ML20204H454
| ML20204H454 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 03/19/1999 |
| From: | Dave Solorio NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9903290075 | |
| Download: ML20204H454 (140) | |
Text
__
March 19,1999 3
t LICENSEE:
Baltimore Gas and Electric Company FACILITY:
- Calvert Cliffs Nuclear Power Plant, Unit Nos.1 and 2
SUBJECT:
SUMMARY
OF FEBRUARY 16 AND 18,1990, MEETINGS WITH BALTIMORE GAS AND ELECTRIC COMPANY (BGE) REGARDING LICENSE RENEWAL ACTIVITIES FOR CALVERT CLIFFS NUCLEAR POvVER PLANT (CCNPP), UNIT NOS.1 AND 2 On Fet.-'ary 16,1996, the Nuclear Regulatory Commission (NRC) staff held a public meeting with representatives of Baltimore Gas and Electric Company (BGE) at the CCNPP Education Center in Lusby, Maryland, to begin a 3-day site visit by the staff in order to discuss some of BGE responses to staff requests for additional inforriation (RAls) as outlined in a February 1,1999, letter to BGE. On February 18,1999, the staff held a public meeting with BGE to summarize the results of the 3-day site visit. to this meeting summary provides a list of meeting attendees for February 16,1999, and February 18,1999. Enclosure 2 provides a summary of the RAls that 4
were discussed by the staff with BGE staff and the information obtained from BGE staff to clarify the RAI responses. Because in some cases during the site visit BGE was not able within the 3 days to provide additional clarification to the staff, BGE subsequently provided their clarification by telephone or by facsimile. Subsequent relevant information provided by telephone was.ncorporated into RAI clarifications listed in Enclosure 2. Subsequent relevant information provided by facsimile was also incorporated into the RAI clarifications listed in, and copies of the facsimiles are attached as Enclosure 3. Additionally, BGE provided CCNPP administrative procedure MN-1-319, " Structure and System Walkdowns," to the staff, which is included as Enclosure 4.
The information obtained from the staff's visit (contained in Enclosure 2) was used by the staff in the preparation of the safety evaluation for BGE's license renewal application for CCNPP Units 1 and 2.
WWN David L. Solorio, Project Manager License Renewal Project Directorate Division of Regulatory improvement Programs Office of Nuclear Reactor Regulation Docket Nos. 50-317 and 50-318
Enclosures:
- 1. List of Attendees
- 2. Summary of RAI Clarifications
- 3. Copies of Facsimiles p& ]3 p
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- 4. Structure and System Walkdown w
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PUBLIC PDLR R/F N. Dudley, ACRS T2E26 DjSTRIBUTION: E-MAIL RZimmerman (RPZ)
WKane (WFK)
DMatthews (DBM)
GNewberry (SFN)
CGrimes (CIG)
FAkstulewicz (FMA)
JStrosnider (JRS2)
RWessman (RHW)
GHolahan (GrAH) i GBagchi (GXB1)
RRothman (RLR)
JBrammer (HLB)
CGratton (CXG1)
JMoore (JEM)
M2obier/RWeisman (MLZ/RMW)
SBajwa/ADromerick (SSB1/AXD)
SBarber (GSB)
JYerokun (JTY)
BBores (RJB)
RArchitzel(REA)
CCraig (CMC 1)
RCorreia (RPC)
RLatta (RML1)
EHackett (EMH1)
AMurphy (AJM1)
DMartin (DAM 3)
FCherny (FCC1)
MModes (MCM)
WMcDowell(WDM)
SStewart (JSS1)
THiltz (TGH)
SDroggitis (SCD)
DSolorio (DLS2)
PDLR Staff Slittle (SLL) 20001G
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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 Commissioners Maryland 175 Main Street Engineering Division Prince Frederick, MD 20678 6 St. Paul Centre i
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
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Baltimore, MD 21203 Suite 2102 Baltimore, MD 21202-1631 Jay E. Silberg, Esquire j
Shaw, Pittman, Potts, and Trowbridge Patricia T. Bimie, Esquire 2300 N Street, NW Co-Director j
Washington, DC 20037 Maryland Safe Energy Coalition P.O. Box 33111 Mr. Bruce S. Montgomery, Director Baltimore, MD 21218 NRM Calvert C!iffs Nuclear Power Plant Mr. Loren F. Donatell 1
1650 Calvert Cliffs Parkway NRC Technical Training Center Lusby, MD 20657-4702 5700 Brainerd Road Chattanooga, TN 37411-4017 Resident lospector U.S. Nuclear Regulatory Commission David Lewis P.O. Box 287 Shaw, Pittman, Potts, and Trowbridge St. Leonard, MD 20685 2300 N Street, NW Washington, DC 20037 Mr. Richard I. McLean Nuclear Programs Douglas J. Walters Power Plant Research Program Nucl Energy Institute Maryland Dept. of Natural Resources 1776 i Street, N.W.
J Tawes State Office Building, B3 Suite 400 Annapolis, MD 21401
~ Washington, DC 20006-3708 DJW@NEl.ORG Regional Administrator, Region I U.S. Nuclear Regulatory Commission Barth W. Doroshuk 475 Allendale Road Baltimore Gas and ' Electric Company King of Prussia, PA 19406 Celvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Mr. Charles H. Cruse, Vice President NEF 1st Floor Nuclear Energy Division Lusby, Maryland 20657 Baltimore Gas and Electric Company 1650 Calvert Cliffs Parkway National Whistleblower Center Lusby, MD 20657-47027 3233 P Street, N.W.
Washington, DC 20007
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NRC & BGE PUBLIC MEETING FEBRUARY 16,1999 NAME ORGANIZATION DAVID SOLORIO NRC/NRR/PDLR CHRIS GRIMES NRC/NRR/PDLR SAM LEE NRC/NRR/PDLR JACK STROSNIDER NRC/NRR/DE TED SULLIVAN NRC/NRR/EMCB LEE BANIC NRC/NRR/EMCB BARRY ELLIOT.
NRC/NRR/EMCB ALLEN HISER NRC/NRR/EMCB JIM DAVIS NRC/NRR/EMCB BARTH DOROSHUK BGE JOHN RYCYNA BGE DON SHAW BGE NRC & BGE PUBLIC idEETING FEBRUARY 18,1999 NAME ORGANIZATION DAVID SOLORIO NRC/NRR/PDLR CHRIS GRIMES NRC/NRR/PDLR SAM LEE NRC/NRR'PDLR P'",HARD WESSMAN NRC/NRR/DE E..,RTH DOROSHUK BGE JOHN RYCYNA BGE DON SHAW BGE I
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Summarv of information Obtained l
by NRC Staff Durina Site Visit ta l
Calvert Cliffs Nuclear Power Plant (CCNPP) from 2/16/99 to 2/18/99 l
The following is a list of information obtained by the NRC staff during a site visit to the CCNPP to obtain clarification related to Baltimore Gas and Electric Company's (BGE's) responses to the staff's requests for additional information (RAls). The information obtained is presented in i
the following format: RAI number, Summary of Clarification Requested, Information Obtained, Documentation Needed, and Resolution.
2.1.1 Summarv of Clarification Reauested BGE did not identify fatigue analysis of containment polar crane (PC) as a time-limited aging analysis (TLAA).
Information Obtained BGE stated that the fuel handling polar crane is not designed for a service life of 40 years. The design is based on number of load cycles at or below the rated load, e.g.,
the number of load cycles below the rated load is 200,000.
Documentation Needed None Resoltdion i
Resolved, as a result of the information provided above by BGE staff.
Since the design analysis is not based on a 40-year life, this is not a TLAA according to 10 CFR 54.3.
2.1.1 Summary of Clarification Reauested Why didn't BGE identify metal corrosion allowance as a TLAA?
Information Obtained BGE conducted search of the electronic docket and the CCNPP Updated Final Safety Analysis Report (Rev 19) using key words indicative of time constraints in accordance with Section 8.1 of their integrated plant assessment methodology (which is provided in Section 2.0 of Appendix A to the BGE license renewal application [LRA) for CCNPP).
BGE reviewed the items pertaining to the design of systems that were identified by the search. BGE stated that the original calculation of wall thickness does not have any allowance for corrosion and a 40-year design life was not a consideration in the 1
4 calculation. Therefore, BGE stated that metal corrosion allowance is not a TLAA as defined in 10 CFR 54.3.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
The staff accepts BGE's interpretation that metal corrosion allowance is not a TLAA.
I 2.1.1 Summary of Clarification Reauested BGE did not identify " inservice flaw growth" as a TLAA.
Information Obtained BGE stated that there was no inservice flaw that was analyzed for the 40-year service life of a component. Therefore, inservice flaw growth is not a TLAA. BGE stated that should there be a need for a flaw evaluation, they will comply with the requirements of ASME Code,Section XI, IWB 3600 for analytical evaluation of a flaw.
Documentation Needed None Resolution Resolved, as a result of the info mation provided above by BGE staff.
2.1.1 Summary of Clarification Reauested BGE identified thermal cycling of containment liner as a TLAA. The staff requested details regarding how the reanalysis of the fatigue analysis of containment liner plate thermal cycling would be accomplished.
Information Obtained BGE stated that they would provide an updated fatigue analysis of containment liner plate thermal cycling to the end of the period or extended operation in the year 2003 (as opposed to the year 2012 in the RAi response). BGE stated that the analysis will be based on 1.5 times the number of cycles for the extended term. The staff verifMxi that BGE has placed this item as an action item for license renewal into a site tracking system.
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Documentation Needed None l
Resolution Resolved, as a result of the information provided above by BGE staff.
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2.1.1 Summarv of Clarification Reauested -
Is fatigue of ASME Class 2 and 3 piping a TLAA7
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Information Obtained Class 2 and 3 piping has a stress limit based on 7000 cycles. BGE performed an electronic search of its design basis documentation and did not identify this as a TLAA.
However, the applicetion already discusses the number of expected cycles for some l
systems. The question remains whether this meets the definition of a TLAA. BGE l
stated they will evaluate further whether RAI response 2.1.1 should be supplemented to 1
include a sentence to say fatigue of Class 2 and 3 piping is a TLAA and is evaluated some place in the application.
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During the ag'ing management review process of the Class 2 and 3 systems, BGE considered fatigue a potential ARDM. For the Class 2 and 3 systems where the LRA identified fatigue as not plausible, BGE considered the following to make the not plausible determination:
- 1. The delta-T between system shutdown temperature and maximum operating temperature. If the delta-T was small (say 50 F), fatigue was not plausible.
- 2. If the delta-T was larger, BGE conservatively estimated the number of tnermal cycles for 60 years using plant operating history. If the number of thermal cycles was less than the design number of cycles, fatigue was not plausible.
Documentation Needed l
BGE should supplement its previous response to this RAI with the above information.
Resolution Resolved, as a result of the information provided above by BGE staff.
2.1.1 Summarv of Clarification Reauested is High Energy Pipe Break (HEPB) a TLAA?
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t Information Obtained HEPB inside containment is based on the Gimbosso letter guidance, which directs the use of stress-based criteria and not criteria based on fatigue cumulative usage factor.
No time dependent analysis is used. There is no specific stress or fatigue criteria for inside the containment.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE.
2.1.1 Summary uf Clarification Reauested is high-cycle fatigues of reactor vessel internals (RVis) a TLAA7 Information Obtained High-cycle fatigue design is based on endurance limits and not based on cycles. There is pre-operational testing that should have revealed high-cycle fatigue failures.
Additionally, it is expected that high-cycle fatigues will occur early in plant operation.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
2.1.1 Summary of Clarification Reauested is reactor coolant pump (RCP) flywheel analysis a TLAA7 Information Obtained Fatigue analysis of RCP flywheel is based on endu '
-a limit which is not dependent on cycles. However, BGE stated that it woti.1 confirrn u.at BGE h0P b not designed for finite pump starts. BGE reviewed RCP documentation and verified no specific limit on pump start.
Documentation Needed None 4
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Resolution Resolved, as a result of the information provided above by BGE staff.
3.1.1 Summary of Clarification Reauested NRC asked for clarification in RAI 3.1.1 to identify the scope of comoonent supports for spent fuel storage system (068). BGE responded spent fuel storage (068) component supports are addressed in the cranes and fuel handling commodity evaluation (Section 3.2 of BGE LRA for CCNPP). NRC did not find these supports discussed in Section 3.2.
ln! ormation Obtained rf There are no component supports for the spent fuel storage system (068).
I Documentation Needed BGE needs to revise or clarify previous RAI 3.1.1 response.
Resolution Resolved, pending submittal of above information by BGE. BGE agreed to revise its response.
3.1.3 Summary of Clarification Reauested Provide an explanation for using the generic implementation plan 2 for scoping component supports Information Obtained BGE stated that it did not exclude any supports from aging management programs based on Site Verification program (SVP) walkdown, and baseline walkdowns do not exclude any supports from further aging management.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
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3.1.6(c)
Summarv of Clarification Recueitstd Why aren't structural steel members identified in the BGE license renewal application for CCNPP?
Information Obtained See RAI 3.1.9 response. BGE's RAI response stated that the heating, ventilation and air conditioning (HVAC) duct support frames made of structural steel members should.
have been included within the scope of supports subject to an aging management review (AMR). The HVAC duct support frames made of structural steel members are now included within the scope of components subject to aging management as discussed on pages 3.1-23 and 3.1-29 of Appendix A to the LRA.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.-
3.1.14 Summarv of Clarification Reauested Please clarify which ring foundations are included within the scope of license renewal in the inservice inspection (ISI) program.
Information Obtained Concrete ring foundations, by themselves, are not in the ISI program. However, System 036 (Auxiliary Feedwater System - Condensate Storage Tank 12) and System 37 (Demineralized Water and Condensate Storage System - Condensate Storage Tanks 11 and 21) have their supports inspected as part of the ISI program. This included a check for concrete damage in the vicinity of the supports.
Note that the Condensate Storage Tank No.12, which is evaluated in System 036, rests on the floor slab of the Condensate Storage Tank No.12 Enclosure, which is evaluation in BGE LRA Section 3.3D " Miscellaneous Tank and Valve Enclosures." It is included in the Component Support Group - Ring Foundations for Flat Bottom Vertical Tanks, for similarity purposes, because of its tank chairs and anchor bolts. Reference page 1 of BGE letter to NRC, dated February 4,1999, on changes to the application for license renewal.
Documentation Needed None 6
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Resolution Resolved, as a result of the information provided above by BGE staff.
3.1.15 (b)
Summary of Clarification Reauested Explain apparent inconsistency in treatment of pipe hangers with respect to dynamic age-related degradation mechanisms.
Information Obtained Please see response to RAI No. 3.1.19.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
I 3,1.18(e)
Summary of Clarification Reauested l
How are the effects of loss of fracture toughness resulting from neutron embrittlement 4
managed for reactor vessel supports?
Information Obtained BGE does not believe reactor vessel supports require analysis or managing management program for the effect of loss of fracture toughness resulting from neutron embrittlement. BGE indicated that, in accordance with NUREG-0933, Revision 3, Generic issue 15, " Radiation Effects on Reactor Vessel Supports," this issue was resolved for the 20-year license renewal period (see page 3.15-8 from NUREG-0933)
NUREG-0933 indicates: " Based on the staff's regulatory analysis, the issue was resolved with no new requirements. Consideration of license renewal period of 20 years did not change this conclusion."
In addition, structural analysis of postulated failure supports indicates that supports and piping have sufficient redundancy to ensure the integrity of the support system.
Based on the above analysis, additional analysis is not warranted.
Documentation Needed None 7
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Resolution Resolved, as a result of the information provided above by BGE staff.
3.1.24 Summarv of Clarification Recuested Have bolts been checked for hardness since bolts with hardness greater than Rc32 are susceptible to stress corrosion cracking (SCC)?
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Information Obtained Testing up to and through Generic Safety issue (GSI) 29, and NRC Generic Letter 91-17, demonstrated hardness within specification on installed and warehouse fasteners, l
and subsequent receipt inspections such that aging effect is not plausible. NUREG-1339 suggestions incorporated into the BG&E program. Rededicated all safety-related bolting in the 1990's. One 4140 or 4340 bolt feed reg. Valve -IGSCC A286 in reactor coolant system - numerous failures. Do have receipt requirements and certification for safety-related fasteners.
q The two materials used in anchor bolt applications at CCNPP meeting this criteria are ASTM A354 and A490. A354 bolting is used in the reactor vessel, pressurizer and safety injection tank anchor bolts and A490 botting is used in the steam generator supports. This topic is discussed in Section 3.1 of the LRA under group 7.
Rocumentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
3.2.1 4
Summarv of Clarification Reauested is the spent fuel cask washdown pit (SFCWP) within the scope of license renewal?
[rl ormation Obtained f
The SFCWP is included in Section 3.3E
- Auxiliary Building and Safety-Related Diesel Generator Building Structures;" however, it is not explicit.
The SFCWP is a feature of the auxiliary building and the AMR of the auxiliary building includes the SFCWP; however, the CWP does not have intended function that would differentiate it from the AMR of the auxiliary building.
The SFCWP does not have any intended functions per 10 CFR 54.4 and therefore is not within scope of license renewal and not subject to an AMR.
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Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
3.2.13(1)
Summary of Clarification Requested Fatigue of polar crane rails
. Information Obtained Holes on polar crane rails are flame cut after fabrication. Holes on other rails are included in fabrication and stresses have been relieved.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
3.2.13(2)
Summary of Clarification Reauested How is wear of moving parts, such as sheath, addressed?
Information Obtained Moving parts are active and are excluded from AMR. However, to be conservative, BGE subjected the ropes to an AMR. The active parts are subject to routine maintenance and the staff observed that there are procedures for conducting the maintenance.
Qocumentation Needed information on page 3.2-5 of Appendix A to LRA is sufficient.
Resolution Resolved, as a result of the information provided above by BGE staff.
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3.3.2 Summary of Clarification Reauested What is the basis for concluding that any freeze thaw damage on the concrete dome is i
j not expected to result in any loss of function?
Information Obtained The possible freeze thaw damage on the containment dome is expected to be insignificant. This will be evaluated further as part of the baseline inspection for containment by 2002.
If corrective actions are needed they will be implemented as part of the inspection.
Documentation Needed None Resolution Resolved, as a result of commitment made to inspect in RAI 3.3.2 response.
i 3.3.3 Summarv of Clarification Reauested What is the basis for BGE concluding that it is unaware of trends associated with the incidence of coating degradation or the corrosion of steel?
Information Obtained Based on discussion with Kevin Anstee, coating specialist in the civil engineering unit at CCNPP, the freque.Ly of coating degradation or failure or both and corrosion of steel (if it even progresses to that point) has been fairly constant with no increasing or decreasing occurrences.
And based on this discussion, CCNPP staff concluded that they are unaware of any paiticular trend.
For example, the inspections conducted inside containment each refueling outage (conducted in accordance with site procedure MN-3-100, painting and other protective coatings) generally identify 20 to 35 areas which require coating restoration. This number has been fairly constant over the last several years.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
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3.3.7 Summary of Clarification Reauested What are the key elements of the caulking and sealants management program and what are the inspection frequencies?
i in longer than 6 years inspectii
- .tervals, provide justification.
Information Obtained Inspection program will be equivalent or similar to the existing fire bureau inspections in content, general approaching and detection of suggestion criteria. The inspection interval will be 6 years or less.
Documentation Needed None l
Resolution Resolved, as a result of the information provided above by BGE staff.
3.3.9 Summant of Clarification Reauested If changes are made to the current codes which BGE is committed, what version of the code will BGE make a repair to?
Information Obtained BGE will, at a minimum, perform repairs in accordance with current licensing basis code versions.
Additionally, please see RAI response No. 3.3.9 (by letter dated 12/10/98), which l
explains the process BGE would use should no codes or standards exist for making repairs.
Documentation Needed None l
l Resolution Resolved, as a result of the information proyided above by BGE staff.
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'l 3.3.12 Summary of Clarification Reauested I
is a straight line extrapolation to 60 years drawn on semi-log paper valid for the period of extended operation? If so, why?
Information Obtained Relaxation creep shrinkage are all considered to be time dependant losses following the general form of exponent to the negative "kt" (e*) curve. Therefore, a straight line extension drawn on semi-log paper is appropriate. BGE letter to the NRC dated -
October 28,1997, Appendix B, Summary of Vertical Tendon eft off forces indicated that the average lift off force is on the order of 675 thousand pounds per square inch (kips),
which is above the expected value (about 650 kips) given in the UFSAR figure 15.6.1.2.
r,4 required by 10 CFR 50.55a (b)(vi), BGE plans to revise the tendon surveillance program to conform to the 1992 Edition with the 1992 Addenda of the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWL. The IWL-2520 gives the examination requirements of tendon selection, force and elongation measurement, wire sampling, etc.
10 CFR 50.55a (b)(ix) requires an evaluation of prestressing force trends for each tendon and groups of tendons to ensure that the predicted tendon forces at the next scheduled examination meet the minimum design prestress requirements. If these requirements are not met, an evaluation is required in accordance with the Engineering Evaluation Report as prescribed in IWL-3300. The IWL-3300 evaluation requires determination of the cause of the condition, acceptability without repair, whether repair / replacement is required, if repair / replacement is required the extent, method and required completion date.
The normalized average lift off forces (kips) for hoop and dome tendons measured during CCNPP's 20th Year Surveillance of the Unit 1 Containment Building-Post-Tensioning System were 621 and 632, respectively.
General information about CCNPP's containment design and its tendon surveillance program is provided below:
In accordance with ACI Code 318-63, the containment design provides for prestress losses that can be predicted with sufficient accuracy as described in Calvert Cliffs UFSAR Section 5.1.4.2. The environment of the prestress system and concrete is not appreciable different from that found in numerous bridge and building applications.
Considerable research has been done to evaluate the causes of prestress losses and this information was used to assign appropriate allowances. Puilding code authorities consider it acceptable practice to develop permanent designs based on these allowances. The structural integrity of the containment shall be maintained at a level consistent with the acceptance practice to develop permanent designs based on these allowances. The structuralintegrity of the containment shall be maintained at a level consistent with the acceptance criteria in the verification requirements identified in
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UFSAR Section 15.61. Curves describing the predicted prestress loss behavior are J
contained in UFSAR Figures 15.61.1,15.6.1-2 and 15.6.1-3. These curves will be
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extrapolated out to 60 years for the extended license period using this current licensing i
basis.
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A licensing basis surveillance change will result from the new rule recently listed in 10 CFR 50.55(a), incorporating ASME Section XI, Subsection lWE/lWL requirements.
This is an acceptable inservice inspection and surveillance method for ungrouted tendons in prestressed concrete containment structures as previously acknowledged by the staff, it currently provided adequate aging management of the tendons and will continue to provide adequate aging management into the extended license period.
Documentation Needed None Recolution Open.
3.3.13 Summarv of Clarification Reauested How was the visual examination of the full transfer tube from the refueling canal performed?
Information Obtained When the refueling pool was drained, the examination of the fuel transfer tube was performed in 1995. The inspection verified that there were no indications of damage or Corrosion.
4 BGE concludes that the leakage effects are small enough to state that any reduction of slab strength is negligible.
BGE is not aware of any leakage from the fellows.
BGE staff also pointed out NRC letter to BGE dated June 30,1998, as an additional source of information.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
3.3.16 Summary of Clarification Reauested Elaborate on degradation of hoop rand dome tendons, and how aging for these components will be managed.
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s Information Obtained -
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BGE did a sampling of the hoop and dome tendons, in the same inspection that i
identified vertical tendon degradation.
The vertical tendon rust is only at the top one foot of the tendons. Conditions are different for the anchorages of the hoop and dome tendons.
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BGE staff provided October 28,1998, letter to the NRC " Containment Tendon Engineering Report" to provide NRC with information to support the above conclusions.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
3.3.17 Summary of Clarification Reauested How does BGE account for potential tendon force loss (8 to 14%) due to elevated temperature?
Information Obtained BGE is not committed to NUREG-1611. The effects of elevated temperature on tendon force due to abnormal sun exposure or proximity to hot penetrations will be detected by STP M-663. Operability of the containment, as it relates to tendon force, is govemed by the requirements of Calvert Cliffs Technical Specification 15.6.1. Testing is performed by STP M-663. A reduction in tendon force will be detected during conduct of the test regardless of the cause. If the Technical Specification requirements are satisfed, then by definition the containment is operable.
Documentation Needed Demonstrate that STP-M-663-1/2 bounds this additional effect.
Resolution Resolved, as a result of the information provided above by BGE staff.
3.3.19 Summary of Clarification Reauested j
Under what conditions and circumstances would BGE commit to take actions to inspect potential degradation in inaccessible areas of the containment structure?
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l information Obtained if conditions adverse to quality are identified in accessible areas adjacent to inaccessible areas, BGE wuuld take appropriate measures to identify the scope and extent of the adverse conditions.
BGE's corrective actions would include dispositioning the condition adverse to quality in the inaccessible area as well.
Please refer to LRA Section 5.7, page 3 for an example of a situation where portions of four buried pipe lines were inspected.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
3.3.23 Summary of Clarification Reauested MN-1-319, Structure and System Walkdowns, had not been formally approved as a site procedure when BGE staff provided a presentation on MN-1-319 to the NRC staff on June 26,1998. The NRC staff did not know if BGE had ever formally approved the procedure.
Information Obtained MN-1-319 has since been formally approved as a site procedure.
Documentation Needed l
BGE staff provided a copy of MN-1-319, Revision 2, to the NRC staff.
Resolution Resolved, as a result of the information provided above by BGE staff.
3.3.24 Summarv of Clarification Recuested How does BGE manage potential detraction of the horizontalliner in the containment due to potential water intrusion through concrete basemat cracks?
What is the condition of the 18" concrete basemat?
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information Obtained The containment system engineer performs a waik-down and visualinspection of the containments, including the floors, every outage when they are open. No cracks in the 18-inch thick, steel reinforced, high density concrete floors have ever been discovered that were considered large enough to affect the liner.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
3.3.33 Summarv of Clarification Reauested l
How does CCNPP ensure the operability of the service water (SW) system from an injection temperature standpoint, assuming the failure of the baffle wall and the intake channel?
Information Obtained 01 (Operating instruction) - 29 contains operating temperature limitations for the salt water / service water heat exchangers. Thus, the contribution of the baffle wall intake channel in getting cooler bay temperature water is bounded by these temperatures.
The baffle wall serves to draw water from the bottom strata of the bay for the purpose of minimizing ecologicalimpact.
The NRC staff reviewed Ol -29 and verified that there were controls in placed as outlined in the Of to ensure the operability of the SW system Documentation Needed None Resolution Resolved, based on review of 01 -29 and the information provided above by BGE staff.
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3.3.36
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Summarv of Clarification Reauested
{
Provide bases information conceming RAI 3.3.2 in which BGE stated that there is no conclusive evidence that confirms that ground water chemistry has become aggressive since plant construction.
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' This relates to the expansion joint in the intake structure which occasionally leaks ground water.
Information Obtained BGE provided water sample well sample results from three wells near the intake structure. One of the samples showed abnormally high concentrations of sulfates and
' chlorides. BGE staff stated that they believe these high concentrations are representative of a very local condition and not representative of the groundwater at the site. Specifically, BGE staff stated they believed that the likely explanation for the high reading was from a small amount of circulating water that is occasionally backflowing from the discharge conduit into the subsurface drainage system piping and entering the groundwater.
Documentation Needed None.
Resolution l
Open because BGE staff has not provided conclusive evidence that groundwater is not subjecting the intake structure to chemical attack. The staff believes that an inspection should be performed prior to the renewal period to confirm the above hypothesis.
i 3.3.43 Summary of Clarification Reauested For Section 3.38 " Turbine Building" roof trusses are included in the description of the l
structural components but not listed on the list of components subject to AMR. Are the trusses subject to AMR?
Information Obtained No. As described in Section 3.3B of the Appendix A to the LRA, only components associated with the safety-related auxiliary feedwater (AFW) pump rooms and the electrical duct banks for the salt water pumps are within the scope of license renewal.
[NRC needs to make a determination regarding credible design failures with respect to the design of this building.)
Documentation Needed None; information needed to scope this component is contained in Section 3.3B of the BGE LRA.
Resolution Open. The NRC staff needs to make a determination regarding credible design failures with respect to the design of this building.
