ML20195H795
| ML20195H795 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/18/1998 |
| From: | Dave Solorio NRC (Affiliation Not Assigned) |
| To: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| References | |
| TAC-M99223, TAC-MA1016, TAC-MA1017, NUDOCS 9811240107 | |
| Download: ML20195H795 (23) | |
Text
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.. _o-l November 18, 1998 i
L Mr. Charles H. Cruse, Vice President ;
Nuclear Energy Division Baltimore Gas and Electric Company 1850 Calvert Cliffs Parkway Lusby, MD 20657-4702
SUBJECT:
CLARIFICATION OF NRC REQUEST FOR ADDITIONAL INFORMATION ON CALVERT CLIFFS NUCLEAR POWER PLANT LICENSE RENEWAL APPLICATION SUBMITTED BY THE BALTIMORE GAS AND ELECTRIC '
COMPANY (TAC NOS, MA1016, MA1017, AND M99223)
Dear Mr. Cruse:
On September 22,1998, the Nuclear Regulatory Commission (NRC) staff held s public meeting with representatives of Baltimore Gas and Electric Company (BGE) at Rockville, Maryland, to discuss the progress of the NRC staff's review of BGE's License Renewal Application for its Calvert Cliffs Nuclear Power Plants, Units Nos.1 and 2. During the meeting, BGE requested l clarification on approximately 25 requests for additional information (RAI) that were subsequently expanded to 26. BGE requested clarification in order to meet the November 21, 1998, milestone for submitting its responses to all the staff's RAl.
The purpose of this letter is to clarify RAI 4.1.17 for BGE. The RAI was revised based on the result of BGE's comments, as provided to the NRC in Enclosure 3 of the September 28,1998,.
NRC and BGE meeting summary (which is provided as Enclosure 1 to this letter), and on subsequent discussions between BGE and NRC staff which were held to gain additional detail regarding BGE's needs for clarification. The revised RAI is provided in Enclosure 2.
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Sincerely, m mm PIO
![{h y.,7g Jdual w O*'d David L. Solono, r et Manager License Renewal Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket Nos. 50-317 and 50-318
Enclosures:
- 1. Revised Copy of USNRC Requests for Additional Information Provided to I NRC by BGE
- 2. Revised Request for Additional information cc w/encis: See next page r DISTRIBUTION See next page ,
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I DATE '11/lkg8 11/g/98 11/ } /98 11/ $/98 OFFICIAL RECORD COPY 9811240107 981118 (, * ' . ,
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-} .5 Baltimore Gas & Electric Company Calvert Cliffs Nuclear Power Plant cc: Unit Nos.1 and 2 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 Baltimore Gas and Electric Company Maryland People's Counsel P.O. Box 1475 6 St. Paul Centre Baltimore, MD 21203 Suite 2102 Baltimore, MD 21202-1631 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. Bruce S. Montgomery, Director Baltimore, MD 21218 NRM 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 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 Mr. Richard 1. McLean Nuclear Programs Douglas J. Walters Power Plant Research Program Nuclear Energy Institute Maryland Dept. of Natural Resources 1776 l Street, N.W.
Tawes State Office Building,83 Suite 400 Annapolis, MD 21401 Washington, DC 20006-3708 DJW@NEl.ORG Regional Administrator, Region i U.S. Nuclear Regulatory Commission Barth W. Doroshuk 475 Allendale Road Baltimore Gas and Electric Company King of Prussia, PA 19406 Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway l
Mr. Charles H. Cruse, Vice President NEF 1st Floor Nuclear Energy Division Lusby, Maryland 20657 Baltimore Gas and Electric Company 1650 Calvert Cliffs Parkway National Whistleblower Center Lusby, MD 20657-47027 3233 P Street, N.W.
Washington, DC 20007 i
-l .'s Distnbution:
HARD COPY (w/ Enclosures 1& 2)
Docket Files '- !
