ML080070080
| ML080070080 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 01/02/2008 |
| From: | Kinsey J General Electric Co |
| To: | Document Control Desk, Office of New Reactors |
| References | |
| MFN 06-119, Suppl 4 | |
| Download: ML080070080 (27) | |
Text
HITACHI GE Hi tachi Nuclear Energy James C. Kinsey Vice President, ESBWR Licensing PO Box 780 M/C A-55 Wilmington, NC 28402-0780 USA T 910 675 5057 F 910 362 5057 iim.kinsey~ge.com MFN 06-119, Supplement 4 Docket No.52-010 January 02, 2008 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001
Subject:
Response to Portion of NRC Request for Additional Information Letter Number 16 Related to ESBWR Design Certification Application - Piping Design - RAI Numbers 3.12-11 S01, 3.12-22 S01 and 3.12-27 S01 The purpose of this letter is to submit the GE Hitachi Nuclear Energy (GEH) response to the U.S. Nuclear Regulatory Commission (NRC) Request for Additional Information (RAI) originally transmitted via the Reference 1 letter and supplemented by an NRC request for clarification in Reference 2. The GEH response to RAI Numbers 3.12-11 S01, 3.12-22 501 and 3.12-27 S01 are addressed in Enclosure 1.
If you have any questions or require additional information, please contact me.
Sincerely, Ja7ZesC.
insey*
Vice President, ESBWR Licensing
MEN 06-119, Supplement 4 Page 2 of 2
References:
1. MEN 06-103, Letter from U.S. Nuclear Regulatory Commission to Mr.
David H. Hinds, Manager, ESBWR, General Electric Company, Request For Additional In formation Letter No. 16 Related To ESB WR Design Certification Application, dated March 30, 2006.
- 2. E-Mail from Amy Cubbage, U.S. Nuclear Regulatory Commission, to GE, dated May 20, 2007.
Enclosure:
1,. Response to Portion of NRC Request for Additional Information Letter Number 16 Related to ESBWR Design Certification Application - Piping Design - RAI Numbers 3.12-11 501, 3.12-22 501 and 3.12-27 501.
- 2. Attachment 1 - Proceedings of ASME-PVP 2007: 2007 ASME Pressure Vessel and Piping Division Conference, July 22-26, 2007, San Antonio, TX, USA. PVP2007-26143. "Application of Draft Regulatory Guide DG-1144 Guidelines For Environmental Fatigue Evaluation to a BWR Feedwater Piping System."
- 3. DCD Markups.
cc: AE Cubbage DH Hinds GB Stramback RE Brown eDRF USNRC (with enclosure)
GEHlWilmington (with enclosure)
GEH/San Jose (with enclosure)
GEHl\\Nilmington (with enclosure) 0000-0075-9909 MVFN 06-119, Supplement 4 Response to Portion of NRC Request for Additional Information Letter No. 16 Related to ESBWR Design Certification Application Piping Design RAI Numbers 3.12-11 S01, 3.12-22 S01 and 3.12-27 S01
IVIFN 06-119, Supplement 4 Page 1 of 8 NRC RAI 3.12-11 DCD Tier 2, Appendix 3D, provides a description of the major computer programs used in the analysis and design of safety related components, equipment, and structures.
According to this appendix, the quality of these programs and computer results is controlled. The programs are verified for their application by appropriate methods, such as hand calculations, or comparison with results from similar programs, experimental tests, or published literature, including analytical results or numerical results to the benchmark problems. To facilitate the staff review of the computer programs used in the ESBWR design, provide the following additional information:
(a) Identify which computer programs will be used during the design certification phase and which programs may be used in the future during the COL application phase.
(b) Identify which programs have already been reviewed by the NRC on prior plant license applications. Include the program name, version, and prior plant license application. As stated in SRP 3.9. 1, this will eliminate the need for the licensee to resubmit, in ý a subsequent license application, the computer solutions to the test problems used for verification.
(c) Confirm that the following information is available for staff review for each program: the author, source, dated version, and facility, a description, and the extent and limitation of the program application; and the computer solutions to the test problems described above.
GE Response (a) The programs used in the certification phase are:
PISYS07 It is a computer code for analyzing piping systems subjected to both static and dynamic piping loads.
ANS1713 The program is for calculating stresses and cumulative usage factors for Class 1, 2 and 3 piping components in accordance with articles NB, NC and ND-3650 of ASIVIE Code Section 111. ANS17 is also used to combine loads and calculate combined service levels A, B, C and D load on piping supports and pipe-mounted equipment.
All of the programs in Appendix 3.D.4 may also be used in the future during the COIL application phase.
MFN 06-119, Supplement 4 Page 2of 8 (b) PISYS05 has been benchmarked against NRC piping models. The results are documented in GE report NEDO 24210, dated August 1979 (Reference 3D 1 of Appendix 3D), for mode shapes and uniform support motion response spectrum analysis (USMVA) options. The independent support motion response spectrum analysis (ISMA) option has been validated against NUREG/CR 1677.
The PISYS05 computer program has been reviewed by NRC, and the results are benchmarked with NUREG/CR-6049.
PISYS07 USMVA and ISMA analyses are the same as PISYSO5. It has been benchmarked with NUREG/CR-6049.
(c) The computer programs listed in Appendix 3D are available for staff review.
These programs are Level 2 programs. The author, source, dated version, and facility; a description, and the extent and limitation of the program application; and the computer solutions to the test problems are contained in the design record file of each program.
MFN 06-119, Supplement 4 Page 3 of 8 NRC RAI 3.12-11 S01 The issue involves the validation of the PIS YS computer code used for the piping analysis. GE should verify that the PISYS computer code correctly implements the RG 1. 92 procedure for mode combinations. In addition, GE should provide a technical justification for accepting the results at those locations that exceed the NUREGICR-6049 acceptance criteria in the PISYS comparison with the NUREG/CR-6049 benchmark analysis.
GEH Response GEH has modified the PISYS program to comply with RG 1.92 Rev. 2, 2006. The new version of the program is PISYSO8.
The PISYS08 program has been benchmarked with NUREG/CR-6049. The results are a 100% match with NUREG /CR-6049, except for a few values that are a 99%/ match. There were no locations that exceeded the NUREG/CR-6049 acceptance criteria in the PISYS08 comparison with the NUREG/CR-6049 benchmark analysis. Therefore, the requirements of RG 1.92 Rev. 2 have been met for the double sum of modal results and high frequency modes.
The detailed analysis and comparison are shown in GE-NE-0000-0070-1785-00, (eDRF 0000-0070-1 785) '"PISYS08 for Regulatory Guide 1.9R2 2006 and NUREG/CR-6049," a proprietary document, which is available for viewing in the GEH Washington office."
DCD Impact No DCD changes will be made in response to this RAI.
MEN 06-119, Supplement 4 Page 4of 8 NRC RAI 3.12-22 DCD Tier 1, Section 3. 1, "Piping design, " states that Class 1 piping systems will be analyzed for fatigue with environmental effects. Provide the analysis and design methods that will be used to perform the fatigue evaluation, including the environmental effects, for the ESBWR Class I piping systems.
