ML20029A724
| ML20029A724 | |
| Person / Time | |
|---|---|
| Issue date: | 02/25/1991 |
| From: | Jocelyn Craig Office of Nuclear Reactor Regulation |
| To: | Griffing E NUCLEAR ENERGY INSTITUTE (FORMERLY NUCLEAR MGMT & |
| References | |
| NUDOCS 9103040139 | |
| Download: ML20029A724 (31) | |
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fdnury 25,19)1 Mr. Edward Griffing, Manacer l
Technical Division Nuclear Management and Resources Council I M6 E ve Street, h.W.
Suite 300 Washinaton, D.C. 20006-?496 Dear Mr. Griffino
SUBJECT:
STATUS OF THE RESOLUTION OF NRC COWENTS ON INDUSTRY REPORTS FOR THE BWR REACTOR VESSEL (90-02) AND BWR REACTOR VESSEL INTERNALS (90 03)
A meeting was held between the NRC, NUMARC, et al. on January 22 and 23, 1991 to discuss the NRC's evaluation of NUMARC's response to previously published i
NRC coments on the two subject industry reports (irs). lists the attendees to that meetino.
The original NRC coments were documented in letters dated April 2,1990 and July 6,1990s with the formal NUMARC response documented in letters dated October 26, 1990 and November 20, 1990 for the PM Reactor Vessel (RV) 1R and P,WR RV Internals IR, respectively.
The ARC's preliminary classification of the NUMARC responses was documented in published records of telephone conversations dated January 2, 1991 and January 11, 1o01 for the BWR RV 1R and BWR RV Internals IR, respectively, from those two telephone conversations a meeting agenda was established to address those responses which the NRC had not rccepted and it was published in a revised meetino notice issued on January 18, 1991, in sumary, of the 138 coments originally identified by the NRC, 54 were acceptable without further discussion (as noted in the documerted telephone conversetions) and another 5 were acceptable due to further internal review by the NRC subsecuent to the aforementioned telephone conversations (as noted by their omission from the published agenda dated January 18, 1991).
Therefore, 79 issues were the subiect of the January ?? and 23, 1991 public meetina l
betweea the NRC and NUMARC and are sumarized in Enclosure 2.
l There are additional outstandino NRC comments that were forwarded to NUMARC on November 30, 1990 for both irs which have not been responded to by NUMARC.
These issues will be the subject of a future meetina.
Durino the public meeting a discussion paper 0-fatique was presented and distributed by the staff.
A copy of this drai _ discussion paper on f atique is attached as Enclosure 3 to this letter.
Even thouoh ecuipment ::ualificattor.
A was included as the last item on the published agenda, it was not discussed, Lj~03 g\\
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fdnery 25, 1991 but it will be the subject of a future nietting between the NRC and NUMARC, tentatively scheduled for April 16, 1991.
The reporting and/or recordkeeping recuirements contained in this letter affect fewer than ten correspondents: therefore, OMB clearance is not recuired under P.L.96-511.
If you have any questions concerning this sumary, please contact me or P.T.
Kuo at 497-3147 Si
- erely, Ib John W. Craig, Director License Renewal Project Directorate Division of Advanced Reactors and Special Projects Office of Nuclear Reactor Regulation
Enclosures:
As stated l
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Pr. Eoward Griffing 2
February 25,19?)
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i but it will be the subiect of a future meeting between the NRC and NUMARC, tentatively scheduled for April 16, 1991.
l The reperting and/or recordkeeping reautrements contained in this letter affect fewer than ten correspondents: therefore, OMB clearance is not reauired under P.L. 96 511.
j If you have any cuestions concerning this sumar), please contact me or P.T.
}
Kuo at 492 3147.
Sincerely, f
/S/
i h hn V. Craig, Director License Rerewal Froject Directorate Division of Advanced Reactors and Special Pro.iects Office of Nuclear Reactor Reoulation
Enclosures:
As stated i
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[nclosure 2 (11 Pages)
Surnary of FWR Reactor Vessel IR and BWR Reactor Vessel Internals IR ltems The following items were discussed in a public meeting between the NRC, NUMARC, et al. held on January ?? and 23, 1991 at the NRC offices in Rockville, Maryland. The order in which items were cresented was consistent with the published acendo of January 18, 1991. A brief description of each of the acreement-in principle (AIP) or open items identifying the needed action by NUMARC is provided, items which are not addressed here but are identified in the NUKARC responses deted October 26, 1990 and Novert.ber 20, 1990, are acceptable to the NRC pendino incorporation of the specific responses into the Industry Reports.
Fatigue - BWR RV ltem Number Alp or Open item G-08 Open - These items are held open, G-09 pending NUMARC's review of the NRC's G-10 draft discussion paper on fatique.
G-11 NUMARC was reouested to provide 5 20 concents, if any, at the earliest 5-47 possible date.
G-12 AIP The IR will clearly state that if the temperature limit between the top and bottom of the EWR-2 reactor vessel is exceeded, a plant specific analysis will be reouired.
G-17 A1P - Section 3 2 will be revised to identify time-dependent considerations of the requiatory instruments.
Included in this assessment will be an evaluation of the applicability of the current 151 reouire.aents to detect the specific age-related deoradation mechanisms of concern, as well as, a discussion on environmental assisted fatigue (EAfl.
5-15 AIP - Justificction for limitino thernial f atioue to only the feedwater nozzles wil' bt provided in the IR.
5-17 alp - Wili provide a discussion in Section 6.0 regarding the potential for combining effects of age-related degradation mechanisms.
For example, if IGSCC is present, the
I s
-2 Fatique-BWR RV (continued)
Item Number alp or Open item potential for fatigue to propagate the crack needs to be considered.
S-?)
