Text

Summary of 980928 Meeting with Util in Rockville,Md Re License Renewal Activities for Plant,Units 1 & 2 & to Provide Rept Card for Both NRC & Util
	ML20195J039
Person / Time
Site:	Calvert Cliffs
Issue date:	11/16/1998
From:	Dave Solorio; NRC (Affiliation Not Assigned)
To:	; NRC (Affiliation Not Assigned)
References
	NUDOCS 9811240181
	Download: ML20195J039 (79)
	v • d • e

-_ _ - _ . - - - - - - - - - -

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0% 4 UNITED STATES s

g NUCLEAR REGULATORY COMMISSION j

$ f, g WASHINGTON, D.C. 2055H001

@E November 16, 1998

* * * * ,o i

, LICENSEE: Baltimore Gas and Electric Company l FACILITY: Calvert Cliffs Nuclear Power Plant, Unit Nos.1 and 2

SUBJECT:

SUMMARY

OF SEPTEMBER 28,1998, MEETING WITH BALTIMORE GAS l AND ELECTRIC COMPANY (BGE) REGARDING LICENSE RENEWAL i

ACTIVITIES FOR CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 AND2 l

l On September 28,1998, the Nuclear Regulatory Commission (NRC) staff held a 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 (LRA) for '

its Calvert Cliffs Nuclear Power Plant, Units 1 and 2, and for BGE to provide the report card for both the NRC and BGE staff's performance on the BGE LRA review. A list of meeting attendees is provided in Enclosure 1 and slides used by BGE for the discussion are provided in

! Enclosure 2.

BGE's presentation focused on an overview of performance relative to the areas of schedule adherence, quality of work, communications, cost control, accomplishments, and concerns.

BGE then highlighted potential areas for increased attention being that they had not received a bill for review fees in the prior month and that meeting the November 21,1998, milestone for replying to the staffs requests for additional information (RAI), issued on BGE's LRA, was highly dependent on achieving clarification on about 5 percent of the RAl. BGE also commented in regards to previous desire to develop better performance measures, that for now it was probably sufficient to maintain the high-level milestones previously developed for the staffs review of BGE's LRA. However, BGE also indicated that better performance measures related to RAI were being discussed between NRC and BGE management with some of the interim proposals highlighted on slide 10. Of note during the presentation was that BGE had categorized many of the RAI as sufficiently clear for BGE to respond; although, there were also a significant number of RAI which BGE stated were already answered in its LRA. Additionally, BGE indicated there was a number of RAI which needed further clarification.

Following the performance overview summary, BGE then provided copies to the staff of comments made by BGE relative to the 5 percent of the staffs RAI that it had requested j

clanfication on. Upon request by the staff, BGE then discussed its comments, as documented /

l in the handout, regarding several of the staffs RAI on BGE's LRA. Specifically, the comments ,

for RAI 5.9.47,4.1.7,5.10.6,7.6, and 8.3 were discussed by BGE. Following this discussion, /

the staff requested that if positions on RAI were not converging that BGE elevate issues. BGE then committed to do so and also reiterated concerns related to several of the RAI wiih respect to the value of summarizing Institute of Nuclear Power Operations and audit reports. The staff l

then added that NRC senior management was very interested in convergence of issues within the regulatory framework and requested feedback from BGE about where it felt regulatory boundaries had been exceeded.

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t November 16, 1998 Additionally, for the one-hundred and thirteen responses to RAI that BGE indicated were already included in its LRA, BGE requested that the NRC staff provide early feedback on whether the responses provided by BGE for these RAI were responsive or not.

Toward the end of the meeting, the staff reiterated that the intent of the management meetings with BGE to discuss the review of its LRA was to make sure progress was being made.

Additionally, the staff invited BGE to provide its perspectives on a letter that the staff recently sent to Nuclear Energy Institute (dated September 23,1998) discussing the staff's proposed guidance for LRA submittals on time-limited aging analyses (TLAA) on environmental ,

qualification for license renewal. Following this invitation BGE used two of the RAI related to l TLAA, provided in the second of BGE's handouts, to highlight how it thought that the RAI were technically wrong. Additionally, BGE then added that the staff needed to believe BGE's response to the RAl. BGE closed its presentation with its agreement to hold a future meeting with the NRC to discuss the proposed guidance.

Following the public meeting, BGE identified an additional RAI for which it requested clarification. On October 1,1998, BGE submitted, via electronic mail to the NRC, a revised copy of the RAI for which it had requested clarification, and which is provided as Enclosure 3.

l l

original signed by: l David L. Solorio, Project Manager '

l License Renewal Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket Nos. 50-317 and 50-318

Enclosures:

1. List of Attendees

2. BGE's Presentation Slides

3. Revised Copy of USNRC Requests for AdditionalInformation cc w/encis: See next page

- DISTRIBUTION:

See next page

DOCUMENT NAME:G:\ WORKING \SOLORIO\SEP_28.MTG OFFICE PDLR/DRPM PDLR/DRPM:D LA: PDI-y NAME Slittle7 DSolorio:sg CGrimes [ g DATE 11/@g8 11/$498 11/ \ @8 ' L OFFICIAL RECORD COPY

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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 l Commissioners Maryland i l 175 Main Street Engineering Division j l Prince Frederick, MD 20678 ~ 6 St. Paul Centre L

. 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 i

Baltimore, MD 21202-1631 L , Jay E. Silberg, Esquire l 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 l

- Mr. Bruce S. Montgomery, Director Baltimore, MD 21218 j NRM '

Calvert Cliffs Nuclear Power Plant Mr. Loren F. Donatell i 1650 Calvert Cliffs Parkway Lusby, MD 20657-4702 NRC Technical Training Center j 5700 Brainerd Road Chattanooga, TN 37411-4017

- Resident inspector U.S. Nuclear Regulatory Commission David Lewis l

. P.O. Box 287 Shaw, Pittman, Potts, and Trowbridge '

ll St. Leonard, MD 20685 2300 N Street, NW Washington, DC 20037 l .Mr. Richard I. McLean Nuclear Programs Douglas J. Walters Power Plant Research Program Nuclear Energy institute Maryland Dept. - of Natural Resources 1776 i Street, N.W.