GE Response Requirements contained in ASME Ill NB-3653. The load combinations contained in Table 3.9-9, and the plant event cycles contained in Table 3.9-1 of the DCD, define the design conditions that are inputs to the fatigue analysis. Additionally, GE has additional design criteria for carbon steel and stainless steel materials that are intended to address environmental issues that have been applied to prior BWR applications, and are likewise being applied to the ESBWR piping design. Additionally, class 1 piping using a fatigue limit of 0.1 instead of the ASME Code acceptance limit of 1.0 in conjunction with a stress ratio limit of 0.80 for Equations 12 and 13 of the ASME Code in order to limit the number of pipe whip restraints within the containment. DCD paragraphs 3.9.3.3 and 3.9.3.4 will be revised in DOD Revision 2 to reflect this commitment as follows:
"Additionally, a fatigue usage limit of 0.10 is used as a design criteria for all Class 1 piping."
Evaluations have also determined that the ASME Code has conservative methods that provide additional margins. Specifically, the ASME Code adds stresses that include P, Ma, Mb, Mc, OTi, DT2, and Dtab by absolute sum when in actuality the direction and signs of the stresses are different. Reference (1) has performed a detail finite element analysis to compare against the results of a NB-3600 analysis and found that the fatigue usage based on N B-3600 is about 10 times more conservative.
This design criteria that is being used for ESBWR is consistent with the design methods used on previous BWR product lines that have successfully operated for the last 40 years without piping fatigue issues. Data from fatigue usage monitors from operating plants have also confirmed that the design criteria specified by GE in the original plant design was conservative.
The simplified NB-3600 analysis has been used for last 40 years successfully. If newly developed environmental fatigue curves are used, high fatigue usage factors are predicted and pipe break locations will be postulated throughout the plant. The economical cost to the plant is huge, and any gain of safety is questionable.
It is recommended that the environmental fatigue design curves should not be used without substantial simultaneous changes in analytical methodology and the ASME Code.
MEN 06-119, Supplement 4 Page 5of 8 Ref. 1. "Fatigue Usage Factor Evaluation For An Integrally Reinforced Branch Connection Using NB-3600 And NB-3200 Analysis Methods" by Henry L. Hwang, PE, General Electric Nuclear Energy, Jack R. Cole, PE, David M. Bosi, PE, Design Engineering, Washington Public Power Supply System. PVP Vol. 313-2, page 139 through 156.
MEN 06-119, Supplement 4 Page 6 of 8 NRC RAI 3.12-22 S01 The RG on environmental effects in the fatigue calculations of Class I piping will be issued soon. GE committed to implement the criteria for evaluating environmental effects, but will request some relaxation in the pipe break criterion for fatigue usage. GE will provide the results of a study showing the impact of the new environmental fatigue criteria to support its request to relax the pipe break fatigue usage criterion. This item is open pending staff review of the GE submittal.
GEH Response The environmental effects on fatigue in accordance with DG-1144 and NUREG/OR-6909 has been incorporated in GEH piping program ANS17014; however, this incorporation is conditional to the NRC accepting a change from 0.1 to 0.4 fatigue usage as specified in BTP EMEB 3-1 to exempt piping components from pipe break consideration. Since this change has previously been discussed with the NRC staff, GEH will proceed to change DOD sections. 3.6.2, and Table 3.9-9 to incorporate this change.
GEH's study of the impact of implementing the new environmental fatigue criteria is shown in Attachment 1, PVP2007-26143, "Application of Draft Regulatory Guide DG-1144 Guidelines for Environmental Fatigue Evaluations to a BWR Feedwater Piping System". This paper contains a detailed description of the methodology and output comparisons of fatigue usage factor with and without inclusion of environmental fatigue.
DCD Impact DCD Tier #2, Table 3.9-9 will be revised in Revision 5 as shown in the attached markup 1.
DOD Tier #2, Section 3.6.2, will be revised in Revision 5 as shown in the attached markup 2..
MFN 06-119, Supplement 4 Page 7 of 8 NRC RAI 3.12-27 DCD Tier 2, Section 3.7.3.12, discusses the effect of differential building movement on piping systems that are anchored and restrained to floors and walls of buildings that may have differential movements during a dynamic event. SRP 3.9.2 Section 11.2.g states that the responses due to the inertial effect and relative displacement for multiply-supported equipment and components with distinct inputs should be combined by the absolute sum method. Provide the combination methods that are to be used in the design of ESBWR piping systems for the inertial responses and SAM responses caused by relative displacements for all analysis methods (including ISM).
GE, Response DCD Tier 2, Section 3.7.3.12, discusses the effect of differential building movement on piping systems that are anchored and restrained to floors and walls of buildings that may have differential movements during a dynamic event.
In general, the piping systems are anchored and restrained to floors and walls of buildings that may have differential movements during a seismic event.
The movements may range from insignificant differential displacements between rigid walls of a common building.at low elevations to relatively large displacements between separate buildings at a high seismic activity site, Piping system is different from multiply-supported equipment. For piping system, the induced displacements in compliance with NB 3653 are treated differently than the inertia displacements. The SRSS method is a standard industrial practice to combine the inertial responses and SAM responses caused by relative displacements.
MFN 06-119, Supplement 4 Page 8 of 8 NRC RAI 3.12-27 S01 SRSS combination of the inertial and SAM responses for USM method of analysis is not consistent with the staff position in the Standard Review Plan (SRP). GE should provide additional technical justification for this position.
GEH Response During the NRC audit meeting held between January 9 through Ja nuary 12, 2007 at San Jose, CA (Reference NRC "Audit Trip Report," ML070930012), the NRC staff found that the SRSS combination for the inertial and SAM responses is acceptable for the piping stress analysis, except for piping support designs. For piping support design, the DCD is being revised to show that the absolute sum method (ABS) is used.
DCD Impact DCD Tier 2, Section 3.7.3.12 will be revised in Revision 5 as shown in the attached markup 3.
MFN 06-119, Supplement 4 -
ATTACHMENT 1 Proceedings of ASME-PVP 2007:
2007 ASME Pressure Vessel and Piping Division Conference, July 22-26, 2007, San Antonio TX, USA.
PVP2007-26143, "Application of Draft Regulatory Guide DG-1 144 Guidelines for Environmental Fatigue Evaluation to a BWR Feedwater Piping System"
MEN 06-119, Supplement 4 - Attachment 1 Page 1 of 9 Proceedings of ASME-PVP 2007:
2007 ASME Pressure Vessel and Piping Division Conference July 22-26. 2007, San Antonio, TX, USA PVP2007-261 43 APPLICATION OF DRAFT REGULATORY GUIDE DG-1144 GUIDELINES FOR ENVIRONMENTAL FATIGUE EVALUATION TO A BWR FEED WATER PIPING SYSTEM Hardayal S. Mehta Henry H. Hwang GE Energy Nuclear 6705 Vallecitos Road Sunol. CA 94586 ABSTRACT Recently puiblish-cl Draft Rcutelatory Guide DO 1144 by thec NRC provides -,u I tncc fmt me in tkknnminin the icccptable faisut life of
`'~E roýiur~c budr iontswith cidersion t sh
'Ict Water MeaIoMIL P
em-rnuoument.