Open - NUMARC is to provide supportina information for its position that fatigue is not a significant degradation mechenism for flanges in a BWR environment and is to address the apperent discrepency with the ASME Task Group on fatigue in Operating Plants.
S-??
AIP - Same as 5-17 S-46 Alp - Same as 5-15 Faticue-BWR pV Internals G4 Open - Same as G-08, etc. (Acreed 57 that loading spectrum is a today's problem and need not be addressed specifically for license renewal.)
5-26 AIP - NRC accepts the NUMApC response.
General and Loose Parts.
REP RV Internals G-1 AIP - Section 3 will remove statementt like" repairs could be implemented before plant safety is affected" etc. Other conclusions will be.iustified.
G-2 AIP - It is centrally recognized that the IR needs clear and crisp criteria which an applicant can readily understand.
The type c' detail needed including ample justification for conclusions will be provided.
- 5 AIP - An item by item response to each of the issues raised in the i
original NRC conment and i
justificatior for why they are no longer applicable will be provided.
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-3 rieneral and loose Parts-BWR RV Internals (continued) l Item Number AIPorOpen1j G6 AIP - IR will justify how all components within its scope are e
adecuetely managed such that they do not become loose parts and/or j
a safety concern.
G9 AIP A discussion on common mode failures under accident conditions due to ace-related degradation will be provided including the upper head support and the reactor head bolts.-
i G 1?
AIP. IR will provide details of the inspection prcgram and
- criteria, included will be an evaluation of the applicability of the current 151 recuirements to detect the specific ege related degradation IMchanisms.
G-16 AIP - All attachment welds, including all those listed in the original NEC coment, will be addressed in the BWR RV 1R.
G-17 AIP - Current acing management programs will be evaluated to determine their effectiveness for license renewal egino management.
Sections 3.2 and P will be revised.
$-2 AIP - Same as G-6 5-12 AIP - Seme as G 6 Radiation Embrittlement - BWR RV Item Number AIP or Open item G.04-AIP - Add annealing as an option in Section 6.4
$-04 AIP - Provide'the basis for the bounding analysis which i
demonstrates that the LPCI norries in BWR/5 plants are not the limiting item for the reactor vessel with respect to neutron irradiation embrittlement.
.. _. _., _ _ _ _ __-. _.-_ _._,_ _ _. _ _ _, _,..... _ _ 2 _ _ _
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e4e Fadictinn Ed rittlement-BWF RV (continued)
Item Number AIP or Ooen Item
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S-07 AIP - Provide background statementthatR.G.gthe information regardin 4.99 may need revision.
5-09 AIP - The NRC agrees with NUMARC response. This is a today's operational problem, not e license 1
renewal issue.
5 Air. Revise IR, page 5 13, to note that one die nsional neutron flux calculational methods need to be validated and documented.
i 5-51 AIP - Review of the 10 CFR 50 Appendix H surveillance prooram will be performed for the entire program on a plant-specific basis rather than only reviewino the chances j
to the surveillance program.
5-5?
-AIP - 10 CFR 50 Appendix H cannot be used for the license renewal period because it presumes a 40 year life. License renewal requires changes to the_
surveillance prooram including withdrewal schedule and/or adding i
new capsules.
S-54 AIP - The IR will' clarify that the initial reference temperature and initial Charpy upper shelf energy may be deteimined from statistical analysis-of the test results of weldments made from the same type of-flux and wire as the beltline weld.
The determination of the amount of copper and nickel for missing materials may be
. determined Statistically on1y by usino chemistry results from weldments prepared using the same heat of wire used to fabricate the beltline welds, if the actual heat of wire is unknown or-unavailable, either a bounding 6
analysis or the results from chemical analysis of-sampics from the beltline weldments must be used.
._ _ - _. - _ _ _. - _ _ _... _. _. _ _ _ _ _.. _. _.. - _. _.... _ ~. _ _. _.. _ _ _ _. - -... - _.. _. _ _, _ _ -,
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9 5
i Radiation Embrittlement +BWR RV (continuedl f
Item Number AIP or Open-Item S-60 AIP. The test results and neutror fluence and flux used in the reactor vessel support i-analysis will be provided.
4 L
S.67 AIP Yessel annealing will be i
added as an option for license renewal. Also, flux reduction a
will be addressed for those vessels requiring some mitigating ection to address embrittlemerit.
s Radiation Embrittlement. PWR PV internals i
item Number ATP or Open item S-18 A1P NRC accepts the NUMARC response.
}
Thermal Aging. BWR RV Internals item Number ATP or Open item S.19 A!P - Additional information will be provided to justify that cast l
stainless steel components are not significantly susceptible
- to the combined effects of thermal aging and irradiation.
NilMARC indicated that the orifice fuel support was the only!such component. ' Justification should include a discussion of low stresses which are below the threshold for: crack initiation, material resistant to IGSCC, no residual--
- stresses, inspected every four years, etc.
S.?0 AIP - The extent of-fracture toughness loss wd1 l
be tated and justified, and i fracturc frechanics analysis will be conducted to-determina the flew size that can be detssted-
- reliably.
The technicues to be used for detecting the flaw size of interest must be cualified and proven effective.
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9 6-fracture Mechanics - PWR RV Internals item Number AIP or Open item F-6 alp A comparison of the slip dissolution model and the classical approech to predict crack growth will be providst. Also, Section 6 of IR will include the slip dissolution method as an aging management option but with a caveat that the NRC has not reviewed nor accepted it.
$ 27 alp NRC acceots the NilMADC response.
5-?9 AIP - 1R will be revised to describe the control rod replacement program which is in the 5-8 year internal.
- Also, identify that the control rods useful neutron life oeverns its replacement in lieu of other meterial acing considerations.
1GSCC-PWP RV Item Number AIP or Open item S-7 AIP - NRC accepts the NUMARC response.
5-30 AIP - NRC accepts the NUMADC response.