Tawes State Office Building, B3 Suite 400 j l Annapolis, MD 21401 Washington, DC 20006-3708 i DJW@NEl.ORG i Regional Administrator, Region I '

U.S. Nuclear Regulatory Commission Barth W. Doroshuk l

475 Allendale Road Baltimore Gas and Electric Company

. King of Prussia, PA 19406 Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Mr. Charles H. Cruse, Vice President NEF ist Floor Nuclear Energy Division Lusby, Maryland 20657 Baltimore Gas and Electric Company

.1650 Calvert Cliffs Parkway - National Whistleblower Center i Lusby, MD 20657-47027 3233 P Street, N.W.

Washington, DC 20007 l

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Distribution:

HARD COPY (w/ Enclosures 1,2 & 3)

Docket Files-PUBLIC ,

PDLR R/F i i

OGC MEl-Zeftawy DISTRIBUTION: E-MAIL (w/ Enclosure 1)

FMiraglia (FJM)

JRob (JWR)

DMatthews (DBM) i CGrimes (CIG) i

-TEssig (THE)

Glainas (GCL) i JStrosnider (JRS2)

GHolahan (GMH) _

SNewberry (SFN) j 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)

AMurphy (AJM1) l TMartin (TOM 2) j J

DMadin (DAM 3)

GMeyer (GWM)

WMcDowell(WDM)

SStewart (JSS1) ;

- THiltz (TGH) l SDroggitis (SCD)

DSolorio (DLS2)

PDLR Staff I

+-

NRC & BGE MANAGEMENT MEETING SEPTEMBER 28.1998 NAME ORGANIZATION

1. DAVID SOLORIO NRC/PDLR

2. STEVEN HOFFMAN NRC/PDLR

3. JACK ROE NRC/DRPM

4. RALPH ARCHITZEL NRC/PGEB -

.5. CLAUDIA CRAIG NRC/PGEB

6. JACK STROSNIDER NRC/DE

7. ROB JOLLY NRC/PGEB

8. MARIAN ZOBLER NRC/OGC

9. ..JANICE MOORE NRC/OGC 10 STEPHANIE MARTZ NRC/OGC

11. KIM GREEN - NUSIS

12. TOM HILTZ NRC/EDO 13.~ WINDSEY LAYTON WASHINGTON POST

-14. BOB PRATO NRC/PDLR

' 15. M.S. TUCKMAN DUKE ENERGY

16. GREG ROBISON DUKE ENERGY

17. NANCY CHAPMAN SERCH/BECHTEL

- 18. MED EL-ZEFTAWY ~ NRC/ACRS

19. C.A' NEZIN -

PROJECT MANAGEMENT CORP.

20. DEBORAH STAUDINGER WINSTON & STRAWN

21. LYNN CONNOR WESTINGHOUSE

22. DOUG WALTERS NEl l

23. MIKE SCHOPPMAN - FLORIDA POWER & LIGHT ;

24. BOB WEISMAN NRC/OGC

25. DAVID LEWIS SPPT

26. - NEll HAGGERTY BGE

27. CHUCK RAYBURN BGE

. 28. DON SHAW ' BGE

. 29. R.P. HElBEL BGE

30. B.W. DOROSHUK BGE

~ 31.. ANNE COTTINGHAM WINSTON & SHAWN 32.' JIM CANNON SENATOR DOMENICI I

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l- l Enclosure 1 l

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Calvert Cliffs License Renewal

@j ,j, Application Status Report to the US Nuclear Regulatory m Commission 4

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.i t Dick Heibel, Manager, Nuclear Projects -

[. Barth Doroshuk, Project Manager, LR

- M September 28,1998 Enclosure 2

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" Presentation Outline

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September 1998

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. . W LRA Report Card u.

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i ie Schedule Adherence e Cost Control d Technical responses being Expect third quarter bill next :'

. i. I worked, BGE requires NRC month I H clarification; SAMA questions

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received on or before

go Quality of Work schedule; BGE/NRC r

[ El NRC has expressed willingness to interaction continuing to clarify

[ lI clarify questions that appear to go NRC expectations N) beyond protocols; policy issues

.- I will be ID'd e Concerns 1 5 5 -

Schedule is highly dependent

[e CommunlCations 3: on clarifying ~ 5% of RAls

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iNRC Overall Schedule

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!!1998

. 4

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Application Submitted n

l 5/10/98 -

Acceptance and Docketing Completed f ,

!" 7/8/98* -

Notice of Opportunity for Hearing Filed by NRC I. 7/7 - 7/10 -

i NRC Environmental Site /Public Visit

.,.l lI 7/9/98 ii -

Public Meetings for Environmental impact Statement e

hy Scoping

@ 8/7/98 Filing of Intervenor Petitions Ends .

D 9/7/98 -

NRC Completes Technical Requests for Additional

, E Information (RAls) 1 9/28/98 -

Management Progress Review

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[ CCNPP License Renewal Status Report to the USNRC 6 98-054

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, c NRC Overall Schedule t 7 j .

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" 10/1 Deadline for Whistleblower Cgnter Filing j

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NRC Completes Environmental RAls

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{ 10 /29/98 - Management Program Review; NRC Steering

~! Committee Meeting l i ,i

! . 11/2 -

BGE and NRC Responses to Whistleblower Center

)u ;] Filing, if any e li 11/12 -

Pre-Hearing Conference, if needed i

, } ., 11/21/98 - BGE Completes Technical Responses '

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12/6/98 -

BGE Completes Environmental Responses

j 12/98 -

Management Progress Review

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WRC Overall Schedule *

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f:: 3/6/99 NRC issues Draft Environmental Statement (DES) for Comment i

' l' i. 3/21/99 -

NRC Completes Draft Safety Evaluation Report

I; (SER) and Open items

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Public Meeting for Draft EIS Comments .