The aalytic;al epeso and. furhri daii.:t provided in NUREO;CREQ C0O Ir, th s lpaptC thle euvWome~ntl faiuen rules tire appled toa BW\\R ftcdlvater flie. The p ipen aaterial is cairbon steel ;SA333: Ga; 6) am-d ;te feedwvatcr nozzle amaerial is lows alloy szeel (SASOS Class 2). Tile transfientst used in dhe evaluation ate based on the thermnal cycle diiaeram of thet piving.
The calcutlated f'trigue usage factors inchiding the ennviomunenmal etfects are conpared-with taose obtained using thle cumnent A'M\\IE Code rules.
In both cass te cumnutaive fotisue misac factor Ire shown to be lesý thani 1,0, BACKGROUND & INTRODUCTION Since the early 1980s the cffec-ts of high merperatilue water onm the faitiuei cyclic life of light water mector [LWVR]
comuponenits have "beti me~nmsiely dmasas,,Td by nmnaserolis researchers, Peertencei I thoneli 1-5 are some oftl th cxamples.
TheSuigsu oin Fatigue Swngqli of tithe ASME Boiler & Prvesli
("ATeelC d is Cull nt-iv 'a k-ikmm onl I Code Case that ""010j i deptcoue for ilimomvcstrw the reactor Water env11'oimýetstal eftect,ý in thle fatt~lle evalultionl conducted pcr the 'Audelines ir Seciom !I! Piat eaaphis NB-3200 and NB3-360fl110].
Recently. thle U.S. Nutclear Rea-ulatory Coammission (NRC) hals published for pu~blic comnasent thle Draft Rcenlatory Guidec DG2-ll44 to provide ptaidance ib-o usie in derctaisiaing dic accept~able fatigue-life of ASME pressýurc boiuidavy coipn Nts ith consideration of LAVR emiviro ninit [171]
The associated detailed --nidaricc documaent i's NL REC 'CR-6900) [IS],
The NýRC addre!ssed the public cornanexivs and Vs expected to issue the fnlversion a~s Regurlitory Guide '07
\\WREG!/CR.6909
- adot, theil iote'a fatieul co-retioal faictor mnethod o
,_. mnethod to acc~osu for dlic ens-irounnaental fatigue eiffects, F. is defined as th m atew of tati~gue initiation fife in air at room teriperature to that in reactoriwater at tile service temaperiatutre. The reruators 'nudies are miot mandatory.,
However, the NRC is likely to ask applicants for Cer tification of new reactor &42115~ fortileb technical approach they plain to followN to ~as en-vironmnental fiatiene effects.
This paper d.-scribes the resuilts of tile applicatton of DG-1144, methodology tona BWR plant piping system. The systema chosecn is feedwvaser apiping inside the containment. TI1ais systerm as trPically classified ais Class I per thle ASNIE Code Classification.
DESCRIPTION OF PIPING SYSTEM Figurre I schematically shosws the Fecd-wamer piping system.,
fte pipina2 systema delivers the 4feedwater to the r eactor It also receieves water froms Res-idual Heat Rtimoval (PT{R,) antd Reactor Core Isolarioia Cooling, (RCIC) systems.ý The: portion of the p~iping betweenn tile reacor-nozzle and the header at phe Copyrigb r C 2WX7 by ASINE
MEN 06-119, Supplement 4 Page 2 of 9 containment oitetration is, tesisawd to AS24IE Class I
requiiretants piping thickness is per schedule 80, Thle sýpecified design pressuire and tenmperaturte for this; piping are 1250 psi and 550'F, respectively. The feedwarer temperature during normtal operation is 4-10' F.
PIPING STRESS ANALYSIS BY CURRENT CODE RULES Fltier 2 shows the mlathematical miodel of the feedwater pipiim2 ssteal, Thet piping nonsinal1 di~amletersi ire 22-inlchtes at the containtunent pentltration (Node 26 at the I i ght hand bo:.tolm of Fiogure 2) and 12-inches at. theý point where Itht riselrs connlect thou t end to tlw ee ae noz Xs(odesý 49, 67 Ind S5 iuT Fieure 2),
A complete Class I piping stress analysis, including a fatigue emaluation, was first conducted accordin2 to the rules of Paragraph -NB-3600 of ASME Saction 111 t19],
A GE proprietary comlputer prolgrans. AN'SI
[201 ss as used, in the anlyis.
A key input to the Code fistipts ev~aluatoin is the pressreitepera irv-ut cycles i-oi the vs, ttm, Figture 3 dhows a part of the pressUrel/temperatut e dulty cycle ffor Thle feedsrater system considered its thisý evaluanion.
It defines the expected feedwater temnperature chanotes during the Hot Standby evecnt, The unumber in each of the dianionds onl thrs figttre repvmresntadeidlodtte I
n rIIse load mtate 2S the temiperarture changes fronti 1 26~C (259'F) to 282%C' (54OTF) in 10 minutets.
The load state 29 is defined as a step drop in tenilperature fromll 2S2-C to 126,C (259T). In a singele Standby event these load states occur 2-4 times. S'ince ther'e are 166 htot standby events postuated. 'the snunber of cycles for events' 28 and 29 are (1166x24) or 398)4.
For tile load states that involve aI temperature trani.ent.
one-dimnitsionat heat tranisfer analyses were conducted to define kthe appiropritie valuesý of temtpel attre pat antetr usdIn thle fstiout tvaluzation.
Specifically. thet qna'rtittes calculated were average teniperarures onl each side of a node point JT, onl side A and Tt, onl side B).1 AýTj (linear thern-Ld 21radietit) and AT, (nt-tonlneai, thermail etradienit, Taible I Qhow' t Itattial Vmtin2 ofthe load states (hereinafter called load sets) inforinmaion at Nkvde 048 (at Feed-water nozzle)t. This, infonnataion is used iti developing the load set pair information for fatiwiuetsac calculation.
A pariali listing of the load set pairs information tot the,aftne node is shownr in Table L. The last cohuitn ýhoss s the partial fatifsic usagte factob. For exanipl~c foi thle load set panr 28-29 (indicated by bold itn Ta ble -2), the calcultate da titiuealteinauing stress is 151 MlPa (columln S froml left sýide) anld the coiresponding partial fiatiguei usage factor is 0.0626 based oil thle cun'enat Code faitiue curve for carbon and lowv alloy steels wNithl tltiniatv Tvcinilc stressý lethans 90 k-,i.
