit a revised capsule withdrawal schedule to NRC for approval by 2003.
Pages 4.2-25 and 26 provides a summary description of the Comprehensive Reactor Vessel Surveillance Program.
Page 4.2-25 states that each unit has one standby capsule. Additionally, Enclosure (2) of Reference (5) shows that each unit has four remaining capsules. One capsule for each unit is noted as a standby capsule. Unit I also has two supplemental capsules that are designated for testing in 2000 and 2012.
I NRC Ouestion No. 4.2.7 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 a Duke Energy letter to the NRC, dated April 22,1998, i
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- l. ELEMENT DRIVE MECHANISM / ELECTRICAL SYSTEM BGE Response
, Reference (6) includes data from McGuire capsules U, X, and V. The 40- and 60-year PTS values for l Unit I Weld Seam 2-203-A,B,C were determined using copper and nickel values and the chemistry factor from the McGuire capsules. The NRC issued a Safety Evaluation Report on January 2,1996 approving our PTS submittal. The SER states: "the staff concluded that CCNPP, Unit Nos. I and 2, i i
beltline materials are projected to be below the PTS screening criteria 20 years after expiration of !
l their licenses."
I The LRA was submitted before the April 22,1998 Duke Energy Letter to NRC. Ilowever, BGE has received and evaluated the data from McGuire capsule Y. Table 1 of Reference (2) includes the l capsule Y data. The inclusion of the new capsule Y data did not result in a "significant" change in the '
l adjusted reference temperature and, in accordance with 10 CFR 50.61, no reporting is required.
NRC Ouestion No. 4.2.8 Provide pressure-temperature 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 i there be any limitations in operation of the Shutdown Cooling System due to American Society of i Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Appendix G, pressure- !
j temperature operating limits and the minimum permissible temperature of the reactor vessel? ;
BGE Response Baltimore Gas and Electric Company understands that NRC will issue a replacement question for this item, l
i NRC Ouestion No. 4.2.9 As identified in Section 4.2.1.1 of the submittal under " Unintentional inclusion of Slag Stringer in RPV",
the fabrication flaw in the Unit I reactor vessel weld was stated as acceptable in accordance with the applicable ASME Code,Section XI, during the pre-service and subsequent inservice inspections.
l 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 location of the flaw within the weld. If the location of the l flaw designates it as a surface planar flaw !a 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 temperatures (RT gor), delta RT nor, margin, neutron fluence, and chemistry (Copper and Nickel) of the weld containing the flaw in accordance with Regulatory Guide 1.99, Revision 2.
, RGE Response l '
The weld that contains the slag stringer flaw was examined during the 1998 10-year inservice j inspection (ISI) of the CCNPP Unit i Reactor Vessel. These examinations were performed in accordance with the ASME Section XI,1983 Edition with Addenda through with Summer 1983,
, using procedures that were qualified in accordance with the Performance Demonstration Initiative l program. During the 1998 examinations, the slag stringer did not provide ultrasonic response to 4
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ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATIONt INTEGRATED PLANT ASSESSMENT REPORT FORTHE REACTOR PRESSURE VESSELS AND CO ELEMENT DRIVE MECHANISWELECTRICAL SYSTEM warrant being a recordable indication. Therefore, no evaluation in accordance with IWB-3000 was necessary.
Baltimore Gas and Electric Company is planning to update to a later edition of ASME Section XI sometime in 1999. Later editions of ASME Section XI no longer assume a 40-year life for 1 components requiring inspections. Table IWB-2500-1 now has the requirements for the l "1" Inspection Interval" and " Successive inspection Intervals", as opposed to previous editions that had the requirements for the "1" Inspection Interval" and " Successive Inspection Intervals, 2"d, 3'd, 4*". Additionally,Section XI has addressed flaw evaluations and flaw acceptance criteria for greater than 40-year operations, as evidenced by the changes in Table IWB-3510-1. In future exams, the slag stringer will be inspected and evaluated under the requirements of the later editions of the Code.