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NRC Completes DES Comments

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5 7/19/99 -

BGE Completes Responses to Open items s

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11/16/99 - NRC issues Final SER and ES

; 2/14/00 -

ARCS Meeting

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Commission Decision on Application

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- RAI, policy issue, RAI response and staff closure

_ ! ?! milestones will be used once developed l: - NRC Branch Chief and BGE Project Manager ,

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' LRA Topical Area of Rats RAb MAIResponse RAIanyond nGE nequires

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, e NRC RAI (technical and SAMA) generally observing regulatory protocols, but some
require clarification e All "important" NRC " issues" are hard to keep track of; no list yet (may be too early) 1 i
CCNPP License RenewalStatus Report to the USNRC 17 98-054
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dt ! e Recommend continuing management z<[ reviews; next meeting on October 29,1998
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1I CCNPP License RenewalStatus Report to the USNRC 21
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BGE REQUESTS FOR CLARIFICATION ON USNRC REQUESTS FOR ADDITIONAL INFORMATION Prepared by BGE Calvert Cliffs License Renewal Project SEPTEMBER 28,1998
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e CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS,1 & 2 FEEDWATER SYSTEM INTEGRATED PLANT ASSESSMENT, SELTION 5.9 DOCKET NOS. 50-317 AND 50-318 Aging Management l
l 5.9.47 One of the most effective ways of minimizing erosion / corrosion is to control secondary water chemistry, that is, pH and oxygen concentration. Describe whether pH and oxygen concentration are controlled in the feedwater system and if so, specify the parameter ranges.
l 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 j addition, there is an effort between NEI (NEI-9706) and NRC on chemistry controls which l
is ongoing that provides a significant amount ofinformation in this area. BGE requests NRC evaluate the need for requesting this additionalinformation, given the above described exchanges already underway.
t 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 l for enhancements. Please briefly summarize the results of the INPO assessment and outline the l INPO recommendations for improvements at the Calvert Cliffs plants.
l BGE requests that NRC review the need for INPO reports since they are proprietary and the NRC hu access to them already.
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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLAhT ,
UNIT NOS.1 & 2 1 REACf0R PRESSURE VESSELS AND CONTROL ELEMENT DRIVE )
MECHANISMS /ELECfRICAL INTEGRATED PLANT ASSESSMENT, SECTION 4.2 ;
DOCKET NOS. 50-317 AND 50-318 i
Section 4.2.2 - Aging Management l
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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 l extended licensed term, will there be any limitations in operation of the shutdown cooliug ,
system due to American Society of Mechanical Engineers Boiler and Pressure Vessel Code l
(ASME Code), Appendix G, P-T operating limits and the minimum permissible temperature of !
the reactor vessel?
BCE is requesting NRC clarify #8 above. Current plant practice is to maintain these curves as required, not necessarily at license termination points. No 10CFR50 requirement 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.
BGE has had a TELCON with NRC on this question and is providing response according to NRC clarification.
l 4.2.17 Section 4.2.2 of the LRA states "The threshold for onset of neutron effects for RPV materials is conservatively defined to be a fast neutron fluence that exceeds IE17a/cm2," citing Appendtx H of 10 CFR Part 50. The staff believes that Appendix H cites the indicated neutron fluence as a threshold below which a reactor vessel material surveillance program is not required for the vessel. Appendix H 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 are negligible effects from neutron fluence below IE17n/cm2.
BGE is requesting clarification of #17 above. BGE has had a TELCON with NRC on this question and is answering this question according to NRC clarificadon.
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I REQUEST FOR ADDITIONAL INFORh1ATION CALVERT CLIFFS NUCLEAR POWER PLANT, (TNIT 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 5.1.8.10 Provide a summary description of Calvert Clifts operating and maintenance experience related to boric acid corrosion of carbon steel components. In particular, characterize the extent to which boric acid corrosion of carbon steel components has changed since the initial implementation of the boric acid corrosion inspection (BACI) program. Also, 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 implementation of the BACI program.
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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1 , t 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 l
experience that demonstrates the effectiveness of the programs that are relied upon to manage i these aging effects:
L a.
boric acid corrosion - Technical Specifications (TS) leakage limits, and ASME Section XI, Subsection IWB, examination categories B-P; 1
b.
cracking oflarge bore piping- ASME Section XI Subsection JWB, examination categories B-J and B F, and flaw evaluation criteria IWB-3006; i c. cracking of small bore piping (less than 4 in but greater than 1 in diameter)- augmented i 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; l d- cracking of bolting - programs consistent with ASME Section XI, Subsection IWB, examination i categories B-G 1 and B G 2, and NRC Bulletin 82-02;
c. pressunzer shell, heads, beater belt forgings - ASME Section XI, Subsection IWB, examination l categories B-B and B-P, and primary water chemistry; I
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 ifloconel is used, information resulting fmm IN 90-10;
g. integral attachments - ASME Section XI, Subsection IWB, examination category B-H, and -
primary water chemistry;
h. heater sheaths and end caps - ASME Section XI, Subsection IWB, examination category B-P, and TS leakage limits; l i. loss ofpreload 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 88-05, and TS leakage limits.
l 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 aging effects but not discussed them in the
LRA since they are not plausible.
4.1.12 It appears that BGE used ferrite criteria to screen components subject to thermal embrittlement.
However, the NRC regards fertite content as inadequate criterion for screening as stated in NUREG-1557. Therefore, justify using ferrite content as screening crite-ia.
The use of ferrite criteria to screen components has been a part of an ladustry Position since 1994. It has also been submitted as part of NEI/EPRI efforts to resolve generic aging i
issues. BGE requests NRC explain the ainadequateness" of a position.
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E 4.1.17 Please provide a summary description for the following procedures regarding how their implementation will address the following elements for their related aging management program (s): (a) The scope of structures and components managed by the program; (b)
Preventive actions designed to mitigate or prevent aging degradation; (c) Parameters monitored or inspected relative to degradation of specific structure and component intended functions; (d)
Detection of aging effects before loss of stmeture and component intended functions; (e)
Monitoring, trending, inspection, testing frequency, and sample size to ensure timely detection of aging efTects and corrective actions;(f) Acceptance criteria to ensure structure and component intended functions; and (g) Operating experience that provides objective evidence to demonstrate that the efTects of aging will be adequately managed.
a. Procedure SG-20, " Primary manway cover removal and installation"
b. Administrative Procedure MN-3-110, " Inservice Inspection of ASME XI Components"
c. Technical Procedure FASTENER 01," Torquing and Fastener Applications"
d. Procedure STP-M-574-1/2,"EC Examination of CCNPP % Steam Generators"
e. CASS Evaluation program
f. Alloy 600 program
g. STP-0-27-1/2, "RCS Leakage Evaluation"
h. MN-3-301, "BACl Program"
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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l REQUEST FOR ADDITIONAL INFORMATION l CALVERT CLIFFS NUCLEAR POWER PLANT, UhTF NOS.1 & 2 FIRE PROTECTION SYSTEM l INTEGRATED 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 clarify its intent. BGE Interprets this question as a compilation of CLB and BGE is not clear on that is what SRC 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 l
DOCKET NOS. 50-317 AND 50-318 l 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.
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 CLIITS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 TIME-LIMITED AGING ANALYSES, SECTION 2.1 DOCKET NOS. 50-317 AND 50-318 2.1.3 Page 2.14 of the license renewal application (LRA) indicates that the pressure-temperature (P-T) limits in the Calvert Cliffs 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 defined by the current operating term, for example,40 years." (Criterion 3 ofthe 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. However, Section 2.1 of Appendix A to the BGE application contains future commitments to perform the TLAA evaluations. The following are examples:
, Subsection Heading Statement i j
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."
a 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 function (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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! , r REQUEST FOR ADDITIONAL INFORMATION i CALVERT CLIFFS UNITS 1 AND 2 INTEGRATED PLANT ASSESSMENT l ON METAL FATIGUE !
DOCKET NOS. 50-317/50-318 '
Section 5.2. " Chemical and Volume Control System" !
l 7.6 Section 3.2.3 of EPRI Report TR-107515 contains an evaluation of environmental effects on the j CVCS Charging inlet Nozzle using methodology developed in EPRI Report TR-105759,"An i
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 l TR-105759. Attached are comments on the application of the EPRI methodology for environmental fatigue factors to the Calvert Cliffs plant. Based on these comments, provide the j following