Thme calculated cumm [livev fati..teusae factors are-diwcussed Late" in thlis paperalong2 with thle tinvironmecltitl fatigue us ccg f. etolr ENVIRONMENTAL FATIGUE EVALUATION METHODOLOGY Appendix A of Reference IS provides tie, ertatilons to calculate the enrvironmetntal correction factor F,!. Table 3 extracted from Reference IS shows the equations for carbon and low alloy steels the materials of interest for feedsvater line.
The cninulative failouc usage factor. U,,. considerinz tile effects of reactor coolantt m-istrollients is Calculated as the fokllowing:
where, U, amd Fm are the partial fatigue usage faictor and the enrvironmnental fatigue cotvectiolt factor, resýpectively, for the "i"ih load set paitt Thle partial fat~igu usage factor Ui is to be based onl air fatigue curves at roomn temmipet antic.
The Fat is defined as the ratio of fatigue life its air at ioomn temperature to That ill waqter at tilse service tenipe anttre N:)
Reference 1S tlso provides alternating siress (S) versits curves for carbon, low alloy andi stainliess steels, Thesie Curv%,es ate dfiffrent than the current S-N cum'--, itt the ASNIE Code.
For covnec.Table 4 gives the slict.ired v alues for the curtent Code carves and thle Refeiecite 1 IS i F.. Calculation for a Load Set prair A review of tile equations in Table 3 indicates, that F,~ is a ftinctiots of four paraineters; S¶: T5". 01 and et".
For the pttrpose of this evaluiation, mlost. Conservative viluies of Slý
(=0.01:51, W (=Iý1n15,1,) and c5 (1{000) were assumed.
The.Appendix A of Reference IS altov.ys th. utse-of average of dth maxitnuims andi ininitmuns tempernatures iii a trainsient Itn The desernlittation of the appropriate tenipemarure T for the calculation of parameter T*. This approach witas followed in this cvahussion as illststratcd by a sample calcutlationi dtesciibed
- next, A part of tile pyartial ctiumulative taugute ts age factor caclt onra Node 0,48 usitig the, csulent A.SN-.E fattigue cturv-es, dissown in Table 2. Tetnmearatre T for itlte calcuilation of F, o the load set pair 282S92 (indicated by bold in the table) was deItelrinited as follows, As seen in F&igure 3. the inaxununan atad iutittimun templjeraitures duringL these nvo load sets aie 2S2"'C (5401F) aid 126'C (259WF). resp-ectively. Therefore. the average tevaperartue dusing thle rrtmnsient is' [
18'26)721 or 20PVC (399'F), Thusý. T* = (204-150) or 544, For tme nozzle side at node 04S. thec iraterial is low alloy steel. The F,ý, for this load set pair is calculated am:
eF,
?2-.~x.055.xlt(2.)xn(.0) tx epf0 702~0101x.
x4h25576Oýi 8,409 Cooy-rittht t?( 2007 by ASME
MFN 06-119, Supplement 4 Page 3 of 9 Thle partial fatigute usage fietoi for load set pair '-S-29 i~s shossu -ts 0 On o-` inl Tablel It is rioted that this is based oil the cutttrett Code tts.curse Fot the samec level of altvumating strss ii-Pil e
ll t'c umbrif vCycle on the l alloysteei S \\cuse2 i e inl NU-RECkCR-6909(se olm 4 in Table 4) otild be I SOS 8-"O This; vwould gsuve air curve partial fatigue u-.age of (39641 5982u)) r 0,0248, It is seen that the ulse of NURE&C-R-6909 ait ftan'ue curve rcesults in a reduction of better than factor of two reiduction, in the partial fatigue us~age value for this load set pair. The partial fatigue mtage for this load set pair considering environniental fatiorle, effects is (0. O24SXSAOP9') or 0,20&.
A siintihr calculation for F,, onl the safe end side, that is carbon steel, Lives a value of '7 84 1, For the alterntatilag stress level of 15 1 MPai for this load set pair. the allowable number of cycles onl the low alloy steel S-N7 curve ziven in \\UJRFG/CRý 6909 (see coluimn ' in Table 4' ivould bec 570820. This would give air cur-ve partcld ftigtieu usage of (396415 768-10 1 or0,0069.
The use of N7UREGCvC P6909 asir fativue curve for carbon steel results in a reduiction of an ahuost an order of vna~intilde inl the partial fatig-ue uisage valuec for this load set pair. Thet parial faigue usage for this load set pailr considering ens iloui..naetal fatig-li cffects is (0t 0u69x`.S41') or 0,054.
It is seen thait at least for thsloAd set Panr thei tcio inl partial anr tan'nge usage thoghne uise of (URG/R-5909 cuirve more than offset the increase dule to F,.
A subrovtinie that calculates. cumullative fitattee usiage inchiding2 reactor water effects accordfing to DRC.-l 144. uvas added to the ANS17 comlptter code used in the, pipuing stress
- analyses, The calctslaticon resuflts for the subject feedwak~ter piping are discussed in the next section.
ENVIRONMENTAL FATIGUE EVALUATION RESULTS Table. 5 provides a stuninary of the calculated vailue~s of ctullularive fatigue us~age iactors at twvo locationis.
For thle fecedwa'ter nozzle location. ulsa~ge factors are providedI fol. both the nozzle side (low Alloy steel) aiid the safe end side (Carbon steel). It is seen tiat there is a miodest impact Onl the calculiated fatigue isagec factors whenl reactor water envir1ouamental effects ire iaciored inl At the safe end location, the redluction is air faitiguec usaze thr-outub the usie of NUREG-I/CR-6909 S-N cturves.
essentially offset the increase due to the usec of F,,
DISCUSSION The inrae inaculated f-aitige usag~e -when cnviromninvtal ftimgnei criecu Ire taken Into accoklitý -was modest for the feedsvater ~it., considered inl this es iluationl, One, of the reasons is that tile normnal operating teltupeftislre f 1or Thle fecdwati ilce tnat40ýC) is comaparative lv lowsei than thce typical operating tensperi'urei for the prisnars piping in LWRs. In the case of carbon and iloný alloy steel. piping systems, tIle increase
<Iue, to thle tise of F_ ik sign ficaaflv offset by thle advantage gained througah the uisc of air S-N curves Provided in NUP.T/CR-909.Ti' us suouLI not bic the, casýe for tils steel pipinst systemns "!.ere the iin S-N curves in NURETCi-!
6909 predict, higher tisage fiicuot than the Code cre In general one would expect sev eral fold increase in the calculated fantige usage factor when) F,,, is used. This wvould have implications; in terusrs of nuimber of licatious wvhere hypothetical pipe breaks need to be postulwitM C urrently. rthe.
NeRC Branch Technical Position M-ED 3- [21 is uised for po~stulation of breaks inl high enecrgy lines, MEB,-I r-equires postulation of a break at an intermiediate: locstions if the fatigue usage at a locatioil exceeds 0A1 or thle prinnary plus seconda-ry stress range exceeds 2.A S,.