5-36 Open - NUMARC nes t to provide additional justift = tion for their position that the capred CRD Return Lines are not susceptible to 500 ir.cludino consideration of their service history, residuel stresses, current inspection recuirenents, stagnant squeous environment, etc.
F 42 AIP - NRC withdrew the commer t.
5-45 AIP - Same as 5-6 (under fracture Mechanics - BWR PV Internals).
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i 7
SCC-RWR RV liem Number AIP or Open item 5-01 A!P NUMARC will revise the IR and its response to address corrosion of closure studs.
Recent documented corrosion of RV closure studs at Dresden Unit 2 demonstrates that they are susceptible to SCC.
5-11 Alp - Rather than identifying _
furnace sensitized CR0 stub tubes as an item reouiring an Age-reisted deGtrdation management program, since only two plants are involved, NUFARC decided to eliminate them from consideration due to budgetary constraints. 1herefore, Section 2 of the IR will be revised accordingly.
5-12 AIP - same as 5 11 5-27 AIP - NUMARC agreed to augment their respnnse to justify limited crack propagation ir,to base metal from stainless steel cladding. Weld repaired areas will be addressed and ictntified as an emerging issue.
5-28 AIP - Same as S-27.
5-29 AIP - Providt discussion on circumferential cracks in IR, includine ecceptability of current aging management programs.
S 33 AIP - Remove first sentence from NL' MARC's response.
5-37 AIP - Since SCC has been observed in 7 out of hundreds low alloy steri nozzles the issuc will be resolved in Section 5 of IP rather then dismissed in Section 4 5-38 alp - Same as S-01.
Its? Number AIP or Open item S-65 AIP - Volumetric inspection and water chemistry control procrams will be more fullv described in the revised IR.
In addition, the resoonse on volumetric examination will be expanded.
5-66 AIP - NRC accepts the NUMARC
'esponse.
,5CC - BWF RV Internals item Nun.ber AIP or Open item S-17 AIP - The IR will address stress levels for ail components not just the
- ring". Also, the IR will cuantif y stress levels and demonstrate that the resulting stress levels are below the threshold for crack initiation, inspection - BWR RV liem Husher Alp or Open item G-05 AIP - The IR will recognize that not all welds can be inspected and will provide justification f or not inspettino cil welds.
Open - As a minimum, 151 reouirements should be eauivalent to those of ASPE Section *,1, 1989 Edi+. ion including Appendices V11 and Vill. Minor exceptions tc accessibility will be reviewed by the NRC on a case-by-case basis.
G-18
.a!P/Open - Seme as G-05.
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9-1 inspection -BWR Ry (continued)
- Item Number ATP or_ Open item 5 03 Air-Sameas5-01(SCCBWRRV) 1 and IR will provide references identifying acceptable inspection enethods applicable to long, i
slender studs.
S 0F AIP - NRC accepts the NUPARC response..
S-41 Open_. NUMARC was requested to I
consider use of ASME Section XI, 1989 Edition (includingAppendices Vl! and Vill) as a minimum standard for 151 recuirements.
i S-53 Open - See G-5 and S.41.
- 5 56 AIP - The NUP. ARC response is not consistent between its stated incustry response and what is proposed to change _in the 1R.
Also, NUMARC recognizes this as a generic comment applicable to other items and will revise the 1R accorEloly.
5 Inspecticn - BWR RV Internals _.
ftem Number ATP or Open item S-03 AIP - Exemptions ere to be evalueted by NRC as part of the LR application on a case by-cese basis. Adecuacy of-AStiE Section-XI is discussed.in G 05 end S-41 (see Inspection BWR RV).
S-04 A1P - NUPARC will revise the IR
format to usist tne applicant in.
determining-its compliance w'.th-the 1R's general conclusion, j
S 05 Open - The aging mananmen' programs addressed in Secti0n 6 neeo *o have some inspection-provisions.te ensure thet they i
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e 10-Inseection BWR RV Internals (continued 1 Item Number AIP or Open item S-05 c;ntinue to be effective subseouent
^'
to their implementation (e.g. water chemistry control and refurbishment should both make use of inspection to ensure their continued effectiveness). NUMARC is to discuss this at its working group meeting.
-S-10 AIP - See comments G-12 (under General and loose Parts - BWR RV Internals) and 5-19 (under Therral aging - BWR RV Internals)
S 11 AIP
'?ction 6 will be revised to orv: ne a general methodology for volumetric inspection of core shroud, top guide and core plate.
Details are only available from industry for the top quide which will serve as en illustrative example for the applicants w.r.t.
the level of detail expected.
5-15 AIP - Same as G-12 (under General
& Loose Parts - BWR RV Internals 1 S-25 AIF - Same as G-12 (under Ger.'ral
& Loose Parts - BWR RV Internals)
S 32 AIP - Same as G-12 (under General
& Loose Parts - BWR RV Internals)
Erosion / Corrosion - BWR RV Item Humber AIP or Open item A-40 AIP - Justification for creas where erosion / corrosion is not a concern will be provided, (e.g. use of steinless steel, addition of oxygen to feedwater, low velocity, etc). Also,.311 erea.s composed of carbon steg1 and subjected to hydrogen water chemistry will be evaluated for erosion / corrosion.
l; i
-11 Ch,emistry - BWR RV Item Number alp _orOpenitem S-26 AIP - NRC accepts the NUPARC
- response, S.32 AIP - NRC accepts the NUMARC response.
1 S-34 AIP - The IR will provide justification as to why oxidation eccumulation is not a serious l
concern, specifically during abnormal chemistry operation and as a hindrance to reactor water I
cleanup.