17 l
l
l 4.1.1 Summarv of Clarification Reauested Why is the pressurizer spray nozzle not included in the scope of license renewal?
Information Obtained
- The pressurizer spray nozzle has no safety-related function and, therefore, does not 1
meet the scoping requirement of 10CFR54.4. It is not credited for the mitigation of any accidents addressed in Calvert Cliffs Nuclear Power Plant Updated Final Safety Analysis l
Report Chapter 14, accident analysas. Therefore, the pressurizer nozzle was not i
included in the written scope of license renewal.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
4.1.7 Summary of Clarification Reauested Discuss several aging effects associated with certain RCS components (reference RAI for complete list) that were not described in the LRA.
i information Obtained BGE did not provide clarification.
Documentation Needed Not applicable i
Resolution Open.
4.1.8 Summarv of Clarification Reauested Explain why its heater bundles are not within the scope of license renewal and not subject to an AMR.
1 1
Information Obtained Resolved this question by looking at drawings on site.
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s s
Documentation Needed None.
Resolution Resolved, as a result of the drawings provided above by BGE staff.
4.1.9 Summant of Clarification Reauested l
Describe in summary form, the extent to which BGE relies on certain programs for aging l
management and provide examples of any operating experience that show the effectiveness of the programs (the staff provided a listing of the components, aging effects, and programs that were not described in the LRA in the RAl).
Information Obtained BGE did not provide clarification.
Documentation Needed Not applicable Epsolution Open.
4.1.11 Summarv of Clarification Reauested What does BGE mean by " damage" in RAI response and is erosion / corrosion plausible for secondary steam generator (SG) manway?
Information Obtained BGE staff stated that SG secondary manway plates are removed every outage and any l
leaks from manway that might be caused by erosion / corrosion would be repaired in 3
accordance with the site Appendix B progrem for corrective action.
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Additionally, BGE stated that erosion / corrosion is not plausible per their definition of j
plausibility in Section 2.0 of the BGE LRA.
j Documentation Needed None i
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i 3
D Q
l Resolution i
f The staff considers erosion / corrosion plausible and will be managed by the SG-20 procedure.
4.1.15 i
Summary of Clarification Reauested j
l BGE response to RAI states that seal leak "off-line" will be blown with air to remove any debrief during outages; however, the staff does not believe that doing so will verify the j
integrity of the " leak-off" line.
Information Obtained None Documentation Needed Not applicable Resolutie' Open 4.1.17 Summarv of Clarification Reauested How do weld examinations of the nozzles capture erosion / corrosion?
Information Obtained Erosion / corrosion of main steam outlet nozzles is captured by regular ISI inspection of main steam outlet nozzle welds because the ultrasonic testing of the welds also requires ultrasonic testing of the surrounding area.
The ISI inspection also looks at the base metal adjacent to the welds and the nozzle inner radii, and this will detect the occurrence of erosion / corrosion.
Documentation Needed None.
Resolution Resolved, as a result of the information provided above by BGE staff.
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-: 13.
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l 4.1.18 Summarv of Clarification Reauested Why isn't ARDI used for RCS?
l l
Information Obtained ARDI is for those parts of the systems (CVCS, SI) without H overpressure.
2 Documentation Needed None. The above information along with slides on CVCS and SI from February 10, 1999, public meeting is sufficient information for reviewer to answer the reviewer's
. question and allow reviewer to write SE conclusion for RCS.
l Resolution Resolved, as a result of the information provided above by BGE staff.
4.2.5 1
l Summarv of Clarification Reauested BGE does not believe radiation embrittlement as measured by the drop in Charpy upper shelf energy (USE) is a TLAA because it does not satisfy the TLAA definition in 10 CFR 54.3., The NRC staff believes USE is a TLAA in accordance with 10 CFR 54.3.
Information Obtained l
BGE and NRC agreed that revisions to Charpy USE at expiration of the renewal process may be performed using current BGE process since BGE must satisfy the requirements in Appendices G and H,10 CFR Part 50.
Documentation Needed None Resolution l
Resolved, as a result of the information provided above by BGE staff.
4.2.8 Summarv of Clarification Reauested l
Reactor vessel (RV) surveillance capsules 21
Information Obtained BGE identified the projected peak neutran fluence at inside surface of the RV at 60 years (end of license renewal period) as 4.95x10 n/cm for CCNPP Unit 1 and 2
5.77x10 n/cm for CCNPP Unit 2.
2 BGE indicated that the current surveillance program consists of a plant-specific program
)
in accordance with ASTM E185, supplementary capsules, and capsules withdrawn from McGuire. The plant-specific program consists of six capsules in each unit. In the current plant-specific program two capsules have been tested, three are to be tested, and one capsule is standby. The peak fluence to be received by a capsule is 4.31 x10" r
2 n/cm for CCNPP Unit 1 and 3.88 x10 n/cm for CCNPP Unit 2. BGE indicated that the surveillance capsule withdrawal schedule will be revised in 2003.
BGE will revise the surveillance capsule withdrawal scheduled from 40 to 60 years (48 EFPY). The new schedule will include the withdrawal of at least one capsule from each unit that will provide data at a neutron fluence equal to or greater than the projected peak neutron fluence at 60 years (48 EFPY) lf BGE withdraws the last capsule from either reactor pressure vessel (RPV) prior to year 55, BGE will establish the neutron irradiation environment (fluence, spectrum, j
temperature and flux) applicable to the surveillance data and pressure-temperature j
limits for the affected RPV. If the RPV(s) operates outside these limits, BGE will inferm the NRC and determine the impact of the condition on the RPV(s)integr ty.
1 If BGE withdraws the last capsule from either RPV prior to year 55, BGE will provide additional dosimetry for the affected RPV.
i Documentation Needed Commitment regarding RV capsule surveillance and reporting to NRC.
Resolution
. Resolved, as a result of the information provided above by BGE staff.
4.2.20 Summary of Clarification Reauested How does BGE decide fatigue analysis proposed is bounding?
Information Obtained When Combustion Engineering (CE) reviewed and developed critical monitoring locations for the fatigue monitoring program for BGE, CE did not include certain systems within these study scopes, such as control element drive mechanism, reactor coolant
. pump, and motor-operated valve because they were supplied by different firms. To ensure that the CE study is bounding, BGE will review the components not included in the CE study. The staff requests BGE to provide a summary description, similar as above, in writing.
22
The NRC staff requested a summary description from BGE to elaborate on the above information. Subsequently, BGE provided the following information:
CE determined the bounding components and locations and the controlling transients that BGE used to develop the Fatigue Monitoring Program (FMP). The process CE used is as follows:
1.
CE identified those systems that the UFSAR and TS include for maintaining the safe operations and achieving safe shutdown of the plant. The result of this review was the " identification of critical systems."
2.
For each of the critical systems CE reviewed all components within the system to identify those components with controlling fatigue usage limits. The critical system review included a review of industry data and experience as well as the original design documents associated with each system and/or component. The result of this review was the " Identification of Component Critical Locations."
3.
CE then performed a fatigue evaluation for each component identified as a criticallocation. This evaluation included an evaluation of the original stress reports to identify the controlling design basis transients with respect to fatigue usage. The results are " Fatigue Evaluation of Critical Locations."
4.
CE then develop a logging and procedure guideline for tabulating cumulative fatigue usage for BGE to use in the development of the FMP.
BGE used the results of the CE work to develop and implement the FMP.
The following systems or components were not included within the CE scope of work:
NSSS Sampling Sy',tein Pressurizer safety valves Power operated relief valves AFW isolation and check valves MFW check valves MFW isolation valves 1
Reactor coolant pumps j
Additionally, CE did not include potential thermal stratification loadings identified in NRC Bulletins 88-08 and 88-11. However, BGE has incorporated components that do' experience thermal stratification loading in the FMP. BGE committed in the LRA to perform an evaluation of the SI system with respect to thermal stratification loadings.
Documentation Needed BGE should provide written submittal containing this information.
Resolution
- Resolved, as a result of the information provided above by BGE staff.
23
4.2.23 and 7.11 Summary of Clarification Reauested
'An evaluation will be performed in the future to determine if additional locations need to I
be added to the fatigue monitoring program.
Information Obtained The staff is comfortable with the methodology. The staff believes it needs the i
evaluation to be completed to determine whether additional locations need to be added to the fatigue monitoring program.
Documentation Neaded To be determined.
Resolution Open - This is a " policy" question on how aging management is demonstrated by a future evaluation.
New item (4.3)
Summarv of Clarification Reauested Low-cycle fatigue analysis of RVis will be performed information Obtained See response to RAI 4.2.23 and 7.11 Documentation Needed To ba determined Resolution Open - This is a " policy" question on how aging management is demonstrated by a future evaluation.-
4.3.7 Summary of Clarification Reauested BGE described the RV internal components by clarifying the FSAR figures and using the information contained in EPRI Report No.103838 Information Obtained Sufficient information was provided to explain the location and configuration of the RV intemal components.
24
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i Documentation N?eded None Resolution 3
Resolved, as a result of the information provided cbove by BGE staff.
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4.3.9 Summary of Clarification Reauested Licensee will provide reasons for non-plausibility aging effects for compone.nts and aging effects listed an NRC RAI 4.3.9 for RV intemais.
Information Obtained BGE showed staff intemal documentation that provided the basis for considering some aging effects of RV intemal component 3 nonplausible. BGE will summarize documentation to identify the reason the aging effect is not plausible and references which document the bases for the conclusion. BGE indicated the information will be provided by February 18,1999.
The staff informed the applicant tilat the response should include the rmal embrittlement CEA shrouds. This was left off of thic original RAl.
Documentate: Neded Summary documentation on RV internals nonplausible aging (or if plausible description of AMR). BGE provided summary documentation as attached.
BGE subsequently provided the following non-plausibility determinations which the staff found to be acceptable:
ARDM: S,QL Components: Lower Support Structure Beam Assembly (ASTM A240 Chromium and
- Chromium -Nickel Stainless Steel), Core Support Barrel Upper Flange (ASTM A182 Alloy Steel), Core Support Barrel (ASTM A182 Alloy Steel, ASTM A240 Chromium and Chromium -Nickel Stainless Steel), Core Support Plate (ASTM A240 Chromium and Chromium -Nickel Stainless Steel), Fuel Alignment Pin (ASTM A276 Stainless Stect),
Core Support Columns (ASTM A351 Austenitic Stainless Steel), Core Shroud Tie Rods and Bolts (ASTM A479 Stainless Steel, ASTM A193 Alloy and Stainless Steel, ASTM A194 Carbon and Alloy Steel), Upper Guide Structure Support Plate (ASTM A240 Chromium and Chromium -Nickel Stainless Steel, ASTM A276 Stainless Steel), CEA Shrouds (ASTM A479 Stainless Steel, ASTM A240 Chromium and Chromium -Nickel Stainless Steel, ASTM A276 Stainless Steel, ASTM A269 Austenitic Stainless Steel, ASTM A451 Austenitic Stainless Steel), Fuel Alignment Plate (ASTM A240 Chromium and Chromium -Nickel Stainless Steel), Fuel Alignment Plate Guide Lugs (ASTM A276 Stainless Steet/ Stellite Faced),
25
Rationale: The ARDM is not plausible for these device types due to non-susceptible material (either Alloy Steel or Nickel Based Stainless Steel), lack of high tensile stresses and a benign operating environment.
References:
NUMARC 90-05 (also known as EPRI TR-103838, PWR RPV Intemals LR Industry Report)
ARDM: lASCC Components: Lower Support Structure Beam Assembly (ASTM A240 Chromium and Chromium -Nickel Stainless Steel), Core Support Barrel Upper Flange (ASTM A182 Alloy Steel), Core Support Barrel (ASTM A182 Alloy Steel, ASTM A240 Chromium and Chromium -Nickel Stainless Steel), Core Support Plate (ASTM A240 Chromium and Chromium -Nickel Stainless Steel), Fuel Alignment Pin (ASTM A276 Stainless Steel),
Core Support Columns (ASTM A351 Austenitic Stainless Steel), Core Shroud Tie Rods and Bolts (ASTM A479 Stainless Steel, ASTM A193 Alloy and Stainless Steel, ASTM A194 Carbon and Alloy Steel), Upper Guide Structure Support Plate (ASTM A240 Chromium and Chromium -Nickel Stainless Steel, ASTM A276 Stainless Steel), CEA Shrouds (ASTM A479 Stainless Steel, ASTM A240 Chromium and Chromium -Nickel Stainless Steel, ASTM A276 Stainless Steel, ASTM A269 Austenitic Stainless Steel, ASTM A451 Austenitic Stainless Steel), CEA Shroud Bolts (ASTM A182 Alloy Steel, ASTM A286 Iron Base Super Alloy AMS 5735), Fuel Alignment Plate (ASTM A240 i
Chromium and Chromium -Nickel Stainless Steel), Fuel Alignment Plate Guide Lugs (ASTM A276 Stainless Steel / Stellite Faced),
j Rationale: The ARDM is not plausible because of the low levels of dissolved oxygen in a 1)WR environment and the low applied strain of the RV intemals components. This mechanism has been observed in BWRs where oxygen levels are considerably higher than in PWRs. A similar ARDM has also been observed ir; PWR CEDM tips where very high strain is applied at very low strain rate in a high fluerv.t field. However, lASCC has not been observed for device types with the temperature, - ygen and radiation levels present for the RV intemals either in operating plants or laNratory tasts.
References:
EPRI NP-5775 (Environmental Effects on Components: Commentary for ASME Section 111), NUREG CR-6048 (PWR Reactor Intemals Aging Degradation Study)
ARDM: Corrosion Components: Lower Support Structure Beam Assembly (ASTM A240 Chromium and Chromium -Nickel Stainless Steel), Core Support Barrel Upper Flange (ASTM A182 Alloy Steel), Core Support Barrel (ASTM A182 Alloy Steel, ASTM A240 Chromium and Chromium -Nickel Stainless Steel), Core Support Plate (ASTM A240 Chromium and Chromium -Nickel Stainless Steel), Fuel Alignment Pin (ASTM A276 Stainless Steel),
Core Support Columns (ASTM A351 Austenitic Stainless Steel), Core Shroud Tie Rods and Bolts (ASYM A479 Stainless Steel, ASTM A193 Alloy and Stainless Steel, ASTM A194 Carbon and Alloy Steel), Upper Guide Structure Support Plate (ASTM A240 Chromium and Chromium -Nickel Stainless Steel, ASTM A276 Stainless Steel), CEA Shrouds (ASTM A479 Stainless Steel, ASTM A240 Chromium and Chromium -Nickel Stainless Steel, ASTM A276 Stainless Steel, ASTM A269 Austenitic Stainless Steel, ASTM A451 Austenitic Stainless Steel), CEA Shroud Bolts (ASTM A182 Alloy Steel, ASTM A286 Iron Base Super Alloy AMS 5735), Fuel Alignment Plate (ASTM A240 Chromium and Chromium -Nickel Stainless Steel), Fuel Alignment Plate Guide Lugs (ASTM A276 Stainless Steel / Stellite Faced),
26 i
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Rationale: The ARDM is not plausible due to resistant materials and a benign operating environment. Austenitic Stainless Steel, Alloy Steel and Nickel-based alloys are quite resistant to general corrosion in a benign operating environment.
1
References:
NUMARC 90-05 (also known as EPRI TR-103838, PWR RPV intemals LR Industry Report), NUREG CR-6048 (PWR Reactor Intemals Aging Degradation Study).
ARDM: Neutron Embrttlement Components: Core Support Barrel Upper Flange (ASTM A182 Alloy Steel)
Rationale: The ARDM is not plausible because fluence level in the area of the CSB flange are not sufficient to cause appreciable neutron embrittlement. Neutron embrittlement is not considered plausible for regions above the vessel nozzles. The CSB flange is made of similar material and is located much further away from the core than the vessel nozzles.
References:
NUMARC 90-05 (also known as EPRI TR-103838, PWR RPV Intemals LR Industry Report)
ARDM: Stress Relaxation Components: Fuel Alignment Pin (ASTM A276 Stainless Steel),
Rationale: The ARDM is not plausible because stress relaxation could not affect the component's intended function. Because the fuel alignment pins are pre-shrunk to fit into the core support plate, some industry documentation considers that stress relaxation could affect this component's function. BGE disagrees with these references.
Unlike a bolt or other threaded faste: er, the preload stresses on an insert pin are compressive and thus would not result in plastic deformation of the pin. Further, if any plastic deformation of the pin did occur, subsequent stress levels on the pin would decrease substantially, resulting in insufficient stress for further relaxation. Therefore, it is not possible for stress relaxation to progress to the point where this ARDM could result in loosened, resultant vibrations and early fatigue of the pins.
References:
NUMARC 90-05 (also known as EPRI TR-103838, PWR RPV Intemals LR Industry Report), EPRI NP-5775 (Environmental Effects on Components: Commentary for ASME Section Ill), DG-1009 (Draft Reg Guide, Standard Format and Content of Technical Information for Applications to Renew Nuclear Power Plant Operating Licenses)
ARDM: Stress Relaxation Components: Core Shroud Assembly Bolts (Core Shroud Tie Rods and Bolts) (ASTM A479 Stainless Steel, ASTM A193 Alloy and Stainless Steel, ASTM A194 Carbon and Alloy Steel)
Rationale: This ARDM is considered plausible (See response to question 4.3.9 and pages 4.3-21 through 24 of the LRA).
27
ARDM: Wear Components: CEA Shrouds (ASTM A479 Stainless Steel, ASTM A240 Chromium and Chromium -Nickel Stainless Steel, ASTM A276 Stainless Steel, ASTM A269 Austenitic Stainless Steel,' ASTM A451 Austenitic Stainless Steel),
Rationale: The ARDM is not plausible for this device type because there is no potential for relative motion between adjacent surfaces.
References:
NUMARC 90-05 (also known as EPRI TR-103838, PWR RPV intemals LR Industry Report), NUREG CR-6048 (PWR Reacto r Intemals Aging Degradation Study).
Components: Fuel Alignment Plate (ASTM A240 Chromium and Chromium -Nickel i
Stainless Steel)
Rationale: This ARDM is considered plausible. Refer to the response to question 4.3.9 and pages 4.3.10 through 4.3-13 of the LRA.)
ARDM: Thermal Embrittlement Components: Fuel Alignment Plate (ASTM A240 Chromium and Chromium -Nickel Stainless Steel)
Rationale: This e.T " :s not plausible because the component is not fabricated of Cast Austenitic Stainless Steel. The operating temperature of the RV intemals system is not
}
sufficient to cause thermal aging for the component material.
j
References:
NUMARC 90-05 (also known K EPRI TR-103838, PWR RPV Intemals LR Industry Report), EPRI NP-5775 (Environmental Effects on Components: Commentary for ASME Section Ill), NUREG CR-6048 (PWR Reactor intemals Aging Degradation Study).
ARDM: Thermal Embrittlement Components: CEA Shroud Tube (ASTM A451 Cast Austenitic Stainless Steel)
Rationale: This ARDM is considered plausible for this subcomponent of the CEA
(
Shroud. See the response to question 4.3.9 and pages 4.3-19 through 21 of the LRA.
Components: CEA Shroud (All other subcomponents) (ASTM A182 Alloy Steel, ASTM A286 Iron Base Super Alloy AMS 5735, ASTM A479 Stainless Steel, ASTM A240 l
Chromium and Chromium Nickel Steel, ASTM A276 Stainless Steel, ASTM A269 Austenitic Stainless Steel)
Rationale: This ARDM is not plausible because the component is not fabricated of Cast Austenitic Stainless Steel. The operating temperature of the RV intemals system is not sufficient to cause thermal aging for the component material.
{
i
References:
NUMARC 90-05 (also known as EPRI TR-103838, PWR RPV Intemals LR Industry Report), EPRI NP-5775 (Environmental Effects on Components: Commentary for ASME Section lii), NUREG CR-6048 (PWR Reactor intemals Aging Degradation Study).
28
Resolution j
Resolved, as a result of the information provided above by BGE staff.
I 4.3.11 and 4.3.18 Summary of Clarification Reauested BGE will evaluate the susceptibility to lASCC and neutron embrittlement of RV intemal components by performing enhanced VT-1 examination of limiting component or by developing data through industry research.
1 Information Obtained BGE is working to develop data through industry research to determine the susceptibility of RV intemals components to irradiation-assisted stress corrosion cracking and neutron i
embrittlement [ awaiting fax from BGE to confirm the acceptability of adding these words). Until the data and analyses that indicate IASCC is not a potentially relevant degradation mechanism and neutron embrittlement is not a concem, [ awaiting fax from BGE to confirm the acceptability of adding these words] BGE will perform enhanced VT-1 inspections to detect cracks (if any occur) in the components believed to be potentially most susceptible to IASCC as well as neutron embrittlement. The inspections will be performed as part of the 10 year ISI inspection program during the license renewal term. Plant-specific justification will be provided to the NRC in the event the analyses and data support elimination of the inspection.
The items selected for enhanced VT-1 inspection are the re-entrant comers of the core barrel inside surfaces (the core barrel surface that faces the core). These comers are constructed by welding annealed 304 stainless steel plate. The residual stresses due to welding, while limited to the low yield strength of the annealed plate, are potentially higher than at any other stainless steel location on the inside surface of the core barrel.
In addition to potentially being the highest fluence, this qualitatively determines the re-entrant comers project in towards the core, between two adjacent fuel bundles, so receive neutron exposure from 270*. Being closer to the fuel than any other stainless steel components, the core barrel plates and comers are exposed to hot leg temperatures on one side. On the other side the environment is cold leg temperature.
Due to the proximity of fuel, gamma heating is also expected to be higher at these locations than at other potentiallocations. Due to the combination of highest stress, fluence, and temperature, the re-entrant comers of the core barrel, intended for enhanced inspection, are the most likely location for IASCC to occur, if it occurs at all and for embrittlement to be manifested, if it occurs at all.
Documentation Needet None Resolution Resolved, as a result of the information provided above by BGE staff.
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4.3.14 Summarv of Clarification Reauested Regarding Screening Criteria used for inspection of cast stainless steel piping, valves, and reactor vessel internal components.
Information Obtained BGE agreed to the following revisions to their screening process for cast stainless steet:
(a)
Statically cast components with a molybdenum content exceeding CF3 and CF8 limits and a delta ferrite content exceeding 10 percent will be subject to ISI in accordance with ASME Code Section XI.
(b)
Ferrite levels will be calculated using Hull's Equivalent Factors or a method producing an equivalent level of accuracy (* 6% deviation between measured and calculated values).
(c)
Cast Stainless Steel components containing niobium will be subject to ISI.
BGE does not agree to the following additional requirements proposed by the staff.
(a)
ISI procedures and techniques must be qualified using methods consistent with those of Appendix Vill of Section XI of the ASME Code. This additional requirement pre-supposes that an ultrasonic exam will be required.
It is possible that attemative examination techniques would be sufficient. The proposed qualification requirements are currently impossible to achieve. BGE would commit to using the ultrasonic examination qualified to Appendix Vill providing it is eventually possible to qualify any UT technique for CASS.
(b)
Flaws in cast stainless steel components with ferrite levels exceeding 25 percent or with niobium will not be evaluated using ASME Code IWB-3640 procedures.
In these instances, fracture toughness data will be prp/ided on a case-by-case basis.
Instead, flaws could be evaluated using IWB-3640 but specific fracture toughness information would be required.
The above discussion applies only to piping, valves, and other affected components in general. It does not apply to the RVis since the EPRI screening criteria do not consider the effect of neutron radiation as applicable to high fluence components. For these components, mechanical loading during ASME Code A, B, C, and D conditions must be low enough or compressive to ensure that the affected component will not fracture.
1 For the Reactor Vessel Intemals. BGE will (a)
Determine whether the components are susceptible to a loss of fracture i
toughness by evaluating whether the components receive a fluence above the threshold of 10"n/cm. If the fluence is below the threshold, the screening i
2 criteria described above for pipes and valves in general will be applied by performing a flaw tolerance evaluation specific to reactor vessel intemal. If the fluence is above the if ceshold, BGE will...
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(b)
Evaluate the mechanical loading to determine if it will remain low enough or compressive to ensure that the affected component will not fracture during ASME Code A, B, C, and D conditions, if this cannot be determined, BGE will...
(c)
Perform enhanced VT-1 inspection of the components.
Documentation Needed None Resoluting Resolved, as a result of the information provided above by BGE staff.
4.3.15 Summary of Clarification Reauested Section 4.3.2 of the LRA indicates that a stress analysis will be performed to specifically evaluate the potential for SCC of control element assembly (CEA) should bolts. BGE could not identify irradiated data to justify the criteria to be used to evaluate the potential for SCC of CEA should bolts.
Information Obtained BGE indicated that they would not be performing the stress analysis because the CEA should bolts are not within the scope of license renewal. They are not within the scope of license renewal because they do not perform a function in accordance with 10 CFR 54.4.
Documentation Needed i
BGE must describe function of CEA should bolts and explain why they do not meet the criteria in 10 CFR 54.4.
Resolution Open.
4.3.16
$_ummarv of Clarification Reaueste_p BGE clarified their response to the RAl.
Information Obtained BGE indicated that they had not had any long outages without chemistry control such that pitting crevice corrosion could occur in RV intemais.
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Qgppmentation Needed None
^
Resolution Resolved based on the applicant response that pitting / crevice corrosion is not plausible for RV intemal components.
4.3.17 Summary of Clarification Reauested Why aren't Alloy 600 or nickel base alloy components in RV internals within the scope of license renewal?
Information Obtained BGE indicates that all Alloy 600 or nickel bases alloy components in RV intemals are subject to replacement on a qualified life or a specified time period.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
4.3.18 Summary of Clarification Reauested j
Resolved by response to 4.3.11.
Information Obtained i
i Resolved by response to 4.3.11.
)
Documentation Needed Resolved by rerponse to 4.3.11.
Resolution Resolved by response to 4.3.11, 32
i l
4.3.19 Summarv of Clarification Reauested Resolution of RAls A.3.11,4.3.18,4.3.14 will result in the resolution of RAI 4.3.19.
. Information Obtained The resolution described in the RAls 4.3.11,4.3,18, and 4.3.14 is sufficient to resolve this RAI 4.3.19.
Documentation Needed Documentation necessary will be provided with additional clarification to RAI responses 4.3.11,4.3.18, and 4.3.14.
Resolution Resolved, as a result of the information provided above by BGE staff for RAls 4.3.11, 4.3.18, and 4.3.14.
4.3.21 Summarv of Clarification Reauested in NRC RAI 4.3.21, the staff requested a description to account for the accuracy required in the use of visual indications of detectable wear to reliably determine changes in the hold down ring (HDR) load developing capability.
The BGE response did not address this part of the staff's request.
BGE is requested to expand upon the description provided in their response. The description should address: the depth of wear required to be detected before the effects begin to compromise the structural integrity of the RVl; the capability of visual examinations to reliably detect the amount of wear; and what are the wear surfaces that are examined and involved in these age-related effects of wear.
Information Obtained Discussion with ISI personnel revealed that wear measurements are not made during Isis. Isis only look for indications of wear.
Discussion with mechanical department leader revealed that the measurement of component wear on the RVI or PV RVi support ledge are not made during the 10-year ISI.
~ Documentation Needed To be determined.
Resolution Open.
33
l 4.3.22
)
Summarv of Clarification Reauested
{
BGE's response to this RAI did not contain the basis requested. Table 4.3-2 in Appendix A to the LRA is a matrix of device types and potential and plausible age-related degradation mechanisms (ARDMs) considered for the RVI. BGE is requested to expand upon its response to the RAI to provide the basis for not considering stress relaxation plausible for the HDR, or demonstrate how these aging effects are or need to be managed.
Information Obtained Loss of pre-stress is not p'8lausible ARDM for the HDR based on the fluence level being in the range of 1012 to 10 n/cm for 60 years. in-pile testing indicated that stress 2
relaxation occurs as operating temperatures, under constant high stress levels and high radiation field (approximately 5 X 102 n/cm and energy greater than one million 2
electron volts) for 60 years. Operating stress of the HDR is less than two-thirds yield.
Documentation Needed BGE provided matrix code list dated 12/10/98.
Resolution Resolved, as a result of the i 'ormation provided above by BGE staff.