The calculated priimary pinls secondary stress rangeL is t.ot n'mpactedt by the use of Fe~ but the, fatigue usage ifactor is. The use of F, results itnmore locations whviere cumiulative fatigule usaoe fac.tor wvould exceed 0. 1. More break-locationta mecans titole p ipe wship restraint to mecet thle recluireninent of General Design Ciittrioii 4 of 10CFR5O ['22].
Howe."ver. the presenice of mlore ptpe whip re~straints adversely affects the ability to condutct piping in-servýice inspections; and thins have I negantvc ifmpat Oil piping reliability during operation.
The 0.1 fatiguie usage threshold was based onl enugaeering jidgmeilet and perhapas can be revised upuvards to say 0.4 or 0.6 to avoid this situation.
Tile revision could be.
jitstitied through a piping reliability analysis; somewvhat similar to that conducted in su~pport of revised Appendix L in ASMIE Section NI Code [231.
SUMMARY
AND CONCLUSIONS This paper prten~t5s the resuilts of a Cliass I stress analysis of a 3\\VR Fcedsývat.-r piping system in which. the etivillomninltal fatigute effects due to reactor wyater per DG-1144 were incluided. The materials consýidered were carbon and low alloy steels. The tesults showed that there isa a modest inreasme in the calcuIlated fatigue usage factois but thle values svere fotund to be acceptable (i.e., less than 1L0). The increase inl fiatige nsage may resutlt inl more locations where CUT-exceeds5 0.1 thereby resulting in more locations with break postulations atid requirement for installation of pipe -sship r estratints. Ani upwavrd revision of 0. 1 fatiguec "sisag threshold is recuoimllneided througýh a piping reliability study, This papel-lid not incltude stainiless steel pipin g systemn evaluation wherc the iLSI a ct of DO-1144 procedures may result inl a significant, increase in the calculated cun'tlhatts e fatigue usage tactor.
REFERENCES
[I]
Hale, D.A.,
ital. "Low Cycle fatigule of Comlmercial Piping Steels In a BWR P'ianairy Environinnrt," J.
Enanittering Materials Technology.Vol 103. pp. IS-2ý5. 1981.ý Copyright ' 2007 by ASME
MFN 06-119, Supplement 4 Page 4 of 9
[2]1 Ranvilath. `-, JNL Kiss, and S.D. Heald. "Fifiii-Behavior of Carbon Steel Coinnt i
Hig~h-Tiepeature Water Envirounments." AST\\'1 STP 770ý American Society for Testing and NMatert'ilsý pp, 4-',q 459,1982,
[3]
Hiyiichi. M. and KR lida. "FatiLgue Srrenjthý Correctin Factors foi Carbon and Low-Ailloy Steels; 'in Oxven.
Cotaining Hieli-Tcniperatue kWtr.t "I Nulear Emrneerienn Designi. VOL. Ii9 pp,23-306 199 1.
[4]
S, M'siumd'ir. O.K. Choprat anid X J, Shack, "'interim Fatigue Designi Curves for Carbon. Low-alloy. and Atistcuntic Stainless Steels i s LXX "R Env'otnnents, NZUREG'CR-55999. April 1993
[5]
\\'le1'ta H S. and S.R. Gos'eli
- n.
onin Factor A~pproasch to Account for XX 71ter Effet,, in PressuIre Vesiel aind piping Fitigue Evaitatons'" Nucleac Engint~eriviandlDe'ian. Vol 11p-1'.998lO
[6]
Melita. i-is.; -An Update on the Consideraton, of Reactor XWater Effects in Code Fatizue Isniatnton Evaluations, for Pressure Vicsse.ls aud Pipin%. AS\\IE PVP1,ohnne 410-2, pp. 4 5-5, 2i000.
7] Higuchi. M_
"Fatigue Cuvvteý an'd Fatigue Design Criteria for Carbon -and Lowx alloy Steels in High Temaperature Water," AS. IE P P Vol ume "86, pp, 161-169. 1999
[S]
Higuchi, NL
-Reviý-e Propoial ot Fatigue Life Conrectiona Factor F, for C arbon and Low alloy Steels in L\\V'R Enyuronin'ent A.SME PVPT Volumce 410-2.
2000.
[9]
Hiuetisi. M, Iia. K. Hi-uano' A, Tsunsumi, K, and Sakazuchi. K. "A Proposal of F'ntiue Life Correction Factor F,~ fol Ansteninc StnesSteel,, in L.WR XWater Ens ironnicints. ASMEr P\\ P Volumse 4.R Ppp, 10()-117. 2 00Ž.
[10]
Chopra. O).K, and XVJ. Shack, Envirnomuental Effects on Fatigue Crack Initiation ini piping and Pressu;re Vessel Steel-,.'* NIJREGICR-6717 \\IMav 200 1.
[11]
Van Der Shiy s.
WA,,
'P\\PC *s Position onl Environmntaill Effects on ftwine Life in LUT, Applictok "Xedo Research C ouncil Bulletin 487. Deceinblý 'i03.
[12]1 H'ipuchi.
M_
- Hirano, T
and Sakagichi.
K..
"Eautonfo Fatieue Dain-az Onl OperaIting plant Components in LWVR Water, ASME PVP Volume 4S0, pp. 129-138,
- 004,
[1.3)
O'Donnell, XX' J WI 0'Doninell. and T.P O'Donniell.
'Propeo.,e 1 Ne.w Fatigue Desdign (inrve~s for Axeii Stainlessý Stetlsý Alloy 6000 andt Alox S00.' Proc. Oif the 200,5
.ASIF Pre,,sin-V'essels and Piciino Cnenc.July 17-21.1 2005. Densver C0, Paper PVP2005-'7 140..
[14]
O'Donnell. XNU_ WX.J.
O'Donnell, and T.p. 0 Doninell,
- 'Propo~sed New Fatigue Desrusn Cutves fo ro C arbon and Low,-Alloy Steels in High Tenmperature XWateiý Proc. of the -2005 ASME pressure X es5,els and Pipin-Conference. July 17-2L. 2005 Denver, COý Paper PXP-005-141I0.
[15]
Nal 'muain' T, M, Higuchi. T. Kusunoki and Y Sugieý "MME C dson Enviv-roinmental Fatirtie Ev aluation.
Proc. of the 2006 ASME pressure Vsesand Piping.
Conterence, July 2.3-277. 2006. \\"iancouiver, Canadal Paper PX P2006-lCPVTl 11-93305.
[16]
Lcner tusin Subcommrittee on Nuclear Powxei Bill O DonnellI to Richard Barnes.
Chiknman Subcommitteie III, doated February 9. 2007, Sub~ject:
Implemtentation of Febrtuary 1, *07 Plin to Resolve Environmental Fatigue Issue for.Nuvclear Powver Plants.
[1!]
Draft Regulatory~ Guide DO-I 144; Guidelines5 for Evaluiating-Fa'r te Analvslv; Incorporating the Life Reduction of Metal Components Due to the E ffects of Fhe LiI:ht WXater Rector-,
Envirks)untent for I New Reactors; US Nucleinl C~~ltr'(owisin July
- 2006,
[1S]
Chopra.