5-35 AIP - Same as 5-34 l
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En cle,sace 3 64 gn3,3),
1 01/23/91 DRAFT DISCUSSION PAPER ON PATIGRI 1.
cENEP.AL APPROACH The staff approach in evaluating the potential fatigue damage to reacter be within the proposed license eenponents for license renewal is to (1) specifically, the adequacy of the current licensing basist (2) renewal rule, be sensitive to the ongoing concern regardirg the adequacy of the code in addressing environmental effects; (3) ensure remedies for fatigue damage experienceds (4) reconstruct actual transient load cycles (5) permit fatigue and evaluation based on rnalysis or a combination of analysis and inspection:
have augmented inspection as a part of the f atigue mariagement program.
(6)
Paragraphs 1.1 through 1.6 provide brief discussions of elements of the staff approach.
Details of the specific evaluation procedures depend on the ASME Code class of the corponent.
For recent vir.tage plants, safety-related components sre classified as code Class 1, 2, and 3 according to Regulatory Guide 1.26.
For older vintage plants constructed to an industry standard other than ASME safety-related components are reclassified as code Class 1, 2,
Section III, for the purpose of inservice inspection, specified in ASME section XI, and 3 Because of different code requirements, according to Regulatory Guide 1.26.
separate procedures for ASME Class 1 compotients and for ASME Class 2 and 3 ccmponents are provided.
flaws detected and evaluated according to Section XI in the In addition, shall be reevaluated for Section XI acceptability to the end current license, of the license renewal term.
Further, it is noted that in the present discussion, the wording "ccaponent" also means "part of a component".
1.1 ggrrent Licensinn Basis The proposed 10 CFR Part 54 license renewal rule states that the current licensing basis (CLB) is acceptable.
It also states that the CLB must be maintained during the license renewal term, in part through an effective program for managing the age-related degradations.
As relating to the fatigue damage, the CLB is the number of design transients or load cycles originally assumed in the licensing of the plant, as supplemented by licensee commitments during the original license term either through plant specific actions or generic actions.
The specific numbers of design transients are listed in the plant Final Safety Analysis Report (FSAR).
Technical Specifications, or other docketed documents.
There are also additional numbers of certain transients assumed in the fatigue evaluatien because of generic actions, such as resulting from NRC Bulletins and Generic Letters.
1
r In the original licensing of a plant, the statt made a determination that there was adequate safety relating to the fatigue lives of reactor components This number if the plant operates within the number of transients postulated.
An applicant for of design transients is the CLB for the f atigue issue.
license renewal shall demonstrate that the number of actual plant transients, projected to the end of the license renewal term, shall be bounded by the corresponding numbers in the CLS.
1.2 onceino concern The staff has a concern that existing code fatigue evaluations may have less cargin than that originally intended in the code when considering effects en the fatigue life by factors such as environment.
Specifically, the onvironmental ef fset on the f atigue life has been identified as a potential generic issue and vill be addressed according to the current regulatory Should the resolution of this potential generic issue result in process.
backfit requirements, plants with approval for license renewal vould also be offected.
if a f atigue recalculation is required f or license renewal before the
- However, staff resolves the potential generic issue, the environmental effects on the fatigue life shall be included in the analysis.
1.3 RePedies for Fatieue Darace For components with a history of fatigus damage, the cause of the damage shall This may result from plant specific be remedied by corrective actions.
actions or generic actions.
An augmented inspection and/or monitoring shall be performed to assess the effectiveness of the fatique mitigation procedures.
1.4 Reconstruction of Actual Transients An applicant for license renewal shall reconstruct the actual transients experienced by the component to the extent practicable.
The plant Technical Specifications require tracking of certain transients.
The tracking If can be requirements are more extensive for recently licensed plants.
the applicant may use industry experience of similar plants or data justified, f rom actual transient tracking over a limited period of time to conservatively estimate the-actual plant transients.
It is also noted that transients shall be tracked from this time on.
1.5 Faticue Analysi_s A f atigue analysis of components shall be performed if the projected number of transients to the end of the license renewal term exceeds that in the CLB.
The fatigue analysis may be an analysis alone or a combination of analysis and inspection.
A bounding f atigue analysis may be performed for a group of components if sufficient justification is provided.
Also, similarity arguments may be used to demonstrate similar fatigue behavior in different components if sufficient justification is provideJ; i.e., addressing similarity in geometry, leading transients, stresses, materials, environment, operating history, etc.
2-
^^
- ~ - ~ _ - _ _
However, based on past staff experience, similarity arguments are coeplex and will be subjected to extensive staf f review.
1.6 Austented Insoectiqn An augnented inspection performed on a periodic basis to detect the presence The of fatigue flaws shall be a part of the fatigue management program.
extent and f recuency of the augrented inspection shall depend on' operating oxperience, fatigue evaluation methodology, confidence of the input to the the result of the analysis, and engineering judgement.
An
- analysis, inspection performance demonstration according to Appendices VII and VIII of Section XI shall be performed to provide added assurance of flaw detection and sizing capabilities.
Section XI inservice inspection schedules are based on an inspection interval of a 10 year duration, which is further divided into 3 periods.
The augmented inspection may follow a similar schedule and may be integrated into the However, augmented i'nspection locations shall Section XI inspection program.
not be limited to locations specified in Section XI.
2.
CLAS S 1 COMPONENTE Figure 1 shows a flow chart for the fatigue evaluation of ASME Code Class 1 components for license re' VS' It considers (1) history of f atigue damage, (2) reconstruction of act,
- l. t'snsients, (3) documented usage factor, (4) CLB, (5)Section III analysis, action XI analysis and Inspection, and (7)
Compw..nts shall be replaced, or committed to be augeented inspection.
replaced at a time shortly after license renewal approval, if they are fcund unacceptable by the fatigue evaluation.
2.1 Eistory of Faticue Darace The history of f atigue damage in plant specific cases and generic cases shall be reviewed.
If there is a history of damage, the f atigue mechanism shall te mitigated and remedial actions performed.