The flaw is located in the Unit 1 Reactor Vessel Weld 2-203C in the middle shell course at 270 and 143 inches below the RPV flange surface. The pre-service examination and the 198610-year ISI ultrasonic examination determined the flaw is a subsurface flaw. The flaw is not a surface flaw on the inside surface of the RPV. No specific PTS analysis is required for this flaw.
l NRC Ouestion No. 4.2.10 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.
BGE Response The Alloy 600 Program is described in Section 4.2.2, Group 5 (Stress Corrosion Cracking [ SCC)), of the BGE LRA. The existing Alloy 600 Program will continue to be utilized during the period of extended operation.
NRC Ouestion No. 4.2.11 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 nozzles? In particular, j discuss how BGE intends to extend its commitments to Generic Letter 97-01, " Degradation of Control Rod (Element) Drive Mechanism Nozzles & Other Vessel Closure Head Penetrations", over the j proposed extended term of operation for the CCNPP units. Include in the discussion an assessment and reanalysis of the CCNPP CEDM nozzles using the latest crack initiation and growth model that was l developed for. the Combustion Engineering Owners Group to assess postulated flaws in Combustion !
Engineer-designed CEDM penetration nozzles. With respect to this reanalysis, provide what the probability will be for cracks to have initiated in the CEDM penetration nozzles 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 flawr in the CEDM penetration nozzles at the end of the current license and at !
the end of the proposed extentled operating' terms. Identify any volumetric examination of Calvert Cliffs
} or of other plants CEDM peretration nozzles that will confirm your susceptibility analysis that crackmg l will not occur during the lice ase renewal term. I i
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RESPONSE'TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR PRESSURE VESSELS AND CONT ELEMENT DRIVE MECHANISM / ELECTRICAL SYSTEM i
RGE Response f
- Baltimore Gas and Electric Company's commitment to Generic Letter 97-01 was not limited to the current license period. The probabilistic inspection timing model described in Reference (7)(which formed the basis for LBGE's response to Generic Letter 97-01) was not limited in its end date to the end of the current licensing period.
Baltimore Gas and Electric Company has received a request for additional information (Reference 8) pedaining to our response to Generic Letter 97-01, which is redundant with this question. Baltimore <
Gas and Electric Company's response to this request will address the subject matter of this question.
NRC Omention No. 4.2.11 Table 4.2-2 of the submittal identifies SCC of the RPV flow skirt as being a plausible age-related degradation mechanism (ARDM). Discuss how BGE intends to 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 weld? i BGE Response The flow skirt is welded to nine core stop lugs. The core stop lugs are circumferentially equally :
spaced around and welded to the RPV interior surface. These welds are visually inspected during each 10-year ISI inspection outage. Flow induced vibration could cause high cycle fatigue consequences if the frequency and amplitude of the vibrations are the same as the natural harmonic of the component. If flow-induced vibration were to affect the integrity of these welds, the effects would have manifested themselves early in plant operations. The results of past visual inspections have not shown any degradation resulting from flow skirt vibration. 1 NRC Question No. 4.2.13 In Section 4.2-2 of the submittal, you have identified the ARDMs for various RPV components. Based on these aging mechanisms, how will your ISI program be tailored to monitor age-related degradation due to these mechanisms for these components? Is there any weld 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 welds during the renewal term.
l BGE Response Baltimore Gas and Electric Company will not have to tailor the CCNPP ISI Program to monitor age-l '
. related degradation of the.RPV components. The existing ISI Program is already credited with the
! discovery, per ASME XI, and management of the effects of mechanical wear (Group 2, Section 4.2.2), general corrosion (Group 1, Section 4.2.2), and SCC (RPV anchor bolts in Group 5, j Section 4.2.2) on those RPV components susceptible to these ARDMs. With respect to the RPV, )
I there are no welds for which BGE cannot perform an examination due to physical constraints. '
However, there are welds for which a 100% weld examination cannot be performed due to physical constraints. For welds that could not be examined 100% during the latest 10-year ISI exam on Unit 1 l (1998), BGE will apply for relief from the Code-required examinations by June 2000. The results from the next 10-year ISI examination of the Unit 2 RPV will determine if any further relief from Code requirements will be necessary.
