ML20197H961
| ML20197H961 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/10/1998 |
| From: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9812140219 | |
| Download: ML20197H961 (5) | |
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CHA~t,ES II. CEUIE Baltimore Gas and Electric Company Vice President Calven Cliffs Nuclear Power Plant Nuclear Energy 1650 Calven Cliffs Parkway Lusby, Maryland 20657 410 495-4455 December 10,1998 U. S. Nuclear Regulatory Commission Washington, DC 20555
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ATTENTION:
Document Contal Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Supplemental Response to NRC Question No. 4.1.26; Integrated Plant Assessment Renort: License Renewal Aeolication
REFERENCES:
(a)
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated April 8,1998," Application for License Renewal" (b)
Letter from Mr. D. L. Solorio (NRC) to Mr. C. H. Cruse (BGE), dated September 3,1998," Request for Additional Information for the Review of the Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Integrated Plant Assessment, Sections 4.1,4.2,5.2,5.7,4.1, and 5.16" (c)
Letter from Mr. C. H. Cruse (BGE) to NRC Document Control Desk, dated November 19,1998, " Response to Request for Additional Information for the Review of the Calvert Cliffs Nuclear Power Plant, Units 1 & 2, Integrated Plant Assessment Reports for the Reactor Coolant System and for the Reactor Pressure Vessels and Control Element Drive Mechanisms / Electrical System" Reference (a) forwarded the Baltimore Gas and Elect.ric Company (BGE) license renewal application (LRA). Reference (b) forwarded questions from NRC staff on certain sections of the BGE LRA.
Reference (c) provided BGE's response to Refercuce (b), indicating for Question No. 4.1.26 that BGE and NRC were working toward clarification.
A site visit and public meeting occurred on
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November 23,1998 for NRC Staff to review material applicable to four questions, including Question No. 4.1.26. At the conclusion of that visit, NRC Staff requested a supplementary response, which would provide additional discussions of operating experience with the CCNPP Alloy 600 Program relative to events that have occurred since the LRA was submitted. The NRC explained that this information would f
be helpful in the staff's review. Attachment (1) forwards BGE's supplemental response.
I 9812140219 991210 i
hDR ADOCK 050003 7 k NRC Distribution Code A036D
m Document Control Desk I
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ShouG you h' ave further questions regarding this matter, we will be pleased to discuss them with you.
I Very truly yours, I
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. STATE OF MARYLAND
- TO WIT-
- i COUNTY OF CALVERT
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I, Charles H. Cruse, being duly swom, 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 l
L response on behalf of BGE. To the best of my knowledge and belief, the statements contained in this j
document are true and correct. To the extent that these statements are not based on my personal 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 to be reliable.
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Su scribed and sworn before me a Notary ublic in and for the State of Maryland and County of uAL
, this /db ay of 1/n1M/,1998.
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d chMf WITNESS my Hand and Notarial Seal:
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l-1 My Commission Expires:
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Attachment:
(1) Supplemental Response to NRC Question No. 4.1.26; Integrated Plant Assessment Report; License Renewal Application l
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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. McLean, DNR j
H. J. Miller, NRC J. H. Walter, PSC l
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1 ATTACHMENT m 4
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SUPPLEMENTAL RESPONSE TO NRC QUESTION NO. 4.1.26; INTEGRATED PLANT ASSESSMENT REPORT; LICENSE RENEWAL APPLICATION 1
Baltimore Gas and Electric Company Calvert Cliffs Nuclear Power Plant December 10,1998
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I A'ITACHMENT (1)
SUPPLEMENTAL RESPONSE TO NRC QUESTION NO. 4.1.26; INTEGRATED PLANT ASSESSMENT REPORT; LICENSE RENEWAL APPLICATION NRC.Ouesfion No. 4.1.26 Provide the results of Baltimore Gas and Electric Company's (BGE's) most recent internal audit of the Alloy 600 Program; including areas of strengths and weaknesses, safety implication of findings, and corrective action plans and schedule for implementation.
BGE Supplemental Response A site visit occurred on November 23,1998 to review, panly, information related to this question. At the conclusion of that visit, NRC Staff requested a BGE License Renewal Application (LRA) supplementary response, which would discuss operating experience with the CCNPP Alloy 600 Program and the demonstration that the program is effective for managing aging. That supplementary response is provided below.
In BGE LRA Sections 4.1.2 and 4.2.2, the Alloy 600 Program Plan and the Boric Acid Corrosion Inspection (BACl) Program are described as the programs used to manage aging of Reactor Coo! ant System components and control element drive mechanism penetrations manufactured of Alloy 600.