4.3.23 Summarv of Clarification Reauested The staff requested a description of the portions of the control element assembly shroud bolts (CEASB) that are accessible for visual examination and a discussion of how the observations can be used to reliably demonstrate that neutron embrittlement will be managed.
Information Obtained Portions of the CEASB that indicate precursors are not accessible for visual examination.
Documentation Needed To be determined Resolution Open. Resolution of 4.3.15 will resolve this item.
34
5.1.1 i
Summarv of Clarification Reauested Why is the AFW piping from the turbine exhaust to the roof exhaust not within the scope of license renewal?
Information Obtained The turbine-driven AFW pumps' exhaust piping is Seismic II/I and designed as described in the UFSAR (Section 5A.3.2.2). As such, it does not present a safety threat to in-scope equipment due to failure. Exhaust steam leakage into the room may cause a rise in temperature; however, operating procedures specify that the normal and emergency (safety-related) ventilation systems be operated as required to maintain temperature in the room below 130 F. In addition, the double doors to the Turbine Building may be opened to provide further venting if deemed necessary, l
Documentation Needed l
None Eggolution Resolved, as a result of the information provided above by BGE staff.
5.7.18 Summarv of Clarification Reauested Describe the current cathodic protection program for the diesel fuel oil system underground piping.
Information Obtained BGE staff stated that they currently do not hrve a cathodic protection program for the subject equipment; however, BGE staff stated that they expect to develop a program in the Summer of 1999. They do not have a copy of RP0169-92 which is the NACE recommended practice for corrosion control and monitoring of underground pipelines.
The NRC staff provided a copy of RP016-92 to the CCNPP staff. Above ground piping is monitored using walkdowns.
Documentation Needed BGE committed to provide an outline of cathodic program that will be used to manage underground portions of the diesel fuel oil system.
Resolution 1
Resolved, as a result of the information provided above by BGE staff.
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l 5.9.44 Summarv of Clarification Reauested Feedwater nozzle is projected to exceed a cumulative usage factor (CUF) of 1.0 within 40 years. What are the corrective actions?
Information Obtained (1)
BGE responded that they would implement corrective actions prior to exceeding a CUF of 1.0, as required by the Fatigue Monitoring Program. At this time the corrective action is planned to be replacement of the affected items. However, if in the future acceptable corrective actions, such as acceptance of ASME Section XI Appendix L by the NRC, become viable, BGE may consider applying these as appropriate.
(2)
The projected usage for the feedwater items as presented in EPRI TR-107515 has been identified by BGE as not being the Analysis of Record for these items.
However, the CUF predictions without application of environment effects under GSI-190, are representative of actual projections.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
5.11.14 Summary of Clarification Reauested Corrosion of heating and ventilation duct lap joints.
Information Obtained Evidence of corrosion of HVAC lap joints will be observed before loss of air flow.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
5.17.1 Summare of Clarification Reauested is non-safety-related portion included in safety-related service water age-related degradation inspections (ARDis)?
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Infnrmation Obtained BGE will include non-safety related portion in safety-related portion in the same ARDI for erosion / corrosion. BGE agrees to document this change to the scope of the ARDI program Documentation Needed None Resolytion Resolved, as a result of the information provided above by BGE staff.
7.1 Summary of Clarification Reauested Has fatigue analysis been performed for heat exchangers (HX) and temperature elements (TE)?
Information Obtained Thermal will of TE is included as part of the pipe.
A Combustion Engineering report was provided that indicated that the regenerative and letdown HX are enveloped by monitoring the charging nozzles Documentation Needed BGE should supplement its rennonse to the RAI, which contains the additional information.
Resolution Resolved, as a result of the information provided above by BGE staff.
3 7.3 - 7.5 Summarv of Clarification Reauestqq Did BGE address the additional letdown cycles for the chemical and volume control system based on EPRI report?
Information Obtained l
l The analysis has since been completed. Update to the FMP is still ongoing. All items I
were shown to be within the acceptance criteria (CUF less than 1.0), for all design loadings and cycles (including 200 loss of letdown events).
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Documentation Needed j
'None Resolution Resolved, as a result of the information provided above by BGE staff.
j 7.8 Summarv of Clarification Reouested How is vibration fatigue near the charging pump managed?
Information Obtained BGE redesigned the charging pump block and piston, pipe routing, and pip support.
BGE has now determined that fatigue is 'non-plausible."
The LRA indicates that vibration fatigue is plausible and will be managed by ARDI.
Documentation Needed BGE should supplement the RAI response with the above information.
Resolution Resolved, as a result of the information provided above by BGE staff.
7.9 Summary of Clarification Reau'ested Discuss vibrational fatigue experience of the RCP leak-off line and parameters monitored.
Information Obtained RCP leakage line is not safety-related and has no intsnded function. The discussion in the application is operating experience of the system whether it is within or outside the scope of license renewal.
BGE staff stated that the RCP leak-off lines are not within the scope of license renewal.
Documentation Needed None 38 i
Resolution Resolved, as a result of the information provided above by BGE staff.
7.13 Summary of Clarification Reouested Corrective actions when fatigue CUF exceeds 1.0 Question: When to imitate corrective actions?
Information Obtained '
Corrective actions will be initiated based on engineering Judgement on CUF, transients, and rate of increase. Corrective actions will be initiated before CUF exceeds 1.0 and will be implemented before excedance of the LLB (e.g., exceeding CUF=1.0). The intent of BGE program is to expand monitoring locations if a monitored location exceeds 1.0 and replaced the affected components.
Documentation Needed i
None Resolution Resolved, as a result of the information provided abovo by BGE staff.
7.21 Summary of Clarification Reouested Analysis to determine if additionalitems need to be in the fatigue monitoring program to be performed.-
Information Obtained (See RAI 4.2.23 and 7.11)
Documentation Needed To be determined Resolution Open - This is a " policy" question on how aging manage' ment is demonstrated by a future evaluation.
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GSI-190 Summarv of Clarification Reauested Need to address Argonne stainless steel fatigue data.
Infonnation Obtained NEl is preparing a letter evaluating the Argonne data. A meeting should be set up for NRC and industry to discuss the technical merits. GSI-190 will be an Open item in the SER done in March.
Documentation Needed l
To be determined.
Resolution Open.
11.6 Summary of Clarification Reauested Use of ARDI for identifying extemal corrosion of SW bolts.
Information Obtained ARDI will no longer be used for inspection of corrosion of extemal carbon and steel components previously relying on ARDI. BGE will now rely on MN-1-319 Systems and Structures program to inspect these SW components.
Documentation Needed BGE staff suggested the information could be provided during an errata update, but they were not sure if BGE has any currently planned. Don Shaw stated that there are two positions for getting information docketed: (1) Around March 21,1999, BGE submitting required annual update, and (2) BGE submitting revised Electrical Commodities report and attach the information.
Resolution Resolved, as a result of the information provided above by BGE staff.
New Question " Baseline Walkdowns" Summarv of Cl #tgation Reauested Explain corrective action (CA) process associated with findings from baseline walkdowns.
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l Information Obtained BGE stated that CAs are handled in accordance with site corrective action programs QL-2-100.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
New Question " Sample Expansion" Summarv of Clarification Reauested How will sample size of baseline inspection of component supports be expanded if
. degradation is found?
Information Obtained BGE provided the reviewer meeting minutes showing their plans to implement baseline inspection and mathematical model for determining sampling size and expansion.
BGE also provided copy of TR10714 (referenced by meeting minutes).
Meeting minutes indicate that sampling model in TR10714 will be used.
Documentation Needed None Resolution
' Resolved, as a result of the information provided above by BGE staff.
1 New Question "3.3A-Scoping" Summary of Clarification Reauested-The staff believes that 10 CFR 54.4(a)(2) and 10 CFR 54.(a)(1)(i) require the tendon access gallery to be within the scope of license renewal in order to access the tendons, and provide protection to the tendon anchorages from an adverse environment.
Information Obtained i
BGE's position, to date, is that the tendon access gallery does not meet structural j
scoping criteria. However, BGE recognizes that the presence of the gallery affords l
access to the tendon anchorages for surveillance.
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UFSAR 5.1.3.1 states that the tendon access gallery was designed as a separate structure. The tendon access gallery was not credited in the analysis of the primary containment structure.
BGE and NRC will re-evaluate its position.
Documentation Needed None Resolution Open New Question "Section 3.3A, Group 4" Summary of Clarification Reauested Please explain how BGE manages the aging effect intergranular stress corrosion cracking (IGSCC) of refueling poolliner.
Information Obtained Aging management as desenbed on page 3.3A-28 of Appendix A to the LRA, is through j
discovery. Walkdowns provide for discovery and management of effects of corrosion J
through visual inspections, reporting the leakage detected, and initiating corrective 1
action. Routine inspections (walkdowns) are performed in accordance with CCNPP Procedure MN-1-319. Under this procedure, any evidence of fluid leakage would be considered adverse to quality and, therefore, addressed in accordance with the CCNPP corrective action program.
A similar type of approach for IGSCC of the spent fuel pool liner is used (refer to BGE LRA Section 3.3E, Group 3)
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
New Question "Secti>n 3.3A, Cable Trays" 1
Summary of Clarification Reauested Cable trays don't appear to be covered in Appendix A to the LRA. Also, where are HVAC ducts covered?
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e information Obtained Cable trays are covered in Section 3.1 of Appendix A to the LRA. The cable tray supports include the cable trays themselves.
"Outside containment" covers all areas within the scope of license renewal outside containment.
HVAC systems within the scope of license renewal are covered in BGE LRA Sections 5.11A, 5,11.B, and 5.11.C. These systems include all safety-related ducting.
Documentation Needed None Resolution Resolved, as a result of the information provided above by BGE staff.
New Question " Scoping for Electrical Commodities" Summarv of Clarification Reouested The list of systems covered by the electrical commodities section of the LRA does not include all scoped systems. Does Section 6.2 include the electrical commodities from these systems?
Information Obtained Electrical commodities are sometimes associated in the plant equipment data base with other systems. For example, System 062, " Control Boards," contain panels associated with the component cooling water, service water, and saltwater system, safety injection, l
main steam, and other systems. BGE began the scoping process with the Master Equipment list for all scoped systems. Furthermore, the annual update process involved a check against all reivent equipment types.
BGE reviewed Table 6.2-1 of Appendix A to the LRA and concluded that all electrical I
commodities associated with BGE systems within the scope of license renewal, including ECs from systems 15,23,41,55,61,67 and 83, are included in Table 6.2-1.
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Documentation Needed None i
l Reso!ution Resolved, as a result of the information provided above by BGE staff.
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o New Question "3.2 - Polar Crane" Summary of Clarification Reauested The staff raised a concem regarding the stress cycle for the polar crane system (crane, rails, and supporting structures).
Information Obtained BGE presented the cranes and fuel handling equipment AMR report that stated that the PC is classified as Type "A" crane and allow a stress range up to 40 ksi. Also because the PC system is constructed of A-56 steel with an allowable of 21.6k, which is lower than 40 ksi, there is no limit for the number of cycles.
Documentation Needed The staff requested BGE to justify the stress will not go beyond 21.6 kai or total number of cycles (stress) is within the allowed cycles.
BGE provided the following additional information:
The discussion for the non-plausibility of Fatigue for the Polar Crane in the Cranes and Fuel Handling AMR relies on an comparison of the maximum allowable stress of ASTM A36 steel vs. stress range limits provided in CMAA #70, the goveming standard for crane design. The staff did not agree with BGE's reasoning in this plausibility decision and requested that BGE justify either that the stress will not go beyond the allowable stress of ASTM A36 or that the total number of cycles is within the allowed range.
BGE now chooses to demonstrate the latter and will no longer base the non-plausibility of fatigue on an evaluation of the allowable stress of ASTM A36 steel vs. stress range limits.
CMAA #70 requires that crane members and fasteners subject to repeated load be designed so that 1) the maximum stress does not exceed 17.6 ksi (less than the allowable stress of ASTM A36 used in the above discussion) and 2) that thei stress range for various subcomponent configurations does not exceed the allowable values given in a Table. For cranes of Service Classification "A," the applicable class for in-scope CCNP cranes per Bechtel Specification 6750-C-42, the Table stress range values are predicated on the number of loading cycles being between 20,000 and 100,000.
Design Calculation C-93-164 projected that the CCNPP Unit 1 Polar Crane components had experienced 8460 load cycles from initial installation until 1992. This evaluation contained the conservative assumption that each lift resulted in four stress cycles to the components. Extending the projection to year 2034, it is estimated that the Polar Crane will experience 13,860 load cycles. This is much less that the maximum of 100,000 assumed for the purposes of determining the allowable stress range. Since the crane is not expected to exceed the originally assumed number of loading cycles, the original
-design remains bounding, and fatigue will be considered not plausible. A similar projection could be made for the Unit 2 Polar Crane.
Resolution Resolved, as a result of the information provided above by BGE staff.
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New Question " Preventive Maintenance Checklist" Summary of Clarification Reauested The " Preventive Maintenance Checklists
- program is a modified aging management program and is to be credited for the follow-on activities of containment air cooler fans (metal spring isolators and fixed bases). However, the applicant did not provide any information and/or basis to demonstrate how the modified program will be implemented.
Information Obtained See page 3.1-28 of Appendix A to the LRA where BGE commits to modify preventive maintenance checklists MPM 09150 and MPM 09151. They will also be modified to inspect the spring isolator supports for signs of general corrosion.
Documentation Needed Copies of MPM 09150 and MPM 09151 were provide to the staff.
Resolution Resolved, as a result of the information provided above by BGE staff.
45
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COPlES of FACSIMILES
,,. 3 RAI 2.1.1 During the aging management review process the Class 2 and 3 systems BGE considered fatigue a potential ARDM. For the Class 2 and 3 systems where the LRA identifies fatigue as not plausible, BGE considered the following to make the not plausible determination:
- 1. The AT between system shutdown temperature and maximum operating temperature.
If the AT was small (say 50*F), fatigue was not plausible.
- 2. If the AT was larger, BGE conservatively estimated the number of thermal cycles for -
60 years using plant operating history. If the number of thermal cycles was less than the design number of cycles, fatigue was not plausible.
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1 ENCLOSURE 3 l
EP/EI*d QWAA FAD ATb 9 f M 192% 9 CHrl 2Mr11.11 1 *1MC PC.OT CCCT.CT MHl.1
In response to discussions on RAI number 2.1.1 on the Containment Liner Fatigue, BGE has generated the following evaluation demonstrating that the current analysis remains valid for the period of extended operation. Based on the conclusion of this evaluation, the TLAA AMRR and LRA section 2.13.5 will be revised. Since this evaluation demonstrates that the current analysis remains valid for the period ofextended operation, no further action or analysis is required for this TLAA.
UFSAR Chapter 5.1.43 states that the linear plate was designed with the following considerations:
"The best basis for establishing allowable liner plate strains is considered to be that portion of the ASME, B&PV Code,Section III, Nuclear Vessels, Article 4. Specifically, the following sections have been adopted as guides in establishing allowable strain limits:
Paragraph N 412(m) normalStress Paragraph N-414.5 Peak StressIntensity Table N-413 Classification of Stresses for Some Typical Cases Figure N-414 Stress Categories and Limits of Stress Intensity i
Figure N-415(A).
Design Fatigue Curves Paragraph N-412(n)
Operational Cycle Paragraph N-415.1 Vessels Not Requiring Analysis for Cyclic Operation American Society of Mechanical Engineers design codes require that the liner material be prevented from experiencing significant distortion due to the thermal load and that the stresses be considered from a fatigue standpoint [ Paragraph N-412(m)(2)]. The following fatigue loads were considered in the design of the liner plate:
Thermal cycling due to annual outdoor temperature variations. Daily a.
temperature variations do not penetrate a significant distance into the concrete shell to appreciably change the average temperature of the shell relative to the liner plate. The number ofcycles for this loading is 40 cycles for the plant life of 40 years.
b.
Thermal cycling due to interior temperature variations during the strrtup and shutdown of the reactor system. The number of cycles for this loading was assumed to be 500.
Thermal cycling due to the LOCA was taken very conservatively to occur only c.
once during plant life. Thermal load cycles in the piping systems are somewhat isolated from the liner plate ji..m.iions by the concentric sleeves between the pipe and the concrete. The attachment sleeve was designed in accordance with ASME, B&PV Code,Section III fatigue considerations. All 4,41ons were reviewed for a conservative number of cycles to be expected during plant life.
Thermal stresses in the liner plate fall into the categories considered in Article 4 Section HI, Nuclear Vessels of the ASME, BAPV Code. The allowable stress in Figure N-415(A) is for alternating stress intensity for carbon stools and temperatures not exceeding 700'F. In addition, the ASME code further requires that significant distortion of the material be prevented.
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500 Heat-ups and Cool-downs.
541 Total This is well below the limit of 1,000 shown above. If the seasonal variations are changed from 40 yrs to 60 yrs, the total would be 561 cycles. His very conservatively results in the following usage:
- ofYears a
N U
40 years 541 1000 0.54 60 years 561 1000 0.56 De above approach is approximate and grossly overestimates the usage for the specified loadings. This is because the stress range for the maximum load case is a pplied to all load cases. From the large difference in severity between load cases (i.e., LOCA vs.
normal) this approximation is likely to be conservative by at least a factor of 10.
Regardless, the usage is less than 1.0 which will ensure a leak tight containment. The overall impact of the extended period will be less than a 4% change in usage.
Based on the above, it is concluded that the containment liner plate is acceptable for 60 yrs with-out consequence.
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S In accordance with ASME, B&PV Code, Paragraph 412(m)(2), the liner plate is restrained against significant distortion by continuous angle anchors and never exceeds the temperature limitation of 700*F. It also satisfies the requirements for limiting strains on the basis of fatigue consideration. A typical section showing the anchors is included in Figure 5-1.
American Society of Mechanical Engineers, BAPV Code, Paragraph 412(n), Figure N-415(A), has been developed as a result of research, industry experience, and the proven performance of code vessels. Because of the conservative factors it contains on both stress intensity and stress cycles, and its being a part of a recognized design code, Figure N 415(A) and its appropriate limitations have been used as a basis for establishing allowable liner plate strains. Since the graph in Figure N-415(A) does not extend below ten cycles, tan cycles was used for the LOCA instead of one cycle mentioned above.
Establishing an allowable strain based on ten significant thermal cycles of the LOCA condition would permit an allowable strain [from Figure N-415(A)] of approximately 2%. Maximarn allowable tensile or compressive strain has been conservatively set at 0.5% (compared to 2% shown above). He maximum predicted strain in the liner plate during LOCA conditions has been found to be 0.25% compression.
At the design LOCA condition, there will be no tensile stress anywhere in the liner plate membrane. His is true both at the time ofinitial pressure release and under any later pressure and temperature condition. The purpose of specifying a non-destructive examination temperature requirement is to provide protection against a brittle fracture or cleavage mode of failure. However, this type of failure is precluded by the absence of tensile strenes.
No allowable compressive strain value has been set for the test condition because the value will be less than that experienced under the LOCA condition. The maximum allowable tensile strain will be 0.2% under test conditions; the predicted value is much srnaller.
The maximum compressive strains are caused by LOCA pressure, thermal loading prestress, shrmkage, and creep. The maximum calculated strains do not exceed 0.0025 in/in. and the liner plate will always remain in a stable condition."
Using a maximum strain of 0.0025 in/in (which incluhs LOCA loads), the stress range is found using Hook's law, or:
6 Where: E = Young's Modulus = 30 x 10 pai
- = Stress, psi e = Strain, in/in From this the stress range for the maximum load state is 75,000 psi. His includes LOCA, pressure, thermal loading prestress, shrinkage, and creep. Using 75,000 psi, the number of allowable cycles in Fig. N-415(A) is in excess of1,000.
The number of cycles required to be considered is as follows:
1 LOCA (Although 10 were considered because the fatigue curves start at 10) 40 Seasonalchanges(40 yrs)
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h Group 6 - RAI 3.1.14 Arc all ring foundations within the scope oflicense renewal in the ISI Program?
Response
Concrete ring foundations, by themselves, are not in the ISI Program. However, System 036 (Auxiliary Feedwater System - Condensate Storage Tank No.12) and System 37 (Demineralized Water and Condensate Storage System - Condense Storage Tanks No.11 and No. 21) have their supports inspected as part of the ISI Program. 'Ihis includes a check for concrete damage in the vicinity of the supports.
Note that Condensate Storage Tank No.12, which is evaluated in System 036, rests on the floor slab of Condensate Storage Tank No.12 Enclosure, which is evaluated in BGE LRA Section 3.3D, Miscellaneous Tank and Valve Enclosures. It is included in the Component Support Group - Ring Foimdations for Flat Bottom Vertical Tanks, for similarity purposes, because ofits tank chairs and anchor bolts. Reference page 1 of BGE letter to NRC, dated February 4,1999, on Changes to the Application for License Renewal.
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M Group 13 - RAI 3.1.15.b Explain apparent inconsistency in treatment of pipe hangers with respect to dynamic ARDMs.
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Response: Please see respense to RAI 3.1.19.
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J RAI 3.1.24 - Where does Calve:t Cliffs have high strength fyield strength greater than 150 ksi) anchor bolts installed? )
The two materials used in anchor bolt applications at CCNPP meeting this criteria are ASTM A354 and A490. A354 bolting is used in the reactor vessel, pressurizer and safety injection tank anchor bolts and A490 bolting is used in the SG supports. This topic is discussed in Section 3.1 of the LRA under Group 7.
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i RAI 3.3.12 - TLAA on tendons - Provide NRC with latest STP M-663-1 results which provide information relative to hoop and dome liA off forces for comparison to expected lift off curves in UFSAR Chapter 15.
1
Response
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The normalized average lift off forces (kips) for hoop and dome tendons measured during CCNPP's 20th Year Surveillance of the Unit 1 Containment Building Post-Tensioning System were 621 and 632, respectively.
General information about CCNPP's containment design and its tendon surveillance programis provided below:
In accordance with ACI Code 318-63, the containment design provides for prestress losses that can be predicted with sufficient accurav as described in Calvert Cliffs UFSAR Section 5.1.4.2. The environment of the prestress system and concrete is not arnreciably different from that found in numerous bridge and building applications.
Considerable research has been done to evaluate the causes of prestress losses and this information was used to assign appropriate allowances. Building code authorities consider it amyable practice to develop p
===t designs based on these allowances.
The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in the verification requirements identified in UFSAR Section 15.6.1. Curves describing the predicted prestress loss behavior are contained in UFSAR Figures 15.6.1-1,15.6.1-2 and 15.6.1-3. These curves will be extrapolated out to 60 years for the extended license period using this current licensing basis.
A licensing basis surveillance change will result from the new rule recently listed in 10 CFR 50.55(a), incorporating ASME Section XI, Subsection IWE/IWL requirements. This is an acceptable inservice inspection and surveillance method for ungrouted tendons in prestre; sed concrete containment structures as previously acknowledged by the staff. It currently provides adequate aging management of the tendons and will continue to provide adequate agmg management into the extended license period.
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/O RAI 3.3.12 As required by 10 CFR 50.55a(b)(vi), BGE plans to revise the tendon surveillance progran to conform to the 1992 Edition with the 1992 Addenda of the ASME Beiler and Pressure Vessel Code,Section XI, Subsection IWL.
IWL-2520 gives the exarninarion re. qui:ements of tendon selection, force and elongation measurement, wire sampling, etc.
10 CFR 50.55a(b)(ix) requires an evaluation of prestressing force trends for each tendon and groups of tendons to ensure that the predicted tendon forces at the next scheduled examination meet the minimum design prestress requirements. If these requirements are not met, an evaluadon is required in accordance with the Engineering Evaluation Repon as prescribed in IWL-3300. The IWL-3300 Evaluation requires determination of the cause of the condition, acceptibility without repair, whether repairheplacement is required, if repair / replacement is requhed the extent, method and required completion date.
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- 1. RAI 3.3.36 - Non-aggressive groundwater chemistry.
BGE Action: Provide bases for statement in BGE response to RAI 3.3.2 under section titled aggressive chemical attack on concrete and corrosion of embedded steel /rebar, that "no conclusive evidence that confirms that the groundwater chemistry has become aggressive since plant construction."
Action: During construction, analysis of groundwater at the CCNPP site indicated neutral pH (7.5), with concentrations of aggressive chemical species well below the levels that can lead to these ARDMs (sulfates at 44 ppm and chlorides at 13 ppm).
Twice in 1996 BGE took groundwater samples from four wells near the Subsurface Drainage System. The results of the first test showed three locations which confirmed that the groundwater remains non-aggressive (i.e., pH st 7.5, sulfates ranging from 2 to 21 ppm, chlorides ranging from 8 to 25 ppm). The remaining sample, Well No. I1, showed a neutal pH of 7.2 and abnormally high concentrations of sulfates (1635 ppm) and chlorides (3045 ppm). BOE believes that this sample agiwuts a very local condition and is not representative of the groundwater at the site. The sample location is immediately adjacent to the <Hecharge canal near where the Subsurface Drainage System drains into the discharge canal. He likely explanation for the high readings is that a small amount of circulating (i.e., bay) water is occasionally backflowing from the discharge conduit into the Subsurface Drainage System piping and entering the groundwater.
The second set of samples taken in 1997 showed similar results, although the concentrations ofions was lower. The location with the highest concentrations ofions previously, Well No.11, had a pH of 7.5,640 ppm sulfates, and 987 ppm chlorides.
Based on the above test results, the groundwater at CCNPP is considered to be non-aggressive.
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s Author: ERNIE TAORMINA at -NESDPO Date:
2/26/S9 8:54 AM Normal TO: JOHlf RYCYNA Subjects SG outlet nozzle ISI question
......................,.....------- Message contents
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Forward Header
Subject:
Re another NRC ISI question related to LR Author: KEITH M ROFFMAN at -NSB1PO2 Date:
2/26/99 Bios AM I think we rced to make the addition shown.
The following question was posed by the NRC:
RAI 4.1.1'1 - How will ISI of the SG outlet nozzles (welds) 6etect erosion corrosion.
De*. reposed response:
The ISI inspection also looks at the base metal adjacent to the welds and the nozzle inner radii, and this will detect the occurrence of t
erosion corrosion.
Your input is A; A appreciated.
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hS RAI 4.2.6 BGE will revise the surveillance capsule withdrawal schedule from 40 to 60 years (48 EFPY). The new schedule will include the withdrawal of at least one capsule from each unit that will provide data at a neutron fluence equal to or greater than the projected peak neutron fluence at 60 years (48 EFPY)
If BGE withdraws the last capsule from either RPV prior to year 55, BGE will establish the neutron irradiation environment (fluence, spectrum, temperature and Dux) applicable to the surveillance data and P-T Limits for the afected RPV. If the RPV(s) operates outside these limits, BGE will inform NRC and determine the impact of the condition on the RPV(s) integray.
If BGE withdraws the last capsule from either FJV prior to year $5, BGE will provide additional dosimetry for the affected RPV.
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RAI 4.2.20 Combustion Engineering (C-E) determined the WA: components and locations and the controlling
. transients that BGE used to develop the Fatigue Monitoring Program (FMP). The process C-E used is as follows:
- 1. C-E identified those systems that the UFSAR and Technical Specifications include for maintaining the safe operations and achieving a safe shutdown of the plant. The result of this review was the "Ide stification of Critical Systems "
- 2. For each of the critical systems C-E reviewed all components within the system to identify those components with controlling fatigue usage limits. The critical system review included a review ofindustry data and experience as well as the original design documents associated with each system and/or component. The result of this reyww was the " Identification of Component Critical Locations."
- 3. C-E then performed a fatigue evaluation for each component identified as a witical location. This evaluation included an evaluation of the original stress repotts to identify the controlling design basis i
transients with respect to fatigue usage. The results are " Fatigue Evaluation of Critical Locations."
- 4. C-E then developed a logging and procedure guideline for tabulating cumulative fatigue usage for BOE to use in the development of the FMP.
BGE used the resuhs of the C-E work to develop and implement the FMP.
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cN RAI 4.2.20 Combustion Engineering (C-E) determined the bounding components and locations and the controlling transients that BGE used to develop the Fatigue Monitoring Program (FMP). The process C-E used is as follows:
- 1. C-E identified those systems that the UFSAR and Technical Specifications include for maintaining the safe operations and achieving a safe shutdown of the plant. The result of this review was the " Identification of Critical Systems."