PUBLIC l PDLR R/F 4
.OGC' MEl-Zeftawy DISTRIBUTION: E-MAIL (w/ Enclosure 2) l FMiraglia (FJM) . 1 JRoe (JWR) j
- DMatthews (DBM) l CGrimes (CIG). l TEssig (THE) !
GLainas (GCL)
JStrosnider (JRS2) i GHolahan (GMH)
SNewberry (SFN)
GBagchi(GXB1)
RRothman (RLR)
JBrammer (HLB)
CGratton (CXG1)
JMoore (JEM)
MZobler/RWeisman (MLZ/RMW)
SBajwa/ADromerick (SSB1/AXD)
LDoerflein (LTD) 6 Bores (RJB)
. SDroggitis (SCD)
, RArchitzel(REA)
CCraig (CMC 1) -
LSpessard (RLS)
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
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BGE REQUESTS FOR CLA.RIFICATION ON USNRC REQUESTS FOR ADDITIONAL INFORMATION Prepared by BGE Calvert Cliffs License Renewal Project October 1,1998 l
Enclosure 1
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CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2
- FEEDWATER SYSTEM INTEGRATED PLANT ASSESSMENT,SECTION 5.9 DOCKET NOS. 50-317 AND 50-318 Aging Management 5.9.47 One of the most effective ways of minimizing erosion / corrosion is to control secondary water
! chemistry, that is, pil and oxygen concentration. Describe whether pit and oxygen concentration are controlled in the feedwater system and if so, specify the parameter ranges.
BGE believes that this question has been answered in the LRA as well as has been discussed at length in meetings. BGE has provided the NRC with copies of the procedures. In
, addition, there is an effort between NEI(NEl-9706) and NRC on chemistry controls which is I
ongoing that provides a significant amount ofinformation in this area. RGE requests NRC evaluate the need for requesting this additionalinformation,given the above described exchanges already underway.
5.9.54 Page 5.9-20 of the application indicates that the Institute of Nuclear Power Operations (INPO) has ;
performed assessment of the BGE erosion / corrosion program and provided recommendation for enhancements. Please briefly summarize the results of the INPO assessment and outline the INPO j recommendations for improvements at the Calvert Cliffs plants. l l
HGE requests that NRC review the need for INPO reports since they are proprietary and the NRC has access to them already.
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l REQUEST FOR ADDITIONAL INFORMATION )
CALVERT CLIFFS NUCLEAR POWER PLANT '
UNIT NOS.1 & 2 REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE ]
i MECil ANISMS/ ELECTRICAL INTEGRATED PLANT ASSESSMENT, SECTION 4.2
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DOCKET NOS. 50-317 AND 50-318 i Section 4.2.2 - Aging Management l
4.2.8 Provide pressure-temperature (P-T) limits for the extended operating term and identify the i operating window relative to pump operation for the shutdown cooling system. During the extended licensed term, will there be any limitations in operation of the shutdown cooling system due to American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Appendix G, P T operating limits and the minimum pennissible temperature of the reactor vessel?
BGE is requesting NRC clarify #8 above. Current plant practice is to maihtain these curves as required, not necessarily at license termination points. No 10CFR50 requirsment exists for such submittals other than to maintain them current and not violate them. The 2nd part ~
of this question is hypothetical and based upon 60 year curves.
HGE has had a TELCON with NRC on this question and is providing response according to NRC clarification.
4.2.17 Section 4.2.2 of the LRA states "The threshold for onset of neutron effects for RPV materials is l l
conservatively defined to be a fast neutron fluence that exceeds lE17n/cm2," citing Appendix 11 of 10 CFR Part 50. The staff believes that Appendix H cites the indicated neutron fluence as a threshold below w hich a reactor vessel material surveillance program is not required for the I
vessel. Appendix H thereby creates in efTect a " regulatory threshold" for neutron fluence, but l
clearly not a mechanistic threshold below which neutron effects do not occur. Please provide l your basis for concluding that there are negligible effects from neutron fluence below IE17n/cm2. l
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BGE is requesting clarification of #17 above. BGE has had a TELCON with NRC on this question and is answering this question uccording to NRC clarification. i l
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l REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 SPENT FUEL POOL COOLING SYSTEM INTEGRATED PLANT ASSESSMENT,SECTION 5.18 DOCKET NOS. 50-317 AND 50-318 Section 5.18.2 - Aging Management !
