ML20151Y138
| ML20151Y138 | |
| 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-M99210, TAC-M99591, TAC-M99592, NUDOCS 9809180120 | |
| Download: ML20151Y138 (5) | |
Text
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s September 3, 1998 Mr. Charl::s H. Crusa, Vice Pr:sident Nuclear Energy Division 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 1 & 2, INTEGRATED PLANT ASSESSMENT REPORT FOR THE SERVICE WATER SYSTEM (TAC NOS. M99591, M9V92, AND M99210)
Dear Mr. Cruse:
By letter dated August 21,1997, Baltimore Gas and Electric Company (BGE) submitted for review the Service Water System (5.17) integrsted 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) st.1ff review the Service Water System (5.17) integrated plant assessment technical report to determine if the report meets 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 renewalif BGE applied in the future.
J By letter dated April 8,1998, BGE formally submitted its license renewal application.
1 i
The NRC staff has reviewed the Service Water System (5.17) integrated plant assessment technical report 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 i
10 CFR 54.21(a)(2). Based on a review of the information submitted, the staff has identified in the enclosure, areas whero additional information is needed to complete its review.
Please provide a schedule by letter or telephonical!y for the submittal of your responses within 30 days of receipt of this letter. Additionally, the staff would be willing to meet with BGE prior to
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the submittal of the responses to provide clarifications of the staff's requests for additional information.
I Sincerely,
}g) g David L. Solorio, Project Manager bIcense Renewal Project Directorate 9809180120 980903 PDR ADOCK 05000317 Division of Reactor Program Management P
PDR Office of Nuclear Reactor Regulation Docket Nos. 50-317 and 50-318
Enclosure:
Request for Additional Information y-Q" Q h k cc w/ encl: See next page DISTRIBUTION.
See next page DOCUMENT NAME G:\\ WORKING \\SOLORIO\\SRW RAI LTR
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FFICIAL RECORD COPY 40
9 Mr. Charles H. Cruse Calvert Cliffs Nuclear Power Plant Baltimore Gas & Electric Company Unit Nos.1 and 2 cc:
President Mr. Joseph H. Walter, Chief Engineer Calvert County Board of Public Service Commission of 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 i
Baltimore Gas and Electric Company Maryland People's Counsel i
EO. Box 1475 6 St. Paul Centre Baltimore, MD 21203 Suite 2102 Jay E. Silberg, Esquire Shaw, Pittman, Potts, and Trowbridge Patricia T. Birnie, Esquire 2300 N Street, NW Co-Director Washington, DC 20037 Maryland Safe Energy Coalition P.O. Box 33111 Mr. Thomas N. Prichett, Director Baltimore, MD 21218 NRM i
Calvert Cliffs Nuclear Power Plant Mr. Loren F. Donatell 1650 Calvert Cliffs Parkway NRC Technical Training Center Lusby, MD 20657-4702 5700 Brainerd Road Chattanooga, TN 37411-4017 Resident inspector i
U.S. Nuclear Regulatory Commission David Lewis J
P.O. Box 287 Shaw, Pittman, Potts, and Trowbridge St. Leonard, MD 20685 2300 N Street, NW Washington, DC 20037 Mr. Richstd 1. McLean Nuclear Programs Douglas J. Walters Power Plant Research Program Nuclear Energy Institute Maryland Dept. of Natural Resources 17761 Street, N.W.
Tawes State Office Building, B3 Suite 400 Acnapolis, MD 21401 Washington, DC 20006-3708 Regional Administrator, Region I Barth W, Doroshuk U.S. Nuclear Regulatory Commission Baltimore Gas and Electric Company 475 Allendale Road Calvert Cliffs Nuclear Power Plant King of Prussia, PA 19406 1650 Calvert Cliffs Parkway NEF ist Floor Lusby, Maryland 20657
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Distribution:
HARD COPY
, Docket Filed PUBLIC PDLR R/F MEl-Zeftawy DISTRIBUTION: E-MAIL:
FMiraglia (FJM)
JRoe (JWR)
DMatthews (DBM)
CGrimes (CIG)
TEssig (THE)
Glainas (GCL)
JStrosnider (JRS2)
GHolahan (GMH)
SNewberry (SFN)
GBagchi (GXB1)
RRothman (RLR)
JBrammer (HLB)
CGratton (CXG1)
JMoore (JEM)
MZobler/RWeisman (MLZ/RMW)
SBajwa/ADromerick (SSB1/AXD)
LDoerflein (LTD)
BBores (RJB)
SDroggitis (SCD)
RArchitzel(REA)
. CCraig (CMC 1)
LSpessard (RLS)
RCorreia (RPC)
RLatta (RML1)
EHackett (EMH1) i AMurphy (AJM1)
TMartin (TOM 2)
DMartin (DAM 3)
GMeyer (GWM)
WMcDowell(WDM)
SStewart (JSS1)
THiltz (TGH)
SDroggitis (SCD)
DSolorio (DLS2)
PDLR Staff TMarsh (LBM)
GHubbard (GTH)
WLefave (WTL1)
Yeuh-Li Li (YCL)
KF-arczewski(KIP) 1
REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCI FAR POWER PLANT UNITS NOS.1 & 2 SERVICE WATER SYSTEM lNTEGRATED PLANT ASSESSMENT. SECTION 5.17 DOCKET NOS. 50-317 AND 50-318 l
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i 1.