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ELEMENT DFIVE MECHANISM / ELECTRICAL SYSTEM NRC Ouestion No. 4.2.14 Based on its evaluation of operating experience, the NRC has determined that potential aging effect mechanisms in components of pressurized water reactor vessels are as indicated in Table 3.1-3 of the NRC Draft Standard Review Plan for License Renewal (Reference 9). Table 3.1-Y identifies components that are considered part of the RPV and identifies the associated aging etTects 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.
BGE Response The following table maps the components and aging mechanisms in Table 3.1-3 from Reference (9) to Section 4.2 of the BGE LRA for the Calvert Cliffs RPVs. This table does not contain Creep and Hermal Aging since Table 3.1-3 of Reference (9) listed these two aging mechanisms as being non-significant for the listed RPV components.
Legend:
N: Non-significant; S: Significant; U: Unresolved IGSCC: Intergranular Stress Corrosion Cracking Note 1: Erosion corrosion was determined to not be an aging mechanism for the closure head and RPV Flanges. Any leakage in the flange region will be detected and corrective actions taken to repair any damage before plant power operation would resume.
Note 2: The control rod drive mechanism.(CRDM) housings are fabricated from austenitic stainless steels which are not susceptible to general corrosion.
note 3: Calvert Cliffs does not have a shroud support ring. There is a core Shroud that sits on a 1-core support plate, which is part of the core support barrel. These components are evaluated for aging management in Section 4.3 of the BGE LRA.
Note 4: The closure head studs and nuts are fabricated from low alloy materials with a specified maximum hardness of 42 Rc or less, which makes them not susceptible to SCC.
Note 5: Neutron embrittlement is only plausible for the lower portion of the nozzle shell courses. Baltimore Gas and Electric Company analysis of the nozzle shell courses l- shows the material in this shell will not embrittle to the degree that PTS or upper-shelf i- issues become relevant.
l-Note 6: The inlet and outlet nozzles are not susceptible to neutron embrittlement because they are located in the nozzle shell courses. See Note 5 above.
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ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR TIIE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISM / ELECTRICAL SYSTEM l NRC Equivalent BGE ARDMs Neutron SCC /lGSCC Corrosion Erosion Wear j ' RPV Component . RPV Component . Embrittlement - (4.2 Group 5) (4.2 Group 1) (4.2 Group 2)
. (Note 1) '
' (4.2 Group 4) -
Closure ficad Closure ficad NRC N U S N N Dt,me BUE Not Plausible - Not Plausible Plausible Not Plausible . Not Pisusible
- CRDM CRDM Upper NRC N U S N N l flousing flousing BGE Not Plausible Not Plausible Not Plausible Not Plausible Plausible (Note 2)
Refueling Seal Permanent Cavity NRC N U N N N Ledge Seal I;ing l See Section 3.3A BGE Not Plausible . Not Plausible > Plausible Not Plausible Not Plausible - )
! (Group 4) .