5.18.10 Provide a summary description of Calvert Cliffs operating and maintenance experience related to !
boric acid corrosion of carbon steel components. In particular, characterize the extent to which I boric acid corrosion of carbon steel components has changed since the initial implementation of I
,, the boric acid corrosion inspection (BACl) program. Alsc, describe the extent to which carbon steel components in the spent fuel pool cooling system have had to be repaired or replaced because of boric acid corrosion, since the i.i'plementation of the BACI program.
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This question is too broad. Similar questions were withdrawn by NRC and refocused. BGE requests the NRC either conduct a site visit and discuss OE with plant personnel or re-phrase question such that it is more focused.
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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 REACTOR COOLANT SYSTEM INTEGRATED PLANT ASSESSMENT, SECTION 4.1 DOCKET NOS. 50-317 AND 50-318 4.1.9 For the following aging effects and components, summarize the extent to which BGE relies upon the associated programs for aging management, and provide examples of any operating experience that demonstrates the effectiveness of the programs that are relied upon to manage these aging effects:
- a. boric acid corrosion -- Technical Specifications (TS) leakage limits, and ASME Section XI, Subsection IWB, examination categories B-P;
- b. cracking oflarge bore piping -- ASME Section >'!, Subsection IWB, examination categories B-J and B F, and flaw evaluation criteria IWB-3000;
- c. cracking of small bore piping (less than 4 in but greater than 1 in diameter)-- augmented volumetric inservice inspection; and, because some safe ends and welds on small bore piping are ofInconel, information resulting from the assessment of NRC Information Notice (IN) 90-10;
- d. cracking of bolting -- programs consistent with ASME Section XI, Subsection IWB, examination categories B-G-1 and B-G-2, and NRC Bulletin 82-02; e, pressurizer shell, heads, heater belt forgings -- ASME Section XI, Subsection IWB, examination categories B-B and B-P, and primary water chemistry; i
l f. pressurizer nozzles -- ASME Section XI, Subsection IWB, examination categories B-D, B E, B-F, and B-P, TS leakage limits, primary water chemistry, augmented inspection of small bore piping; and if Inconel is used, information resulting from IN 90-10;
- g. integral attachments -- ASME Section XI, Subsection IWB, examination category B-li, and primary water chemistry;
- h. heater sheaths and end caps -- ASME Section XI, Subsection IWB, examination category B-P, and TS leakage limits;
- i. loss of preload in bolting -- ASME Section XI, Subsection IWB, examination categories B-G-1, B-G-2, and B-P, response to NRC Bulletin 82-02 and Generic Letter 88-03, and TS leakage limits, i
BGE believes this question is too broad and requests NRC clarify its intent. An option for
- disposition is meetings either at Calvert Cliffs or NRC Offices to review the site l documentation that may have addressed these issues.
, 4.1.12 It appears that BGE used ferrite criteria to screen components subject to thermal embrittlement.
llowever, the NRC regards ferrite content as inadequate criterion for screening as stated in NUREG-1557. Therefore, justify using ferrite content as screening criteria.
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The use of ferrite criteria to screen components has been a part of an Industry Position since ;
1994. It has also been submitted as part of NEI/EPRI efforts to resolve generic aging issues. I BGE requests NRC explain the " inadequateness" of a position.