The primary aging management effect being managed by the Alloy 600 Program is primary water stress corrosion cracking (PWSCC).
In short, the Alloy 600 Program Plan currently identifies all pressure boundary applications of Alloy 600 exposed to reactor coolant (except for steam generator tubes and plugs, which are evaluated under the steam generator programs). Each component is evaluated with respect to product form, fabrication history, installation technique, and operating temperature and history. Each component is evaluated using algorithms to predict time to cracking and time to through-wall cracking. The algorithms used in the Alloy 600 Program Plan were derived by nuclear industry
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vendors who have developed them based on laboratory analysis and years of operating experience.
I Mitigation, preventive repair, or proactive replacement are scheduled to be performed prior to predicted times for cracking.
Augmented inspection may be scheduled for particular components that may have earlier times to j
crack, or for which industry ar d plant experience indicates augmented inspection is appropriate.
The BACI Program provides an independent verification of the Alloy 600 Program Plan by providing inspections of all Alloy 600 pressure boundary components for leakage once each refueling cycle.
All anticipated or observed types of PWSCC degradation have been evaluated and determined to not represent a near-term challenge to pressure boundary structural integrity. To ensure the Alloy 600 Program Plan continues to refine itself based on actual plant data, one of the provisions of the Program includes monitoring and incorporating industry (and Calvert Cliffs) operating experience.
Any future discoveries bearing on the nature and progression of PWSCC (resulting mainly from operating experience, but also from laboratory investigadon) will be considered; and inspection, replacement, repair and mitigation activities will be rescheduled, as appropriate.
1 To illustrate the nature of this experience-driven program, operating experience is currently bemg factored into the Alloy 600 Program Plan in our response to the recent 1998 leak in the upper level tap on the Unit 2 pressurizer (Reference 1). Prior to this event, PWSCC of weld metal had been identified in the laboratory, but was expected to lag the cracking of the wrought material connected to I
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ATI ACHMENT (1)
I SUPPLEMENTAL RESPONSE TO NRC QUESTION NO. 4.1.26; INTEGRATED PLANT ASSESSMENT REPORT; LICENSE RENEWAL APPLICATION l
it.. Accordingly, the Alloy 600 Program Plan concluded that weld metal was of low susceptibility, and all wrought Alloy 600 would lead all weld metal in time to cracking. As part of an Alloy 600 nozzle repair in the early 1990s, an Alloy 690 nozzle was welded into the Reactor Coolant System using Alloy 600-type weld metal. In that particular application, it was not possible for the wrought material to crack before the weld, because the wrought material was not susceptible to PWSCC. A BACI Program inspection verified a leak, an issue report was generated, and an investigation was performed, which eventually concluded that the Alloy 600 weld material had cracked. Subsequently, l
it was determined that the Alloy 600 Program Plan needed to be revised to re-evaluate the susceptibility of weld metal, including scheduling for augmented inspections, proactive replacement, or preventive repair, as appropriate. The required changes were identified and are scheduled to be incorporated into the Alloy 600 Program Plan.
The Alloy 600 Program Plan effectively manages PWSCC of Alloy 600 pressure boundary components. In the future, the Alloy 600 Program Plan will be revised to also include evaluation of non-pressure boundary applications of Alloy 600. The plan uses algorithms to predict time to cracking for all susceptible components, and provides for actions to ensure discovery and remediation of PWSCC before leakage occurs. In addition, the Alloy 600 Program Plan ensures that leakage, ifit j
should occur, would not represent a challenge to the structural integrity of the pressure boundary.
The plan documents that leakage does not constitute a challenge to the structural integrity ofine pressure boundary. The B.ACI Program inspections independently verify that the Alloy 600 Program Plan predictions are valid. Where industry experience or BACI Program results indicate that the Alloy 600 Program Plan predictions may be in error, the Alloy 600 Program Plan is revised to reflect this new knowledge. As previously stated in Section 4.1.2 of the BGE LRA, some RCS components have already been replaced and some have been nickel plated as a result of the Alloy 600 Program Plan.
We believe, therefore, that the Alloy 600 Program is adequate for managing the aging mechanisms l
for which it is credited.
Reference 1.
Letter from Mr. P. E. Katz (BGE) to NRC Document Control Desk, dated August 24,1998, Licensee Event Report 318-98-005, " Plant Cooldown Due to Reactor Coolant System Pressure Boundary Leakage" l
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