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' (a) Discuss the impact of the current Argonne National Laboratory (ANL) statistical l
correlations of environmental test data on the Calvert Cliffs fatigue evaluation.
1 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. l L i (b) The technical basis for the assertion that the American Society of Mechanical Engineers (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 acc2 pts the ASME code as endorsed by 10CFR50.55a as part of our CLB.
(c) The technical justification for the strain threshold values.
BGE 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).
Section 3.3.3 of EPRI Report TR 107515 contains an evaluation of the Surge Line using 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).
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 contest conclusions.
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7.16 Section 3.3.3 of EPRI Report TR-107515 contains an evaluation of the Surge Line using 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).
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.
1 7.17 Section 3.3.3.2 of EPRI Report TR-107515 indicates that the procedure in Section 3.1.3.2 of the l
EPRI report was used to develop the environmental factor used in the evaluatiou. 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 Repon 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. I Compare the results of the statistical assessment with the ANL assessment contained in l 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 l
comparison, indicate whether the use of the " weighted average" approach will produce an adequate margin to account for test data scatter.
I 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.
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 t
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 conclusion (see attachment).
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, i
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,, i REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 INTEGRATED PLANT ASSESSMENT ON GENERIC SAFETY ISSUES l DOCKET NOS. 50-317 AND 50 318 i 8.3 In a letter dated June 2,1998, the staficoncluded 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 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 snanage the EQ analyses for the period of extended operation.
BGE has provided all the requested information in the LRA section on EQ except for the EQ Master List. BGE requests to discuss this with NRC since compth ' ace with :
10CFk50.47 as requirement that will carry forward as part of the CLE a accordance with the rule. The EQ Master List is maintained an<l available on site. Pr, . 'ag such a list would be redundant, require additional regulatory controls beyond 10, i "'50.49 and will unnecessarily burden BGE.
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l REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLAhT, UNIT NOS.1 & 2 REACTOR VESSEL thTERNALS j INTEGRATED PLAST 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 LSSBA) are susceptible to neutron embrittlement, which generally results in loss of fracture toughness in the material composing the component. This loss of fracture toughness is a reduction in resistance to crack growth, which could mean that parts that are macroscopically degraded (through wear or some sort of cracking mechanism such as SCC or fatigue) may fail (fracture) at load levels and/or degradation (i.e., smaller crack sizes) that are lower than those if the part was not in an embrittled condition. Identify for each component that is susceptible to neutron embrittlement, the peak neutron fluence at the end of the extended period of operation, and the materials used to fabricate the specific component. For the limiting component (considering the neutron fluence, material fracture toughness and operating stresses in devermining the limiting component), provide a fracture mechanics analysis to determine the critical flaw size during normal operation and emergency and faulted conditions. Provide data to justify the fracture toughness assumed in the analysis. Identify the inspection procedure and the capability of the inspection to detect flaws smaller in size than that of the critical flaw.
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 effects are somehow unique to license renewal in addition, BGE believes the overall inspections proposed are more conservative than using an analytically Imunding location approach.
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REQUEST FOR ADDITIONAL INFORMATION CAINERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 COMPONENT SUPPORTS AND PIPING SEGMENTS THAT 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 3.1.18 Please clarify the following concerns regarding the information described in Table 3.1-3:
a.
The loading due to rotating / reciprocating machinery has the potential to affect 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 fixed base supports potentially affected by loading due to hydraulic vibration or waterhammer.
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-coolant accident restraints potentially affected by loading due to thermal expansion of piping and/or components, d.
Provide the basis for the "not plausible" determmation for supports potentially affected by stress corrosion cracking of high strength bults.
c.
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.
This question is too broad. Similar questions were withdrawn by NRC and refocused.
BGE requests the NRC either conduct a public meeting or a site visit and discuss details of aging effect plausibility calls 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 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 3.3.9 Provide the details of specific national codes and standards (e.g., ACI, AISC, etc.) including their editions 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 pan of the initial licensing basis, describe plans for incorporating these changes in the CCNPP Updated Final Safety Analysis Repon.
BGE requests the NRC clarify this question. BGE finds it difficult to identify specific codes and standards that would be used in corrective actions for unidentified or hypothetical deficiencies. BGE also finds the request to reconcile changes to thelicensing 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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l j REQUEST FOR ADDITIONAL INFORMATION i CAINERT CLIFFS NUCLEAR POWER PLANT UNIT NOS. I & 2 EMERGENCY DIESEL GENERATOR SYSTEM l
INTEGRATED PLANT ASSESSMENT, SECTION 5.8 DOCKET NOS. 50-317 AND 50-318 l Section 5.8.2 - Aging Management 1
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.
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.
i 5.8.8 Page 5.8-1 of the report states that operating experience relevant to aging was obtained based on Calvert Cliffs Nuclear Power Plant specific information and past experience. Describe the basis upon which Baltimore 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 efTects 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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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT, UNfr NOS.1 & 2 INTEGRATED PLANT ASSESSMENT, SECTIONS 4.1,4.2,5.2,5.7. 5.15, AND 5.16 DOCKET NOS. 50 317 AND 50-318 i
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.
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.
Section 5.2," Chemical and Volume Control System" 5.2.3 Provide the results of BGE's most recent intemal audit of the BACI Program; including areas of strengths and weaknesses, safety implication of findings, and corrective action plans and schedule for implementation.
BGE requests the NRC clarify this question. It is not common practice to docket licensee j