4.I.17 Please provide a summary description for the following procedures regarding how their l implementation will address the following elements for their related aging management l program (s): (a) The scope of structures and components managed by the program; (b) Preventive j actions designed to mitigate or prevent aging degradation;(c) Parameters monitored or inspected i relative to degradation of spe-ific structure and component intended functions;(d) Detection of l
aging effects before loss of structure and component intended functions; (e) Monitoring, trending, inspection, testing frequency, and sample size to ensure timely detection of aging effects and corrective actions;(f) Acceptance criteria to ensure structure and component intenden functions; and (g) Operating experience that provides objective evidence to demonstrate that the effects of aging will be adequately managed.
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- a. Procedure SG 20," Primary manway cover removal and installation"
- b. Administrative Procedure MN-3-il0," Inservice inspection of ASME XI Components"
- c. Technical Procedure FASTENER-Oi," Torquing and Fastener Applications"
- d. Procedure STP-M 574-1/2,"EC Examination of CCNPP % Steam Generators"
- e. CASS Evaluation program i
- f. Alloy 600 program
- g. STP-0-27-l/2,"RCS Leakage Evaluation"
- h. MN-3-301, "BACI Program" l i. EN-1-300," Implementation of Fatigue Monitoring" This question is too broad. Similar questions were withdrawn by NRC and refocused. BGE requests the NRC either conduct a site visit and discuss details of plant programs with plant personnel or re-phrase question such that it is more focused.
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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 FIRE PROTECTION SYSTEM INTEGRA'I ED PLANT ASSESSMENT,SECTION 5.10 DOCKET NOS. 50-317 AND 50-318 Section 5.10.1 - Scoping 5,10.6 Summarize the changes to the post-fire safe shutdown analysis and the fire hazards analysis that have been implemented since plant licensing and briefly discuss how the analyses, including changes, were addressed in the system level scoping process.
BGE believes this question is too broad and requests NRC elarify its intent. BGE interprets this question as a compilation of CLB and BGE is not clear on that is what NRC intends nor to its contribution to scoping results.
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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNITS NOS.1 & 2 SERVICE WATER SYSTEM INTEGRATED PLANT ASSESSMENT,SECTION 5.17 DOCKET NOS. 50-317 AND 50-318 5.17.7 The rate of corrosion of the components in the SRW system can be mitigated by proper 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. j BGE believes this question is too broad and requests NRC clarify its intent. An option for disposition is meetings either at Calvert Cliffs or NRC Offices to review the site documentation that may have addressed the attributes or details of the program but not discussed them in the LRA since they were referenceable.
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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 TIME-LIMITED AGING ANALYSES,SECTION 2.1 DOCKET NOS. 50-317 AND 50-318 l 2.1.3 Page 2.1-4 of the license renewal application (LRA) indicates that the pressure-temperature (P-T) limits in the Calvert ClitTs Technical Specifications are valid for Units I and 2 for 48 and 30 effective full power years, respectively. Section 4.2 of Appendix A to the BGE application indicates that the Unit 2 reactor vessel is less susceptible to neutron embrittlement. Discuss why the P-T limits for Unit 2 are valid for a shorter time period than for Unit 1. Also, discuss whether the existing P T limits "[i]nvolve time-limited assumptions dermed by the current operating term, for example,40 years." (Criterion 3 of the definition of TLAA in 10 CFR 54.3(a))
See #8 in Reactor Vessel /CEDM Section. This is a duplicate.
2.1.4 10 CFR 54.21(c) requires an evaluation of TLAAs as part of the contents of an LRA. Ilowever, Section 2.1 of Appendix A to the BGE application contains future commitments to perform the TLAA evaluations. The following are examples:
Subsection lleading Statement 2.1.3.2 Irradiation Embrittlement " .. will continue to be updated.. "
2.1.3.5 Containment Liner Plate "This review . . will be projected Fatigue Analysis . by the year 2012."
2.1.3.6 Containment Tendons . recalculated by the year 2012.. "
Prestress Loss 2.1.3.7 Poison Sheets in Spent "This analysis is currently being Fuel Pool updated.. "
In accordance with 10 CFR 54.21(c)(1)(iii), describe how BGE will ensure that the effects of aging on the intended functioa(s) will be adequately managed for the period of extended operation.