Closure ficad integral part of the NRC N U N N N Lifting Lugs Closure llead BGE . Not Plausible - Not Plausible Plausible Not Plausible Not Plausible Shroud Support See $cction 4.3 NRC N U N N N Ring (Note 3) BOE N/A N/A N/A. .N/A N/A -
Closure ficad Closure ficad NRC N U S S S Flange Flange BGE. Not Plausible Not Plausible Plausible Not Plausible l
. Plausible Closure Stud Closure licad Studs NRC N S S N S q Assembly Nuts, and Washers BGE. . Not Plausible Not Plausible Plausible Not Plausible flausible (Note 4)' l Vessel Flange RPV Flange NRC N U S S S BGE Not Plausible Not Plausible Plausible Net Plausible Plausible Leakage RPV Leakage NRC N U N N N Monitoring Tube Monitoring Tube BGE Not Plausible . Plausible . Not Plausible Not Plausible Not Plausible Upper (Nozzle) Nozzle Shell NRC S U N N N Shell Courses BGE Plausible-Note $ Not Plausible Not Plausible Not Plausible Not Plausible Primary Coolant inlet and Outlet NRC S U N N N Nozz!cs Nozz!cs BGE Not Plausible Not Plausible Not Plausible Not Plausible Not Plausible (Note 6)
Intermediate and intermediate and NRC S U N N N Lower Shell Lower Shell BGE Plausible Not Plausible Not Plausible Not Plausible Not Plausible Core Support Pads Core Stop and NRC N U N N S (Lugs) Stabilizing Lugs BGE Not Plausible - Plausible Not Plausible Not Plausible Plausible Bottom llead Lower licad Plates NRC N U N N N Dome BGE Not Plausible . Not Plausible Not P!:.usible . Not Plausible Not Plausikla instrumentation Instrumentation NRC N U N N N Tube /Penetratisns Tubes / Nozzles BGE Not Plausible Plausible Not Plausible Not Plausible Not Plausible NRC Ouestion No. 4.215 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 irs, where boric acid j 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 irs?
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ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR Tile REACTOR PRESSURE VESSELS AND CONT ELEMENT DRIVE MECilANISM/ ELECTRICAL SYSTEM BGE Respanac Baltimore Gas and Electric Company uses the Boric Acid Inspection Program with ISI personnel to visually examine (VT-2 [a type of visual examination described in ASME XI, IWA-2212])
containment after attaining hot standby, and a second time prior to plant startup. During the first inspection after shutdown, the Boric Acid Inspection Program is used by ISI personnel to examine all of the areas / components where boric acid has the potential for leaking and causing corrosion. If leakage is found during this initial walkdown, an IR is written and corrective actions are taken. A second ISI walkdown is performed prior to plant startup to examine those areas where boric acid leakage was found (if any) during the initial walkdown. The ISI must ensure that all components that were the subject ofIRs where boric acid leakage has been found, are examined in accordance with the requirements of the program.
NRC Ouestion No. 4.2.16 Section 43 of the LRA indicates that the core support barrel snubber and snubber bolts are addressed in 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.
BGE Response Baltimore Gas and Electric Company evaluated the core support barrel snubber (referred to as snubber spacer blocks) and snubber bolts (referred to as capscrews) for aging managemem review in Section 4.2 of the LRA. In particular, these items are mentioned on pages 4.2 8,4.2-9, and 4.215 (Group 2 -wear). The only plausible aging mechanism for the snubber spacer blocks was wear, which ;
is managed by the ISI Program. There were no plausible aging mechanisms for the snubber spacer block capscrews.
1 NRC Ouestion No. 4.2.17 '
Section 4.2.2 of the LRA states: "The threshold for onset of neutron effects for RPV materials is 2
conservatively defined to be a fast neutron fluence that exceeds IE17n/cm . citing Appendix II of 10 CFR Part 50. The staff believes that Appendix 11 cites the indicated neutron fluence as a threshold below which a reactor vesse! material surveillance program is not required for the vessel. Appendix l-I l thereby creates in effect a " regulatory threshold" for neutron fluence, but clearly not a mechanistic threshold below which neutron effects do not occur. Please provide your basis for concluding that there j are negligible effects from neutron fluence below lE17n/cm2 , ,
BGE Response The first sentence on BGE LRA page 4.2-24 should read: "The effects of neutron fluence on RPV 2
materials at a level less than 1E17n/cm are negligible."
NRC Ouestion No. 4.2.18 Inconel alloy and stainless steel components become susceptible to irradiation assisted stress corrosion 2
cracking (IASCC) at neutron fluence greater than SE20 n/cm . Since the flow skirt or flow baffle is located between the core and the RPV, the component would be expected to 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 IASCC or SCC 7 Are there any 9
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FORTHE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISM / ELECTRICAL SYSTEM other Inconel alloy componer.!s (such as the surveillance capsule holders) that receive a sufficiently large 2
neutron fluence (greater than or equal to SE20 n/cm ) that are potentially susceptible to IASCC7 In such cases, 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).