An augmented inspection and/or monitoring shall be performed to assess the effectiveness of the fatigue mitigation procedures.
Furthermore, generic actions relating to the potential for unanticipated fatigue loading, such as NRC Bulletins 88-08, " Thermal Stresses in Piping Connected to Reactor Coolant Systems," and 88-11,
" Pressurizer Surge Line Thermal Stratification," shall be performed as requested.
For example, if a thermal sleeve had become detached resulting in unanticipated cyclic thernal loading of the pressurizer surge line to hot leg:n nozzle, a corrective action may be to install a redesigned thermal sleeve.
this case, augmented inspection shall be performed to monitor the integrity cf the thermal sleeve attacheent veld to ensure the proper function of the thermal sleeve in mitigating thermal f atique of the nozzle.
NRC I&E Information Notice 82-30, " Loss of Thermal Sleeves in Reactor Coolant Syste-Piping at Certain Westinghouse PWR Power Plants," provides discussien rela:;
to the integrity of thermal sleeves.
3
Peeenstruction of Actual Transients 2.2 The applicant shall reconstruct the actual transients experienced by the plant Because the transient tracking requirements are to the extent practicable.
plant specific, the extent of transient reconstruction will vary from plant to Also, if can be plant depending on plant specific record keeping practices.
the applicant may use industry experience of similar plants or Justified, actual transient tracking over a short period of time to conservatively estimate the actual plant transients.
In any case, the confidence of the actual transient reconstruction affects the uncertainties in further ovaluations.
2.3 Docurented Usace Factor Tor plants designed to ASMI section III requirements, a fatigue cumulative is documented for each class 1 component to the and of the usage factor (CUT) original license.
.A screening criterion based on the documented CUT is The applicant shall review the documented Curs.
However, the developed.
applicant shall not recalculate the Curs in applying this screening criterion.
A recalculation of the CUT nay reduce the existing ccuservatism in the In view of the ongoing concern on the fatique original calculations.
calculation, it is prudent not to diminish < xisting conservatisms.
If the documented CUF is less than 0.4 for the component, the component is acceptable for license renewal without further analysis, provided a certain requirement relating to the number of transients as discussed below is satisfied.
However, further analysis is required if the documented CUT or if there is no CUT available such as for components designed exceeds 0.4, to B31.1 requirements.
The value of 0.4 is an acceptable threshold for further analysis.
The number 0.4 is consistent with that used in Table IWB-2500-1 in Section XI for inspection sampling, specifically, pressure retaining welds in Class 1 piping with a CUT exceeding 0.4 are required to be inspected as part of the 25% sampling of welds every inspection interval.
If an applicant has an original license for 40 years and is requesting license renewal for 60 years, the applicant is requesting to extend the plant operating license life by a factor of 1.5.
If the documented CUF for a component is less than 0.4, the CUT for 60 years shall be estimated from the CUF f or 40 years by increasing it by a f actor of 1.5.
Similarly, the number of design transients for 60 years shall be estimated from the number of design transients for 40 years by increasing it by a factor of 1.5.
For a component to satisfy this screening criterion, the actual transients projected to 60 years shall be less than the number of design transients for 60 years as calculated above.
For components with a low CUT, a conservative assessment may suf fice in reconstructing actual transients in Paragraph 2.2.
There are other regulatory requirements that depend on the CUF, such as the postulation of high energy pipe ruptures and the Section XI sampling as discussed above.
With regard to Section XI inspection sampling, the code has a further requirement that the initially selected weld samples shall be inspected in successive inspection intervals.
Thus, the weld samples initially selected during the original license term shall continue to te
- However, inspected every inspection interval during the license renewal tern.
the inspection shall be augmented with examination of welds having a pre]erte:
-4 l
I
~
CUF to the end of the license renewal torn exceeding 0.4, if not already in the saeple.
34 C1B The CLB is the number of design transients originally assumed in the licensing as supplemented by licensee commitments during the original of the plant, license term either through plant specific actions or generic tetions.
The specific numbers of design transients are listed in thc plant FSAP, Technical Specifications, er other docketed documents.
The plant Technical Specifi.ca'tions also require the tracking of some design transients.
In general, the listing of design transients is more extensive for more recently There are also additional numbers of certain transients licensed plants.
assumed in the fatigue evaluation because of generic actions, such as resulting from NRC Bulletins 88-08 and 88-11.
for license renewal shall demonstrate that the number of actual An applicant transients experienced by the component, projectsd to the end of the license shall be bounded by the corresponding numbers listed in the renewal term, If the FSAR, Technical Specifications, and docketed licensee commitments.
applicant cannot demonstrate to the staff that the projected transients would be bounded by the numbers in these documents, even in a single transient category, further f atigue analysis is required.
Similarly, if the applicant cannot demonstrate that the relevant transients for the component can be derived from those listed in these documents, further fatique analysis is reqJirsd.
il the original TSAR postulated 200 heatup cycles, the applicant For example, has to demonstrate that the component would not experience in excess of 200 heatup cycles to the end of the license renewal term.
Otherwise, further fatigue evaluation is required.
If the projected transients to the and of the license renewal term is within the CLB, further f atigue analysis of the component is not required.
- Hcwever, augmented inspection shall be performed based on (1) the extensiveness of the design transients in the CLB, and (2) the confidence in the reconstruction of actual transients.
In any case, augmented inspection shall be performed at every inspection interval on components with projected transients close to the CLB snd a CUT close to unity.
2.5 section III Analysis If further f atigue analysis is required because the CLB in Paragraph 2.4 is exceeded and there is adequate reconstruction of actual transients in Paragraph 2.2, a fatique analysis in accordance with Section III shall be However, because of the ongoing concern as discussed in Paragraph pe rf o rm ed.