BGE is responding to this question by referring to its commitment management procedure.
This has been discussed with NRC Branch Chief.
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- l REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS UNITS I AND 2 INTEGRATED PLANT ASSESSMENT ON METAL FATIGUE DOCKET NOS. 50-317/50-318 Section 5 2," Chemical and Volume Control System" l l
7.6 Section 3.2.3 of EPRI Report TR-107515 contains an evaluation of environmental effects on the CVCS Charging Inlet Nozzle using methodology developed in EPRI Report TR-105759,"An Environmental Factor Approach to Account for Reactor Water Effects in Light Water Reactor Pressure Vessel and Piping Fatigue Evaluations," dated December 1995. The attached evaluation summarizes the staffs technical concerns regarding the methodology in EPRI Report TR 105759.
l Attached are comments on the application of the EPRI methodology for environmental fatigue i factors to the Calvert Cliffs plant.13ased on these comments, provide the following:
(a) Discuss the impact of the current Argonne National Laboratory (ANL) statistical correlations of environmental test data on the Calvert Cliffs fatigue evaluation.
IIGE does not beliese the subject research project can be commented on in a timely nor reasonable fashion. BGE requests NRC withdraw this question since commenting on execrpts from research activities can produce out of context ,
conclusions. l (b) The technical basis for the assertion that the American Society of Mechanical Enginects (ASME) Code stainless steel fatigue design curve contains sufficient margin to accommodate moderate environmental effects. Include a discussion of the factor required to adjust the laboratory test data for size and surface finish effects and the margin necessary to account for scatter of the test data.
BGE requests the NRC withdraw this question. BGE accepts the ASME code as endorsed by 10CFR50.55a as part of our CLB.
(c) The technicaljustification for the strain threshold values.
BCE will provide this answer.
Section 4.1," Reactor Coolant System" 7.15 Section 4.1 of the application Indicates that environmental effects do not apply to the RCS components because of the low oxygen concentrations and because the RCS carbon steel interior surfaces are clad with stainless steel. Discuss the applicability and impact of the latest stainless steel fatigue correlation from ANL on this conclusion (see attachment).
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BGE does not believe the subject research project can be commented on in a timely nor reasonable fashion. BCE requests NRC withdraw this question since commenting on excerpts from research activities can produce out of context conclusions.
7.16 Section 3.3.3 of EPRI Report TR-107515 contains an evaluation of the Surge Line using i methodology developed in EPRI Report TR 105759. Discuss the applicability and impact of the latest stainless steel fatigue correlation from ANL on this evaluation (see attachment).
BCE does not believe the subject research project can be commented on in a timely nor reasonable fashion. BGE requests NRC withdraw this question since commenting on '
excerpts from research activities can produce out of context conclusions.
7.17 Section 3.3.3.2 of EPRI Report TR-107515 indicates that the procedure in Section 3.1.3.2 of the EPRI report was used to develop the environmental factor used in the evaluation. Indicate whether the factor was calculated based on a " standard" treatment or " weighted average" approach as discussed in a June 1,1998, letter from the Nuclear Energy Institute to the NRC regarding EPRI Report TR 105759. If the " weighted average" approach was used, provide the test data used to develop the approach. Include a statistical assessment of the test data scatter. ,
Compare the results of the statistical assessment with the ANL assessment contained in '
NUREG/CR-6335," Fatigue Strain-Life Behavior of Carbon and Low-Alloy Ferritic Steels, Austenitic Stainless Steels, and Alloy 600 in LWR Environments." On the basis of this comparison, indicate whether the use of the " weighted average" approach will produce an adequ'a te margin to account for test data scatter.