BGE Response In accordance with Section 43.2 of the LRA, BGE does not consider IASCC to be a plausible degradation mechanism for stainless steel or Inconel components. The response to NRC Question No. 4.2.11 provides the basis for this detennination.
No Inconel alloy component is predicted to receive a fluence equal to or greater than SE20 n/cm2, either at the end of the current licensing period, or at the end of a 20-year extended license period '
NRC Ouestion No. 4.2.19 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.
l BGE Response The following RPV components are not routinely inspected as part of the ISI Program or any other program:
e RPV Leakage Monitoring Tube; e CEDM Thermal Sleeves; e RPV Flow Skirt; e RPV Surveillance Capsule Holders; e RPV Core Stop Lugs;
- CEDM Upper and Lower End Fittings;
- CEDM Motor Housing; e CEDM Upper Pressure Housing Tube; e Reactor Vessel Level Monitoring System Upper and Lower End Fittings; and o' Reactor Vessel Level Monitoring System Upper Pressure Housing Tube.
NRC Ouestion No. 4.2.20 Section 4.2 indicates that the locations ofinterest for low cycle fatigue are the RPV main 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 l criteria used to determine that the other RPV components and/or subcomponents have a low susceptibility to low-cycle fatigue.
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ATTACHMENT (1) i RESPONSE TO REQUEST FOR AD91TIONAL INFORMATION; 1 INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECIIANISM' ELECTRICAL SYSTEM j MGJuana Pag: C-21 states: "The bounding locations for low-cycle fatigue are the RDV closure studs arJ RPV outlet coolant nozzles." The RPV closure studs and outlet coolant nozzles bound all other RPV corrgonents and subcomponents for which fatigue was determined to be plausible, except for the i CEDM and Reactor Vessel Level Monitoring System (RVLMS) housings. (See response to Question l No. 23) The criteria for determining the bounding location (e) is the location (s) that will have the high%t cumulative usage factor (CUF) based on the cycle definitions and fatigue analyses contained in the analyses of record (AOR) for each RPV. 1 NRC Oues. tion No. 4.2.21 I Section 4.2 indicates that the Fatigue Monitoring Program (FMP) monitors and tracks low-cycle fatigue I usage for the selected components of the nuclear steam supply system and the steam generators.
Decibe the parameters that are monitored by the FMP that are applicable to the RPV. Also describe
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how the monitored parameters are compared to the fatigue analysis of record.
BGE Response Page 4.2-20 and 21 describe the FMP and the specific RPV transients that are monitored.
The critical transient for the RPV outlet nozzles is RCS cooldown from Mode 1 operation l (2 5% power). The specific parameters monitored are RCS cold leg temperatures, pressurizer !
pressure and reactor power. The number of RCS cooldowns is compared to the number of transients in the AOR every six months.
The critical transient for the RPV closure studs is RCS heatups. The specific parameters monitored are RCS cold leg temperatures, pressurizer pressure, and reactor power. The nuuber of heatups is compared to the number of transients in the AOR every six months.
For both RPV locations, all other plant transients analyzed in the AOR are assumed to have occurred and the corresponding fatigue contribution is accounted for as " initial" fatigue usage in the FMP.
The total CUF for each Iccation is computed as the sum of initial CUF and incremental CUF arising from each of the cv r.ted transient (s) for the monitored location.
NRC Ouestion No. 4.1.22 1
Section 4.2 indicates that in order to stay within the design basis, corrective action is ini:iated well in advance of the CUF approaching one or the number of cycles approaching design allowable. Describe the criteria used to determine when corrective actions will be initiated.