- 2. For each of the critical systems C-E reviewed all components within the system to identify those components with controlling fatigue usage limits. The critical system review included a review ofindustry data and experience as well as the original design documents associated with each system and/or component. The result of this review was the " Identification of Component Critical Locations."
- 3. C-E then performed a fatigue evaluation for each component identified as a critical location. This evaluation included an evaluation of the original stress reports to identify the controlling design basis transients with respect to fatigue usage. The results are
" Fatigue Evaluation of Critical Lccations."
- 4. C-E then developed a logging and pcedure guideline for tabulating cumulative fatigue usage for BGE to use in tie development of the FMP.
The following systems or components were not included within the C-E scope of work:
NSSS Sampling System PZR safety valves Power operated Reliefvalves AFWisolation r.ud check calves MFW check valves MFWisolationvalves Reactorcoolant pumps Additionally, C-E did not include potential thermal stratification loadings identified in NRC Bulletins 88-08 and 88 11. However, BGE has incorporated components that do experience thermal stratification loadmgs in the FMP. BGE committed in the LRA to perform an evaluation of the SI system with respect to thmnal stratification loadings.
BGE used the results of the C-E work to dewlop and implement the FMP.
ir A,e /Mrt4L 27l RAI 4.3.14 For Pinina Valver etc.
BOE agrees to the following revisions to the screening process for cast stainless steel.
a) Statically cast components with a molybdenum content exceeding CF3 and CF8 limits and a delta ferrite content excaadina 10 percent will be subP.at to inservice inspection in accordance with ASME Code Section XI.
i b) Ferrite levels will be calculated using Hull's Equivalent Factors or a method producing an equivalent level of accuracy (+/- 6% deviation between measured and calculated values),
c) Cast stainless steel components containing niobium will be subject to inservice inspection.
BGE does not agree to the following additional requirements proposed by the StafE a) Inservice inspection procedures and techniques must be quali5ed using methods consistent with those of Appendix VIII of Section XI of the ASME Code. This additional requirement pre-supposes that an ultrasonic exam will be required.
It is possible that altemative ernminatian techniques would be sufHelent. The proposed qualification requirements are currently impossible to achieve. BGE would commit to using the ultrasonic examination qualified to Appendix VIII providing it is eventually possible to qualify any UT technique for CASS.
b) Flaws in cast stainless steel components with ferrite levels exceeding 25 percent or with niobium will not be evaluated using ASME Code IWB-3640 procedures. In these instances, fracture toughness data will be provided on a case by case basis.
Instead, flaws could be evaluated using IWB-3640 but specific fracture toughness ' formation m
would be required.
The above discussion applies only to piping, valves, and other affected components in general it does not apply to the Reactor VesselInternals since the EPRI screening criteria do not consider the effect ofneutron radiation as applicable to high fluence components. For these components, mechanical loading during ASME Code A, B, C, and D conditions must be low enough or compressive to ensure that the affected component will not fracture.
For the Rametor Veael Intemmic BGE will a) Determine whether the components are susceptible to a loss of fracture toughness by evaluating whether the components receive a fluence above the threshhold of 10" n/cm'. If the fluence is below the threshhold, the screening criteria described above for pipes and valves in general will be applied by performing a flaw tolerance evaluation specific to reactor vessel internals. If the fluence is above the threshhold, BGE will...
b) Evaluate the mechanical loading to denine ifit will Innain low enough or compressive to ensure that the affected component will not fracture during ASME Code A, B, C, and D conditions. If this cannot be determined, BGE will...
J c) Perform enhane>A VT-1 inspection of the components.
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v Regarding RAI 4.3.21:
As is noted in the LRA on pages 4.3-10 through 4.3-13, HDR wear is managed by the ISIprogram.
Specifically for the HDR procedure NDE-5715-CC is used. A VT-3 exam is performed on the surfaces. BGE is not committing specifically to NDE-5715-CC, but rather to our ISI program as documented in the LRA.
There is a discussion of the effectiveness ofISI in detecting RVI wear on page 4.3-13 in the second paragraph.
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.3 0 RAI 4.3.21 The following refemces, from the dockets of the St. Lucie and Palisades plants, support the contention that wear of the Hold-down Ring, although plausible, is not ew-d to be significant. This ring, formerly referred to as the expansion compensating ring, was redesigned to eliminate excessive wear ofintemals caused by a loss of core barrel 4
clamping force at the Palisades Plant in 1973. The new design was installed at CCNPP, Palisades and St. Lucie.
Ref(1) - Safety Evaluation by the Office of Nuclear Reactor Regulation related to Admendment 80 to Facility Operating License No. DPR-67, Florida Power & Light Company, St. Lucie Plant, Unit No.1, Docket No. 50-335 of May 1987.
This reference relates the NRC's concurrence that Tech Spec monitoring of Core Support Barrel motion was no longer required since "the redesigned core barrel hold-down ring has elimi=ted the possibility of excessive core barrel movement such as that occurred at Palisades."
Ref(2) - Proposed Tecnical Specification Change Request - Delete Technical Specification 4.13, Reactor Intemals Vibration Monitoring, Docket 50-255, License 1
DPR-20, Palisades Plant, dated March 29,1985 This reference relates that regarding reactor vessel internals vibration that: " measured values have been... consistently far below the established limits for vibration."
Regarding the hold-down ring, this rrference 4tes that: " Consequently, the modiScation which increased the core barrel clamping force has ben proven r"sctive over a num-n af years and core.jcles." NRC review of the Palisades proposal is documented in Federal Register Volume 50 No. 78 of Tuesday, April 23,1985, page 16002.
CCNPP uses the same hold-down ring design as Palisades and St. Lucie. BGE reiterates j
it's contention that Section XI Visual inspection of th-tiold-down Ring for wear is adequate to manage this plausible but insignificant aging effect.
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NRC Request:
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For programs that are yet to be developed, please provide infonnation on those elements that are j
available. For the remaining elements, please provide the deta.iled process, including the basis, that BGE will use to develop them and a schedule for when the proyam development will be completed.
i Discussion-I During the site visit on February 16, NRC indicated that an acceptable alternative to providing the above information for the Buried Pipe Inspection Program (now referred to j
as the Buried Pipe Condition Monitoring Program (BPCMP)) would be for BGE to commit to adopting NACE Standard Recommended Practice RP0169, Control of External Conosion ofUns.y od or Submerged Metallic Piping Systems.
BGE obtained RP0169 and found it to contain significant useful information and references. Nevertheless, the preponderance of the standard addresses design and installation for new construction. Seven of the 11 sections of the standard (Determination of the Need for Corrosion Control, Piping System Design, Criteria and Other Consideraticms for Cathodic Proteeion, Design of Cathodic Protection Systems, 3
Installation of Cathodic Protection Systems, Corrosion Control Records) have little bearing with regard to BGE's current needs. While useful, much of this and the remaining guidance duplicates or parallels what is already provided in BGE design standards and guidance.
BGE commitment to adopt the NACE standard without exception could therefore result in the need to identify, reconcile, and justify differences between existing BGE guidance and the standard. The standard, however, discourages using selected parts. It states that it must be used in its entirety; "using or referencing only specific paragraphs or sections can lead to misin%d.Gon and misapplication." Therefore BGE will not formally viopt RP0169 and will instead provide additional information that responds to the NRC
< xtuest in the first paragraph above.
Bfs; Asponse for Euried Pine Conditian Manhorine Prnerum BGE's current ne.h revolve primarily around integrating various design, operational, and environmental factors to identify locations where buried piping degradation is occurring so that timely conective and preventive action can be taken. De BPCMP is being established to address this need. He following discussion consists of two major parts. He basis for the BPCMP is first discussed, including the cunent status of program development activities. Dis is followed by a discus.sion of the applicable attributes of the program and the schedule for establishing the program.
P Basis for the Buried Pipe Condition MonitorinF rogram He Buried Pipe Condition Monitoring Program will be based on an ongoing anae== ment that includes the following elements:
Review of existing cathodic protection FM results Review of previous buried pipe inspection reports Inspections of select sections of buried pipe e
Detailed cathodic protection surveys e
This assessment includes piping within and beyond the scope ofLicense Renewal.
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availatie on the site records imaging system from about 1991. nree PMs apply:
Clean and inspect circuit breakers - annually Check tap settings and take voltage and amperage readmet - monthly e
Conduct cathodic protection potential survey - quarterly The cathodic protection system is tested quarterly and adjusted to provide the optimal protection for the serviced components. Checks on the system operation and performance are included in the site preventive mamtenance program and the system is normally not permitted to be out-of-service for more than 60 days. In general, surveys have found present protection levels to be satisfactory. Where protection levels have been found to be hig'aer or lower than desired, the system has been or is being upgraded, ne failure of the system to operate will not cause or allow significant corrosion of serviced components in the shoit term. Since it is a design feature, modifications to the system and its operation are subject to the provisions of 10 CFR 50.59. Current site processes will be adjusted, as needed, to ensure that the role that this design feature plays is considered during future modification and maintenance activities.
Review or nreviaan aia* innaa**'a=
_ =. This eetivityidentified two recent excavations.
Diesel Fuel Oil (DFO) piping was exc ivated in 1994 to tie in the new diesels.
Coating flaws were noted and stones were noted in the original backfill. No corrosion damage was found on the piping and cathodic protection was found to be present DFO piping was excavated adjacent to 21 Fuel Oil Storage Tank (FOST) in 1996 to e
investigate oil found in the soil. Coating flaws were noted and stones were noted in the original backfill. No corrosion damage was found on the piping and cathodic protection was found to be present.
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Six sections of pipe have been inspected during the development phase of the buried pipe condition monitorieg program. Hese se:tions were either selected as worst case areas based on historical information, configurations with the highest likelihood of corrosion problems, or were excavated during other maintenance.
Two sections of DFO piping were excavated in 1997. Coating damage resulted from the excavation and was wd. No corrosion was found; soil composition and cathodic protection were good.
One section of tube oil piping was inspected using remote field eddy current
. techniques in 1997. The examination was from inside the pipe and found the pipe to
., be in good condition.
Two AFW lines were excavated h 9 ning in 1996 to replace heat tracing. Coating damage was noted where the plastic wrap was cut for the heat tracing and was corrected. Significant corrosion was noted on the piping but did not require pipe replacement. Pipe insulation was considered a primary contributing factor based on Etv9E*d 9p69 G6p Plte SINIWla it c:HO QNnut1*1HR of tpIt AAAT-AT-NHU
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similar experience with buried insulated piping, including a failed plant heating pipe.
As a result, the initial focus of the program is expected to provide close attention to buried insulated pipe.
DFO piping from the 11 FOST unloading station was inspected in 1998 where the piping passes through the dike around the tank. This inspection was prompted by weak cathodic protection levels and significant corrosion damage where rocks damaged the pipe coating where the pipes exit the soil. The pipes were found to be coated but not wrapped and areas with lifted coating showed significant corrosion.
De**IInd endiadie Brotection surveys:
A survey was performed in 1997 following replacement of west-road anodes and found some areas of weak protection. Upgrades were installed in the fall of 1998 and will be followed by another detailed survey. Survey results will be factored into the buried pipe I
condition monitoring program.
Conclnsions The periodic cathodic protection surveys, the results of prior inspections-of-opportunity, the results ofinspections conducted to develop the BPCMP, and the provisions of the corrective action program provide the basis for confidence that worst-case locations can be identified and monitored. Inspections more than 25 years after installation have identified some minor coating damage, confirmed the effectiveness of the combined design features of coatir;s and cathodic protection, and identified piping insulation as a significant face-M" ied pipe degradation. Inspection results have been consistent with cathodic proter. ion survey results and resulted in appropriate improvements to the cathodic protection system. De BPCMP discussed below will provide for additional inspections through the period of the current licenses and, as appropriate, during the period ofextended operation.
Attrihntes of the Buried Pine Candi*ian Mamitarina Pmerani Espas: Buried and transition portions ofpiping in DFO and AFW systems.
Methods Visual Inspection by competent personnel for evidence of coating perforation, holidays, and other damage; for evidence of pipe degradation, and for evidence of the continued effectiveness of cathodic pmtenon system BGE may also employ examinations from inside pipe to assess pipe wall condition (i.e.,
Periadiciev==A tic;< afinspections: In addition to inspection opportunities afforded by excavations st mu maintenance, subsequent inspections will be performed consistent with :. hs t f previous inspections and other operating experience. He periodicity and tin.a.6 of these inspections will be such that there is reasonable assurance the pipe wall condition will be found acceptable at the next scheduled inspection.
Cunent plans include some inspections during the last 5 years of the current operating licenses.
Variables and experience considered in the scope and timing ofinspections will include Design factors (i.e., cathodic protection proximity, materials, coatings, insulation) 7:e s o,,co ec,, m,,
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Environment (i.e., soil composition & conditions, wall or surface penetration, other structures in the area that affect the pipe)
Results of previous inspections Results of cathodic protection system PMs, including tap settings, voltage and current readings, and quarterly cathodic protection potential surveys Occurrence of nearby events that are likely to have damaged pipe coating e
Acceptance Criteria Evidence of coating perforation, holidays, or damage, and evidence of pipe damage will be evaluated in accordance with the CCNPP Corrective Action Program. This program responds to 10 CFR 50 Appendix B and includes provisions for repair or replacement, determining root cause, assessing generic implications, and establishing action to prevent recurrence.
A draft of the administrative procedure for this program is currently in review. The current BGE schedule is to have the necessary procedures in place by the end of 1999. This schedule is based on current site priorities and workload and is therefore subject to change.
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' CLR400 NUCLEIS 02/26/1999 Page 1 of w
CHECKLIST SHEET Requested By:
E43053 JOHN
. RYCYNA
-Dest Name:
n888 Num Copies:
1 Checklist ID:
i
NCLR400 NUCLEIS 02/26/1999 Ch2cklict Ch3cklist ID: MPM09150 Class Type: R
--.--------- MAINTENANCE REQUIRBEENTS AMD SPECIAL INSTRUCTIONS -----------
Dstcr: INSPECT CONTAINMENT AIR COOLERS TOOLS FLASHLIGHT 2 10' STRAIGHT LADDERS DROP LIGHT AND EXTENSION CORD INSPECTION MIRROR
- NOTE *****
OBSERVE REQUIREMENTS OF CH-1-102 AND MN-1-109 1.
PICKUP TAGOUT.
/
2.
INSPECT THE CONTAINMENT AIR COOLER BOOT LOCATED BETWEEN
/
THE FAN COIL AND THE AIR CGOLER ENCLOSURE.
' INSPECT FOR TEARS, HOLES OR DETERIORATION.
NOTE RESULTS IN THE REMARKS.
3.
INSPECT THE CONTAINMENT AIR COOLER BOLTS (THERE ARE
/
72 BOLTS, 6 PER COIL).
INSPECT TO ENSURE BOLTS ARE INSTALLED AND NOT LOOSE.
NOTE RESULTS IN THE REMARKS.
4.
INSPECT THE CONTAINMENT COOLER COILS TO ENSURE THEY
/
ARE RESTING FLAT ON THE COOLER FRAME. NOTE THE RESULTS IN THE REMARKS.
5.
CONDUCT A~ COMPLETE INSPECTION OF THE CONTAINMENT AIR
/
COOLER FOR CLEANLINESS, CORROFION AND LEAKS.
RECORD OBSERVATIONS IN THE REMARKS SECTION.
6.
INSPECT ALL SERVICE WATER CONNECTIONS FOR LEAKS.
/
7.
INSPECT FAN SCREEN COVERS AND SCREEN FASTENERS.
/
8.
INSPECT FAN BLADES FOR ANY OBVIOUS DEFECTS IN THE
/
MOTOR'NOUNT AND BLADES.
1 9.
CLEAN WORK AREA AND TDRN IN TAGOUT.
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4 NCLR400 NUCLEIS 02/26/1999
- Paga, 3 of 4
i l'.hscklist ID: MPM09150 Class Type: R l
10.
GENERATE AN IR TO CORRECT ANY PROBLEMS NOTED.
/
11.
SEND A COPY OF THIS COMPLETED CHECKLIST TO THE
/
SYSTEM ENGINEER.
10.
GENERATE AN MO TO CORRECT ANY PROBLEMS NOTED.
/
11.
CLEAR TAGOUT.
_/
COMPLETED BY g gg DATE REMARKS i
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NCLR400 NUCLEIS 02/26/1999
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...........................-- OTHER REFERENCES.------ -.........-.-.......
Reference TD Tvoe Descrintion Latest Revision Rrsp Name: HARRITT Minr Chng Dt:
Rev:
0 Date: 11/22/1994
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NCLR400 NUCLEIS 02/26/1999 Paga 1 of 3
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E43053 JOHN
. RYCYNA Dest Name:
n888 Num Copies:
1 MPM09151 Checklist-ID:
~
6 NCLR400 NUCLEIS 02/26/1999 Checklist Chscklist ID: MPM09151 Class Type: R
MAINTENANCE REQUIREMENTS AND SPECIAL INSTRUCTIONS -----------
D:2cr: INSPECT CONTAINMENT AIR COOLERS TOOLS FLASHLIGHT 2 10' STRAIGHT LADDERS DROP LIGHT AND EXTENSION CORD INSPECTION MIRROR
- NOTE *****
OBSERVE REQUIREMENTS OF CH-1-102 AND MN-1-109 1.
PICKUP TAGOUT.
/
2.
INSPECT THE CONTAINMENT AIR COOLER BOOT LOCATED BETWEEN
/
THE FAN COIL AND THE AIR COOLER ENCLOSURE.
INSPECT FOR TEARS, HOLES OR DETERIORATION.
NOTE RESULTS IN THE REMARKS.
3.
INSPECT THE CONTAINMENT AIR COOLER BOLTS (THERE ARE
/
72 BOLTS,-6 PER COIL).
INSPECT TO ENSURE BOLTS ARE INSTALLED AND NOT LOOSE.
NOTE RESULTS IN THE REMARKS.
4.
INSPECT THE CONTAINMENT COOLER COILS TO ENSURE THEY
/
ARE RESTING FLAT ON THE COOLER FRAME. NOTE THE RESULTS IN THE REMARKS.
5.
CONDUCT A COMPLETE INSPECTION OF THE CONTAINMENT AIR
/
COOLER FOR CLEANLINESS, CORROSION AND LEAKS.
RECORD OBSERVATIONS IN THE REMARKS SECTION.
6.
INSPECT ALL SERVICE WATER CONNECTIONS FOR LEAKS.
/
7.
INSPECT FAN SCREEN COVERS AND SCREEN FASTENERS.
/
9.
INSPECT FAN BLADES FOR ANY OBVIOUS DEFECTS IN THE
/
MOTOR MOUNT AND BLADES.
9.
CLEAN WORK AREA AND TURN IN TAGOUT.
/
a NCLR400 NUCLEIS 02/26/1999 Paga 3 of 3 :
Checklist Chacklist ID: MPM09151 Class Type: R 10.
GENERATE AN IR TO CORRECT ANY PROBLEMS NOTED.
/
11.
SEND A COPY OF THIS COMPLETED CHECKLIST TO THE
/
SYSTEM ENGINEER.
I I
COMPLETED BY
/
SIGNATURE DATE REMARKS
............................. OTHER REFERENCES ----------------------------
Reference ID Tvoe Descrintion Latest Revision RKp Name: MARRITT Minr Chng Dt:
Rev:
0 Date: 11/22/1994 EP/0I *cl 9P69 S6P OTP DIM 133n31500 3 MOW 11108 EE:0I 6661-6i-80W
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N discussion for the non-plausibility of Fatione for the Polar Crane in the Cranes and Fuel Handling AMRR relies on an comparison of the maximum allowable stress of ASTM A36 steel vs. stress range limits provided in CMAA #70, the goverrung standard for crane design. N Staff did not agree with BGE's reasoning in this plausibility decision and requested that BGE justify either that the stress will not go beyond the allowable stress of ASTM A36 or that the total number of cycles is within the allowed range.
BGE now chooses to demonstrate the latter and will no longer base the non-plausibility of Fatigue on an evaluation of the allowable stress ofASTM A36 steel vs. stress range limits.
CMAA #70 requires that crane members and fasteners subject to repeated load be designed so that 1) the maximum stress does not exceed 17.6 ksi (less than tbc allowable stress of ASTM A36 used in the above discussion) and 2) that the stress range for various suMaaaent configurations does not exceed the allowable values given in a Table. For cranes ofService Classification "A", the applicable class for in-scope CCNPP cranes per Bechtel Specification 6750-C-42, the Table stress range values are predicated on the number ofloading cycles being between 20,000 and 100,000.
Design Calculanon C-93 164 projected that the CCNPP Unit 1 Polar Crane components had experienced 8460 load cycles from initialinstallation until 1992. ' Ibis evaluation contained the conservative assurnption that each lift resuhed in four stress cycles to the components. Extending the projection to year 2034, it is estimated that the Polar Crane will experience 13,860 load cycles. This is much less than the maximum of 100,000 assumed for the purposes of d**Ialag the allowable stress range. Since the crane is not expected to exceed the onginally assumed number ofloadmg cycles, the original design remains bounding, and Fatigue will be considered not plausible. A similar projection could be made for the Unit 2 Polar Crane.
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RAI 4.3.11 and 4.3.18 BGE is working to develop data through industry research to determme the susceptibility of RV internals components to irradiation assisted stress corrosion cracking. Until the data and analyses that indicate IASCC is not a potentially relevant degradation i
mechanism become available, BGE will perform enbeed VT-1 inspections to detect i
cracks (if any occur) in the components believed to be potentially most susceptible to IASCC as well as neutron embrittlement. The inspections will be performed as part of l
the 10 year ISI inspection program during the license renewal term. Plant specific justification will be provided to the NRC in the event the analyses and data support elimination of the inspection.
The items selected for anhed VT-1 inspection are the re-entrant corners of the core i
barrel inside surfaces (the core barrel surface that faces the core). These corners are constructed by welding annealed 304 stainless steel plate. The residual stresses due to weldine while limited to the low yield strength of the annealed plate, are potentially higher than at any other stainless steellocation on the inside surface of the core barrel. In l
l addition to potentially being the highest stressed location, the re-entrant corners are believed to also receive the highest fluence. This is qualitatively determmed; the re-entrant corners project in towards the core, between two adjacent fuel bundles, so receive neutron exposure' from 270'. Being closer to the fuel than any other stainless steel components, the core barrel plates and comers are exposed to hot leg temperatures on one I
side. On the other side the environment is coldleg temperature. Due to the proximity of fuel, gamma heating is also expected to be higher at these locations than at other potential I
l locations. Due to the combination ofhighest stress, fluence, end temperature, the re-entrant corners of the core barrel, intended for enhaned inspection, are the most likely location for IASCC to occur, ifit occurs at all and for embrittlement to be manifested, if l
it occurs at all.
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Regarding SVP baseline walkdowns and additional baseline walkdowns:
q BGE did not exclude any supports from aging management programs based on SVP e
Walkdowns. Baseline walkdowns do not exclude any supports from further aging management.
l Table 3.1 4 identifies the aging management programs for each support with plausible aging effects. Table 3.1-4 clearly does not credit SVP Walkdowns as an aging management program.
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' _ EP'f T101 7, g 3 1-9 re:
Temperature monitoring of reactor cavity Reactor cavity temperatures are currently monitored by thennocouples. These provide inputs to alarming recorders in the control room. A point which goes out of spec high produces a main annunciator alarm to alert the operator to an abnormal condition. In this way, CCNPP personnel are alerted to higher than normal reactor cavity temperatures and an appropriate response considering the affected equipment can be developed and implemented in a timely manner. However, cavity cooling is not in the scope oflicense renewal and BGE does not credit this tempetotuu monitoring as an aging management program.
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FEB-24-1999 12:49 BALTIMORE GAS & ELECTRIC 410 495 6946 P.03/04 2/23/99 ik
/F f 1. Group 1 - RAI 3.1.6.c Where are structural steel members for HVAC ducting supports identified?
Response: BGE provided clarification by referring to RAI respnse 3.1.9. Additionally, Errata to Section 3.1, Component Supports, dated November 19,1998, provides amplifying information in the third, fourth, and sixth bullet. Specifically, under the component support 5
y category of HVAC Ducting Supports are two Component Support Types. They are HVAC
- j Ducting Supports Inside Containment and HVAC Ducting Supports Outside Containment.
Both of these groups include structural steel members for HVAC ducting supports. As I
indicated in response to RAI 3.1.9, aging management for structumi steel for HVAC ducting '
supports is included with the Group 2 discussions on pages 3.1-23 through 3.1 29.
J tatus: BGE thinks this item is closed.
I
- 2. hroup 1 - RAI 5.17.1 TPR 01 vas wrinento include NS SR piping in the S SR I ortion of ARDI for e/c?
is will I e reflected inthe ann update.
Action: Don Shaw H willthis eange in program be uni ed toNRC7 Response:Through e AnnualUp ate.
Status: BGE to ao unicate ourr sponse.
i 3.
3roup 2 RAI.3.36 Why do we mileve that dwater chemintry has rem '
stable
- ince plant etion?
'on: Bob alk Draftresponsei s el
- cation and additiona site ews. Alsc e lain why centrations of sulf s and orides in fourth well so ighrelativ to or ' nary sa water concentrations.
Statu -
C 4. Group 6-RAI3.1.14 Are allring foundations WSLRinISIProgram?
Response: LRA Table 3.1 2 identifies the associated systems for component support group ring foundations for flat bottom vertical tanks. System 036 (Auxiliary Feedwater System -
Caad-anate Storage Tank 12) and System 37 (Demineralized Water and Condensate Storage System - Condensate Storage Tanks 11 and 21) have their supports inspected as part of the
)
ISIVicy Status: R==aae pr= ped. Communicate tn NRC, 5.
'O 3.3.15 Exp de
- - M sand dome r
IL talfu. doc W
e em -
s, l-a s
- \\
(
RAI 3.3.17 - Additional tendon force loss due to elevated temperatures g
BGE Action: Explain how STP M-663 bounds any types of additional tendon force loss (8 to 14%) due to elevated temperature as described in NUREG-1611.
BGE is not committed to NUREG-1611. The affects of elevated temperature on tendon force due to abnormal sun exposure or proximity to hot penetrations will be detected by STP M-663.
Operability of the containment, as it relates to tendon force, is governed by the requirements of Calvert Cliffs Technical Specification 15.6.1. Testing is performed by STP M-663. A reduction in tendon force will be detected during conduct of the test regardless of the cause.- If the Technical Specification requirements are satisfied, then by definition the containment is operable.
(
r J.
4
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RAI 3.3.24 - Containment horizontal liner BGE Action: Describe the material condition of the concrete floor of the containment, relative to cracks, in order to validate the assertion that it will prevent water from coming into contact with the inaccessible horizontal steel liner.
The containment system engineer performs a walk-down and visual inspection of the containments, including the floors, every outage when they are open. No cracks in the 18 inch thick, steel reinforced, high density concrete floors have ever been discovered that were considered large enough to affect the liner.
I.
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<4 e
e ARDM: Thermal Embrittlement
(.
Components: CEA Shroud Tube (ASTM A451 Cast Austenitic Stainless Steel)
Rationale: This ARDM is considered plausible for this subcomponent of the CEA Shr,ud. See the respora to question 4.3.9 and pages 4.3-19 through 21 of the LRA.
i Components: CEA Shroud (All other subcomponents) (ASTM A182 Alloy Steel, ASTM A286 Iron Base i
Superalloy AMS 5735, ASTM A479 Stainless Steel, ASTM A240 Chromium and Chromium Nickel Steel, ASTM A276 Stainless Steel, ASTM A269 Austenitic Stainless Steel)
Rationale: This ARDM is not plausible because the component is not fabricated of Cast Austenitic Stainless Steel. The operating temperature of the RV internals system is not sufficient to cause thermal aging for the component matenal
References:
NUMARC 90-05 (also known as EPRI TR-103838, PWR RPV Internals LR Industry j
Report), EPRI NP-5775 (Environmental Effects on Components: Commentary for ASME Section III),
NUREG CR 6048 (PWR Reactor Internals Agmg Degradation Study).
i l
C
ATTACHMENT (1)
APPENDIX A - TECHNICAL INFORMATION
[
4.3 - REACTOR VESSEL INTERNALS SYSTEM
(
Cm. hug experience relanve to the ISI 1"ic iiuu at the CCNPP has been such that no site specific problems or events have occurred that required changes or adjustments The program has been effective in its function of perfonnmg exammations required by ASME Code Section XI.