BGE does not believe the subject research project can he commented on in a timely nor reasonable fashion. BGE requests NRC withdraw this question since commenting on excerpts from research activities can produce out of context conclusions.
i Section 5.15. " Safety injection System" 7.22 Section 5.15 of the application indicates that environmental effects do not apply to the Si components because of the low oxygen concentrations and the stainless steel components materials used in fabrication of the affected piping and valve subcomponents. Discuss the applicability and impact of the latest stainless steel fatigue correlation from ANL on this l conclusion (see attachment).
l l BGE does not believe the subject research project can be commented on in a timely nor reasonable fashion. BGE requests NRC withdraw this question since commenting on excerpts from research activities can produce out of context conclusions.
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l REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 INTEGRATED PLANT ASSESSMENT ON GENERIC SAFETY ISSUES i DOCKET NOS. 50-317 AND 50-318 8.3 In a letter dated June 2,1998, the staff concluded that license renewal applicants can address GSI-168. " Environmental Qualification of Electrical Equipment," by providing a technical rationale demonstrating that the current licensing basis for EQ pursuant to 10 CFR 50.49 will be maintained in the period of extended operation. The NRC staff has not completed guidance on the information necessary to demonstrate adequate aging management for the EQ time limited aging l analyses (TLAAs). Until that matter is resolved, please provide the EQ Master List of electrical equipment and indicate which of the TLAA categories in 10 CFR 54.21(c)(1) apply to each of the electrical equipment groups. In addition. summarize the procedures that are used to maintain compliance with the requirements of 10 CFR 50.49, and justify that those procedures will adequately manage the EQ analyses for the period of extended operation.
HGE has provided all the requested information in the LRA section on EQ except for the EQ Master List. BGE sequests to discuss this with NRC since compliance with 10CFR50.49 is requirement that will carry forward as part of the CLB in accordance with the rule. The EQ Master List is maintained and available on site. Providing such a list would be redundant, require additional regulatory controls beyond 10CFR50.49 and will unnecessarily burden BGE.
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REQUEST FOR ADDITIONALINFORMATION 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 4.3.2 - Aging Management 4.3.18 Table 4.3 indicates that many components (CEASB, CS, CSTR, CSB, CSC, CSP, FAPFP, and LSSB A) are susceptible to neutron embrittlement, which generally results la loss of fracture toughness in the material composing the component. This lors 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 :omponent that is susceptible to neutron cmbrittlement, the peak neutron Cuence at the end of the extended period of operation, and the materials used to fabricate the specific component. For the limiting component (considering the neutron fluence, material fracture toughness and operating stresses in determining the limiting component), provide a fracture mechanics analysis to determine the critical Daw 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 Haws smaller in size than that of the critical Daw.
BGE requests clarification from NRC on this question. BGE's LRA already provides for
( inspections for these aging effects. The proposed analysis appears to assume these aging l effects are somehow unique to license renewal. In addition, BGE believes the overall i inspections proposed are more conservative than using an analytically bounding location
! approach.
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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 I COMPONENTSUPPORTS AND PIPING SEGMENTS TIIAT PROVIDE STRUCTURAL SUPPORT COMMODITY REPORTS, SECTIONS 3.1 AND 3.1 A DOCKET NOS. 50-317 AND 50-318 Section 3.1.2 Aging Management Review 1
3.1.18 Please clarify the following concerns regarding the information described in Table 3.1-3:
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- a. The loading due to rotating / reciprocating machinery has the potential to afTect many of the supports listed in the table. Provide the basis for the "N/A" and "not plausible" determination for supports other than electrical raceways, electrical cabinets and instruments, and tanks potentially affected by rotating / reciprocating machinery loads.
- b. Provide the basis for the "not plausible" determination for piping frame and stanchion supports and for metal spring isolators and fhed base supports potentially affected by loading due to hydraulic vibraticn or waterhammer. l
- c. Provide the basis for the "not plausible" and "N/A" determination for piping frame and stanchion supports, for metal spring isolators and fixed base supports, and for loss-of- l coolant accident restraints potentially affected by loading due to thermal expansion of piping and/or components. '
- d. Provide the basis for the "not plausible" determination for suppons potentially affected by stress corrosion cracking of high strength bolts.