BGE Response Every six months the FMP reviews plant data, the resulting CUFs and cumulative critical transient usage for all components and transients monitored by the FMP. Engineering judgment is used to evaluate the rate of increase of both indicators, CUF and cumulative critical transient usage, and the approximate time until the limit (s) will be reached. An IR is initiated years before reaching the limit (s). This allows sufficient time to strategically plan the corrective actions. There is no defined
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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISM / ELECTRICAL SYSTEM limit (s) percentage that triggers an IR. The decision to write an IR is bar.d on how quickly the limit (s) is(are) being approached and will vary for each monitored component and transient.
NRC Ouestion No. 4.2.23 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 CEDM/RVLMS components. Please indicate the reason the FMP is performing an engineering evaluation on these components.
BGE. Response The CEDM/RVLMS housings were designed to the 1968 edition of the ASME Boiler and Pressure Vessel Code,Section III for Nuclear Vessels. The design of these housings included a N-415 fatigue analysis. The evaluation performed to determine the bounding locations for the FMP did not include the CEDM and RVLMS housings. The engineering evaluation will determine if these housings are bounded by the existing RPV bounding components. If they are not bounded, they will be added to the FMP.
NRC Ouestion No. 4.2.24 Section 4.2 indicates that the CUFs for the critical RPV components are well below one. Provide the usage factors projected for the critical RPV components at the end of the proposed extended period of operation including a summary discussion of how they were derived.
BGElesponse For the RPV outlet nonle:
The CUF for the RPV outlet nozzles is determined from the following equation:
CUF = {[0.2475)(b)]/500} + 0.0028 s; 0.2503 where: 0.2475 is the CUF computed in the AOR.for 500 cycles of RCS cooldown, (b) represents the cumulative sum of RCS cooldowns, initiated from Mode 1 operation (2 5%
power),
500 is the assumed number of RCS cooldowns in the AOR, and 0.0028 is the " initial" fatigue usage, the usage that is attributed to all other transients included in the design fatigue analysis.
As of June 30,1998:
Allowable RCS cooldowns Total RCS CUF (2 5% power) Cooldowns Unit 1 500 88 96 0.04636 Unit 2 500 63 70 0.03399 12
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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION;
. INTEGRATED PLANT ASSESSMENT REPORT FOR THE REACTOR PRESSURE VESSELS A ELEMENT DRIVE MECHANISM / ELECTRICAL SYSTEM l
T lt is predicted that Calvert Cliffs will not sustain the 500 AOR-allowed RCS cooldowns during 60 years of operation. The CUF for the RPV outlet nozzles at 60 years of operation will be less than 0.2503.
For the RPV closure stude i
The CUF for the RPV closure studs is determined from the following equations: !
CUF(unit :) = 0.0012632(a) + 0.0027785 + 0.199585 s 0.834 CUFmnu) = 0.0012632(a) + 0.0572397 + 0.16069 s 0.849 l l
where: 0.0012632 is the CUF computed in ;he AOR for 500 cycles of RCS heatups, 1 (a) represents the cumulative sum of RCS heatups, and 0.0027785 and 0.0572892 is " fixed" fatigue usage for each unit respectively, the usage that is attributed to all other transients included in the design fatigue analysis after installation of the permanent cavity seals, and 0.199585 and 0.16069 is " initial" fatigue usage for each unit respectively, the usage that is ;
attributed to all other transients included in the design fatigue analysis before installation of the ;
permanent cavity seals. !
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' As of June 30,1998: )
RCS Heatups Allowable CUF i Unit 1 97 500 Unit 2 0.3249 )
71 500 0.3077 1 It is predicted that Calvert Cliffs will not sustain the 500 AOR-allowed RCS heatups during 60 years of operation. The highest CUF for the RPV closure studs at 60 years of operation will be less than I 0.849.
NRC O-*6 No. 41M '
Section 4.2 of the LRA indicates that the licensee in conjunction with the Electrical Power Research i Institute has initiated an additional study to evaluate the effects of low-cycle fatigue on various fatigue !
critical plant locations. Provide a description of this study and describe its applicability to the Calvert I Cliffs RPV and CEDM/RVLMS components.