laternal audits. Rather, these audits are available for NRC inspection on site and are typically summarized in monthly resident inspector reports. '
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BGE REQUESTS FOR CLARIFICATION ON USNRC REQUESTS FOR ADDITIONAL INFORMATION l
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1 Prepared by BGE Calvert Cliffs License Renewal Project October 1,1998 t
Enclosure 3
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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, pH and oxygen concentration. Describe whether pH 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 (NEI.9706) and NRC on chemistry controls which is ongoing that provides a significant amount ofinformation in this area. BGE requests NRC evaluate the need for requesting this additional information, 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 per'ormed 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 recommendations for improvements at the Calvert Cliffs plants.
BGE 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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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECHANISMS / ELECTRICAL INTEGRATED PLANT ASSESSMENT,SECTION 4.2 DOCKET NOS. 50-317 AND 50-318 Section 4.2.2 - Aging Management 4.2.8 Provide pressure-temperature (P-T) 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 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 permissible temperature of the reactor vessel?
BGE is requesting NRC clarify #8 above. Current plant practice is to maintain these curves as required, not necessarily at license termination points. No 10CFR50 requirement 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.
BGE 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 !
conservatively defined to be a fast neutron fluence 'that exceeds IE17n/cm2," citing Appendix H of 10 CFR Part 50. The staff believes that Appendix H cites the indicated neutron fluence as a
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l threshold below which a reactor vessel material surveillance program is not required for the l vessel. Appendix H 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 are negligible effects from neutron fluence below 1 E17n/cm2. l BGE is requesting clarification of #17 above. BGE has had a TELCON with NRC on this question and is answering this question according to NRC clarification.
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Enclosure 3
REQUEST FOR ADDITIONAL INFORMATION i
CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 '-
SPENT FUEL POOL COOLING SYSTEM INTEGRATED PLANT ASSESSMENT, SECTION S.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 boric acid corrosion of carbon steel components has changed since the initial implementation of the boric acid corrosion inspection (BACl) program. Also, 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 implementation 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 CLii/FS NUCLEAR POWER PLANT UNIT NOS, I & 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 l
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 of large bore piping -- ASME Section XI, Subsection IWB, examination categories B-J I i
and B-F, and flaw evaluation criteria IWB-3000; i
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;
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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;
c. pressurizer shell, heads, heater belt forgings - ASME Section XI, Subsection IWB, examination categories B-B and B-P, and primary water chemistry;
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; l g. integral attachments -- ASME Section XI, Subsection IWB, examination category B-H, and pnmary water chemistry; t
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 88-05, and TS leakage limits. ,
1 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 these issues.
4.1.12 It appears that BGE used ferrite criteria to screen components subject to thermal embrittlement.
However, the NRC regards ferrite content as inadequate criterion for screening as stated in l NUREG-1557. Therefore, justify using ferrite content as screening criteria.
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s 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.
BGE requests NRC explain the " inadequateness" of a position.
4.1.17 Please provide a summary description for the following procedures regarding how their implementation will address the following elements for their related aging management program (s): (a) The scope of structures and components managed by the program; (b) Preventive actions designed to mitigate or prevent aging degradation;(c) Parameters monitored or inspected relative to degradation of specific structure and component intended functions;(d) Detection of 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 intended functions; and (g) Operating experience that provides objective evidence to demonstrate that the effects of aging will be adequately managed.
a. Procedure SG-20, " Primary manway cover removal and installation"
b. Administrative Procedure MN-3-110, " Inservice inspection of ASME XI Components" c,
Technical Procedure FASTENER-01," Torquing and Fastener Applications"
d. Procedure STP-M-574-1/2,"EC Examination of CCNPP % Steam Generators"
e. CASS Evaluation program
f. . Alloy 600 program
g. STP-0-27-1/2, "RCS Leakage Evaluation"
h. MN-3-301, "BACI Program"
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 INTEGRATED 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 clarify 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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l REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNITS NOS.1 & 2 l 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 l 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. l l BGE believes this question is too broad and requests NRC clarifyits intent. An option for i disposition is meetings either at Calvert Cliffs or NRC Offices to review the site i 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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. e 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 2.1.3 Page 2.1-4 of the license renewal application (LRA) indicates that the pressure-temperature (P-T) limits in the Calvert Cliffs 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 defined by the current operating term, for example,40 years." (Criterion 3 of the c'efinition 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. However, Section 2.1 of Appendix A to the BGE application contains future commitments to perform the TLAA evaluations. The following are examples:
Subsection Heading 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 function (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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CALVE.RT CLIFFS UNITS 1 AND 2 INTEGRATED PLANT ASSESSMENT ON METAL FATIGUE DOCKET NOS. 50-317/50-318 Section 5.2," Chemical and Volume Control System" 7.6 Section 3.2.3 of EPRI Report TR-107515 contains an evaluation of environmen:al 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 staff's technical concerns regarding the methodology in EPRI Report TR 105759.
Attached are comments on the application of the EPRI methodology for environmental fatigue factors to the Calvert Cliffs plant. Based 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.
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.
(b) The technical basis for the assertion that the American Society of Mechanical Engineers (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.
BGE will provide this answcr.
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. BGE 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 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).
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.