1.2, the effects of environment on the fatigue life shall be addressed.
(!ne staff can provide technical references for the latest fatigue test data.)
Furthermore,thefatigueanalysisshallincludee((ectsoflow-amplitude high-cycle fatigue, i.e., f atigue cycles up to 10 The CUT calculated t:
the end of the license renewal term shall not exceed unity.
Ter In performing the Section III analysis, the Code of Record shall be used.
components constructed to B31.1 requirements, the 1971 edition of Section III 5-s
.~
shall be used.
for components constructed to the 1969 edition of B31.7, the Code of Record shall be used instead of Section Ill.
Tne CUT calculated to the and of the license renewsi torn shall also be used With in other related regulatory requirements as discussed in Paragraph 2.3.
regard to Section XI inspection sampling, the we*,d sarples initially sele:ted during the original license tern shall continea to be inspected every inspection interval during the licensa renoval term.
However, the inspection shall be augmented with examination of pressure retaining welds in piping having a CUT to the end of the license rrnewal term exceeding 0.4, if not already in the sample.
Depending on the uncertainties in the reconstruction of actual transients, an augmented inspection of selseted fatigue critical locations shall be The augmented inspection shall include inspection of at Isast lot performed.
of evaluated component locations with high Curs at every inspection period.
/
It may also include inspection of pressure retaining. velds in piping with Curs slightly less than 0.4 at every inspection interial.
2.6 section XI Analysis and Inneeetion
)
If f urther f atigue analysis is required because the CLB in Paragraph P.4 is exceeded and there is inadequate reconstruction of actual transients in Paragraph 2.2, a fatigue crack grovth analysis in accordance with Section XI shall be performed as follows:
(a)
Perform a volumetric inspection of the fatigue critical locations, which may be determined based on similarity arguments.
Verify the absence of a flaw exceeding a certain size that can be reasonably detected, e.g.,
90%
confidence that the detection probability for a flaw of this size is 95%
or better and would not be undersized by more than 10%.
An inspection performance demonstration similar to Appendices VII and VIII of Secticn XI shall be perf ormed.
(b)
Postulate a flav, having a site as discusred in (a) above, to exist at the critical location at the date of the inspection.
Perf orm a f etique crack growth calculation from then to the end of the license rene.w term, using the curvet in Section XI accounting for the effect of environment.
(The staf f can provide technical references for the curve for stainless steel in a watsr reactor environment.)
(c)
The projected flaw site at the end of the license renewal tern, shall be demonstrated to be structurally stable.
If there is an evaluation criterion in the code, it shall be used, e.g., IWB-3640 of Section XI shall be used for evaluating austenitic stainless steel piping.
(d)
The critical locations shall be monitored by augmented volueetric inspection every period in each subsequent Section XI inservice inspection interval.
6-
j 1
l -
2.7 Aucrented Inacection As discussed in Paragraph 1.6, an augmented inspection performed on a periodic of the fatigue basis to detect the presence of fatigue flaws shall be a part The extent and frequency of the augmented inspection management program.
fatigue evaluation methodology, fatigue chall depend on operating experience, cnalysis, confidence of the input to the analysis, and the result of the Specific discussient have been provided in Paragraphs *2.3, 2.4, onalysis.
However, engineering judgement is requirad to select the most 2.5, and 2.6.
of f ective augmented inspection program.
3.
c1 ASS 2 AND 3 COMPONENTS Figure 2 shows a flow chart for the fatigue evaluation of ASMS Code Class 2 It considers (1) history of f atique and 3 components for license renewal.
damage, (2) operating temperature, (3) reconstruction of actual thermal cycles, (4) '." LB, (5) Verification of Allowable Stress, (6)Section XI analysis augsented inspection.
Components shall be replaced,
~
and Inspection, and (7) or committed to be replaced at a time shortly after license renewal approval, if they are found unacceptable by the fatigue evaluation.
Because there is no code f atigue design requirements for components other than piping, the evaluations according to Paragraphs 3.2 through 3.7 apply only to piping.
For vessels designed to the alternative design rules in accordance with NC-3200 of Section III, the methodology of Class 1 components in Faragraphs 3.1 through 2.7 shall be used, with the exception that NC-3200 shall be used instead of Section III in Paragraph 2.5.
Further, components not cover by the code fatigue requirements shall be evaluated according to If there is a history of fatique damage, the fatigue mechanism Paragraph 3.1.
shall be mitigated or minimized and an engineering evaluation shall be performed to demonstrate either structural integrity will be maintained throughout the license renewal term or the component shall be replaced at c.n acceptable replacement schedule as justified.
An augmented inspection using volumetric methods shall also be performed on a periodic basis.
3.1 Histo ' of Fatieue Damace similar to Paragraph 2.1, the history of fatigue damage in plant specific cases and generic cases shall be reviewed.
If there is a history of damage, An the fatigue mechanism shall be mitigated and remedial actions performed.
augmented inspection and/or monitoring shall be performed to assess the effectiveness of tha fatigue mitigation procedures, rurthermore, generic actions relating to the potential for unanticipated fatigue leading such as NRC IEE Bulletin 79-13, " Cracking in Feedwater System Piping," shall be perf ormed as requested.
if a component had failed in fatigue due to excessive vibratory For example, stresses induced by a nearby positive displacement pump, a pulsation danper may be installed to mitigate the fatigue mechanism.
A periodic inspecticn shall be performed to ensure the proper function of the pulsation darper. -______-_
- ~..
3.2 coeratina' Temperature.
Because thermal cycles are perceived as the dominant f atigue mechanism on Class 2 and 3 components, code requirements on f atigue are addressed through the thermal expansion stress allowable limit.
It would be consistent to develop a screening criterion for fatigue analysis based on operating
-terperature.
If the operating temperature is less than 120'T, which is basically an upper bound on the ambient temperature, the component is acceptable for license renewal without further analysis.