BGE Response The purpose of Electric Power Research Institute Report TR-107515, " Evaluation of Thermal Fatigue
. Effects on Systems Requiring Aging Management Review for License Renewal for the Calvert Cliffs Nuclear Power Plant," December 1997, specific to Calvert Cliffs was to provide evidence that the effects of reactor water environments for ASME Class I components are already compensated for by X the portion of the design fatigue curve margin factor of 20 that is ascribed to moderate environmental effects (approximately 4). This study was specifically discussed in the LRA and at several public meetings in early 1998.
'lhe study results do not represent Calvert Cliff's fatigue AOR nor fatigue design basis for any 1 component. The study was intended to be representative of conditions for typical older vintage 1
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Combustion Engineering PWRs, and intended to be compared to the results reported in NUREG/CR-6260, " Application of NUREG/CR 5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components," for such plants, but was not intended to be a Calvert Cliffs licensing basis calculation. There is no direct applicability of this study to the RPV or CEDM/RVLMS fatigue AOR or our FMP.
NRC Ouestion No. 4116 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 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 I 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 strucmre and component intended functions; and (f) Operating experience that provides objective evidence to demonstrate that the effects of aging will be adequately managed.
1 BGE Respony i Baltimore Gas and Electric Company can access all RPV components if required.
References
- 1. Letter from Mr. G. C. Creel (BGE) to NRC Document Control Desk, dated June 30,1992,
" Response to Generic Letter 92-01, Reactor Vessel Structural Integrity,10 CFR 50.54(f)"
- 2. Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated July 1,1998, l " Response to Request for Additional Infonnation Regarding Reactor Pressure Vessel Integrity L at Calvert Cliffs Nuclear Power Plant, Unit Nos. I and 2 (TAC Nos. MA0532 and MA0533)"
- 3. Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated August 30,1994, Generic Letter 92-01, ' Reactor Vessel Structural Integrity,' Closcout Letter / Upper-Shelf j Energy for Weld Seams 3-203-A,B,C (TAC No. M83446)"
- 4. Letter from Mr. R. E. Denton (BGE) to NRC Document Control Desk, dated July 20,1995, l Generic Letter 92-01, .' Reactor Vessel Structural Integrity,' Close-out Letter / Upper Shelf l Energy for Weld Seams 2-203 A,B,C (TAC No. M83447)"
- 5. Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (BGE), dated October 2,1997,
" Issuance of Amendments for Calvert Cliffs Nuclear Power Plant Unit No I (TAC No.
j M95181 and Unit No. 2 (TAC No. M95182)," and amended by Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (BGE), dated December 15,1997, "Calvert
- Cliffs Nuclear Power Plant, Unit Nos. I and 2 - Correction to Safety Evaluation - Amendment 4
Nos. 222 and 198" 14
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RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION; INTEGRATED PLANT ASSESSMENT REPORT FOR Tile REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECIIANISM/ ELECTRICAL SYSTEM
- 6. Letter from Mr. R. E. Denton (BGE) to NRC Document Comrol Desk, July 21,1995, " Request for Approval of Updated Values of Pressurized Thermal Shock (PTS) Reference Temperatures (RTm) Values (10 CFR 50.61)"
- 7. ABB-Combustion Engineering Topical Report No. 4.2.CE NPSD-1085, "CEOG Response to NRC Generic Letter 97-01, Degradation of CEDM Nozzle and Other Vessel Closure Head Penetrations," submitted by CEOG to NRC by a letter dated July 25,1997
- 8. Letter from Mr. A. W. Dromerick (NRC) to Mr. C. H. Cruse (BGE), dated August 24,1998,
" Generic Letter (GL) 97-01, ' Degradation of CRDM/CEDM Nozzle and Other Vessel Closure Head Penetrations' Responses for Calvert Cliffs Nucler.r Power Plant, Unit Nos. I and 2 and the Relationship of the Responses to Topical Report No. CE NPSD-1085"
- 9. NRC " Standard Review Plan for the Review of License Renewal Applications for Nuclear Power Plant, Working Draft," September 1997 i
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