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" l approach as discussed in a June 1,1998, letter from the Nuclear Energy Institute to the NRC regarding EPRI Report TR.-105759. If the "weignted 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 LWP, Environments." On the basis of this comparison, indicate whether the use of the " weighted average" approach will produce an i adequate margin to account for test data scatter. I i
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 j excerpts from research activities can produce out of context conclusions. '
- 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 I
conclusion (see attachment). l l
BGE does not believe the subject research project can be commented on in a timely nor l
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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Enclosure 3
REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 INTEGRATED PLANT ASSESSMENT ON GENERIC SAFETY ISSUES 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 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 maintzin 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.
BGE has provided all the requested information in the LRA section on EQ except for the EQ Master List. BGE requests 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 l DOCKET NOS. 50-317 AND S0-318 Section 4.3.2 - Aging Management 4.3.18 Table 4.3 indicates that many components (CEASB, CS, CSTR, CSB, CSC, CSP, FAPFP, and LSSBA) are susceptible to neutron embrittlement, which generally results in loss of fracture toughness in the material composing the component. This loss of fracture toughness is a reduction in resistance ta crack growth, which could mean that parts that are macroscopically degraded (through weai or some son of cracking mechanism such as SCC or fatigue) may fail (fracture) at load levels and'or degradation (i.e., smaller crack sizes) that are lower than those if the part was not in an embrittled condition. Identify for each component that is susceptible to neutron embrittlement, the peak neutron fluence at the end of the extended period of operation, and the materials used to fabricate the specific component. For the limiting component (considering the neutron Guence, material fracture toughness and operating stresses in detent ining the limiting component), provide a fracture mechanics analysis to determine the critical flaw size during normal operation and emergency and faulted conditions. Provide data to justify the fracture toughness assumed in the analysis. Identify the inspection procedure and the capability of the inspection to detect flaws smaller in size than that of the critical flaw.
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 i effects are somehow unique to license renewal. In addition. BGE believes the overall l inspections proposed are more conservative than using an analytically bounding location
approach.
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.* l REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 COMPONENT SUPPORTS AND PIPING SEGMENTS THAT PROVIDE STRUCTURAL SUPPORT COMMODITY REPORTS, SECTIONS 3.1 AND 3 l A DOCKET NOS. 50-317 AND 50-318 Section 3.1.2 - Aging Management Review 3.1.18 Please clarify the following concerns regarding the information described in Table 3.1-3: I
a. The loading due to rotating / reciprocating machinery has the potential to affect many of the supports listed in the table. Provide the basis for the "N/A" and "not plausik. P.ermination for supports other than electrical raceways, electrical cabinets and l
instrumeno and tanks potentially afTected by rotating / reciprocating machinery loads.
b. Provide the basis for the "not plausible" determiration for piping frame and stanchion supports and for metal spring isolators and fixed base supports potentially affected by loading due to hydraulic vibration or waterhammer.
c. Provide the basis for the "not plausible" and "N/A" determination for piping frame and stanchion suppens, for metal spring isolators and fixed base suppons, and for loss-of-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 suppons potentially affected by radiation embrittlement of steel l f. Provide the basis for the "not plausible" determination for supports potentially affected by grout / concrete local deterioration.
l g. Provide the basis for the "not plausible" determination for suppons potentially affected by lead anchor creep.
This question is too broad. Similar questions were withdrawn by NRC and refocused. BGE i
requests the NRC either conduct a public meeting or a site visit and discuss details of aging j effect plausibility calls with plant personnel or re-phrase question such that it is more
! focused.
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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUL EAR POWER PLANT, UNIT NOS.1 & 2 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 3.3B,3.3C,3.3D,3.3E and 6.2 3.3.9 Provide the details of specific national codes and standards (e.g., ACI, AISC, etc.) including their editions 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 licensing basis, describe plans for incorporating these changes in the CCNPP Updated Final Safety Analysis Report.
BGE requests the NRC clarify this question. BGE finds it difficult to identify specific 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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CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 EMERGENCY DIESEL GENERATOR SYSTEM 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.
BGE is answeririg this, as weli 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 aging was obtained based on :
Calvert Cliffs Nuclear Power Plant specific information and past experience. Describe the basis '
upon which Baltimore Gas and Electric Company concluded that cavitation corrosion, intergranular attack, stress corrosion cracking, and thermal damage were not plausible aging i effects for EDG systems in relation to any industry-wide experience with these aging effects in l EDG systems.
This question is too broad. Similar questions were withdrawn by NRC and refocused. BGE I 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 INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1,&J INTEGRATED PLANT ASSESSMENT DOCKET NOS. 50-317 AND 50-318 Water Chemistry Proaram 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.l. Control of the secondary water chemistry plays an important role in ensuring that steam generators I 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 pH 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 i chemistry parameters specified in the BGE CP-217 procedure and the values recommended I 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 the NRC either conduct a site visit and discuss these questions with plant personnel or re-phrase question such that it is more focused.
Enclosure 3
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REQUEST FOR ADDITIONAL INFORMATION l CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 I 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 CED is/ 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. '
BGE requests the NRC clarify this queuion. 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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i Section 5.2, " Chemical and Volume Control System" l
l 5.2.3 Provide the results of BGE's most recent intemal audit of the BACI Program; including areas of strengths and weaknesses, safety implication of findings, and corrective action plans and schedule for implementation.
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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BGE REQUESTS FOR CLARIFICATION ON USNRC REQUESTS FOR ADDITIONAL INFORMATION Prepared by BGE Calvert Cliffs License Renewal Project October I,1998 i
Enclosure 3