Hogever, further analysis is required if the operating temperature exceeds 120 T.
Thus, systeus like the service water system which operates at the ambient temperature may be acceptable for license renewal relative to the fatigue concern.
However, augmented inspection shall be performed on components with a relgtively high thermal stress even if the operating, temperature is less than 120 T.
3.3 Reconstruction of Actual Thermal Cveles l
If further f atigue evaluation is required pursuant to paragraph 3.2, the applicant shall-reconstruct the actual thermal cycles.orperienced by-the class The extent of thermal cycles
'2 and 3 components to the extent practicable.
reconstruction will vary from plant to plant depending on plant specific record keeping practices.
Also, if can be justified, the applicant may use industry experience of:similar plants or actual thermal cycle tracking over a J11mited period of time to conservatively estimate the actual plant-thermal cycles.- In any case,-the confidence of the actual thermal cycle
. reconstruction affects the uncertainties in further evaluations.
3.4 CLR l
l As discussed in Paragraph 2.4, the CLB is the number of design thermal cycles originally _ assumed in the licensing of the plant, as supplemented by licensee commitments during the original license term either through plant specific actions or generic-actions, (An_ applicant for license renewal shall demonstrate that the number of-actual j
thermal cycles, projected to the end of the license renewal term, shall be l
bounded by the corresponding numbers originally assumed-in the fatique
- analysis, i.e., in calculating the thermal expansion stress allowable-limit for both Section III and 831.1; plants.
If the applicant cannot demonstrate to
- the staff-that the projected thermal cycles would be bounded by the numbers in these documents, further fatigue analysis is required.
For example, if-the original thermal expansion stress calculation postulated less than 7,000 thermal cycles in selecting the stress allowable limit, the applicant has to demonstrate that the numbers of thermal cycles on the component would not exceed 7,000.to the and of the license. renewal term.
Othe rwise, further fatigue evaluation is required.
l If the number of projected thermal cycles to the end of the license ranewal-i term is within the CLB, further fatigue analysis of the component is not l- - - -
1 However, augmented inspection shall be performed based on the re quired.
confidence in the reconstruction of actual transients.
In any case, augmented volumetric inspection shall be performed at every inspection interval on components with projected thermal cycles close to the CLB and a thermal expansion stress close to the allowable limit.
3.5 verification of A11ovable stress is If further fatigue analysis is required because the CLB in Paragraph 3.4 exceeded and there is adequate reconstruction of actual thermal cycles in the thermal expansion stress shall be verified to continue to Paragraph 3.3, satisfy the thermal expansion stress allowable limit to the end of license renewal tern according to B31.1 or Section III.
Furthermore, because earlier versions of the code had limited number of stress intensification f actors avcilable for calculations, stress intensification factors not available were estimated by various means.
Thus, the stress intensification factor used in the original calculation shall be verified to be approximately similar to those in current codes.
if the original thermal expansion stress calculation assumed For example, 7,000 thermal cycles and the component is projected to experience 14,000 thermal cycles to the end of the license renewal term, a stress range reduction factor of 0.9 corresponding to 14,000 cycles according to NC-3611 of Section III shall be used.
Thus, the thermal expansion stress shall be demonstrated to remain within the allovable limit using the reduction factor Fu rthe rmore, if the stress intendification factor for the component of 0.9.
was not available from the earlier code and was estimated, the value used shall be comparable with that in the current codes.
An augmented inspection shall also be performed.
Consistent with the discussion in Paragraph 1.5, pressure retnining welds in piping with a calculated thermal expansion stress within 80% of the allowable shall be volumetrically inspected every inspection interval.
The 80% on stress is estimated from the inspection requirement at a CUF of 0.4 and the code fatique design curve (see page 9 in NUREG/CR-3243)
-0.2 i Sd = 245,000 N is the design where i is the fatigue-based stress intensification factor, Sd stress range in pai, and N is the number of allowable fatigue cycles.
For example, assuming fatigue is due to cycles having the same stress amplitude with a stress range (i S,) of 20,000 pai, the design curve above gives the number of alloweble f ati@e cycles (H) of 280,000.
A CUF of 0.4 corresponds to 40% of the number of allowable fatigue cycles, i.e.,
fatigue cycles of 110,000.
Using, 110,000 for N in the above design curve gives an allowable stress range of 24,000 psi.
The ratio of 20,000 to 24,000 is 0.8 ind ice.t ing that 80% of the allowable stress is equivalent to 40% of the number of
-allowable fatique cycles.
Furthermore, depending on the uncertainties in the reconstruction of actual thermal cycles, an augmented inspection of selected f atigue critical locat::rs In any case, augmented volumetric inspection shall be may be performed.
performed at every inspection interval on components with projected thermal 9
l cycles clogs to the cycle limit assumed in selecting the stress range reduction factor and a thermal expansion stress close to the allowable limit, i
3.6 Section XI Analysis and Insoection is If further fatigue analysis is required because the CLB in Paragraph 3.4 oxceeded and there is inadequate reconstruction of actual thersal cycles in Paragraph 3.3, a fatigue crack growth analysis in accordance with*Section XI shall be performed as follows:
I (a)
Perform a volumetric inspection of the fctigue critical location, which J
may be determined based on similarity arguments.
Verify the absence of a flaw exceeding a certain site that can be reasonably detected, e.g.,
90%
confidence that the detection probability for a flav of this size is 95%
i or better and would not be undersized by more than 10%.
An inspection j
performance demonstration similar to Appendices VII and VIII of Section XI shall be performed.
(b)
Postulate a flaw, having a size as discussed in (a) above, to exist at the critical location et the date of the inspection.
Perform a fatigue crack growth calculation from then to the end of the license renewal term, using tte curves in section X,1 accounting for the ef f ect of environment.