Specific to the visual exammations for wear of the RVI, the ISI Program has been used in previous refueling outages to examme accessible surfaces This er.uug expenence shows a consistent ability to identfy surface indications, e.g., scratches and souges, for evaluation agamst =rPe-cntena For example, the 1991 examination records for the Unit 2 CSBA idennfy that gouges found dunng the visual exammations were ^=r-+Moned as mmor adications such that keyway serviceability is not affected
[ Reference 12]
The ISI Picy.m records show that the gouges were again found in the subsequent required exammations performed in 1993, and the report states that the gouges were noted in previous exammations and accepted as is. [ Reference 13] The minor nature of these indications, and the repeatability in findmg them, provides evidence of the effocavness of the ISI Piu,-u to identify the effects of wear.
Group 1 (Wear)- Demonstration of Aging Management Based on the factors presented above, the followmg conclusions i===i.w management of the effects of wear en the RVI 9- ;==:
k
'Ibe RVI --..;----+ provide structural support to the fuel assemblies, CEAs, and ICI, and their e
configuration must be mamramed under CLB conditions.
Wear is plausible for RVI e- ;=== and resuhs in matenal loss which could lead to loss of
=tanded function.
The CCNPP ISI Pic r-u provides for penodic visual exammations of accessible surfaces of RVI u
u.....= -..
Fv==ia** ions will be pr.ifosoed, and appropnate corrective acnon will be taken if significant wear is discovered Therefore, there is reasonable assurance that the effects of wear will be nanaged in order to maintam the structural mtegnty of RVI consistent with the CLB durmg the period ofenanded +A Group 2 (Neutron Embrittlernent)- Device Types, Materials, and Environment Table 4.3-2 shows that neutron embrittlement is plaustile for nine device types in the RVI. This group of device types includes: CEASB (except the spanner nuts and tabs), CS, CSTR, CSB (except the upper flange), CSC, CSP, FAP, FP, and LSSBA.
lhese components are constructed of vanous namwe and alloy steels (ASTM A-182, A-193, A-194, A-240, A-269, A-276, A-351, A 451, and A 479; and Aerospace Matenal Specificarice (AMS) 5735 iron base superalloy A-286). [ Reference 3, CEASB, CS, CSTR, CSB, CSC, CSP, FAP, FP, and LSSBA, Aarh=aa' 3s]
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Application for License Renewal 4.3-13 Calvert Cliffs Nuclear Power Plant I
.t.
)
Matrix Code List (Revision 3)
SYSTEM NUMBER:
084 SYSTEM NAME: Reactor Vessel Intemals Date: 12/10/98
(
DEVICE TYPE: Hold Down Rirw EQUIPMENTTYPE: Structural Sunnorts l CODE l-DESCRIPTION -
1-SOURCE l
l 10 Resistant materials.ed proper chemistry control make general corrosion not NUMARC 90-05 plausible for this device type. Austenitic stainless steel, alloy steel and nickel-based alloys are quite resistant to general corrosion in a benign operating NUREG CR-6048 l
enviro unent. A benign operating environment is assured by proper water f
chemistry control through CP-204-2.
BG&E CP 204-2 11 The amount of oxygen chlorides in the surrounding fluid is NOT sufficient to NUREG CR-6048 cause Pitting / Crevice Corrosion. Pitring is not a concem in the RV internals environment unless the reactor has a history of long outages without proper NUMARC 90-05 chemistry control. Water chemistry is controlled for the RV Internals during plant operation and during outages by chemistry procedure CP-204 2.
BG&E CP-204-2 Derefore, Pittina/ Crevice Corrosion is not plausible for this device type.
13 The high operating pressure, relatively low fluid velocity and low level of NUMARC 90-05 particulates in the RCS fluid ensure that erosion is not plausible for nickel-based alloys, high alloy steels and stainless steel components in the PWR NUREG CR-6048 environment. De operating pressure,2250 psia, precludes cavitation erosion, and the purity and particulate' control of the reactor coolant eliminate the BG&E CP-204 2 possibility of particulate erosion. Derefore, erosion is not a plausible ARDM for this device type.
14 De high operating pressure, relatively low fluid velocity and low level of NUMARC 90-05 particulates in the RCS fluid ensure that erosion / corrosion is not plausible for nickel-based alloys, high alloy steels and stainless steel components in the PWR BG&E CP-204-2
)
environment. Derefore, Erosion / Corrosion in not a plausible ARDM for the device type.
l 16 De operating temperature of the RV Intemals system is NOT sufficient to NUMARC 90-05 l
f cause thermal aging for the materials being used for this Device Type. Stainless
\\
steel, nickel-based and steel alloys are not susaepGle to thermal aging in the NUREG CR-6048 temperature range of the RV Internals System.
EPRI NP-5775 17 The amount of available hydrogen in the surrounding fluid is NOT sufficient to EPRI NP-5775 cause hydrogen damage for austenitic stainless steels, low alloy steels and nickel-based alloys. Above 4000F hydrogen diffuses rapidly in steel and is BG&E CP-204-2 eliminated by off gassing. Hydrogen damage is not plausible since the J
temperature of the RCS is greater than 4000F and the hydrogen can easily diffuse. Tight water chemistry control of hydrogen at lower temperatures through CP-204-2 also helps to assure that hydrogen damage does not occur durina shutdown periods.
19 His Device Type does not depend on pre-load for functionality. Stress NUMARC 90-05 relaxation is not plausible for components that do not depend on pre-load for functionality.
19.2 Radiation levels in the area of the hold down ring are not sufficient for this DG-1009 ARDM to occur. Because the hold down ring is preloaded during installation of the reactor vessel head, stress relaxation needs to be considered for this NUMARC 90-05 component. In-pile testing of stainless steel materials has shown that l
substantial loss of pre-load is possible at PWR operstmg temperatures in a high EPRINP-5775 radiation field (~5x10" n/cm', E>l MEV) when the matenals are stressed at or above yield stress. However, extrapolating fluence values from a CE CE Memorandum dated memorandum dated August 4,1977 " Relaxation of 13Cr-4Ni Hold Down Ring August 4,1977 l
Material", fluence levels of the hold down ring will be in the range of 10" to 10 n/cm' for 60 year life. Consequently, loss of pre-stress is not a plausible ARDM for this component.
20 SCC /lGSCC is not plausible for this device type d.ie to non-susceptible material NUMARC 90-05 (Alloy Steel and Nickel Based Stainless Steel), lack of high tensile stresses and tight control of water chemistry through CP-204-2.
BG&E CP-204-2
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Page1of2 i
N LA5VUttoAshcnSB935W105 N
v'.VS112 l
4 Author: MARV BOWMAN at -NSSDP0 Date:
2/23/39 4:07 FM g 'ormal f
S
{
DONIS L SEAN, JOHN RYCYNA g
( sject: NRC Question on Secondary CPemistry
.................................--- Message contents Don: Bere is the answer to the " late" question. The addition is g
downstream of the condensate domineralizers since they remove the volatile chemicals used. The chemicals make their way through the condensate piping, feed piping, feedwater heaters, etc to the s/G. Since the chemicals are volatile, they are carried through the s/g, steam lines, turbine, condenser, and back around.
Forward Esader
Subject:
NRC Question on Secondary Chemistry
{
f Author: BRIAN C BATES at.NESDPO Date:
2/23/89 9:50 AM l
Regarding the steam generator chemistry iser The normal method for addition of the all. volatile chemical additives
)
into the secondary is by pumping them into the condensate system downstream of the demineralizers. There are also alternatives to line I
up the chemical addition system to the aux feedwater system or directly to the steam generators, however, these methods are used very infrequently.
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Calvert CliNs NuclearPowerPlant Administrative Pacedure STRUCTURE AND SYSTEM WALKDOWNS
\\
MN-1-319 Revision 2 ENective Date
~/d ~ih Tech Spec-Related Managensent Related X
Writer: T. W. Singer Sponsor: E. "R" P*
En eer-Malatenance Comiponent Engineering Unit x /
/0 97 Approved:
n =_--
n,,.
Snclosure 4
MN-1-319 Structure and System Walkdowns Revision 2 Page 2 of 50 RECORD OF REVISIONS AND CHANGES Revision Change Summary of Revision or Changes 2
Section 1.2.E. Added reference to Attachment 14, I
Refueling Equipment Walkdown Report - Refueling Machme Walkdown Report.
Sectxm 7.0. Added A*=4w 14 as a record Added A*a4* 14.
e Struaure and System Walkdowns MN-1-319 Revision 2 Page 3 of 50 TABLE OF CONTENTS SECTION TITLE FAGE 1.0 INTROD UCTION..........................................................................
1.1 Purpose............................................................................................6 1.2 Applicabihty/ Scope.................................................................................. 6 4
j 2.0 C
I
................................................................................................7
\\
2.1
.....................................................................................7 2.2 Performance.........
.................................................................... 8 3.0 DEFINITIONS.........................................................................................................8 4.0 RESPONS IBILITIES.................................................................................
5.0 P
ESS..............................................................
.................................8 5.1 Walkdown L yo.w...............................
........................................ 8 5.2 Walkdown Objectives..........................................................................s..... 10 5.3 Stmeture Walkdown.............................................................................. 12 5.4 System Walkdown................................................................................. 13 5.5 Piping Inspection Notes...........
...........................................................14 5.6 Outage Work Sccpe Changes........................................................................ 15 6.0 BASES...............................................................................................................15 7.0 RECORDS......................................................................................................16 An=d __* 1, Mechanical S Walkdown l
...........................................................17 Ar ' - =t 2, Elecencal System Walkdown Report.................................................
............... 18 1
Anwhment 3, Structure MM;u.
Walkdown; C4 d.ui Structure...............................................19, Structure Monitoririg Walkdown; Concrete Stmetures Other'Ihan Co
.m.............. 22 Attadiment 5, Structure Monitoring Walkdown; Masonry Walls................................................... 27, Structure Monitoring Walkdown, Intake Structure....................................................*a Anachment 7, Structure Monitoring Walkdown; Buried Anchnrages, Pipe Supports, Ushipvuod Cathodic Protection System, And Buried Piping.............................................. 33 A*ew 8, Structure Mesuring Walkdown; Steel Structures And C+- =++:........................... 35
o Structure and System Walkdowns MN-1-319 Revision 2 Page 4 of 50 TABLE OF CONTENTS SECTION TITLE PAGE A*=^=t 9, Structure Monitoring Walkdown; Storage Tanks............................................ 3 A*=Amt 10, Structure Monitoring Walkdown; Dams, Embaabnanu, Canals, And R*=iaing Wa 1, Structure Momtonng Walkdown;1arge Equipment Supports And Anchorages Seismic Gaps................................................................................
A*=4mant 12, Walkdown Report G "
" ion Sheet..................................
......................43 A*=Ama'* 13, Pipe Support Inspection Guidelines........................................................... 4 A*=d-at 14, Refuchng Equipment Walkdown Report............
....................................... 48 l,
f MN-1-319 Structure and System Walkdows Revision 2 Page 5 of50 LIST OF EFFECTIVE PAGES Page No.
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MN-1-319 Structure and System Walkdowns Revision 2 Page 6 ofD
1.0 INTRODUCTION
1.1 Purpose The purpose ofthis procedure is to walkdowns and reporting /&+-
provide dirwtice for sfu.auig structure and system
---g the walkdown results. ' Ibis procedure meets the sq' uu. ;. for evaluatmg structure and system matenal condition m a.w d.sce wi:h the h L;"=e Rule at Calvert Clifis Nuclear Power Plant (CCNPP).
1.2 Applicability / Scope This procedure applies to all structure and system #t ; fvio.ed by Plant hgi=-sh.s Section g-- -- s 'Ibc level ofdotad may vary based upon the purpose of the walkdown, bat the general mtemt and method of reposting walkdown results are standardized withm this procedure.
'Ibe assessment of the condition ofstructures and systems that are in the scope ofthe Mastenance Rule shall be pio.ud using this procedure.
A**-h==h to this Procedure A.
A*=4-* 1, Mechanical System Walkdown Report, and A*=4m=* 2, Electncal System Walkdown Report, are ineandad to capture the results of walkdowns on _=---L=hl and electricalsystems
\\
B.
A*=hmaats 3 through 11 are intended to capture the resuhs of both basehne and ongoing w=*c.; on structures at CCNPP that have been included in the scope of the M-..... r-v Rale. 'Ihe attachments are applied on the basis of structure Maa_=
(functional ='
i) as follows.
=-
j i
1.
Containment Structure - applies to the interior and extenor of the ocatainment
{
building (Attachment 3, Structure Mo iu.ii., Walkdown; C=*=6=aat Structure).
2.
Concrete Structures other than C-+=6-aa'-
lies to Auxdiary Buildmg, Fuel Handling Building, Spent Fuel Pool Areas, Generator Buikhngs, etc.,
(Attachment 4, Structure Monitonns Walkdown; Concrete Structures Other than C~+= 6=aa').
3.
Mascary Walls - applies to any walls that serve to separate equipment important to safety or that support equipment imyu.uun to safety (Attachment 5, Struaure Menitoring Walkdown; Masonry Walls).
t 4.
Intake Structure - applies to the structure thet houses the equipment that serves as the source of water for the ultunate heat sink (Sak Water Pumps and Salt Water Bays) as well as the source of.L...t-..og water for the =wlanaar (Attachment 6, Structure Mo c.oih4 Walkdown; Intake Structure).
5.
Buried P Pipe Supports, and Eqmpment Anchorages - applies to all buried i
Pping as Lormi.st to safety, cathodic protection systems, pipe supports that are attached to equipment and piping classi6ed as io ees.ui to safety, and any anchara8es that are used to secure equipment
,~.i.ut to safety (A*-h=aat 7, Structure Monitormg Walkdown; Buried Piping, Pipe Supports, and Anchorages).
6.
Steel Structures and C-a~* ions - a Steel Strucsures and C panels (A* pplies to cranes, cra structures, and blowout h_maat 8, Structure Monitormg W a~+icas).
3 A
s.
Structure and System Walkdowns MN-1319 Revision 2 Page 7 of 50 1.2 Applicability / Scope (Continued)
- 7. Water Storage Tanks - applies to any fluid reinmg tanks that serve pment
)
t to safcty (A*aamaat 9, Structure Monitoring Walkdown; r Storage R. Dams, Emba=%, Pei=ing Walls, and Canals - applies to any features of the facilit a&+y intended to contam, restram or channel water, other fluids, or soil from ag the functions of equipment important to safety (An -t =^ 10, Structure Monitoring Walkdown; Dams, Emba=baa* R*iain-Walls, and Canals).
- 9. large Equipment Supports and Anchorages and Seasmic Gaps - applies to supports for such equipment as steam generators, reactor vessel, and NSSS piping. Sessmic gaps refers to those features of the civil constmetion that allow for sway and movement during seismic events (A*=4maat 11, Structural Mce.iim g Walkdown; Large Equipment Supports and Anchorages and Seismic Gaps).
C.
Att=A aaat 12, Walkdown Report Continuation Sheet, is a enahanahan siret that can be used with the Walkdown Record aheets for A*=^=a* l-11.
D.
An=ehmant 13, Pipe Support inspection Guidahnee, provides guidance based on a CCNPP enpneermg standard (ES 002) for the assessment of pipe support functionality. This guide can be used in conjunction with any of the walkdown attachments to damaa* ate the capability of the pipe supports associated with a specific walkdown.
E.
A*=4mant 14, Refueling Equipment Walkdown Report - Refueling Machme Walkdown Report. Pmvides gaiA== for p.fvuulug a visual or video inspection of Fuel 14andhng Equipment
2.0 REFERENCES
2.1 Developmental A.
MN-1-101, Control of Maintenance Activities B.
MD-1-100, Tennporary Alterations C.
OM 1-300 Octage Ma=-; --=
D.
MN-1-112, M--ing System Performance E.
PR-1-100, Preparation and Control of Calvert Cliffs A?
4.Jve Procedures F.
A +- ' t=-.Gve Pmcedures Writer's Manual G.
10CFR50.65, Requirements for Monitoring the Effectiveness ofM=intan== At Nuclear Power Plants
- f.
Reg. Guide 1.160, Monitoring the Effectiveness of M=L
= e At Nuclear Power Plants I.
NUREG-1526, I.esson 1 *=raad from Early Implementation of the Maintan== Rule at Nine Nuclear Power Plants J.
NUREG-1522, Assessment ofInservice Conditions ofSafety-Related Nuclear Plant Structures
F~
f MN-1-319 i
Structure and System Walkdowns Revision 2 i
Page 8 of 50 f
2.1 Developmental (C-- tmued) l K.
NRC Inspection Procedures #62706 - Maintenance Rule, #62707 - Maintamance Observation, and #62002 -In==*= of Structures, Pr.ssive Components, and Civil F=p=e mg Features at Nuclear Power Plants L.
NUMARC 93-01, Industry Guideline fer MME-Q the Effectiveness of M--
=== at Nuclear Power Plants M.
NE! 96-03, Guideline for Morutonng the Condition of Structures at Nuclear Power Plants 2.2 Performance A.
EN-1-100, Fm Service Process Ovemew B.
ES-002, Pipe Support inspection Stanclard C.
ES-005, Civil / Structural Design Criteria D.
ES 040, Piping Design Cnteria E.
ES-042, Piping Functional Evaluation Criteria F.
ES-201, BGE Environmental Standard G.
MN-1-112, M=aapag System Pw'ormance H.
NRC Region IIntegrated Pwfus ws A==se Team Inspection Report 50-317/91-82 and 50-318/91-82. Dated January 1992. (CT9009554) 1.
OM-1-100, Managmg Outages J.
PEG-13, System Turnover K.
QL-2-100, Issue Reportmg and A=====^
L.
QL-3-102, Program Deficiency Report Program 3.0 DEFINITIONS None.
4.0 RESPONSIBILITIES Responsibilities are as directed in the Promss section of this procedure.
5.0 PROCESS 5.1 Walkdown Purposes A.
System Engmeers shall periodically walkdown their structures and systems as manslated by structure or system performance, plant operating conditions, scheduled material condition assessments for the Mai==-:e Rule, or as required by the responsible Principal Fmgia**r.
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,e Structure and System Walkdowns MN-1-319 Revision 2 Page 9 of 50 5.1 Walkdown Purposes (Continued)
B.
Structure and system walkdowns shall be conducted every refueling outage to assess the matenel condition of those structures included in the scope ofthe M=ia'aama~ Rule program at CCNPP. 'Ibcpr or system is capable ofpene. pose of these walkdowns is to establish whether the s
--og its intendad function. The structural system engmeer (System 102) is responsible for gifvio.ing/monitonag these walkdowns and ensunng that the requirements of the Maiataa a~ Rule are followed.
C.
An Le of 11 FOST (Fuel Oil Storage Tank), shall be r=incted at least once per menth aw.d-s to ES-201. The dLsel oil system engmeer (System 023) is i-yonsivic for gifvio 'us this walkdown and demnaahag it on Attachment 12, Walkdown Report Continuation Sheet.
D.
During the performance of a walkdown the matenal madi*iaa of structures, systems, and sie--4 (SSCs) should be assessed such that any degraded condition will be id= Mad,
& m==% cause L.
4, and corrective action initiated before the 4. inion proceeds to failure of the SSC to perform its intendad fua~
E.
'Ibe scope of a structure or system walkdown may vary based on the condition of the plant and the mtentions of the System Engmeer. The duration and level ofdetail of a walkdown will depend on the intent. Followmg are reasons for which a walkdown could be conducted:
1.
MmE=e Rule (SSC) Material Condition A=-==-ts i
The Maintaanne* Rule requires naadWa monitonog to be p. fun.d to ensure that the material condition of a stmeture or system is maatained For structures i
that are in the scope of the Maintenance Rule at CCNPP, a walkdown that hmises the status of the materiel condition of the structure or system may be theonly racticalwaytoar- =.L5 that the structure is capable ofperforming its int -
functions.
2.
System Pa dH Review A walkdown should typically be gifv ud toward the end of a major outage to venfy the readinant of the system to function. Some uipment may bein service, but equipment is typically not at normal operatmg tions.
3.
Start-up Review A walkdown should be gifvi.id when the system is initially pressurized, energized or placed into normal service. This is typically gu,ve.d at Mode 5, but may vary.
4.
System =p=i Fannhanzation r
A--:+; =-v of a new assignment acconhng to PEG-13, System Turnover, shouldinclude a of the system This will aid in familianzation with equipment locations, arrangements, operating temperature, noise levels, process poir.a.c rs, and matenal condition l
o MN-1-319 Structure and System Walkdowns Revision 2 Page 10 of 50 5.1 Walkdown Purposes (C-dag 5.
Pre-Outage Review Subsequent to a major, planned outage, a walkdown typically will help venfy tim completeness of the h.ceted system problems. An in-plant review of both system and major equipment perfvie.esce char-iA will help to identify additional outage work scope and may preclude subsequent forced outages 6.
Job-Specific Walkdowns Tob S c walkdowns can be performed as requimi, to witness work or cations being - for. d. on the s a
process as detailed in EN-1-100, !=
- ystem, or as part of the modification r
w Service Process Overview. This walkdown shouki be namm in scope and focused on the work or maMcaten at hand.
7.
Peruviic Walkdown At regular intervals (typically monthly, or as negotiated with the responsible Principal F=k=)the System h i-shall a walkdown of the s The extent olthe walkdown shall be withthePrincipalh=a.ystem.
"Ihe intervals and extent of the periodic eballbe documented to the GS-PES in a memc,. Specific areas walked down shall b noted on Attachmaat 1, Mechamcal System Walkdown Re Walkdown Report, as applicable. port, or Attachment 2, Electncal Sy record anynew or condition that could prmimt the system or a componet from - -,mg its '
Aeiaa. Any or all of the above mentioned (steps 5.1.B.1 through 5.1.B.5) can be accomplished n past of the Penodic Walkdown 5.2 Walkdown Objectives A.
'Ibe objectives of Structure or System Walkdowns include.
1.
Fei==%
The primary objective of a structure or systet walkdown is to assess the
= W of SSCs, ensure the safety and power generation functionahty of the SSCs (i.e., the system will wform when called
),and ensure that the SSCs.
'Ihis includes identifymg missmg components, tags, Issue Report tags, incomplete mamtenaner, temporary r.-: U-Miaa=, unavailable support systems, dessmod egmpment, or other tags that are potential indicators of A=c+iaa-E*y con arns.
2.
SSC Stress or Abuse
- Ihermal insulation damage, bent or broken hangers, distress to equipment anchorage, excess piping motion or vibration, and damaged tstag or flex conduits are examples of unusual noises, paea*i=! stress or abuse to the system. Excessive vibrations, excessive temperatures, discolored fluids, relay chatter, idW of flow through closed valves, external leakage of fluids, corona discharges, or are paths are some examples of aa'aad=1 equi purpose of this objective is to assess the condition of SSCs. pment stres= The I
a
u Structum and System Walkdowns MN-1-319 Revision 2 Page 11 of 50 5.2 Walkdown Objectives (Contmuod) 3.
Safety / Fire Hazards Broken doors orhardware; *
,ropriate breaches of fire or flood barriers; broken, loose, or missing devices; personnel injury hazards; and missing equipment guards are examples of safety or fire hazards 4.
GeneralHad~ ping Condition Debris, c'=2:-- s, condition of painted surfaces, lighting, inapprepnate use or storage of materals or tools, covers not in place, and unreadable, out of date or nussmg labels and signs are examples ofhoue + Exclusion (
-d-concerns. Poor housekeeping can lead to loss of Foreign Matenal 5.
Cw%= Adverse to Quahty Conditions adverse to quahty are conditions outlined in QL-3-102, Program Defkisy Program, and are first @==*M by Issue Reports mi4 to QL-2-100, Reportug and A==~ --*
6.
th =' -d-iTanporary Alterations An==ih Temporary Alterationis any configuration change that has not been evaluated for operational tab an approved process. Ifsuch conditions are identified during the following course of action shall be followed.
Verify that the condition is an unauthonzed t ; i.iy alteration by a.
consulting MD-1-100, Temporary Alterations.
Design Fh orpnintion should be consulted for guidance b.
If the condition is concluded to be an unautbrized temporary alteration, A- - =; t'ne condition on an Issue Report asia-g to QL-2-100.
After the condition is A - =Al on an hoe Report, perform one of the c.
following, as applicable:
(1)
If the condition is not required for continued safe operation, IMMEDIATELY have the unauthonzed temporary ahoration removed.
(2)
Ifthe r=deion is for =*i=M safe operation and is i.yer y in nature, IATELYinitiate a tanporary alteration accordung to MD-1-100.
(3)
If the <=d% is required for <=+i=M safe operatxa and is permanent in nature, IMMEDIATELY initiate an Engmeenng Service Process (ESP) accordmg to EN-1-100, Fg=;g Service Process Overview.
Structure and System Walkdowns MN-1-319 Revision 2 Page 12 of 50 5.2 Walkdown Objectives (Continued) 7 Structure Degradation Ifstructure 3n'-.non is observed during the pefoi.. r.ce of a structure walkdown, a should be issued to ensure that cause is L-J-- ' and corrective action is initiated. Addstionally, the assessment of matenel condition performed after the walkdmvn is completed will serve as the L =% tion of
= m ble p i' the Maintaannem Rule and will be used to disposition
- - = =
the structure as either 1)or(aX2).
a.
Ifa structureis '~
- -=-i to be (a)(1) all of the req 6
.-s of MN-1-112 apply.
and matenel e-I* inn ass)essments to ensure that the co I
b.
i effective and that the graded matenal conditinn of the structure is arrested
)
or ehnunatad and the structure will be capable ofp fornung its intandM function.
5.3 Structure Walkdown A.
The intent of the structure walkdown is to perform a visual inspection of the structure for the purpose of deterrmamg if any s i.4id.on that is noted could ~*aa+ially challenge the u
abihty of the structure to perform its intended AW=. These w'alkdowns and the subsequent assessments ofstructure Wh meets the intent of condition monstonng as idaa*WA in NUMARC 93-01 for compliance with the requirement of the Mastenance Rule.
B.
The walkdowns shoukt be as intrusive as - - y to ensure idmtified deficiencies or structure degradataan will not result in structure failure or loss cf structure i=*aa.
n Followmg completion of the walkdowns the system shnli make a '--'- - = Mon of the structure's status from a gife... r.ce basis such t tim.tucture may be d imhi as either (aXI) or (a)(2) per the Maintaaa_=_Rukt C.
Ifstructure degradation or e=r== fadures are noted during the tvalkdown, the system engmeer should involve the Principal Faaiaaar - Maintanance C-aa en "== ~=-4 Unit (PE-MCEU) to ensure that the structure is capable ofperforming its intandeIfunction.
Degradation of structures or structure 9 ---g- =- :Jelements should be hmaatad and resolved through the Issue Report process.
D.
Ameasements of structure degradation should be based on previous Wh. This should consist of a compenson ofthe last walkdown with the results of the curret walk down to Am.-c if the degradation is dynanne or static. Dynanne degradatian should be the subject of an IR and shouki have a plan of corrective action. Typically degradation will proceed at a slow rate that over time will result in a failure or loss of structure Ai-*iaa. Therefore, if a dynamic=='=2== of dogmdation is involved, the structure should be consadored for (aXI) c'==g- :-w
a MN-1-319 Structure and System Walkdowns Revision 2 Page 13 of 50 l
5.3 Structure Walkdown (Contmued) 1 l-An example ofdynamic degradation of a structural element would be a safety-related pump pedestal that has extensive crackmg and the pump anchor / mounting bolts are pulling loose such that tbc pump is exhibiting abnonnal vibration and movement dunng surveillance tests. Static degradation of a structure would be on thatis arrested or that is pr~~% at such a slow rate that the =* a==1 ility ofthe structure n
would not be challenged. Crackmg of a concrete wall that shows no other indications of failure and is not a threat to other SSCs meetmg intended functions wocid be an example.