- e. Provide the basis for the "not plausible" determination for supports potentially affected by radiation embrittlement of steel.
- f. Provide the basis for the "not plausible" determination for supports potentially affected by grout / concrete local deterioration.
- g. Provide the basis for the "not plausible" determination for supports potentially affected by lead anchor creep.
j This question is too broad. Similar questions were withdrawn by NRC and refocused. IIGE requests the NRC either conduct a public meeting or a site visit and discuss details of aging l
cffect plausibility calls with plant personnel or re-phrase question such that it is more focused.
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REQUEST FOR ADDITIONAL INFORMATION l CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 l PRIMARY CONTAINMENT STRUCTURE, SECTION 3.3A TURBINE BUILDING STRUCTURE, SECTION 3.3B INTAKE STRUCTURE, SECTION 3.3C MISCELLANEOUS TANK AND VALVE ENCLOSURES,SECTION 3.3D ELECTRICAL COMMODITIES,6.2 DOCKET NOS. 50-317 AND 50-318 General Questions Related to Sections 3.38,3.3C,3.3D. 3.3E and 6.2 2.; ? Provide the details of specific national codes and standards (e.g., ACI, AISC, etc.) including their co..:nns that will be used to determine repairs and acceptance criteria. If there are changes with ,
respect to specific national codes and standards previously committed to as part of the initial l
licensing basis, describe plans for incorporating these changes in the CCNPP Updated Final Safety l Analysis Report.
BGE requests the NRC clarify this question. BGE finds it difficult to identify specifie codes and standards that would be used in corrective actions for unidentified or hypothetical deficiencies. BGE also finds the request to reconcile changes to the licensing basis that may have involved codes and standards, or changes to these codes incorporate into the CLB difficult to respond to.
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[ REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 EMERGENCY DIESEL GENERATOR SYSTEM i INTEGRATED PLANT ASSESSMENT, SECTION 5.8 DOCKET NOS. 50-317 AND 50-318 Section 5.8.2 - Aging Management 5.8.7 Discuss the corrosion allowances in the design of EDG system components that are subject to corrosion, and how they will be addressed as part of the aging management program.
l BGE is answering this, as well as similar RAls, but suggests discussions with NRC to clarify
- any concerns it has. It is not apparent to BGE the significance of corrosion allowances in any of the CCNPP LRA findings.
5.8.8 Page 5.8-1 of the report states that operating experience relevant to ag;ng was obtained based on Calvert Cliffs Nuclear Power Plant specific information and past experience. Describe the basis upon which Haltimore Gas and Electric Company concluded that cavitation corrosion, intergranular attack, stress corrosion cracking, and thermal damage were not plausible aging effects for EDG systems in relation to any industry-wide experience with these aging effects in EDG systems.
This question is too broad. Similar questions were withdrawn by NRC and refocused. BGE requests the NRC either conduct a site visit and discuss OE with plant personnel or re-phrase question such that it is more focused.
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REOUEST FOR ADDITIONAL INFORM ATION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 INTEGRATED PLANT ASSESSMENT DOCKET NOS. 50-317 AND 50-318 Water Chemistry Program The following questions apply to the secondary water chemistry as discussed in Section 5,12," Main Steam and Blowdown System," and Section 5.9,"Feedwater System," of Appendix A to the Baltimore Gas and Electric Company (BGE) license renewal application:
9.1. Control of the secondary water chemistry plays an important role in ensuring that steam generators and other components exposed to secondary water will not be damaged by corrosion and will preserve their integrity. Please include the following information on your secondary water chemistry control program:
a) What amine is being used for controlling pil in the secondary water system?
b) Specify major differences in the secondary water chemistry (feedwater and/or steam generator) for power operation, startup, and shutdown.
c) Describe and provide technical bases for any significant differences in secondary water chemistry parameters specified in the BGE CP-217 procedure and the values recommended by the Electric Power Research Institute (EPRI) in their guideline reports, referenced in Section 5.12 of Appendix A to the BGE license renewal application.
d) Specify the upper limits of the major chemistry parameters and the allowable time period to restore chemistry parameters to acceptable limits.