OXK.
atid INJ.
Shack-. -Effect of LINVR Cwoolat Etivironnients on The Fatieue Life of Reactor Materials." NURECI/CR-6909, Draft PRepomt February 2006; Final Report, February 2007.
[19]
"Rute for Construction of Nuclear Power Plant C~omponsents,~ Section III, Division 1, ASME Boiler-and Pressure VseCodie. American S.ocietv of Mechanical Engineers.
[20]
piping Stress A~nalysis; Cotinputer Code, ANS17 (GE Proprietary).
[2))
Standard Review Plan Section 3,6. of \\UREC-IOSM0 Branch Technical Position \\IEB 31. 'Postulted Ruipture Locations, in Fluid Ssstensi Piping. hinide and Outsidle Coittaiinment.- Revis~ion, 2. June 1.9S7.
[22]
10 CFR Part 50. appendix A. Geneial Design Criterion i
- 4. Evrneta n
isl Des;ign Basi's"
[23]
A'Inrerirds Reh/abiiits 1-i sogi'aui:
Recommeunded Miproa'enjents to AS21IE Se.ction.X7A4ppendtd~ L i'MRP-62), EPPJ. Palo Alto. CA and U$.S departenett of Einegy. Washington. D.C.-. 2002.
4 4~CoDovri.jt C: 2007 b 'ASME
MEN 06-119, Supplement 4 Page 5 of 9 Table I Example of Load Set Inforination for Node 048 TRANSITION NEAR NODE 048.
'24 28 29 31 10.
3984 3984.
33, za) 7,
-27788604
- 7.
-20168404 7.7
- 7.
-25143124
- 1.
-2013447E
- 1.
372692016
-272175S 324 7752 2
-177447ZS 2243589 42316516 19521136 2243509 E9<197272
-MŽ71240 E7952 10034U42
-49962788
-7175757 215 ZA) 133.00 255.00
- 116 Wc 7neatLur'a IN2)
TES DTI 144. Q7 14.79 132.00 261.00 1".
01 15S.
1.
74.600
-43.00 7M 1.60
-8. 90 1.47,
- 0. 47 Table 2 Fatiggue U-sage Calculation Process at Node 048 Using, Currenmt Code Curve TA1t13 7O3
'2 I.
_:ý.-D
-SIRES'. RANGE AND 'tATIGUr VUSAGE 45 42 303.
85.
'-7.
1%
215 12.
a K
UA4 1z
$I 1 "
z~Ž FE 147.
44.
71.
k1.00
.42 C.S'B5 0.4S4 5 a5a42 151. 0.231 0,094 3934.
5.40.
j 41 HE.
212237
- 2 74 7.
62894.
28 55 29 28
- 0. 062t 71470.
BMW.
MUT.
45 (Copyri ght -3 2007 by~ A~SM'E
MEN 06-119, Supplement 4 Tible 3 F,. Equarious froin Appetidis A of NUREGX'C"R69O9 Page 6 of 9 Th oiin en p0vnv y22w oo.101io fatIO tnol for CabI I see
- Z2;,i~
~~~i-(C,6
-0 10~ak WS Tii 4 0'
= x(702 - -,101 Sý 74Q T
T C)~ ~Ind tra nsfori~me~d S content, ieilpnpea reý DO
-.111, ndstIn ne..
nvl S n 00 m
< 0.00 1 wt.%)
S, S
_0 0 i5 vi 00
~DO
<0 O04 popni C)- '
DV 04)
~
o4 Ippj DO 0: pp ii S 0
> 1%:'ý)
~~~:~~
1ý,)
000 t
lf(I000 )
0 iR A ý 2,ý (Al:
6 6 Copri~h ©~ 007 by ASME
MFN 06-119, Supplement 4 Page 7 of 9 Table 4 S-N V'alues in Current Code and INIREG/CR-6909 Cycle~s CS/,LAS CS Air LAS/Air SS SS C~urrenit Code
'NUREG/CR-6909 NUTREG!'CR-6909 Cuirent Code NU,7REG/CR-6909
[TUTS<8OKsij
("MRh (MIPa)
CMPa)
(MPa) 10 3 9 99.2 5357.5 5467.8 4881.7 599&8.
2 0 2827..
3833.7 3882.0 3530.3 4302.6 50 1896.2 2509,8 2)440.9 2378.8 2 7 51.2 100 1413,5 1820.3 1758.3 1799.6 17.
201068,77 1358.3 1303.2 13591441,1 500 724,0 937.77 903.3 1020.5 97 2.2 1000 572.3 7730.9 717.1 820.5 744.7 2000 441,3 584.O 577.8 668.
590,2 5000 331,0 453.O 435.1 524.0 450.3 10000 262.0 373,0 348.2 441.3 368.2 2.OOE+04 213.7' 304.8 277.9 382.729.
5.00E-04 158.6 237.9 210.3 319.2 235.1 1.OME-05 13-7.9 201.3 171.7 281.3 195. 8 2.00E-104 113,8 1
75.8 142.0 247.5 168&2
ýý.OOE-05 93.1 153.8 115.8 213.7 142.0 1.00E+06 86.2 142.7 106.2 195.1 126.2 2.0.E-'06 83,.2 138.0 102.5 157.2 113.1 5MOE+06 7 9.3 131,9 97.8 126.9 102.0 1.OOE+07 76.5 127.6 94.5 113.1 99.3 2.00E+07 73:9 123.2 91.2 104.8 98.71 5.00E-07 7/0.7
- 117, 87.1 98.6 97.8 1.00F+08 68,3 113.8 84.1 9
7.2 97.2 1.OOE-09 60.;
101.
75.2 95.8 95.8 1.OOE+10 54.5 89, 66.9 94.5 94.5 L.00E+11 48.3 80.0) 59.31 93.81 93.s Table, -5 Current Code and Environmental Fatigue Usage Factors Node 'Locat ion/l.ateria Fatigue Usage by Current Code Fatiaue Usape by I
NT-TREC~i/CR-6909 Node 26/Header/CS 0.083
.0.1171
- Node, 48/"NozzleiLAS 0.085 0.302 Node 48/Safe En&CS 0.05
.06 7
C Copyright 10 2007 by ASVEE
MEN 06-119, Supplement 4 Page 8 of 9 r
WJnC Figure 1. Scheinatic of Feedwasvet Pipitig Systemn FWCASý- INZ%250 ý 7FF-13 T-721 Figure 2 Matrheinaticad Model of Feedwater Piping Systein Iq S. Copvright !C' 2067 by ASME
MEN 06-119, Supplement 4 Page 9 of 9 3ittoA An Extract frcoin Prs itref/Tcinperti't re Cycle Dingrinr for feedwatei' Systei 9
9Copyriglitý Cý -007 by ASMEF
DCD Tier 2 Revision 5 Markups
MFN-06-1 19, Supplement 4 - DOD Markups (No.1)
Page 1 of 5 26A6642AK Rev. 05 ESBVR Desigin Control DocnmneattiTier 2 Table 1.9-9 Load Combinations and, Acceptance Critex-ia for C lass I Piping Systens; Conditio Load Comnbiniation for all terins' h j
Acceptance
________________j J
Crite~ria De,,itr PD WT Eq 9:L ý5 Sm NB-3652 Serviceý Level PP. TE, A-T I.