In either case good engmeerug practice and reviews of each individual mstance of structural degradation that has been idenh6ed should be applied.
E.
The performance of a structure walkdown will be <taenmanted by the use of the appropriate chackhet in the attachments section. The structure walkdown chackhets are Attachmante 1 through 11 and should be used in system engmeer to record the condition of the structures as noted during the F.
Upon completion of the structure walkdown, the system engm' eer and the PE-MCEU will determme if the performance of the structure is =~ap*=ble by reviewmg any irlenhEM agamst =~~p*_=hle limits cantamM in inthwhy standards, industry
, or design /hcense basis &-==M. This evaluation will provide the basis for the assesrment i
that the structure is either (a)(1) or (a)(2) within the context of the Mn
=e Rule.
l G.
'Ihe completed structure walkdown package ird" the observations and performance assessment will be forwarded to the Maintan=nceM Coordmator (MR.). The MR. w 1
schedule the package and the performance assessment results for Expert Panel review and approval. The system engineer and the PE-MCEU should attend the expert panel meetag and provide technical input during the review process H.
Dc->
- ::on associated with the disposition of the structure walkdowns will be retained by the MR for future reference and the results will be included in the (a)(3) periodic j
assessment report required by the Mh- -e Rule.
\\
1.
Structure walkdowns should be >fm 4 every refuehog outage and =chMidM to ensure i
that every structure at CCNPP will receive a walkdown u a nummum every third refuehog outage. Hence, a stmeture performance assessment will be gfmo.ed on each i
l structure once every 6 years (muumum). Each stmetural element that is a e=r=d of a given structure should be reviewed during the 6 year interval such that items such as expansionjoints, moisture seals, metal coatings and insulation that can degrade the overall perfonnance of a structure are assessed in a timely manner. Any extension of the
'Ibe attachm%.er.cy beyond the 6 year interval requires an approval from the GS-PES.
M
ents to this procedure provides a sampling of the types of structural elements l
to be==*=M 5.4 System Walkdown l
A.
The intest of the walkdown is to perform a VISUAL INSPECTION, not nare==rily a comprehensive physical inspection. If dunng the ineaartiaa. deviations from design l
conditions are noted or suspected, the System Fi=i-r shall further investigate the l
conditico. 'Ihe detail of the inspection and the mvestigation is lea to the discretion of the System Fi=ia B.
'Ibe System Fe=i-should normally be aware of changes to his system; if aare==ry, he should perform a review ofopen de uostation for the system prior to the Seid walkdown.
This should include the latest " system report card" (SSIP), open MOs, PDRs, approved temporary modifications, malor and minor moddications in progress, deferred PMs, previous System Walkdown Reports, and other open items or issues.
I tL-
MN-1-319 Structure and System Walkdowns Revision 2 Page 14 of50 5.4 System Walkdown (Continued) i C.
Conditions adverse to functionality, indicat ons of system or equipment stress or abuse, safety or fire hazards and housekeeping deficiencies shall be documented on A*=rh= ant 1, Mechanical System Walkdown Report; An=rkmant 2, Electrical System Walkdown Report; (and A*=rk-at 12, Walkdown Report r'aa+iaa=+iac Sheet), as reqmred, and an Issue Report,ifrequired.
D.
The System F=% may crganize a walkdown team as appropriate to report to the area to be visited. 'Ifiis may melude personnel from; PES, Maintenance, Operations, Fire and Safety, Bu' ' Services, Radiation Protection, ('hamimy, or other interested groups.
Craft support Id be considered appropriate to facilitate access to system equipment or to perform minor tasks authorized by rover MOs. All craft actions shall be according to stationprocedures Or.;;
practicsble, System F= j+.oca walkdowns are not prohibited, however, yhm mdividual can be anotlir System Fmei ged to do walkdowns in pairs. The second
- s are encoura doing a walkdown in the same area, or the Work Group leader. Plannmg for a walkdown should include special work permits and consihs;cis of ALARA, as well as an elevated awareness ofany trip-sensitive equipment to be cam =W. Walkdowns should bs scheduled for plant conditions that provide good indications of system Fear *iaa=lity.
E.
The walkdown team should visit as much of the system as was negotated with the I
Principal Fm iaaar (see step 5.1.B.7), thatis reasonably accessible. Consideration should be given to ALARA, safety, plant conditions, and other inspections of the same areas. All j
unusual conditions should be recorded on Attachment 1,2, or 12, as applicable.
F.
All conditions requinng corrective action nest be a~ a-a=ai~i by an " Action Taken" statement and 4:-:
^: I on the System Walkdown Regiort (Attachment 1 or 2) in the
" NOTES" section. Examples of Action Taken are ististion of an Issue rt and a GOLD card. Detaded 4:-: - - : :iaa such as an Issue Report number s also be hazards prior to walkmg away from the are,a.secorded on the attachments 24y action should be taken for safety G.
Areas visited during the walkdown shall be 4:-w=ted on the System Walkdown Report.
H.
Completed System Walkdown Reports shall be provided to the unit Principal Fn-i Work Group leader, and the system file (Section I).
5.5 PipingInspection Notes A.
While perfu,..ar, system walkdowns, it is important to look for problems that can lead to fatigue failures of piping. These failuns can cause personnel and plant safety risks and the forced outages that accom them. It is -=11y '
wt to watch forproblems on umsolable lines, or lines that, pressure is lost, a tnp wYoocur such as Instrument Air, lines are particularly."
vus),to plant pm cl. ISI(in-service ia a~+iaa)inspectsEHC (electr piping bangers and performs UTS (ultrasonic testag to ??
!= if wall rhianing is taking place on selected lines. Systcm Fa-iaaars noo)d to look at the overall sy notes are meant to be a guide of things to Took for and think about durmg walkdowns B.
Most piping failurcs are due to FATIGUE. Imok for the following e-Eaae, which can set up vibration or harmonic motion of the piping:
1.
Valve cycling orchatter.
2.
Cavitation noise.
o j
Structure and System Walkdowns MN-1-319
)
Revision 2 I
Page 15 of 50 5.5 PipingInspection Notes (CMm_'ad)
' 3.
Inadequate support of a smaller branch line of of a larger pipe (also cantilevers).
4.
Discharge of a positive E=r2===st pump 5.
Vibration of another piece of equipment or rotatmg nMma to which the pipe is attached or by whichitis supported.
6.
Piping or piping.w.i. exhibiting excessive motion.
C.
Other items that can contribute to fatigue failure or result in overstress of pipes inclode:
1.
f'ar*ilevers (relatively heavy e hangmg off others with little support).
2.
Piping luk Lw with other piping or components.
3.
Evidence of rubbing or corrosion causing wall loss.
4.
Hagers that are broken or out of adjustment.
5.
Pipe supports welded daractly to the pipmg D.
Other cr==wlerations in piping problems are:
1.
Failures often occur at weldjoints. Socket welds provide high stress risers.
2.
Pipe wall thickness usually thick-walled piping is less susceptible to fatigue than thm-walled pipe.
3.
Piping grows thermally: an inspection of cold piping often will not reveal a problem that is occurrmg in the hot position 4;
Addmg supports may increase stress (not all shaking piping is bad).
E.
ifa
' I problem exists, but additional guZ-s is needed, c<w*md supervision or sgrgied member of PES or DES to aid in the Am! n on. Reference piping and support drawmss during the nmew. Involve DES immediately and dc+
=e the conditxm in an issue Report when a problem is w-i --+i 5.6 Outage Werk Scope Changes i
A.
As a result ofwalkdowns, System F= iaa~s may identify deficiencies that,in their opinion, reduce a system's capability to cany out its f=W. 'Ibese
=== may also require the system to operate without desued mdications and annunciation. Along with the action noted on Mahs 1, 2, or 12, the Systeen F==i-may include equipment m ies in a planned or forced outage work list weig to OM-1-100, Outage Management. On the other hand, the System Engmeer may identify activities that should not be incloded in planned or forced outages.
6.0 BASES None.
6-MN-1-319 Structure and System Walkdowns Revision 2 Page 16 of 50 7.0 RECORDS A.
Records generated by this procedure include:, Mechanical System Walkdown Report, Electncal System Walkdown Report A*=ehenant 3, Structure Monitoring Walkdown; Cantninment Structure A*=rk'aaat 4, Structure Monitoring Walkdown; Concrete Structures Other than Cantainment, Structure Monitonna Walkdown; Masonry Walls A*=eh aa
- 6, Structure Monitoring Walkdown; Intake Structure, Structure Monitoring Walkdown; Buried Piping, Pipe Supports, and Ant-A*= ' - =t 8, Structure Monitonng Walkdown; Steel Structures and Connections A*=ek aant 9, Structure Mosutoring Walkdown; Storage Tanks An=eh==-at 10, Structure Monitoring Walkdown; Dams, Embankments, F*=ia% Walls, and Canals Attmehment 11, Structural McainW Walkdown; Large Equipment Supports and Anchorages and Seismic Gaps A*=eha=nt 12, Walkdown Report Cnnemnation Sheet 4, Refueling Equipment Walkdown Report - Refuchng Machina Walkdown Report B.
UEpon completion of a walkdown, leted forms and instructions shall be forwarded to
- -:===t and Infonnanna Retrieval acco.d-g to PR-3-100, Records
- t. A copy of the 4:-:==% should also be sent to the Maintaamare Rule C.
The Mah= ------?
Rule Comt tw should record the[- - im..a of all structure system walkdowns and include the results as part of t Penodic Annan=n=t required by Paragraph (a)(3) of the Maiaran==ca Rule.
4
e Structure and System Walkdowns MN.I.319 Revision 2 -
Page 17 of 50 j
ATTACHMENT 1, MECHANICAL SYSTEM WALKDOWN REPORT MECHANICAL SYSTEM WALKDOWN REPORT NAME:
DATE:
/
/
SYSTEM:
U1 MODE: 12 3 4 5 D U2 MODE: 12 3 4 5 D N/A N/I SPECIFIC !1NDINGS// CTION TAKEN INSP DESCRIPTION (R / MO NUDERS)
, 4.0 LVALVES anemw%wNOTs59%*eswwM^o+Ams h. >
+
lastausd a properSow drecties?
In conestpenitionfor merentmods?
.Propertylabelsd?
- An'y umusualpadrang lankage?
.Best seem er mining hand wheel?
3esusthseeds properly hahr== sad?
Amy air, steese er trydredse leaks assed?
Fa.*M Yes No 4<2:0 EPUMPS2WBWWWA:M li^wtNOTES:94MMMMOM%%%MMNi@hWC Amy unumaalnoisas asend?
.Properlylabeled?
.preparty lubnsated?
.thamaat peekmserasalleakass?
Mesor amperage aermal?
.Ducharar sresswo nenmal?
ICenemuod? Yes No
)
33.07tCONDUIT A PIPESUPPORTSnWMOTES:aMRimsss>MWFswrm an:es4u
. Spreg cams ! ^^
",, _" aut?
Propereyelmit sagassmet em yserut?
. Good bearing suriner aestact on base pleess?
Do smubben have sufficiset Suid hi reservoir?
Strucewal cemcrees spalles/demage?
.Budag? Properalgunent? Tooloose AresMPreftePenned?
Vamts/ drain lines properly.. _f.J7 SEFA?T. 4 e n FOR GUIIWCE ra-aM Yes No W4.05 DEIRUMENTATION & BRPAKERS HMNOTESfi!MdW6N!MSMMW4EMEW@
.N,
- ,mataDedandfunctiesel?
I a=Idag things?
. Amymissing aa-Ta===== on tubing supports?
ca-ra""* Properly labeled?
Breakers in properM6?
Fa=*=mana? Yes No H25.0 P TEMPORARY: ALTERATIONS 4N9NOTESi%MinhEMMWlEWaitaMW
. Are castagtsey altsi,.w, p.47
. Are diere any -sharized temp shs?
W' metallaties, remeval er
=addamenaa e(plant osmSaaration)
SEESECT.12.A.6 fiOR OUEWCE arw? yes No E6.07^. GENERAL Mi@^#4f+:6dW3 EMENOTEStEMiW:iBWinMM4WMWWM huulationproperty munaussi
. Heat tacing insentled and operational?
seuseast preams paramsest vakses esasussut 4 surrent stods?
.?
__ adequalslycooledAustilated?
. Operaler aids, signs, tamp masas and label Pisees concedyposted andlogged?
.IR tagsin tbs Said stiD appropnate?
. ls gemaralI
, a widnin set adandanis?
. Are disre any safety concerns?
. Are coetmas applied and intact?
ICoramuod? Yes No N/A-Not Applicable AREA OF WALKDOWN:
N/I:
Not = ;-+=Me I
INSP: Inspected SHEET OF
MN 1-319 Structure and System Walkdowns Revision 2 Page 18 of 50 ATTACHMENT 2, ELECTRICAL SYSTEM WALKDOWN REPORT ELECTRICAL SYSTEM WALKDOWN REPORT NAME DATE:
/
/
SYSTEM:
Ul MODE: 12 3 4 5 D U2 MODE: 1 2 3 4 5 D N/A N/I SPECIFIC FINDINGS / ACTION TAKEN rwsp DESCRIITION (IR/ MO NUMBERS)
- mL0% TRANSFORMERS 9fatW WB2MNOTES:T5?@!sWRPn#6 Mis @af@E9M&
Oillsaks?
- Fass opersang?
- 00 pg opensing?
-Taupersawes oserect?
.Prussusis corren?
- Coldicipensk a good con &hom?
.ABhydwaggesa ed?
= Paind W r n=e
,e? Yes No U2.0 WMOTORS MhM!r4MNW;unasneMSNOTESmMMMkNRPN 'taf(W'mMenMW#
. Amy unusual noises notee
- Amy usualal YElfatjens?
Motor maparage monnal?
- Properlylabeled?
lramannsd7 Yes No 3*;3.0 EBREAKERSiW44isi*NM*PnMheNOTES!niWMCiR#MMWVWM% WW1
- Cabstats art istad-Do 10000 paits?
No brakan ia6casors?
.No dropped Bass on relaying?
. Breakers psoportyinstalled?
Ifbreaker out ofoutsele, is k restremad fross roningaromul?
- Af* 8uPPorts Pamaad?
. Are venes and drain lines properly supportsd?
Phu osadmit --e?
E^
"7 Yes No 24.0 & INSTRUMENTATIONiFMWPiNiiNOTESr9tf+iin MeOSM*MdWaiNM5MmmiW I
. No broken parts?
In entibratice?
pra=,
,e7 Yes No
@ic ? TEMPORARY ALTERATIONS GA%NOTESMMMMWBMWAdipM@MeitM+it:WJr Armatesinne aks Properly pos.c
. Are there any uneuduriand tsemp aks?
(Wevabnesede=a.n Founovelor modincation of plant conBauration)
SE SECT.12A.6 IOR GUIMVCE r'
" "7 yes No
- %6.0WGENERAL@M@iW: uWW.!YNOTESiytbEt4&di&M4%MutNTMem%W
.c
.-ads,iewysociarvietoes d?
. Operasar aids, signs, tseg notes and label i
F nnes correaty poseed and logged?
.Doatagsinthetwidid sity ximing d.rh?
Is geosral bossekeeping wieben set mandards?
Are there any safety conceras?
IComunued? Yes No N/A: NotApplicable AREA OF WALKDOWN:
N/l:
NotInspectable INSP: Inspected i
SHEET OF I
l I
o Structure and System Walkdowns MN4319 Revision 2 Page 19 of 50 ATTACHMENT 3, STRUCTURE MONITORING WALKDOWN; CONTAINMENT STRUCTURE (Page 1 of 3)
UNIT CONTADmIENT STRUCTURE PLANT MODE PERFORMED BY:
DATE:
N/A N/I SPECIFIC FINDINGS / ACTION TAKEN INSP STRUCTURAL ELEMENT ASSESSED GR/MO NUMBERS) m:g;.^.igip's:eMfumw9ew.r.m:st..u;gd.tv. @<mga.
kFRESDfr NyMIR##$ 0:
f,ht.0M
.a:: NOTE e r + w. ' e.,
FM CONTAINMENTSUPPORTS'ANDiM W PRESENTL D
@SM
?RESTRESSING LLEMENTSTIENDONSD
':':eWi f ' NORN/A4 i #@^~. '
- W^F e
L Excessive environmental degradatson (exterior andinterior):
A.
Concrete e =- *ing?
B.
Concrete Spelling?
C.
Visible Signs of waterintrusion?
D.
Mineral deposits?
II.
Un wytable Corrosion ofemheMed steel /rebar A.
Hairline cracks?
B.
Rust Staining?
C.
Evi&== ofincreased creking?
III.
U== - rpA21e Corrosion of attehed metal components:
A.
Conosion of bolts?
B.
Ceiredon of metal e,-ait plates?
n C.
Corrosion of anchornga=%aaarts?
IV.
Uance?A;le Degiadadon of Tendons and Associated Equimt:
A.
Buttress De=='=90n?
B.
Tendon Grease leakage?
V.
Misc. C6vations:
Camanuart YES NO-AmmhNotestothisfwm 91@$ %g22d BM@b94WWI:
Sgr. pjy?gESETILEMENTOF:(mgN % g 'j ] 9 4 (F4W.mL 3taNOTm1 ggjmqppi pgg.s t
!7{2.0$
% tPRESENT;:
%e PRESENT.'O R N/A d
@$$$4M$i@r
':ra.%,JR#es:p 62M 4%31!@CONTAINMENTSTRUC'IURE $# @ M Wit I.
CMived settlement of the containment structure-A.
Excessive maahgnment of pipes and Fr*n hardware?
B.
Excessive cracking and stress fractures ofconcrete?
C.
Excessive warping of steel parts /lia*4 D.
Excessive misalignment of major equipment?
E.
Test resultsindicate excessive settlernent?
II.
Misc. 06ivations:
ra*w? YES NO - Anah Nous to this fwm N/A:
Not Applicable AREA OF WALKDOWN:
N/I:
Not Inspectable INSP: Inspected SHEET OF
MN.10319 Structure and System Walkdowns Revision 2 Page 20 of 50 ATTACHMENT 3, STRUCTURE MONITORING WALKDOWN; CONTAINMENT STRUCTURE (Page 2 of 3)
CONTAINMENT STRUCTURE N/A N/I SPECUTC FINDINGS / ACTION INSP STRUCTURAL ELEMENT ASSESSED TAKEN(IR/MO NUMBERS) hrh@gynamn;rkn::;e egen.dsnn.:aww$@dk:: 4PRESENTi y$7RESENTf g+ ityOUNDATIONDRADIAGE'8YSTEM ung$
1.m;we
& hun nt;
?q$g}.wn;Jav !w?
+-
gI9IR#$33h 99 d?M4 :n + w S w e: M M t; w M i4 R ei! @ @ e t! w: rise ;:w O R N/A d '!!^ 4 ^*WNd
)
IV.
Dramage system c
- ,Is af transferrag
{
water from structure-Siridng water or water marks at base ofbuildinn?
Standing waterin buildag extenor pits and below grade equipment areas?
Erh or presence cfground waterintrusion?
Ex-vc Eiging waaer or evidence of reenthnf water (rust stains)in undesirable areas??
Degi.d.iion of concrete, groot, or metal components (corrosion)due to h nonce?
Ms eg moisture exclusion, gap seal and or expansionjoint material?
D.
Misc. Observations:
communert ns NO-Amash Notes to this rana p+5m sme:evageiwye w pawwam:se ne; ri9stera.cNOTe;p g; im a m y;N;e Me% tINTEGRTIV OF 70UNDATION STRUCTURRS O PRESENT - QPRESENTE R
BIRA A
non& WW neswsr. uMmceimmM&;4mewNew iug;*ce&; :410 R N/A 4 mud.I &
IV.
Equipment support capable of transferring allloads:
Gaps between WA--- at and base?
"G lose or not secured to base?
Base cracked or dew?
V.
Equipment anchorage sy='a= functional:
Grout support chair notintact or desr dad?
Base Plate entantn.e or da==4=d?
Anchor Bolts minena or degr= dad?
Nuts mi- =5 or degradad_?
contamert nS No-AmeshNotestothisinna III.
U- =---j : le corroman of metal
-a===ts:
Cuciv.;on of anchor bolts?
Corrosion of =~hara_aaa?
C-.
vs of ?==t plates?
IV.
Misc. Observations:
cuenned? nS NO-Amed Notesto1 bis Sons N/A: Not Applicable AREA OF WALKDOWN:
N/I:
Not Leble INSP:
Ta==+M SHEET OF
s.
MN 1319 Structure and Systetn Walkdowns Revision 2
{
Page 21 of 50 ATTACHMENT 3, STRUCTURE MONITORING WALKDOWN; CONTAINMENT
{
STRUCTURE (Page 3 of 3)
CONTAINMENT STRUCTURE l
N/A i
N/I j
SPECIFIC F1NDINGS/ ACTION INSP STRUCTURAL ELEMENT ASSESSED TAKEN GR/MO NUMBERS) gg pSntuCrURAL STEELCOMPONENTS@ r,:.lWW @NOTW gg';g4SNW !
@iBt#is@Qp!
M:f E5.0lji j (INCLUDES PIPING / EQUIPMENT.S UPPORTS ;; ;PRESENT5 PRESENT
&JP Mysi 9 SEM:N9Mt NDLINER)Ji#f@8MM InfMFi:b WOR'N/AD TN
)
"^
1.
Ua==*=hle corrosson of stre=1 steel:
A.
Cvav. ion of snetal structural elemente?
B.
Corrosion ofinctalbeams?
C.
Corrosion ofsteel support members?
II.
Dw.dation due to corrosion:
A.
Rust stains?
B.
Flakmg/bubbhng ofprotective enatmss?
III.
Misc. Observations NM YES NO-AnadNeesstothisform
!gg/p*d$FZ'f t at.* m D:tW.MfMP!4%%DHTJQat.%rG;y Jdi<uM*9M M MUI.c?
J.Fy$IR#Mhip:
9tkFMM 56.0 % q E n@% REVIEW 0FINTECRATEDJb!2hg HPRE GIT!!...PRESENTo t.w:
- j.j 39LEAKMTEhasG RESULTSWauen NGesFni MnelerN/AV
&M wMd8be I.
ILRT results acceptable?
II.
Misc. Observations Cantamed? YES NO-AmadNeemstothisfann
?P M gggy(44**M8M4?/MJ:W *rN M' 9;pe@Jb T Mnie4sm* +il
!ytt:NOT;r+g 47.0 i ggbM01STURE' BARRIERS; SEA 13,WM;. @
$6;p@i;EGt#djjh:tp@c gigi t dves,tr.* !
TRESENT PRESEITE 4::g4 v.::dge. AND EXPANSIONJOINTS M:m WWM@ 40RN/A4 :14W'M* T M*"
1.
Loss or damage of barriers to moisture intruston?
A.
Absence of apahng matertal or damage to exanneon joints?
B.
husce of FA water or accumulated moisture?
C.
Conson of metal structural elements?
D.
Corrosion of metal beams?
E.
Corrosion of steel support members?
2 Couv. ion of steel plates?
II.
4.dation ofEquipment enclosures (Roof / Ceiling):
A.
Standing water or waterintrusion from ceiling or walls?
B.
Water marks on ceilings, floors, or eQuiDinent surheme?
III.
Misc. C6v=*iaae C"-M YES NO-Ama& Notes toihis form N/A:
Not Applicable AREA OF WALKDOWN:
N/I:
NotInspectable INSP: Inspected SHEET OF
MN-1-319 Structure and System Walkdowns Revision 2 Page 22 of 50 ATTACHMENT 4, STRUCTURE MONITORING WALKDOWN; CONCRETE STRUCTURES OTHER THAN CONTAINMENT (Page 1 of 5)
CONCRETE STRUCTURES OTHER THAN CONTAINMENTI WIT PLANT MODE PERIORMED BY:
DATE:
N/A N/I SPECIFIC FINDINGS / ACTION INSP STRUCTURAL ELEMENT ASSESSED TAKEN (IR/MO NUMBERS)
N;yg%gW. cEnrien%ghMyMisOn45@tF#mSQ 1WMnW 'nNOTM lt ' jyrgs! -
c ies; ?f4 CONCRETE SLABSJ BEAMSfCOLUMNS$$ PRESENT-PRESENT1 jif pIRE R@W tsis.tBASE'PLATESlAND POUNDATIONS ~ & M #ts SiORN/A - A IMM A
M I.
Excessive enviE+=- =t:1 deWdiaa*
A.
Concrete cracians?
B.
Concrete Saalling?
C.
Visible Signs ofwaterintrusion?
D.
Mineral deposits?
II.
U==~~/=We Corrosion of N steel /reber-A.
Hairline cracks?
B.
Rust Staining?
C.
Evidence ofincreased cralag?
III.
Un=-N -1le Corrosion of attached metal components:
A.
Corrosion of bolts?
B.
Corrosion of metal support =1d=7 C.
Corrosion of anchorages / supports?
IV.
U==-- +iesle settlement of foundations and slabs:
A.
Misalignment ofmajor c- =g==s in the structure?
{
B.
Cracking or warpingofstructural
)
elements due to settlement?
V.
Misc. Observations' Conhaued? YEr NO-AnneNoustothisform 111ds category of structure is ameant to nacinde Audiary Buildings, Fuel Handling Buildings, Spent Fuel Fool Areas, Diesel Generator buildings, Speat Fuel and refueling Pool, etc.
N/A: Not Applicable AREA OF WALKDOWN:
N/I:
NotInspectable INSP: Inspected SHEET OF
l MN-1-319 Structure and System Walkdowns Revision 2 Page 23 of 50 ATTACHMENT 4, STRUCTURE MONITORING WALKDOWN; CONCRETE STRUCTURES OTHER THAN CONTAINMENT (Page 2 of 5)
CONCR's 9, STRUCTURES OTIIER THAN CONTAINMENT N/A Nil SPECIFIC FINDINGS /ACT10N TAKEN INSF STRUCTURAL ELEMENT ASSESSED (IR/MO NUMBERS)
- nOR;% p S N i#jh y ?id!! M M W i$i%
M WJ E4NOTgg idgh@@dhMAND MOUNTINGME@jpatW%i
- ygd#gi@fQ
$2.01 EQUIPMENTSUPPORTS Nks iPRESFNT PRESENT bdi!!RIR#;f i
- iwsQi RAY f!IG MFMID WOR N/AE SDWh;@M 7Mii$
I.
Equipment support @le of transfemng allloads:
A.
G-bi-o equipment and base?
B.
Se= art lose or not secured to base?
C.
Base cracked or demM7 II.
Equipment anchorage ;ggm functional:
A.
Grout support chair notin tact or degr M B.
Base Plate = =-- = or h 'a?
C.
AarharBolts miring or degraded?
D.
Nuts missing or h :'='?
III.
UhA corrosion ofmetal couponents:
A.
Cvaudon of anchor bolts?
B.
Couveon of anchorages?
C.
Corrosion of =mt r'==7 IV.
Misc. Observations:
h=M YES NO-Amach Nasas totbs form
?dnd!? $ h$ kt;@ @ FA R W R M!s. R $1M[y$ M;H@g@ pWMOT M W S3 3QQ
];GO;iXR###;p%M h
$3.0; 5tS$eiiWATER RETADUNG STRUCTURESl;jgk
- PRESENTr ;PRESENTI; ff^
Mik W@m%gK49@#4s#;Me@l%Gbeg6 3%@WM NORM /N8 'i<tGdf4Widd+i L
Condition of metalhc and non-ind.the liners:
(
A.
Exmssive corrosion and cracking?
B.
Excessive differential settlement?
C.
E-4 degradation or corrosen ofunderwater parts?
D.
Excessive Leaktge from tell-tale W,0.
drams?
N/A: Not Applicable N/I:
NotInspectable AREA OF WALKDOWN:
INSP: Inspected SHEET OF
4*
t MN 1-319 Structure and System Walkdowns Revision 2 Page 24 of 50 ATTACHMENT 4, STRUCTURE MONITORING WALKDOWN; CONCRETE STRUCTURES OTHER THAN CONTAINMENT (Page 3 of 5)
CONCRETE STRUCTURES OTHER THAN CONTAINMENT N/A N/I SPECIFIC FINDINGS / ACTION TAKEN INSP STRUCTURAL ELEMENT ASSESSED GR/MO NUMBERS)
!!10@@L NhMWATER RETAINING STRUCTURES;i!I4!