. e CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 FEEDWATER SYSTEM INTEGRATED PLANT ASSESSMENT,SECTION 5.9 DOCKET NOS. 50-317 AND 50-318 i Aging Management 1
5.9.47 One of the most effective ways of minimizing erosion / corrosion is to control secondary water chemistry, that is, pH and oxygen concentration. Describe whether pil and oxygen concentration are controlled in the feedwater system and if so, specify the parameter ranges.
j BCE 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 ongoing that provides a significant amount ofinformation in this area. BGE requests NRC i evaluate the need for requesting this additionalinformation, given the above described exchanges already underway. l 5.9.54 Page 5.9-20 of the application indicates that the institute of Nuclear Power Operations (INPO) has 1 performed assessment of the BGE erosion / corrosion program and provided recommendation for enhancements. Please briefly summarize the results of the !NPO assessment and outline the INPO recommendations for improvements at the Calvert Cliffs plants.
BGE 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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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 REACTOR PRESSURE VESSELS AND CONTROL ELEMENT DRIVE MECilANISMS/ ELECTRICAL INTEGRATED PLANT ASSESSMENT, SEC'llON 4.2 DOCKET NOS. 50-317 AND 50-318 Section 4.2.2 - Aging Management 4.2.8 Provide pressure-temperature (P-T) limits for the extended operating term and identify the operating window relative to pump operation for the shutdown cooling system. During the I 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 permissible temperature of the reactor vessel?
BGE is requesting NRC clarify #8 above. Current plant practice is to maintain these curves as required, not necessarily at license termination points. No 10CFR50 requirement 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 cunes.
BGE 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 conservatively defined to be a fast neutron fluence that exceeds IE17n/cm2," citing Appendix H of 10 CFR Part 50. The staff believes that Appendix H cites the indicated neutron fluence as a threshold below which a reactor vessel material surveillance program is not required for the
. vessel. Appendix H 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 are negligible eff: cts from neutron fluence below 1E17n/cm2.
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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 according to NRC clarification.
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CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 SPENT FUEL POOL COOLING SYSTEM i 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 C::! vert Cliffs operating and maintenance experience related to boric acid corrosion of carbon steel components. In particular, characterize the extent to which boric acid corrosion of carbon steel components has changed since the initial implementation of the boric acid corrosion inspection (BACI) program. Also, 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 implementation of the BACI program.
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 S0-318 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 !WB, examination categories B-P; b,
cracking oflarge bore piping -- ASME Section XI, Subsection IWB, examination categories B-J j
and B-F, and flaw evaluation criteria IWB-3000, l
c. cracking of small bore piping (less than 4 in but greater than I 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. l cracking of bolting -- programs consistent with ASME Section XI, Subsection IWB, examination categories B-G-I and B G-2, and NRC Bulletin 82 02; l
e. pressurizer shell, heads, heater belt forgings -- ASME Section XI, Subsection IWB, examination categories B-B and B-P, ed primary water chemistry;
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 ifInconel is used, information resulting from IN 90-10;
g. integral attachments -- ASME Section XI, Subsection IWB, examination category B-H, 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 88-05, and TS leakage limits.
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 these issues.
4.1.12 It appears that BGE used ferrite criteria to screen components subject to thermal embrittlement.
l However, the NRC regards ferrite content as inadequate criterion for screening as stated in l
NUREG-1557. Therefore, justify using ferrite content as screening criteria.
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.s 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 r ging issues.
BGE requests NRC explain the " inadequateness" of a position. I 4.1.17 Please provide a summary description for the following procedures regarding how t ':ir i implementation will address the following elements for their related aging manageoent j program (s): (a) The scope of structures and components managed by the program: b) Preventive actions designed to mitigate or prevent aging degradation;(c) Parameters monitort i or inspected ,
relative to degradation of specific structure and component intended functions;(d Detection of l
aging effects before loss of structure and component intended functions; (e) Morr oring, trending, inspection, testing frequency, and sample size to ensure timely detection of agirg effects and corrective actions;(f) Acceptance criteria to ensure structure and component ir ce ided functions; ,
and (g) Operating experience that provides objective evidence to demonstrate t at the effects of I aging will be adequately managed.
a. Procedure SG-20," Primary manway cover removal and installation" l
b. Administrative Procedure MN 3-110 " Inservice Inspection of ASME XI I Components"
c. Technical Procedure FASTENER-01," Torquing and Fastener Applications" l
d. Procedure STP-M 5741/2,"EC Examination of CCNPP % Steam Generators" i
e. CASS Evaluation program
f. Alloy 600 program j
g. STP-0-27-1/2,"RCS Leakage Evaluation" I
h. MN-3-301,"BACI Program"
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 cither 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 ADDITIONALINFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 FIRE PROTECTION SYSTEM INTEGRATED 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 NitC clarify 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 iesults.
. C 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.
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 ADDITIONALINFORMATION l
CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 TIME-LIMITED AGING ANALYSES,SECTION 2.1 DOCKET NOS. 50-317 AND 50-318 2.1.3 Page 2.1-4 of the license renewal application (LRA) indicates that the pressure-temperature (P-T) limits in the Calvert Cliffs Technical Specifications are valid for Units 1 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 defined 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. However, Section 2.1 of Appendix A to the BGE application contains future commitments to perform the TLAA evaluations. The following are examples:
Subsection Heading 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 function (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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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS UNITS 1 AND 2 INTEGRATED PLANT ASSESSMENT ON METAL FATIGUE DOCKET NOS. 50-317/50-318 Section 5.2, " Chemical and Volume Control System" 7.6 Section 3.2.3 of EPP,1 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 staff's technical concerns regarding the methodology in EPRI Report TR-105759.
Attached are comments on the application of the EPRI methodology for environmental fatigue factors to the Calvert Cliffs plant. Based 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 Cahert ClitTs fatigue evaluation.
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.