T2.TA-TB. RVI. RVd1 RV.D.
A S, B T-SV. SSEL SSED Eq 12 &, 13 S 24 A~
Fatigutc - NB-3653-S~ervi.:e Level B PP --
WVT 4-(TV)
Eq 9 < 1.
1,,
buth not PP - WT + (RVI)
~zetrthal 1,5 Q_.
PP -- WT -4 (RVi)
PressAitt n1or to exzceed Servic~e Level C PP +- WT -
(C:HUGI)ý 4- (RV 1) 2 ]1"2 Eq 9 :ý2. 25 9,,. bt nott PP +WT
- VCHUGI) 2 ;- (R'V'2 )2 ]pý2 gre~iter rhani 1.8 'Sy pressure njot to exceed I,1.5 Pý (NB -:3 654)
,Sevicc Level D PP WT - [(ssEDk1 +T'
~
PP, - WT - l(S"SEIY -
(CHt0)Cr)
(RV0 1)" 1' PP W
\\T
-~ [`SSEI)
(C j-i GD
- (RV -2I) 2fr2 PP
- wI-T (SE (C O)NDI)2 (R\\ Il 2]'.ý P P -
WV T + RSS.,E T (C ONDI)-
(RX t2]
PP -
WI - [(SSEI)2
(.API)2f (I) R;iia T'SV tloads are used for NIS Lines~ only Eqlý)!<3.0 S_ but not greate titan 210 S,.
P1 esstre n10t to exceed 2.9O Pý (NB-3654)
(2)
RV2 represents RV2 ALL (all valves). RV\\,2SV (Automatic Depressurization operation)
(Sine le Valve) and RtV2 AD (3)
Fo,-r the SRV discharge pipiun~. all direct loads for SRV anid LOCA loads are evaluiated for submerged piping.ý (4)
In conjunction with compliance -with, RG 1.207. the farigrue visage limrit of <0.40 will be used as the criteria for pipin~g locations exempt fromt pipe break consideration.
Where:
API = Aunulutis Pressurization Loads (Inertia Effect)
CHUCH = Chugging Load (Inertia Effect)
ONDI =Condensation Oscillation (Inertia Effect)
PD =Design Pressure PP Peak Pressure or the Operating Pressure Associated with that transient RX71 = SRV O~peniing Load~s (Acoustic Wave) 3.ý9-103
MFN 06-119, Supplement 4 Page 2 of 5 - DCD markups (No. 2) 26A6642AJ Rev. 05S ESB%%`R Design Controt Doctimient/Tiet, 2 0
The pressure. 'Water level, and flow; senlsor instrumientation for t hose. sa fety-rela ted systenis-which are required to finiction following a pipe rulpture, are protected.
Hi'Th-energy fluid systemn pipe whip res.traitits and protective mleasures are des&igned so that a postulated break in one Pipe could nor. in tRnI lead tO a rupture-Of Other nearby p~ipes or comlponlents. if the secondary rupture couild resulit in consequences that. would be consýidered unacceptable for the initial postuflated break.
0 For any p~ostulated pipe rupture, the structural integrity of the containmient structure is maintained.
In addition. for those 'postula0ted ruptures classified as a loss of reactor Coolant, the design lea kti~ghtnes,.s of the Containment fissioni product barrier is maintained.
Safety relief valves. (SRVs) are located and rest.-rainied -so that a pipe failutre would not prevent depressuriza tion.
Protection~ for the FMC:RD scramn insert lines,' is not required. because the miotor operation of the FMvCRD canl adequately insert the control rods even with a complete loss of insert lines (Sutbsection 3,6.2, 1 3).
Thle escape of steani, water. comibustible o1' corros-ive fluids.. gases. and heat in the ev ent.
of a pipe rupture do not preclude:
-accessibilit-y to any areas required to Cope w~ith thle posctullated pip.e rulpture:.
-habitability of the control roomn: or the ability of safety-reIa ted instrumentation, electric power supiplies. colmponents'. and controls. to perform their -sa fety-r-el~ated funciitionl.
3.6.2 Deterinaiii~tion of Bic~k Loca.tionis adDimicEfetAsoitdwith the PostulAted.1Rup1ture of Piping Information concernin2 break and, crack location criteria and methods of analysis for dynamic effects are lis'cussed ink this, Sutbsection in accordance with NNUREG-0S800 Draft Rev, 2.
April 1996. SRP 3.6.ý2, This includes location criteria and methods of anialysis, needed to ev. ah1Iate thle dynlamlic effects Associated wkith positulated breaks and cracks in htln'L and rtodeiate-energyv fluid systemn piping inside and Outside of the prinlary containment.
This minormation pio\\ ide's the basis for the requliertemeIt's for the. protection of safetx -related. structures. s\\ stenils
,and comnponenits dlefined inl the introduction of SecTion 36. -xwhich include-, meetirig the requirements, of CTDC* 4 as it relates to safety-related stritucres. systems and Components ($sc )
bein!
leZ~ nc to accommrodate the dynamic effects ofj01 potulatecl pipe rtipnnlr ilchiding postiulatnonl of pipe rupture locations: break anid crack charzactem istics: dytiarniL analy'Ssi' of pipe-
'Xx hip: and jet imlpinigemnent loads.
The plant mneets the relevant requirements of GDC 4 as follows:
(1)
Criteria defining pos;tulated pipe rupttiue locationl anld confilgurlationls inside Conlta linment are in accordanlle xvith Branch Technical Positioni (BTP) EMEB 3-t. For the pipinig systeml wvith reactor wa (ter, if the enviromnlenital fati~zue is included inl accordance with RG. 1,.207, thle fatigue. usagt. Iflmit should be < 0.40 as tite criteria instead of <'0, 10 for deterimlirng pipe break locations.
MEN 06-119, Supplement 4 Page 3.of 5 - DCD markups (No. 2) 26A6642AJ Rev'. 05 ESBN-8R Design Control Doctutent/Tier 2
()Protection agi:t ostulated pipe rutipures outside contaimnment is provided in accordance with BTP ESMEB 3 -1.
(3)
Detailed acceptance criteria covering pipe-whip dynamic analysis. including determination of the forcing fuinctions of jet thrust and jet impingemient are in accordance with Section HI of SRP 3,6.2, The general bases and assumptions of the analysis are in accordance with BTP EMEB 3-1.