$s QI^9 @!N1MfU KfdAiGIO!5fd3 M 8 h ; M i!#fdsj WSis#%n giNOTRy ggpgF#iliBC9]!,
.PRESENT PRESENTn !*PJFsR#7NNNIIF k%5 SIF2%i@$1MnMMWisSdMMM5!Q##adQ %MGm3
- iORN/AI M55hdWif5NI!!
4 II.
Condition of concrete sumps and fluid liner structural supports A.
Crackmg and e'Wa+ ion of concrete?
B.
Degradation of reinforcing steel?
C.
Separation or differential settlemant between structuralsupport and liner?
ID.
Misc. Observanons commmari YES NO-AnnahNeenstothisfann
- jjjjsg W$i%f8IW8FetB#$672DR$@@@g 3}hyjgiifgggANOTBqg5tp'JRht*M$jg.h a4.92 'A FOUNDATIONSf UNDERSIDE OF CROUND bg tlFRESENT?
FRESENT, $ inn @NN;Qif y!Qe
^iBe$ GE7LOOR'SLkB2AND" ROOFS /CERJNGS WH TMi!!N#dMI EOR N/Af N
I.
FM eenV1ronmentaldegradation V
A.
Concrete crarting?
f B.
Concrete Ra=11 ins?
C.
Visible Signs of water intrusion?
D.
Mineral Wits?
D.
Umpid,le Corrosion of errdwided steel /rebar-A.
Haartme cracis?
B.
Rust Staining?
C.
Evidence ofincreased cracking?
III.
U=-:+;:nle Corrosion of st*=cl=I metal co -- = =v A.
Corrosion of bolts?
B.
Corrosion of metal support ala***?
C.
Corrosion of sneharma+2/ supports?
l N/A: Not Applicable AREA OFWALKDOWN:
N/I:
Notinspectable INSP: la ;-*ad SHEET OF
c
=
4 MN-1-319 Structure and System Walkdowns Revision 2 Page 25 of 50 ATTACHMENT 4, STRUCTURE MONITORING WALKDOWN; CONCRETE STRUCTURES OTHER THAN CONTAINMENT (Page 4 of 5)
CONCRETE STRUCTURES OTHER THAN CONTAINMENT N/A N/I SPECIFIC FINDINGS / ACTION TAKEN INSP STRUCTURAL ELEMENT ASSESSED GR/MO NUMBERS)
I$j$ $ssMWEMiediff+$fdEMis!@!d!Bf85WW :JfiifREE 4*
!!iiJfMOTW NM@jgggg y;:4.9;gMFOUNDATIONS/ UNDERSIDE 0F GROUNDit:' !! PRESENTS iPREitENT W JA. n #
4i:M rp.b FWR'3LABPAND ROOFS /CEILDIGSM^ WGMnWE NOR N/F }MNE@ffrqEMId IV.
U== -941ewaterintrusion:
A.
Rainwater /groundwaterintmsion through roofor ceilin-=7 B.
Groundwater antmsson through floors, structuraljoints or walls C.
Failure of the groundwater drainage system?
D.
Failure of the floor drain system?
V.
Misc. Observations cassumrt YES NO-AnasNetstoeninne
$d4 $ijfM d!M Wi$ $ # @ @i!d@ @ MN M @j$.:l
!!!! $!!W2@40 - *ygNOTsy ag$hlR#f::M tiMQQyff
$5.01 ;,$eSTRUCI'URAI;8TEEllCOMPONENTS i Fjg :WBJgh$
FWFAENT PRRKENT!.
3 DEW : :(INCLUDES FIPING/ EQUIPMENT SUPPORTS)" @SP Wi@
MM4 f0R N/A F 1.
U-=*Mle corrosion of structural steel:
A.
Corrosion of metal structural elements?
B.
Corrosion of metalbeams?
C.
Corrosion of steelsupport inembers?
II Degadation due to corrosion:
A.
Rust stains?
B.
Flalung/ bubbling of protective matingc7 III.
Misc. Observations comunuart YES NO-AnasNeemtoEfann i
i L
I N/A: Not Applicable AREA OF WALKDOWN; N/l:
Notinspoctable INSP: WM SHEET OF l
l i
Structure and System Walkdowns MN-1-319 Revision 2 Page 26 of 50 ATTACHMENT 4, STRUCTURE MONITORING WALKDOWN; CONCRETE STRUCTURES OTHER THAN CONTAINMENT (Page 5 of 5)
CONCRETE STRUCTURES OTHER THAN CONTAIhMENT N/A N/I SPECIFIC FINDINGS / ACTION TAKEN INSP STRUCTURAL ELEMENT AS3ESSED (IR/MO NUMBERS)
??#sih G T!!it. M i R n 9 M W W E ur v;.:
sini Rf.adMtc._
.. FNOT4 4 g1% @47 m UE M3] CONCRETE PRESTRE$ SING SYSTEM 4i '
N
!PRESENTy PPRESENT?
!?
?IRfe
%e6 we my m a, ww3
,c mwm g:
- % + me E O R N/A W s
4
- E>W I.
D.gisdation ofpuuiod gabouug elements:
A.
E Wemoistureintrusion?
B.
High stress cracking / corrosion of anchors present?
C.
Corrosion of e -4ns?
)
II.
Degradation of greased pabonug elements:
)
i A.
Excessrve grease 14=ge?
B.
Excessive Corrosion of metali-i.?
C.
Tcadaa/ conc.i.idasc. Lor degradation?
III.
Misc. Ct.antions Costanues? YES NO-Anach Notes tothis form
@@ig d N! G Grsb @ d n H s WiiiE*.g n;t@ g g.; j piry h g ENOTr4,; W3M a jf wr fh:7AR ;iCEILINGS, MOISTURE BARRIERSL SEAIAC kPRESENTj 4PRESENT e
MN :f ' y"hEf!
j$iyfIR Ng pum *ANDEXPANSIONJOINTSm WW-Mt@nWN
?ORN/A :.
J.
l.
1.oss or damage of barriers to moisture intrusion?
A.
Absence ofscaling materialor d===*e to mmnsion joints?
B.
Fia ofstandtag water or accumnhH moisture?
C.
Currosion of metal structural elements?
D.
Corrosion of metal ku*?
E.
Coricsion ofsteel suyywi members?
F.
Corrosion ofsteel plates?
II.
Degradation ofRoof7 Ceiling:
A.
Standmg water or waterintrusion from ceiling or walls?
B.
Water marks on ceilings, floors, or equiFnt surfaces?
III.
Misc. 06vations t'aar===d? YES NO - Ana& Notes to this farm N/A: Not Applicable AREA OF WALKDOWN:
N/I:
Not Taphie INSP: Inspected SHEET OF
s
.2 MN-1-319 Structure and System Walkdowns Revision 2 Page 27 of 50 ATTACHMENT 5, STRUCTURE MONITORING WALKDOWN; MASONRY WALLS (Page 1 of 2)
UNIT MASONRY WALLS PLANT MODE PERPORMED BY:
DATE:
N/A j
N/I SPECIFIC FINDINGS / ACTION TAKEN INSP STRUCTURAL ELEMENT ASSESSED (IR/MO NUMBERS)
.. ~. e. u,:.. :s
.a m Qf}%wtw + zy.wo~a.u;w,w, Ar ~:mes <.Ui W >.~;w q,ggq qq w nwa
.. M.. ><'.
W?NOT.;:-:*:ce<
"g y.-
!u s' f.flR # M.e @+.
m-
'* ; A'
_.t....f
.t<.J tm 9PRESENT:?
?
E1.0Q jjM qSURFACE CONDITION Pk PRESENT4 Eu
$&%ih"~pt ffygW q;;;q n % K::ORNik%r O
I Excessive surface degradation of walls:
A.
Deficienciesinjoints?
B.
Differential movement ofjoints?
C.
Evidence ofwarping, bulgmg, or I
saggmg ofcomponents?
II.
Misc. Obsenations:
ca===d? YES NO-Anna Notestothis form
- . 0;.g sgui
.u.~.:
Wh.a;j.t/.Q. ;,M:r :Je%gt:T. >.N s aTip;q%.Qg;x.s.sy 1 a., -
e+a e ggg
+4
.-eM. nu.n.st;.1
' :. NOT+7;N v
y my,.r:p;;+mpa;.4l.:
+
- J
^ +w.,
p nt 13.0j. " l2.tiiK@%c3gMag, WALLS 4196$):
WpTIRM;Wt!OI d b M E g
isj9 JPRESENTE =PRESENTf, rCRAC1GNG
- n@
4:e;Asf iegg;g;$szitd:t M ;g;3ql i;i+;ORN/A'M I.
Craciang of masonry walls:
A.
Presence c' extensive cracking?
B.
Presence of extensive shifhng of basejoint?
C.
Presence of extensive shalhng of ceiling joint?
D.
Imss of grout or mortar?
II.
Misc. Observations:
Coatsmed? YES NO-AnachNotastotPusform y ap; g e m. m. m...c, s a gt.a >.m: t.. seLsp;;:e.m.;w.n, te.PNOTh.
e ;
v w"m*eu ~
amute a
.n.u..w A..n x
nw~~
n t
~.
+94 MASONRYWALLSUPPORTSMM$r.: <A.h PRESENT& t!:
p > wpi.; ggg.gmgytp :yg;ggu pt' PRESEf&s a $IR # NIO'*.*
>m 73.0 P M
wnw gg.yp^
MOR NIAC^
ti T M M "
1.
Support structures that anchor walls to minimim movement:
A.
Lateral supports not anchored properly B.
Excessive differential settlement?
C.
Excessive degradation or corrosion of support elements / parts?
II.
Misc. Observations:
em=W7 YES NO - Anach Notes to this form N/A: Not Applicable AREA OF WALKDOWN:
N/I:
NotInspectable INSP: Inspected SHEET OF
MN-1-319 Structure and System Walkdowns Revision 2 Page 28 of 50 ATTACHMENT S, STRUCTURE MONITORING WALKDOWN; MASONRY WALLS (Page 2 of 2)
MASONRY WAT.rs N/A N/I SPECIFIC FINDINGS / ACTION TAKEN INSP STRUCTURAL ELEMENT ASSESSED (IR/MO NUMBERS)
M@R ih!-SE M M M i M A s & M 3 M f3 6 diinE$5f#%METALCOMPON h
f $$^W9MIJ R^4NOTpg;y %p! DR#DYp
.g fgisG;g d
<PRESENTe &PRESENT$
$4,e N
ANdq J$?db0d9%9 aaNmihe&9s#BFidN20B Rtn@i6F@U ORN/A$
20AUMb L
U==-
- -We Corrosion of emhaMed steel /rebar-A.
Hairkne cracks?
B.
Rust Staining?
C.
Evidence ofincreased Wing?
II.
Unacceptable Corrosion of attached metal components:
A.
Corrosion of bolts?
B.
Corrosion of anetal support phtes?
C.
Corrosion of anchorages / supports?
III.
Misc. Observations i
I r*W YES NO-AnachNeessaEfwm 9)iNis
$:##Mif+MGME!baMa Ui?!,Rii^:AWatdg 889&7sm IggNOT&g pdadsbS14 Wit wieL 06 BOUNDING STRUCFDRAL ELEMENTSi: Ele iPRESENTL APRESENTw-rIW4IR#31$
NdflE M@hizim%%iFGW:09^ Ag@EWadHil" MfSNO OiORN/Af 2 ENNMMS$F
' I.
Unacceptable condition ofjoints at ceiling and floorinterface with walls:
A.
Cracking or excessive gaps between ceilinginterface with wall?
B.
Cracking or excessive gaps between floor and interface with wall?
C.
Absence ofjoint scalant material between wall and ceiling or floor?
i D.
( eking of celhng or floor at wall l
boundaryindicatmg excessive loading?
II.
Misc. Observations em=4 YES NO-Anachth2 Om fem N/A: Not Applicable AREA OF WALKDOWN:
N/I:
Not Inspectable INSP: Tapad SHEET OF
s, MN-I-319 Structure and System Walkdowns Revision 2 Page 29 of 50 i
' ATTACHMENT 6, STRUCTURE MONITORING WALKDOWN; INTAKE STRUCTURE (Page 1 of 4) l UNIT INTAKE STRUCTURE PLANTMODE I
PERFORMED BY:
DATE:
1 N/A N/I SPECIFIC FINDINGS / ACTION INSP STRUCTURAL ELEMENT ASSESSED TAKEN (IR/MO NUMBERS)
$ 4 s A % !M A?0 di2 R d 4 @ 9 5 Q Hi d Qi S i? Q 5%fbW @.RBOTgg 5%%V ZIERLUID RETAINING.. WALLS AND. SLAB.;G3 3EOtdN;1fygfy,Qyfb f"Q
/ LOB
- gs ?PRESENT 1MwIRs
$';FRESENTg hM VahM@N!GMt4NTM$IMasN:Ma#f3fiEt W itneR44 ; TOR N/CI x ME I.
Excessive environmental degradation (salt water corrosion):
1 A.
Concrete wHaf?
B.
Concrete Spalling?
C.
Visible Signs ofwaterintrusion?
D.
Mineraldeposits?
II.
U==- rp-W Corrosion ofm l
steel /rebar:
A.
Hairline cracks?
B.
Rust Stainine?
C.
Evidence ofincreased crar*iaE7 ID.
Unacceptable Corrosion of annehart metal 1.. e-nents.
A.
Corrosion of bolts?
B.
Corrosion of metal== art plates?
C.
Corrosion of anchorancs/===ts?
IV.
Misc. Observations:
th=d? YES NO. AnachNotes tothisfam i
N/A: NotApplicable AREA OFWALKDOWN:
N/I:
Not taehle l
INSP: !==r-di SHEET OF
4 f
MN-1-319 Structure and System Walkdowns Revision 2 Page 30 of 50 ATTACHMENT 6, STRUCTURE MONITORING WALKDOWN; INTAKE STRUCTURE (Page 2 of 4) i INTAKE STRUCTURE l
N/A N/I SPECIFIC FINDINGS / ACTION TAKEN INSP STRUCTURAL ELEMENT ASSESSED (IR/MO NUMBERS)
$% MMOEWi ?@iBI'NW;%@EfiON6SM%
$$@S.$lisNOT5@.%5;p@w%$y
.B Ihih31R#i 510si H EQUIPMENT:SUPPOR15 AND MOUNTING & !PRESENT: %^PRESENTA i 1
h!* r2Es;jtqm - iMW@N!Ma!%;e.93;rl@!t i@;@%g M*!ogy/Atst w M^$ Ode *?
1.
Equipment support capable of transferring allloads:
A.
Gaps between equipment and base?
B.
Support lose or not secured to base?
C.
Base cracked or degr=M7 II.
Equipment anchorage system functional:
A.
Grout support cl. air not intact or degraded?
B.
Base Plate nussmg or degraded?
C.
Anchor Bolts missing or degraM7 D.
Nuts missing or degr M7 III.
UnWahle corrosion of metal components:
A.
Corrosion of anchor bolts?
B.
Corrosion of anchorages?
C.
Corrosion of support plates?
IV.
Misc. Observations:
Ceaunued? YES NO-Ansc6Notsetothisfann di% Wi@MiMP?!NiMMi$diffdifi?rf4ST@f @i{i$
j^ligNOTdyp
- WfNT@O ejPRESENTE i>E IR#0hQgF#ij$gg
% 0ii #MINTAKE AND DISCHARGRTUNNELSMN PRESENT J
2.4 EdWMM4W3tWM:@!!M70 GMT @: 4^cWithO @ OR N/A F B5 2MM" I.
Intake and discharge ennnMs serviceable and functional:
A.
Excessive debris?
B.
Excessive differential settlement?
C.
Excessive Degradation or corrosion of underwater parts?
D.
Corrosion of embedded steel?
E.
Craciung circulating water 108" Prestressed piping? Note size, location, dtrcction of travel (axial, circumferential)in partIIbelow, Misc. Observations F.
Damaged or degr=M pipe joints?
D.
Misc. C6 ations:
c-dM YES NO. AnschNonestoEfann N/A: Not Applicable AREA OF WALKDOWN:
N/I:
Not la=p~+=ble INSP; Inspected 1
SHEET OF
o MN-1-319 Structure and System Walkdowns Revision 2 Page 31 of 50 ATTACHMENT 6, STRUCTURE MONITORING WALKDOWN; INTAKE STRUCTURE (Page 3 of 4)
INTAKE STRUCTURE N/A N/I SPECUIC FINDINGS / ACTION TAKEN INSP STRUCTURAL ELEMENT ASSESSED GR/MO NUMBERS) pi&4@ W RikhMM:iQiMM M Mr@fMt?^R Myp:-
w!!nwJTatip Mifi
$PijNOTti+ic WK@#p!1R#[m.;
$ g W dn MTOUNDATIONS AND.UNDEEEEmE OFNndww ROUNDPLoORsLAswW (m!!
Me iPRESENT/ EPRESENT4 s!.iB ew aroRwF c
W: ww I.
Excessrve envirnamantal degradation (salt j
water corrosen):
1 A.
Concrete cr-+ia_=?
B.
Concrete Raattiar?
l C.
Visible Signs of waterintrusion?
D.
Mineral deposits?
II.
T'==--- -M: Corrosion af *=haMad steel /reber A.
Hairline cracks?
B.
Rust R*=iaiaf?
I C.
Evidence ofincreased er=<+ia!?
III.
UaP Corrosion of attached metal components:
A.
Corrosion ofboks?
B.
Corrosion of metal =taa=t =1='-?
C.
Corrosion of anchor =?='==aarts?
IV.
Misc. Observations i
Can-mr? YES NO. AnatNemtothis$m
$3pi37disf4M?dN@k%I$' COMPONENTS $$M jdS($
4f$$W:9N#ri@%$ #fiiWRis (p$ SNOT #;g
}Egji.fER#ffMpjpg@gg
{
M9 STRUCTURAL $7 EEL
>PRESENTl 416g PRESENTR g iih iDliCLUDES PIPING /EQUIPMENTSUPPORTSf fMSUiW^4 k.NWiM'M i
MOR' WAD I.
U==-- =;Ile corrosion of structural steel:
A.
Corrosion of metalstructural elements?
l B.
Corrosion of metalbeams?
C.
Corrosion of steelsupport 1
members?
II.
Degradation due to corrosion:
4 A.
Rust stains?
B.
Flaking / bubbling ofprotective coatmas?
HI.
Misc Observauons cas ri YES NO. Amm6NemtothisEm N/A:
Not Applicable AREA OF WALKDOWN:
N/I:
Not laca~*=hle INSP:
Ia=a=*~8 SHEET OF
r MN-lo319 Structuit and System Walkdowns Revision 2 Page 32 of'a ATTACHMENT 6, STRUCTURE MONITORING WALKDOWN; INTAKE STRUCT. EE (Page 4 of 4)
INTAKE STRUCTURE I N/A N/I SPECIFIC FINDINGS / ACTION TAKEN INSP STRUCTURAL ELEMENT ASSESSED GR/MO NUMBER $)
.jiE!MMMM5MEdsfiM@Bi-Ri^%i!!Fud@?
B%&h 9%fis.
yy3NOT@@ ;igTj&#pigig j@
l
- UFRESENTE 40E! 9 CEILINGS, MOISTURE BARRIERS; SEAI$t
'.V EIR
,e b Qp A;tAND EXPANSION'JOINTSM-IRG +# DNMI DMORN/AMT I!NNM{eE;pe -
]
TRESE!ftr I.
Ims or damage of barriers to moisture intrusion?
A.
AO~of scahng material or damaae to avaaa*% iai*7 B.
Presence of standag water or accumulated moisture?
C.
Corrosion of metalstructural eternents?
D.
Cormsion of metalbeams?
E.
Corrosion of steel support memhers?
F.
Corromon of Eeeiplates?
II.
Degradation ofRoof/(g A.
Standmg water or waterintrusion from cethng or walls?
B.
Water marks on cethngs, floors, or equipment surfaces?
III.
Misc. Observations N=el? YES NO. AmadNotastothisfann N/A: Not Applicable AREA OF WALKDOWN:
N/I:
Not aphie T
INSP: lae~i SHEET _ OF
t EA MN-1-319 Structure and System Walkdowns Revision 2 Page 33 of 50 ATTACHMENT 7, STRUCTURE MONITORING WALKDOWN; BURIED ANCHORAGES, PIPE SUPPORTS, UNDERGROUND CATHODIC PROTECTION SYSTEM, AND BURIED PIPING (Page 1 of2)
UNIT BURIED PIPING. PIPE SUPPORTS, AND PLANT MODE EOUIPMENTANCHORAGES PERPORMED BY:
DATE:
N/A N/I SPECIFICPINDINGS/ ACTION TAKEN k
INSP STRUCTURAL ELEMENT ASSESSED (IR/MO NUMBERS) i:End NIS@@$3I!iiE!RsM]JSinn2@EfifM!d25MInii, IDS $3% ;/pigNOTN4ii:iggj;ijff@$!j;;jg n$iIT4t:
! BURIED ANCHORAGES ANDFIPE SUPPORE <PRESENT!
o::L
?#0RN/Mn$
TERWi$iMdMS PRESENT
% !?IR# W 69fMMad#ggi$!s%3tch@l52!!!?Mn > M @964a?
I.
Excessive degradarian of anchorages and A.
Separation of groot matenal beneath -: =t =" supports?
B.
Corrosics of anchorbolts and meenciatai hardware?
C.
Corrosion of support a3=*-7 D.
I-e or extensive free motion of anchors or===~ts E.
Evidence of warpag, bulging, or ene ia? cf - ?
II.
Misc. Observations:
c--si YES NO-AmadNeemtoG fann M*
SNE5M!Il5$DdItfd5M@!Nk $?NIDNN?b'- NINMN I E b5Ih I
%2$f FUNDERGROUND' CATHODIC PROTECTIONP PRESENT; jkMM h'ma:q$N31(ag t
k bMd! 2@ $h W 5@M M SYSTEM # M bPR@dfS h
- VdD!sM 43OR N/ASan dMA$IGu I.
Status of U1yv 4 Cathadne Protection System:
A.
Presence of extensive / excessive corrosion of buned metalpipingor structures?
B.
MWOs on cathodic protection open greater than 90 days?
C.
Resultsofsurveinamw testag,or mspections of rashadic Pranae*ina?
D.
Misc. Observations:
1 Pan==#1 YES NO AmashNotestoEfium i
j l
i r
i N/A: Not Applicable AREA OF WALKDOWN:
N/I:
Not kWable j
l INSP:
T= W ad i
SHEET OF t
i
o MN-1-319 Structure and System Walkdowns Revision 2 Page 34 of 50 ATTACHMENT 7, STRUCTURE MONITORING WALKDOWN; BURIED ANCHORAGES, PIPE SUPPORTS, UNDERGROUND CATHODIC PROTECTION SYSTEM, AND BURIED PIPING (Page 2 of 2)
BURIED PIPING. PIPE SUPPORTS. AND EOUIPMENT ANCHORAGES N/A N/I SPECIFIC FINDINGS / ACTION TAKEN INSP STRUCTURAL ELEMENT ASSESSED (IR/MO NUMBERS)
%Wjfpd93 @BURIEDPIPING!%" EMF'fdr dea drMiGitgNjsy
@MM3.0;f w!M!*i5!NisistMI*IG
- ;n@pgMe wjM:NOT4eja
$^;p@pydhg;ti;t7+
7(g gr iPRESENT[ EPRESENTdi 3+ n#
M i
92@@2%
- M die beOR' /AW N
L Results ofvisualexams when pipeis accessible:
A.
Separation of pipingjoints?
B.
Indications of environmental degradation?
C.
Excessrve degradation or corrosion of support elements / parts?
D.
I.cakage out of pipes or pipe seams?
E.
I2akage into pipes or pipe samme?
F.
Degraded coannas?
G.
Wallloss due to corrosion?
H.
Electricaljumper across mechamcal joints missing or d fed?
H.
Results of observations of pipe fill and fill support adequacy A.
Erosion of surrounding / supporting soil?
B.
Excessive settlement ofpiping due to erosion of supporting soil?
C.
Presence of sinkholes?
D.
Water flowing from ground or constant flow ofwaterin drainage systern?
E.
Presence ofemndme water or moisture?
F.
Improperbackfillcornposition and
- ;+:dag that could damap protective coatings orimase corrosion rate, e g., stone or clay clods nuxed with backfill?
III.
Mh. Observations:
c- -==W YES NO. Anad Noam to E fem N/A: Not Applicable AREA OF WALKDOWN:
N/I:
Notlaspectable INSP: rn p aa SHEET OF
O
'J Structure and System Walkdowns MN-1-319 Revision 2 Page 35 of 50
. ATTACHMENT 8, STRUCTURE MONITORING WALKDOWN; STEEL STRUCTURES AND CONNECTIONS (Page 1 of 2)
UNIT _
STEEL $TRUCTURES AND PLANT MODE CONNECTIONS PERFORMED BY:
DATE:
N/A N/I SPECIFIC MNDINGS/ ACTIONS TAKEN INSP STRUCTURAL ELEMENT ASSESSED GR/MO NUMBERS)
$$$P}Q?aWiE!WWMDMiNM$iMiWti$1[)"f@h;2i MiWi!!ig 3;;tNOT@}i PM@%g 'E ygh!I PLO 30!@E STR ETEELFRAMESN PRESEITT! $FRESENT4 vmdWii./WEJW;#1NEls;:@nfit:@@ji$s@itii;9
%[> !bSENs;bgi fi!R#.
4.tQGn*@ @ OR N/S09 n
I.
E.w
,4 degradation of structural steel clacients A.
Degradation due to chemicalor envimamental cormsson?
B.
Ly.d. don due to physical damase?
C.
C _-- +-;on due to fatigue?
D.
Ly.d& des due to excessive ta.aiar conditions?
E.
Dw--+==e= due to improper Wia-or heat treat?
F.
Ly.d. ace due to exposure to extreme heat or Bre?
G.
Evidence of warpmg, bulging, or sa=iar of comaa=ats?
II.
myadation of cranes, hoists, trolleys, monorails and sraaa ts-A.
Excessive corrosion ofmetal comaaaaats?
B.
Signs ofstrain orstressloedmg A
imposed byvibration or movement ofatteched components?
g C.
IM Lh or missing
- gam or securms hardware?
D.
Signs of stress due toimproper or j
excessive in.asar af" c 2--t?
III.
Blowout s,.a.1 functionality:
l j
A.
Blowouti 1;securedwithproper j
anannfors7 i
B.
Blowout f a.:. =whad to the structurein such away that they will not perform their function?
j l
i 1
N/A: Not Applicable AREA OF WALKDOWN:
N/I:
NotInspectable INSP: Inspected SHEET OF
m I
f MN-1-319 Structure and System Walkdowns Revision 2 Page 36 of 50 ATTACHMENT 8, STRUCTURE MONITORING WALKDOWN; STEEL STRUCTURES AND CONNECTIONS (Page 2 of 2)
STEEL STRUCTURES AND CONNECTIONS N/A N/I SPECIFIC FINDINGS / ACTIONS TAKEN INSP STRUCTURAL ELEMENT ASSESSED GR/MO NUMBERS)
$iE.4$L 3yjf MWeM;fd84N"di@AMWer h3O;g 3
IsW5$?$ !Ij;;;;NOT:q, *' 195}iR0;$$y, Wgggjyy
$10R :h:L.JIS@WM I #MMid:IR6W Wzdtie 1:$STRUCTURALSTEEL FRAMESdjjp$g 4
PRESENT! EP3ESENT.4 pj
- gadt WWFMNSU TEOR N/AF PAsidO5Mic IV.
Eh4 corrosion of metalconnectors or metal components-A.
Corrosion ofmetalstructural elements (e.g. I-beams, struts, braces, and etc.)?
B.
Corrosion ofmetal connectors (e.g.
melds, rivets, bolts, rods, studs, and wire ropes)7 C.
Degradation of structural elemente due to loss ofprotecove coatings?
V.