(The staf f can provide technical references for the curve l
.for stainless staal in a water reactor environment.)
1 shall be (c)
The projected flaw size at.the end of the license renewal term, demonstrated to be structurally stable.
If there is an evaluation criterion in the coda, it shall be used, e.g., IWB-3640 of Section XI 4
l shall be used for evaluating austenitic stainless steel piping.
The critical location shall be monitored by augmented volumetric (d) inspection every period in each subsequent Section XI inservice inspection interval.
3.7 Auerented Inseeetion As discussed in Paragraph 1.6, an augmented inspection performed on a periodic basis to detect the presence of fatigue flaws shall be a part of the fatique The artent and frequency of the augmented inspection management. program.
shall depend on operating experience, fatigue evaluation methodology, fatigue analysis, confidence of the input to the analysis, and the result of the analysis.
Specific discussions have been provided in Paragraphs 3.2, 3.4,
- 3. 5, and 3.6.
However, engineering judgement is required to select the most effective augmented inspection program..-
==
Class 1 Components o
History of Tatigue Damage ?
Yes _
Mitigate Tatigue Mechanism No and conti.rm by Auguented Inspection y
o Reconstruct Actual Transients
' t Yes Documented Usage Factor Less Than 0.4 7 No (or No Usage Factor Available) 9 Projected cy'
.s to End of License Yes Renewal Tern Within Limits in Current Licensing Basis 7 No y
Yes Adequate Transient Reconstruction ?
No u
Section III Class 1 Section XI Tatique Analysis Crack Growth Analysis l
Considering Considering Environmental Effects Environmental Effects 4
I n
Analysis Acceptable ?
No _
Component
~
Replacement Yes 7
Augmented Inspection (Volumetric) 9 Tatigue Easistance Acceptable for License Renewal Figure 1
Class 2, 3 Components F
History of ratigue Damage ?
Yes _
Mitigate ratigue Mechanism No and Confits by Augmented Inspection y
o
~
operating Temp *5**ur*
Yes Laes Than 120 F 7 No 9
Reconetctet Actual Thermal Cycles o
Projected Cycles to End of License Yes Renewal Tern Within Limits Originally Atsumed in Thermal Expansion Stress Allowable ?
~
No Yes_ Adequate Thermal Cycle Reconstruction ?
No 1
9 Verify Recalculated section XI Thorwal Expansion Stress Crack Gruvth Analysis Allowable Hot Exceeded considering for Projected Cycles Environmental Effects y
o
<P Analysis Acceptable ?
No _
Component Rep 12coment Yes 1
Augmented Inspection (volumetric) o Tatique Resistance Acceptable for License Renewal Figure 2 !
Class 1 Components o
History of Fatigue Damage ?
Yes.
Mitigate fatigue Mechanism No and Confirm by
~
Augmented Inspection Reconstruct Ac val Trans ents y
Yes Documented Usage Factor Less Than 0.4 ?
No (or No Usage Factor Available) it Projected Cycles to End of License Yes Renewal Term Within Limits in Current Licensing Basis ?
No 9
1r y
Section III Class 1 Section XI Fatigue Analysis Crack Growth Analysis Considering Considering Environmental Effects Environmental Effects 4
<r o
Analysis Acceptable ?
No _
Component
~
Replacement Yes P
Augmented Inspection a
Fatigue Resistance Acceptable for License Renewal Figure 1 01/22/91 - -
Class 2, 3 Components o
History of Fatigue Damage ?
Yes.
Mitigate fatigue Mechanism No and Confirm by
~
Augmented Inspect, ion P
Yes OperatingTempegature Less Than 120 F 7 n
Reconstruct Actual Thermal Cycles o
Projected Cycles to End of License Yes Renewal Term Within Limits Originally Assumed in l
Therral Expansion Stress Allowable ?
No o
e 9
Yerify Recalculated Section X1 Thernal Expansion Stress Crack Growth Analysis Allowable Not Exceeded Considering for Projected Cycles Environmental Effects I
{
9 l
1 Analysis Acceptable ?
Lo Corponent
- - - - - * " Replacement Yes o
Augmented Inspection (Volumetric) l Fat *,ue Resistance Acceptable for License Renewal I
Figure 2 I
01/22/91
_q_
r:
a...
Mr. Edward _Griffing
-2 Fabiery 25, 1931 but it will be the subject of a future raceting 'between the NRC and NUMARC, tentatively scheduled for April 16, 1991.
The reporting and/or recordkeeping reautrements contained in this letter 6ffect fewer than ten correspondents: therefore, OMB clearance is not reauired under P.L. 96 511.
If you have any auestions concerning this suranary, please contact me or P.T.
.Kuo at 492 3147.
Sincerely,
/S/
John V. Craig, Director License Renewal Project Directorate Division of Advanced Reactors and Special Projects Office of Nuclear Reactor Reautation
Enclosures:
As stated DISTRIBUTION J.Partlow,ADPR D.Crutchfield,DAR PDR A.Thadani, DST R.Bosnak,RES R. Jones J. Norbert, ENEB PDLR k F.
- C.Cheng,EMCB G.Baachi,ESGB L.Shao,RES F. Rosa,SICB W.Minners,RES C.McCracken,SPLB R.Borchardt,DARSP Central File J. Richardson,DET.
S.Newberry,SELB S.Treby,0GC
'R.Kirkwood,EIB W.Lefave,SPLB H.Ashar,ESGB W.Norris,SSEB
- See previous concurrence OFC LRPD:PM*
- LRPD:5C*
LRPD:D P 7-(............:............:............:.......
NAME :RParkhill/lb PTKuo
- JCraig DATE :02/07/91
- 02/22/91 02/}f?91 OFFICIAL RECORD COPY
'Documer
.ames-RP GRIFFING LETTER ON IR BWR