This question is too broad. Similar questions were withdrawn by NRC and refocused. BGE requests tne !.dC cither conduct a site visit and discuss these questions with plant personnel or re-phrase question such that it is more focused.
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REQUEST FOR ADDITIONAL INFORMATION ,
CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 l INTEGRATED PLANT ASSESSMENT, SECTIONS 4.1,4.2,5.2,5.7,5.15, AND 5.16 DOCKET NOS. 50-317 AND 50-318 Section 4.1," Reactor Coolant System," and Section 4.2," Reactor Pressure Vessels and CEDMs/ Electrical Systems" 4.1.26 Provide the results of 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.
BCE requests the NRC clarify this question. It is not common practice to docket licensee internal audits. Rather, these audits are available for NRC inspection on site and are typically summarized in monthly resident inspector reports.
Section 5.2," Chemical and Volume Control System" 5.2.3 Provide the results of BGE's most recent internal audit of the BACI Program; including areas of j strengths and weaknesses, safety implication of findings, and corrective action plans and schedule !
for implementation. l BGE requests the NRC clarify this question. It is not common practice to docket licensee internal audits. Rather, these audits are available for NRC inspection on site and are
- typically summarized in monthly resident inspector reports.
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l h**. l Clarification on Staff Reauest for AdditionalInformation Begarding Baltimore Gas and Electric's Anolication for License Renewal )
For the Calvert Cliffs Nuclear Power Plant. Unit Nos.1 & 2 l The following is a revised version of RAI 4.1.17. The previous version of the RAI are given in Enclosure 1.
4.1.17 Regarding Program STP-M-574-1/2 l
Because industry experience to date indicates that eddy current inspection of steam generator (SG) tubes is not enough to manage aging effects before there is a loss of intended function, the staff also relies on licensee primary and secondary side chemistry control programs to mitigate corrosion, as well as, Technical Specification (TS) limits for SG leakage to detect aging effects. Based on the staff's review of Section 4.1 of BGE's license renewal application, it appears that BGE relies solely upon eddy current testing ;
to manage degradation of SG tubes before there is a loss of intended function. The information provided in the application for 'e management of SG tube degradation does not appear to meet template requirt : ents. Please clarify your explanation of the aging management programs for SG tubes. Also, describe the implementation of eddy current procedures in more detail.
Regarding Program MN-3-110 The LRA cites the subject program to manage erosion corrosion of various SG components such as the main steam outlet nozzles. Confirm that the inspection frequency of the various components is on a refueling basis. If the frequency is longer, provide the specific inspection frequency and the basis for the inspection frequency for the following components: main steam outlet nozzles, secondary manway, handhole and associated cover plates.
Provide relevant operating experience to demonstrate the effectiveness of the program in managing erosion corrosion of the following components: main steam outlet nozzles, secondary manway, handhole and associated cover plates.
l Regarding Program SG-20 l
, The licensee cites the subject program to manage general corrosion of the SG primary manway cover. Provide the acceptance criteria for the subject procedure. Provide relevant operating experience to demonstrate the effectiveness of the program in managing general corrosion of the manway cover.
Regarding Program FASTENER-01 The licensee cites the subject program to manage stress corrosion cracking (SCC) of the SG primary manway studs. Staff experience with SCC indicates that visual inspection may not be adequate to detect SCC because of the size and nature and Enclosure 2
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location of such cracks. Provide the basis for concluding visualinspection of the studs is sufficient to detect SCC before there is a loss of intended function.
Provide the acceptance criteria for the subject procedure. Provide relevant operating experience to demonstrate the effectiveness of the program in managing SCC of the SG primary manway studs.
Alloy 600 Program Confirm weld metals 182/82 are implicitly included within the scope of the Alloy 600 program.
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