Misc. Obsemitions:
Contumes? YES NO. Attad Noemstothisfonr l
l N/A: NotApplicable AREA OFWALKDOWN:
N/I:
NotInspectable INSP: leM SHEET OF
4 o
MN lo319 Structure and System Walkdowns Revision 2 Page 37 of 50 ATTACHMENT 9, STRUCTURE MONITORING WALKDOWN; STORAGE TANKS (Page 1 of 2)
UNIT STORAGE TANKS PLANT MODE PERFORMED BY:
DATE:
N/A N/I SPECIFIC FINDINGS / ACTIONS TAKEN INSP STRUCTURAL ELEMENT ASSESSED (IR/MO NUMBERS) ilQDs p.i@ W)w! Q e gsrq m 2W W g g ;1 A g+ ' yg M Q;i g gn.,pSNOT @ Jfp?4Wgp!;.E P[iL95 FFVISUAI;3NSPECTIONOFTANKS 6
TPRESETiTi ? PRESENT:::' ;;My!IR#1: MM i Mi% ?.k15?inO M4 JEi% Ri M Eih!!Mi % L,.
ith W d.WSii %onytkF*'
% A '+L?M 6 6 '
I.
Fyreecive degradation of anchorages and supports:
A.
Degradation / Corrosion of anchor bolt chairs?
B.
Degradation / Corrosion of anchor bolts and associated hardware?
C.
Degr*%WCorrosion of support plates and saddle supports?
D.
Missing anchorage com?-:==' hardware (e.g. auts, bolts. and etc.)
E.
T =aaa+ or extensive free motion ofanchors or supports F.
Evidence ofwarping, bulging, or sagging ofcomponents?
II.
Degradation or interference from attached or proximate metal components:
A.
Excessive corrosion ofmetal attachrrmts to the tank?
B.
Signs of strain or stressloading imposed byvibration or movement of attached components?
C.
Attached metal components are secured to the tank-no signs of damage?
D.
Pipes near the tank are not leaking water or chemicals onto tank?
l l
l 1
N/A-Not Applicable AREA OF WALKDOWN:
N/1:
Notinspectable i
INSP: faW i
SHEET OF 1
I
b MN-1-319 Structure and System Walkdowns Revision 2 Page 38 of50 ATTACHMENT 9, STRUCTURE MONITORING WALKDOWN; STORAGE TANKS (Page 2 of2)
STORAGE TANKS N/A N/I SPECIFIC FINDINGS / ACTIONS TAKEN INSP STRUCTURAL ELEMENT ASSESSED (IR/MO NUMBERS)
M51 MMNS$iif:E!$E9M6@ItN!!di4%@dNE:!S9 Gi$FdVISUAL INSFECTION D@F TANESIN:i$if Os^4 M %p? $
Sili1NOTsish ::MfBiiff!Qr$if i[iik?aWBsiM 3M M W W WW D(We 39WZ9lm$1 EPRESENT 96Bi i mir:!atiiGM 4 R N/A @ ' WWOF G O
III.
Settlement or base mat degr"aa-A.
Cr4 ordegradanaa ofconcree base?
B.
Excessive wettament ofconcree base / tank?
C.
Bulging or depressicas near base of steel tanks orat potats of small radiiofcurvature?
IV.
%Aon ofperuneter and penetration seals:
)
i A.
Caulking orscalant 4 hon l
due to spar, weathermg, etc.?
B.
Absena ofpenmeter or penetration sealing matenals?
V.
Misc. C64tions:
t'-*=W YES NO. Aanch Ness 2& fann 1
1 i
N/A: Not Applicable N/I:
Not wahle AREA OF WALKDOWN:
INSP: In d SHEET OF
d MN-1319 Structure and System Walkdowns Revision 2 Page 39 of 50 ATTACHMENT 10, STRUCTURE MONITORING WALKDOWN; DAMS, EMBANKMENTS, CANALS, AND RETAINING WALLS (Page 1 of 2)
UNIT DAMS. EMBANKMENTS.CANAIS. AND. PLANT MODE _
RETAINING Watt A PERFORMED BY:
DATE:
N/A N/I SPECIFIC FINDINGS / ACTION INSP STRUCTURALELEMENTASSESSED TAKEN GR/MO NUMBERS)
M95& gjfs5Sisi?2@n#jsjisMirMsgNgin W M)M1MR R,aRIENOT odt g ggsi b@iN@!IR#qy;
$sSBN 192 AqPAMSiEMRANKMENT!yhMD CANALS."j yPRESENTN s
Qi%bRiB@EWK4MiirtnI@.@:ida W iPRESENT) 5 SM
{
e f%:miws/
BORN /Ah6 MWWE 1.
Exassive &=="= L=. offluid-raemintr structures:
A.
Settlement -presence oflocalized i
or overall settlement, depressions, or sinkholes?
B.
Slope Stability-presenceof slope irregularities that wouldindicate mternalmovement orstufung of h..t,n,ne.?
C.
Seepage -presence of or indications of seepage or degradation that couldlead to seepage and degradation?
D.
Dratnage Systems iadi->'iane that there have been failures of the drainage system that have resulted in degrada' ion?
E.
Slope Protection -degradation of the slope protection features as evidenced by the formation of gullies, notches, and etc., or loss of vegetative cover?
F.
Differential Set lem:nt -preserse of differential settlement such as concrete maing loss of scalaat or fill = renal, excessivejoint gaps, and etc.?
l N/A: NotApplicable AREA OF WALKDOWN:
N/I:
NotInspectable INSP: Inspected SHEET OF
b MN-1-319 Structure and System Walkdowns Revision 2 Page 40 of 50 ATTACHMENT 10, STRUCTURE MONITORING WALKDOWN DAMS, EMBAhTMENTS, CANALS, AND RETAINING WALLS (Page 2 of 2)
DAMS. EMBANKMENTS. CANALS. AND RETAINING WALLS N/A N/I SPECIFIC FINDINGS / ACTION INSP STRUCTURAL ELEMENT ASSESSED TAKEN (IR/MO NUMBERS)
$1%!' $s;{d@W@f;%RETAININGWAllhg[g j!:ig.;q;g;;mgggyneb,. gyp ri pr i3ffstPE$iUti'JJGij$gy! Es Eis vsed i@ NOTE & Jg,f IRC% y Qgph'M3ty((i A
13.02 9
PRESENTi ?PRESENT:yl b N D' N n* M se + ue gp st n.qt#4 t T O R N/A W I.
Degradation of fluid retauung walls:
A.
FyreetiVe corrDsion of metal components?
B.
Crarting ofconcrete wall structure?
C.
Craciang or excessive clearance at base ofwall(wallto floor interface)?
D.
Missing grout, mortar, or holes in the wall?
II.
Berm retaining w1tll degradation:
A.
Excessiveloss ofberm matenal due to crosson?
B.
Breachesin the berm retauung walls?
III.
Misc. Observations:
i l
Cantmuod? YES NO. AnachNotastoessform i
i N/A: NotApplicable AREA OF WALKDOWN:
N/I:
Not Inspectable INSP: Inspected SHEET OF
n o
MN-1-319 Structure and System Walkdowns Revision 2 Page 41 of 50 ATTACIBiENT 11, STRUCTURE MONITORING WALKDOWN; LARGE EQUIPMENTI SUPPORTS AND ANCHORAGES AND SEISMIC GAPS (Page 1 of 2)
UNIT gggEEOUIPMENT SUPPORTS AND PLANT MODE I
ANCHORAGES AND SEISMIC GAPS PERFORMED BY:
DATE:
N/A N/I SPECIFIC FINDINGS / ACTION INSP STRUCTURAL ELEMENT ASSESSED TAKEN OR/MO NUMBERS) j:^ NN &%RENS%^i %E ~ 2%+hi*W + h%%Q. Gi:t%fhid..350T % M &i k ?QC, 4@ q 4 @QUIPMENTSUPPOR'IS)AND ifH
+!L0y; MLARGE E
.PRESENT: iPRESENTf @N*NIE M E D$9:iIR ANCHORAGESYOWi~T@ME 5'NMS6 LORN /f6 1@
I.
Excessive degradation large equipment supports and anchorages:
A.
Presence ofcracking of concrete
- r. car embedded plates and structural steel?
B.
Presence of support anchorage degradation for supportsfor Class 1,2, and 3 piping and structural supports?
C.
Presence of thermalinducad degradation of con: rete such as generalcracking or spalhng and loss of dynamic stiffness?
D.
Excessive corrosion / degradation of metalcomponents associated with the support structures including the absence ofhardware such as bolts, washers, nuts, and etc.?
Mi@.Sy.pj^ ' e
{ hic:r;[s((h]sM$$M!%!iWiMJfy@$3pSEISMIC GAPS $$ggM.(
fRESENT[ i.PRESENT? de'i$IRMhM
^3%%;n GNOTeq ;ig$4%.gf *.m fj f103
> c *:w
- .. P :7p%@%. + e.w.m ~%
yne
- 'OR NIXo ^MW '
"+7" :u 1.
Degradation of gaps between structums intended to mitigate the effects of excessive motion, thermal expansion, or seismic events:
A.
Exassiveloss or absence ofjoint filler material?
B.
loss of clearance at thejoint as observed by direct measurement of the width of thejoint?
C.
Condition of thejoint and fill materialindicates exmssive movement or distress?
1Large equipment includes steam generators, reactor vessel, and NSSS piping N/A: Not Applicable AREA OF WALKDOWN:
N/I:
Notinspectable INSP: Inspected SHEET OF
e
\\
MN-1-319 Structure and System Walkdowns Resision 2 Page 42 of 50 i
ATTACHMENT 11 STRUCTURE MONITORING WALKDOWN; LARGE EQUIPMENTI SUPPORTS AND ANCHORAGES AND SEISMIC GAPS (Page 2 of 2)
I LARGE EOUIPMENT SUPPORTS AND ANCHORAGES AND SEISMIC GAPS N/A N/I SPECIFIC PINDINGS/ ACTION INSF STRUCTURAL ELEMENT ASSESSED TAKEN (IR/MO NUMBERS) 1, g s g g g*si j
ML iW@gB@ sMMUBMM8:pgdi:ig~gp @PRESENTiME
@;NOTag.
% p %Mxemytt% gig:%%3fb + GiejMs.AgggSEISMIC GAPS 1Et.W~$i@W; p:.;PRREENT@.P QR :4@@p i{!10!.
W WIR# tM.
y dE+
i:si n g g %
n g gjg D.
Degradation of the fill material as observed by the presence of moisture or debris filling the seismic gap?
III.
Misc. Observations:
CM YES NO. AnadNeemstothisform ILarge equipment includes steam generators, reactor vessel, and NSSS piping
(
i N/A:
Not Applicable AREA OFWALKDOWN:
N/I:
NotInspectable INSP: Inspected SHEET OF
s.
MN 1-319 Structure and System Walkdowns Revision 2 Page 43 of 50 ATTACHMENT 12, WALKDOWN REPORT CONTINUATION SHEET WALKDOWN REPORT CONTINUATION SHEET NAME:
DATE:
/
/
SYSTEM:
SECTION/
SPECIFIC SPECIFIC FINDINGS / ACTION TAKEN COMPONENT IR / MWO NUMBERS i
r SYSTEM PERFORMANCE DATA COMPONENT SHORT COMPONENT SHORT ID DESCRIPTION VALUE ID DESCRIPTION VALUE SYSTEM ENGINEER GUIDANCE
- 1. Momtor and investigate system weepage; report any leakage. Secuan XI systems may require operability determination.
- 2. Report ALL safety concerns to the responsible supervisor h==* on Issue Report, Gold Card or Safety Audit Form.
- 3. Verify that an observed condition has not been previously identified and evaluated.
- 4. Investigate and report significant discrepancies and or changes in system parameters N/A-NotApplicable AREA OF WALKDOWN:
N/I:
Notinspectable INSP: Inspected SHEET OF
MN-1-319 Structure and System Walkdowns Revision 2 Page 44 of50 ATTACHMENT 13, PIPE SUPPORT INSPECTION GUIDELINES (Page 1 of4)
"Ihe followmg Fuare su..
M from ES-002 Pipe Support inspection Standard, and are intended to be a reviewed for a quick refresherjust prior to performing a walkdown. If a more co.uygh..dve review is desired, refer to the original ccus.snt. In all cases the reviewer shall defer to the performance p=hrds contained in ES 002 as the ie....iuiug element for concluding that a structural feature does not meet design requirements and is degraded to the extent that the requirements of P h(a)(1) of the Man >=are Rule applies (ref. MN-1-ll2). Determmations should be consistent with intent and cantant of ES-002.
A.
GeneralConditions 1.
'Ibe general condition of the pipe support should be assessed. Problems may include any of the followag Damage deformation or structural degradation of fasms, springs, clamps or a.
other components.
b.
Missing, h% or la==4,..yceests.
Indications of scahng on anyp' support U- -==a, weldmg surface c.
indications or sigr ofscaling to corrosion'that may reduce the load beark capacity of the support.
d.
Arcstnkes weld soonng, roughness, or general corrosion on close tolerance =,=+
or si surfaces.
Dented or cracked spnng hanger housings.
c.
f.
Indications of concrete or grout deterioration such as erosion, corrosion, chipping, crackmg or spalhng within 12 inches from the edges of pipe support base plates.
g.
Fluid loss beyond speci5ed limits or lack of fluid indication (Saubbers only).
B.
Threaded Connections 1.
Threaded connections on pipe support cc. pce,-. should be htea~*M to confirm the engagement and tightness of the ah includmg, but not lirinited to; concrete anchor bolts, rod hanger turnbucides, weldless eye nuts, rod miplin==, sway stnrts and load pins.
For example, the followmg am ways in which the connection can be veri 5ed (refer to ES-002 for applicable standards):
Visible threads can be confirmed by verifying that the male thread is at least flush a.
with the top of the female thread b.
For threadmg with tubular devices, such as sway struts, the prope
- "- t can be F-..^. +1 by the observation of11. A surfaces within the sig
i 4
+
MN-1-319 Structure and System Walkdawns Revision 2 Page 45 of 50 A'ITACHMENT 13, PIPE SUPPORT INSPECTION GUIDELINES (Page 2 of 4)
C.
Structural Frames / Support Steel 1.
This refers to pipe supports==,a=A of structural members such as wide flange, angle, channel, tube sections, plate or bar sections. Usually a member is in direct contact with a pipe (restramt) or in close proxmuty Also referred to as a " box" frame, it may work in conjunction with other pipe support components. Typically, the " box" frame is used to restram piping in more than one direction. As a two-way restraint, this sup movement in the axial swbce. When used in combination with pipe lugs, port allo it can restram movement in the axial duection also. For example, the followung are features of structural frames that should be exammed (refer to ES-002 for applicable standards).
%ese supports typically have gaps in the bonzontal and vertical (upward) a.
diim,dess. The pipe will be restag on a member in the vertical (downward) duection. The required gap will be generally shown on the pipe support drawing.
The two types ofgaps used are "restramt gaps" and " clearance? If unsure what type is bemg av=H refer to the pipe support drawing, which will generally show the requued gap.
(1)
" Restraint gaps" are not greater than 1/16 inch, and are between the pipe and the nearest structural member. His gap is provided for the radial expansion of the pipe and to allow imtial installation.
(2)
" Clearances" are usually greater than 1/16 inch in " cold" conditions and allow pipe movement within the frame.
(3)
Shims may be used to achieve the design gaps. Gaps less than 1/16 inch on cold pipes should be further investigated. Gaps greater than 1/16 inch on hot pipes should also be investigated.
b.
Check general design geometry and weld configuration and size, for co b.m with pipe support drawings (if used).
D.
Rod Hangers 1.
Rod hangers restrain the pipe in the vertically downward direction only and are a- -:=4 ofvarious threaded components attached to the piping system by a pipe ch.p. Rod hangers are designed to articulate, all free motion of the p pung system m the horizontalplane. The assemblyis y limited to a "swmg" angle of 5' total travel.
a.
Check that rod e=nanants are loaded (ti
). His can be w=ua by
{
=*~=*6 to manip'ulate the rod assemb by hand if accessible, or visually l
Mndorloose or vibratmg If system is dramed, venfy any observations afterit's been i
b.
Check that the pipe surface is 9- :--^t the pipe clamp Check to see that the Rod hanger "swmg" angle is less than 5* from vertical.
c.
(Five degrees of swmg is apyivu.uiady 1 inch per foot oflength.)
i 1
1 1
1 a
MN-1-319 Structure and System Walkdowns Revision 2 Page46 of 50 ATTACHMENT 13, PIPE SUPPORT INSPECTION GUIDELINES (Page 3 of 4)
E.
Spring Supports 1.
Spring hangers are similar to Rod hangers, they support the " dead weight" of the system allowmg free movement in the horwantal plane, however, they also allow it limited vertical freedom of motion. Spring hangers are engineered so that they are never t-:=:a+1 or out. Check Sprmg supports in the same manner as for Rod hangers (4) with the additions:
Check for spring settings on the spring can. 'Ihese consist of marks located en a a.
calibrated plate (spring scale) adjacent to the spring tion indir=w. Hot and cold settag marks should be indicated on the plate.
normal operatmg position should indicate between the hot and cold settings if the piping is flooded. If the spring scale is nu'ssmg, Maced or not clearly marked, or the indication is outside of the marks, contact MCEU and ististe and Issue Report.
b.
Verify that travel stops (piu blocks) are not installed dunng normal operanon F.
Constant Support Hangers 1,
Cw= d Support hangers accomplish the same function as hangers and are checked j
according to Rod hanger (4) and Spring hanger (5) guidelmes hangers consist ofa i
mechanical hnkage tied to a spring similar to a spring hanger.
G.
Sway Struts 1.
Sway struts perform a sirmlar function as Rod hangers in that restrict r.otioninone plane while allowmg free movement in other planes. Because of construction, sway struts are not limited to vertically downward loads. They can resist monon in compression as well as tension, regardless of orientation.
Check that the Sway strut ball bushmgs (one at each end of the strut) are:
a.
(1)
Not dislodged, loose, or missing, (2)
Not pon.eding more than one quarter of the ball bushmg thickness from the side of the sway strut assembly.
(3)
Tight and free to rotate.
b.
Spacers must be clamp assembly.present, but not welded to the =~M== bracket and the pipe The pipe clamp assembly must be tight and makmg good contact with the pipe.
c.
d.
Check for thread engagement (see step B.I.b).
H.
Concrete Atta4===t Plates 1.
Concrete =~h=aat plates consist of emhaddad plates or plates attached to concrete surfaces using various types of bohs or embedded anchors. The purpose of these plates is to transfer loads to the concrete structure.
a MN-1-319 Structure and System Walkdowns Revision 2 Page 47 of 50 ATTACHMENT 13, PIPE SUPPORT INSPECTION GUIDELINES (Page 4 of 4)
I H.
Concrete Attache Plates (Contmuod)
Check that bolts or nuts are at least hand tight, properly engaged a.
this =*=>h=t, and that the bolt head, nut and washer are in um,per step B.1.b of form contact with the base plate, b.
Check that the plate is flush with the concrete surface to withm 1/8 inch unless it can be deternu,ned that the gap occurs outsule the penmeter established by the outsule surface ofthe anchor boks.
1.
Piping Weided Attach-ts
)
1.
Piping welded attachments are lugs welded directly to the pipe, intended to transfer axial j
loads to a support structure or pipe clamp. These lugs must be installed trum.dy close to the adjacent pipe support structure or clamp, to maimize dynamic landmf uring a d
seisnue event.
Check that the gap between the lug and adjacent support structure or clamp is less I
a.
than 1/16 inch generally and 0.0 inches for lugs transferring vertically downward i
loadms I
J b.
Check that any shims used to actueve these dimensions are less than 1/4 inch thick I
and wided J.
Viscous Dampers (Soubbers) 1.
Viscous dampers (Snubbers) are used on piping systems to control vibration or rapid movements dynamically. 'Ibey do not transmit static loads or restnet slow movements (thermal expansion). 'Ihey generally resemble shock absorbers with an attached fluid reservotr. These are m=H=8 g-
==a and any unusual appearances should be discussed with MCElf, including any of the following a.
Missing rubber boots, straps or bands.
b.
Ixaking fluid or suspiciously low fluid levels.
Travel-bound conditions, Snubbers that are wayd'y extendad or compressed.
c.
d.
Loose pipe clamps =*=^hia a snubber to the pipe.
I I
i
a
~
MN-1-319 Structure and System Walkdowns Revision 2 Page 48 of 50 ATTACHNIENT 14, REFUELING EQUIPMENT WALKDOWN REPORT A. SPENT FUEL HANDLING MACHINE WALKDOWN REPORT (Page 1 of 3) i NAME:
DATE:
/
/
SYSTEM:
Ul MODE: 1 2 3 4 5 D U2 MODE.1 2 3 4 5 D N/A N/I SPECIFIC FINDINGS / ACTION TAKEN I
INSP DESCRIFTION GR/MO NUMBERS)
- 91.03 SFHMBRIDGB AND TROLLEYm mNUIES ex% ~ memn atme e tvet sta ~ <ae A. Damaged?
l B. Corrosion?
C. Detenonauon?
I l
l Continued? Yes No
- 42.00 SFHMHOISTCABIE ASSEMBLh9xNOTESta cat tvWoaWwmem aw:w Wew: e A. Broken strands or kinks?
1 i
B. Sheaves worn or damaged?
1 C. Load cell worn ordamaged?
I IContinued? Yes No w3.0 i* SFHMFNEUMATIC CCNIRCLPANELN NOTES:ndommmsMem6em.en emrm A. Hydraulic lines worn or demated?
B. Missing or damaged gauge glasses?
C. Looseorleakingfittings?
} Continued? Yes No i
- :t4.0o SPHMLIMITSWI!UES ANDCAMSm;6NOIES:e 4+: e m e**
- N er++
>-.tm
- w
+
l A. Damaged?
B. Obstructions that impede operation?
l Continued? Yes No Bn!5.0. SPHMBRIDGE; TROLLEY;HOISTENOTES:%etydt@@
p!!
uECHEEMM! MIN $$$$$$$$$M
- % gae>M
+ntg+mf+
"^*
A. Lubricantleakage?
B. Worn orloose couplings?
!C=*iased? Yes No W6.0giSFHM RAll; SURFACES AND GEAR qNOTES;q;dgySh AtM?N.N@@:!:pb%gggpWggg:c s
dngg;sa
' i!!NT@% RACK TEETH WdiGb9MtMISSMNFe*M a.
Dtst>w.G A. Clean and free of debris?
K'aaniw? Yes No N7.0 a SFHM ELECTRICAI;ett%e anWNOTESais Neim-w + "rnWG4 +w.%urewme.
A. Motors condaten? Noise? Vibrabon?
1 B. Cables candi'ian? Damaged?
^
Chaffed? Connections tight?
C. Wiring condition? Broken strands?
Chahg? Embrittled in=1stion?
D. Coatil panel condition? Switches?
Wiring? Lights?
E. Rad nenitorsecure?
IContmuod? Yes No i
t N/A-Not Applicable AREA OF WALKDOWN:
l N/I:
NotInspectable INSP: la W M SHEET OF 1
J
MN-1-319 Structure and System Walkdowns Revision 2 Page 49 of 50 ATTACHMENT 14, REFUELING EQUIPMENT WALKDOWN REPORT B. REFUELING MACHINE WALKDOWN REPORT) (Page 2 of 3)
NAME:
DATE:
/
/
SYSTEM:
UI MODE: 1 2 3 4 5 D U2 MODE: 1 2 3 4 5 D N/A N/I SPECIFIC FINDINGS /ACTIONTAKEN INSP DESCRIPTION (IR /MO NUMBERS)
+d.01RFM BRIDG ! ANDTPW W> m* + Nuirs:' m-mt-aa
- w
>*-*waA~
+
A. D a -gedJ B. Corrosion?
C. Deterioriation?
D. Festoon cable worn or damaged?
IContinued? Yes No e 2.0 mRFM HOIST CABIE ASSEMBLY min Nurrx m n*4 " "
2ew * ' ^ * -
+
A. Bidou strands or kiuks?
B. Sheaves worn or d===gM7 C. Load cellworn or damaged?
ICo. - -- *7 Yes No r 3.0, RFk MAST r'm -
ION RINQ w a =Nu an59.*+e
<u-ew s<"
~m<e A. D u ged?
B. Tag lines for limit switches properly installM7 tContinued? Yes No i
m4.0
- RIMPfRNA11CCDNIELPAMLeetaNOirs: 'www-me etg-mmme-Nwem*
A. HyGaulc lines worn or damaged?
B. Missing ordamerM gauge glasses?
C. I.aose or leaking fittings?
IContinued? Yes No
- 5.0-RFM LIMIT SWiiCHES AND CAMS *NOizSMet 4
w~ ^
< *'* * ~ *
- m "* '
A. Damaged?
B. Obstructions that impede operation?
IContinued? Yes No
- 649RFM BRIDGEst1RO11EY; HOIST,;:a:
g& M ' 7ANDMASTROTATE GEARMOTOR Nuszag awrld#wef+xyMqnw " g
- r;g e
0 n$
I
?<
ts "N )M b t 4 d^2 M *;';' *. U d 5d 3 m pM M $ 6 TitiA iDRIVE MECHANISMS MaaW.WWW - -
A. Lubricantleakage?
B. Worn orloose couplings?
IC=6nued? Yes No 3909RFMRAIL: SURFACES AND GEARtwg:.NOsus@%g::
Mn angy
+ M*stacq,g~, gats'<rfig*ggega;$yggpf c.nHRp:RACKTEETH@+3mMWiW2Mb #Fhab rWL-22 A r4A - tSh**
t A. Cleanandfreeofdebris?
ICa* M7 Yes No 5363.0inRFMELiiiumICAbastesmit&Oex NuluS;wWimmm&immeWM:ciw;6@#Nu e
A. Motors condition? Noise?
Vibration?
B. Cablescondition?
Chased? Connections ti ?
C. Wiring condition? Broken strands?
Chaffing? Embrittledinsulation?
D. Control panel condition? Switches?
Wiring? Lights?
lContinuc0 Yes No N/A-Not Applicable AREA OF WALKDOWN:
N/I:
NotInspectable INSP: Ta W SHEET OF
1 MN-1-319 Structure and Systern Walkdowns Revision 2 Page 50 of 50 ATTACHMENT 14, REFUELING EQUIPMENT WALKDOWN REPORT C. FUEL TRANSFER EQUIPMENT WALKDOWN REPORT (Page 3 of 3)
NAME:
DATE:
/
/
SYSTEM:
U1 MODE: 1 2 3 4 5 D U2 MODE: 1 2 3 4 5 D N/A N/I SPECIFIC FINDINGS / ACTION TAKEN INSP DESCRIPTION (IR /MO NUMBERS)
=W1.0 :: FUEL TRANSFER CARRIAGE pair. NOTES %xnter4@Nwhnt - +Ne:m.wwe A. Operates smoothly without binding?
B. Cable has nobroken strands or kinks?
IContinued? Yes No
)
210-FUELUPENDER * *.uwemwomNOTES:vounsie *.3 mMMSmWWWNe' A. Bearings worn or damaged?
B. Sheaves worn or da===A?
C. Sheaves operate smoothly?
D. Welds at the hydrauhc cyhnder gusaets and attachmg pointsintact with no cracks?
Kbatinued? Yes No 43.O r SPELEVATQtS tacre*mTG' ? nwNO1ES- % w-:e ^vrewwnwemwn - ex A. Cables have no broken strands or i
kinks?
B. Rollers worn or damaged?
C. Rollers operate smoothly?
D. Roller tracks worn or da===A?
E. Roller track weldsintact with no cracks?
l Continued? Yes No j
44.0uTRANSFER CARRIAGE ANDMme F w % c:
r.q.p;ib iMUnikELEVATORS ELECTRICALWMi'%sNOIES:Sifgpw*A..5fEEii@ ! ' ${;f g y:Mi'L!r+2
]
2is%4*inedI 4:'x w1 A. Motors condition? Noise?
Vibration?
B. Cablescondition? Damaged?
Chaffed? C=~ians tight?
1 C. Wiring Condition? Broken Strands?
twmag? Embrittled Laaita*iaa.?
D. Control panel condition? Switches?
Wiring? Lights?
iContinued? Yes No N/A: Not Applicable AREA OF WALKDOWN:
N/I:
Not Tap =ble INSP: Inspected SHEET OF i
j