(b) The technical basis for the assertion that the American Society of Mechanical Engineers (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.
i BGE will provide this answer.
I Section 4.1," Reactor Coolant System" l
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. BGE requests NRC withdraw this question since commenting on excerpts from research activities can produce out of context conclusions.
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l 7.16 Section 3.3.3 of EPRI Report TR-107515 contains an evaluation of the Surge Line using methodology developed in EPRI Report TR-105759. Discuss the applicability and impact of the i latest stainless steel fatigue correlation from ANL on this evaluation (see attachment). '
BGE does not believe the subject research project can be commented on in a timely nor
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reasonable fashion. BGE requests NRC withdraw this question since commenting on l excerpts from research activities can produce out of context conclusions.
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7.17 Section 3.3.3.2 of EPRI Report TR 107515 indicates that the procedure in Section .t.l.3.2 of the EPRI report was used to develop the environmental factor used in the evaluation. Indicate j whether the factor was calculated based on a " standard" treatment or" weighted average" l l
approach as discussed in a . lune 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
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comparison, indicate whether the use of the " weighted average" approach will produce an adequate margin to account for test data scatter. 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 j excerpts from research activities can produce out of context conclusions.
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 conclusion (see attaciunent).
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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I REQUEST FOR ADDITIONAL INFORMATION i
CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 INTEGRATED PLANT ASSESSMENT ON GENERIC SAFETY ISSUES 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 analyses (TLAAs). Until the aatter is resolved, please provide the EQ Master List of electrical equipment and indicate wiv . 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 i
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.
l BGE has provided all the requested information in the LRA section on EQ except for the EQ Master List. BGE requests 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 l
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REQUEST FOR ADDITIONAL INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 REACTOR VESSEL INTERNALS INTEGRATED PLANT ASSESSMENT, SECTION 4.3 t
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 LSSBA) are susceptible to neutron embrittlement, which generally results in loss of fracture toughness in the material composing the component. This loss of fracture toughness is a reduction in resistance to crack growth, which could mean that parts that are macroscopically degraded (through wear or some sort of cracking mechanism such as SCC or fatigue) may fail (fracture) at I load levels and/or degradation (i.e., smaller crack sizes) that are lower than those if the part was not in an embrittled condition. Identify for each component that is susceptible to neutron embrittlement, the peak neutron fluence at the end of the extended period of operation, and the materials used to fabricate the specific component. For the limiting component (considering the neutron fluence, material fracture toughness and operating stresses in determining the limiting component), provide a fracture mechanics analysis to determine the critical flaw size during.
normal operation and emergency and faulted conditions. Provide data tojustify the fracture toughness assumed in the analysis. Identify the inspection procedure and the capability of the inspection to detect flaws smaller in size than that of the critical flaw.
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 effects are somehow unique to license renewal. In addition, BGE believes the overall inspections proposed are more conservative than using an analytically bounding location approach.
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REQUEST FOR ADDITIONALINFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 COMPONENT SUPPORTS AND PIPING SEGMENTS THAT 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 3.1.18 Please clarify the following concerns regarding the information described in Table 3.1 3:
a. The loading due to rotating /reciproc: ting machinery has the potential to affect 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 fixed base supports potentially affected by loading due to hydraulic vibration or waterhammer.
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-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 supports 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 suppons potentially affected by grout' concrete local deterioration.
g.
Provide the basis for the "not plausible" determination for supports potentially affected by lead anchor creep.
This question is too broad. Similar questions were withdrawn by NRC and refocused. BGE requests the NRC either conduct a public meeting or a site visit and discuss details of aging effect plausibility calls with plant personnel or re-phrase question such that it is more focused.
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c. - l REQUEST FOR ADDITIONALINFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 PRIMARY CONTAINMENT STRUCTURE, SECTION 3.3A TURBINE BUILDING STRUCTURE, SECTION 3.3B l 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 3.3B,3.3C,3.3D,3.3E and 6.2 l
3.3.9 Provide the details of specific national codes and standards (e.g., ACI, AISC, etc.) including their editions 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 j licensing basis, describe plans for incorporating these changes in the CCNPP Updated Final Safety l A nalysis Report.
BGE requests the NRC clarify this question. BGE finds it difficult to identify specific 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 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 'he aging management program.
BCE 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 aging was obtained based on Calvert Cliffs Nuclear Power Plant specific information and past experience. Describe the basis upon which Baltimore 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 INFORMATION CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS.1 & 2 i INTEGRATED PLANT ASSESSMENT i
DOCKET NOS. 50-317 AND 50-318 '
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Water Chem _istry Proaram The following questions apply to the secondary water chemistry as discussed in Section 5.12," Main Steam l 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.
Coiarol of the secondary water chemistry plays an imponant 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:
i a) What amine is being used for controlling pH in the secondary water system?
! i 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 l requests the NRC either conduct a site visit and discuss these questions with plant personnel l or re-phrase question such that it is more focused.
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CALVERT CLIFFS NUCLEAR POWER PLANT, UNIT NOS.1 & 2 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" t
1 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.
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.
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 strengths and weaknesses, safety implication of findings, and corrective action plans and schedule for implementation.
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, 18 i
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