Section 5.17, indicates that a previously performed evaluation concluded that the non-
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safety-related portions of the service water system (SRW) are adequately rugged to j
withstand a design basis earthquake, which is credited in the design basis for preserving system inventory. The same section also indicates that all safety-related portions of the SRW are within the scope for license renewal. Since the non-safety-related portions of the SRW piping are credited in preserving system inventory during a design basis earthquake, it is not clear why this portion of piping is not within the scope for license renewal. Provide the basis for excluding this portion of SRW from the scooe of license renewal or a cross reference to where it is addressed in the license renewal application (LRA).
2.
Section 5.17.2, indicates that the safety-related SRW system piping will be included in an l
Aging-Related Degradation inspection (ARDI) program to verify that degradation of the l
piping is not occurring, and ine results of that inspection will be evaluated for applicability to the non-safety-related SRW piping. In addition, you state that the non-safety-related portions of SRW piping and the safety-related piping were both originally designed to l
USAS B31.1 and both are subject to the same environmental service conditions and chemistry controls. The applicability evaluation will also caneider, at a minimum, flow rate and configuration differences between safety-reisted and non-safety-related SRW piping. Please clarify how the flow rate and configuration differences between safety-j related and non-safety-related SRW piping will be considered in the applicability l
evaluation, and clarify the basis upon which you concluded that the results of the l
inspection of the safety-related piping are adequately representative of the aging degradation of the non-safety piping.
3.
According to Subsection 5.17.1.1, the SRW piping to the instrument and plant air compressors and aftercoolers is within the scope of license renewal for fire protection.
I However, a failure anywhere in the SRW supply or retum piping to these components (or any connected systems or components) can affect not only the fire protection safe shutdown, but also all other safe shutdown events requiring the operation of the SRW j
system. Clarify the basis for determining why the SRW system piping to the compressors and aftercoolers is within the scope of license renewal for fire protection, but not within the scope for the SRW.
t 4.
In Section 5.17.1.3, you have identified that the only passive function associated with the SRW system not otherwise dispositioned is "to maintain the pressure boundary of the system liquid." In light of your response to Component Cooling Water System RAI No. 2 i
(letter dated August 1,1998), and the air-operated components in the SRW system, 1
Enclosure
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5 i identify how the aging management review has been conducted for the air-operated components in the SRW system.
5.
Are there any parts of the SRW systems, structures or components 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 inaccessible areas based on conditions found in surrounding accessible areas, please prcvide 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 i
management techniques are needed for the inaccessible areas, please provide a summary to address the following elements for the inaccessible areas: (1) preventive actions that will mitigate or p'revent aging degradation; (2) parameters monitored or inspected relative to degradation of specific structure and component intended functions; (3) detection of aging effects before loss of structure and component intended functions; (4) monitoring, trending, inspection, testing frequency, and sample size to ensure timely detection of aging effects and corrective actions; (5) acceptance criteria to ensure l
structure and component intended functions; and (6) operating experience that provides j
objective evidence to demonstrate that the effects of aging will be adequately managed.
l 6.
Section 5.17, indicates that the SRW system was designed to USAS B31.1 Code l
requirements. While B31.1 does not require an explicit fatigue analysis, it does specify allowable stress levels based on the number of anticipated thermal cycles. Table 5.17-3 l
l indicates that fatigue is not a plausible age-related degradation mechanism (/sRDM) for l
the SRW system. Because fatigue is normally treated as a Time-Limited Aging Analyses in accordance with the requirements of 10 CFR 54.21(C), please provide the basis for concluding fatigue is not a plausible ARDM for SRW components.
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7.
The rate of corrosion of the components in the SRW system can be mitigated by proper l
control of the water chemistry. Provide the specifications for the water chemistry in the SRW system. Include the target values for the individual parameters and their monitoring frequency.
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