Piping in Containment Penetration Areas No pipe breaks or cracks are postulated in those portions of piping fromu the containment wall penetration to and inchuding the inboard or outtboard isolation valves which meet the following requirements in addition to the requirement of the ASM_%E Code,,Section III. Subarticle NE-1120:
0The followingz design stress and fatigue limits are not exceeded:
For ASME Code, Section 111, Class I Piping The maximum stress range between any two load sets (including the zero load set) does not exceed 2.4 Sm. and is calculated by Equation (10) in N13-3653. ASIME Code.
Section III. If the calculated iiiaximumtn stress range of Equation (10) exceeds 2.4 5m*
the stress ranges calculated by both Equation (12) and Equation (13) in paragraph N'B-3653 shall meet the limit of 2.4 S_.
-~
The cumiulati-ve usage faictor is less than 0.1.
For the piping system with reactor water, if the enx'irormuental fatigue effect is included in accordance with RG 1,.2077 the fatigne usage limit should be :ý 0.40 as the criteria instead of,-" 0. 10 for determining pipe break locations.
The maxinmun stress as calculated by Equation (9) in NB-365_2 under the loadings resulting fr-om a postulated piping, failure beyond those portions of piping, does not exceed the lesser of 2.2 5 Sý and 1.8 S, except that. following a failure outside contaiinment, the pipe between the outboard isolation valve and the first, restraint may be permitted higher stress. provided a plastic hinge is not formed and operability of the valves with such stresses is assured in accordance with the requiiremnent identified in Subsection 3.9.3.
Primary loads include those that are deflection limited by whvip restraints.
ASME Code Section Ill Class I Piping in Areas Other Than Containment Penetration With the exception of those portions of piping identified above, breaks in ASME Code.
Section HLI Class 1 piping are postulated at the following locations in each piping and branch rttn:
- At. terminal ends.
- At intermediate locations whAere the inaximiunn stress range as calculated by Equation (10) in NB1-3653. ASMIE Code.Section III exceeds 2.4 Sm. and either Equation (12) or Equation, (13) in Paragraph NB-3653 exceeds 2.4 S,.
- At intermediate locations where the cumulative usage factor exceeds 0.1. As -a result of piping reanalysis caused by differences, between the design configuration and the as-built confliguration, the highiest stress or cumiulative usagae fatctor locations may be shifted; 3.6-S
MEN 06-119, Supplement 4 Page 4 of 5 - DCD markups (No. 2) 26A6642AJ Rev. 05 ESBNNR Desigu Control DocuinenilTier 2 however,, the initially determined intermediate break locations need not be changed unless one of the following conditions exists:
Thle dynamic effects fr~omi the new (as-built) intermediate break location,', are not mitigated by the original pipe whip restraints and jet shields.
-A chan'ge is required in pipe parameters. such as major differences in pipe s-,ize, wall thickness, and routing.
For the piping system With reactor water, if the environmental fatigue effect is included in accordance with RG 1.207. the fatigue usage limit should be -- 0.40 as the criteria. instead of K0. 10 for determining pipe break locations.
3.6.2.1 Criteria Used to Define Break and Crack Location anid Configuration The following subsections establish thle criteria for the location and configuration of postulated breaks and cracks.
Definition of High-Energy Fluid Systems Higgl-ener-gy fluid systems.- are defined to be those systems or portions of systems that, (hiring normal plant conditions (as defined in Subsection 3.6.1.1)., are either in operation or are maintained pressurized uinder conditions where either or both of the following are met:
" nmaximum operating tenmperature exceeds 93. 3'C (200'F); or niaxiurnu operating pressure exceeds 1.9 MVPaG (275 psig).
Definition of Moderate-Energy Fluid Systems Moderate-ener~gy fluid.systems are defined to be those systems or portions of systems, that, during-normal plant conditions (as defined in Subsection 3.6. 1. 1). are either in operation or are maintained pressurized (above atmospheric pressure) under conditions where both of the following are met:
0 maximum operating, temperature is 93.3'C (1-00'F) or less; and 0
miaxuiniun operating pressure is 1.9 MPaG (275 psig) or less.
Piping systems are classified as moderate-enemgy systems w.hen they operate as high-energy piping for only short operational periods in perfoirting their system function but, for the niajor operationial period. qualitý7 as mioderate-energy fluid systems.
An operational period is considered short if thle total fraction of time that the system operates within thle pressure-temperature conditions specified for highi-energyfluid systems is less than 2%of the total timie that the system operates as a nioderate-eniergy fluid system.
Postulated Pipe Br-eaks and Cracks A postulated pipe break is defined as a sudden gross failure of the pressure boundary either in the form of a complete circumferential severance (guillotine break) or a sudhen longitudinal -split without pipe severance,. aiid is postulated for high-energy fluid systems only.
For moderate-energy fluid systems-. pipe failures are limited to postulation of cracks in piping and branch rims; these cracks affect the surrounding en-virounmental conditions only and do not result in whipping of the cracked pipe.
High-energy fluid systems, are also postulated to have cracks for 3.6-9
MEN 06-119, Supplement 4 Page 5 of 5 - DCD Markups (No. 3) 26A6642AJ Rev. 05 ESB'%NVR Desi-gn Contr~ol Doeminen,itTiei' 2 In place of the response s;pectrium analysis. the ISMV timle history mnethod of -analys4is is used for mnulti-supported systemns sub~jected to distinct support motions. in which case both inertial and relative displacement effects are already inc ludled.
- 3. 17.3. 10 Use oif Equivalent Vertical Static Factors Equivalent vertical static factors are used when the requiremients, for the static coefficient mnethod in.Subsection 3.7.2.1.3 are satisfied.
- 3. 7.3.11 Torsionial Effrcts of Eccentric.3Masses Torsional effects of eccentric miassesi are included for subsystemis similar to that for the piping systemlS dli'scussed in Subsection 3.7.3.3.1.
- 3. 7.3.12 Effect of Differentialt Buildingq Movemients In miost case~s, subsystems are anchored and restrained to floors anid walls of buildings that may have differential movements clurinz a s'eismnic event.
The movements inra range froml insielnificant cliffecrential disp laceinents between rigcid walls of a commnon building at low elevations to relatively large displacements between separate buildings at a high seismlic activity
'site.
Differential endpoint or restraint deflections cause forces and mioments to be induced into the system. The stress thus produced is a secondary stress. It is justifiable to place this stress, which results' froml restraint, of free-end displacement. of the system., in the secondary stress category because the stresses are self-limiting and,. when the stresses exceed yield strength. mninor distortions, or de formlat ions wxithlin the systeml satisfy the condition wvhich caused the stress to
- occur, For the piping stress analysis-. SRSS combination for the inertial and the SAM (Seismnic Anchor Motion, incluiding Effect of Differential Building Movements) responses is, acceptable. For the piping. support design. the absolute suini mnethod (ABS) is used.
- 3. 7.3.13 Seismic Catiegory I Buried Piping, Conduits and Tunnels There is no directl1y buried Seism.iic Category I (C-1.) piping or conduits that are dlirectly buried underarotunl.
Fire Protection Systemn (FPS) yard piping with a C-I classification are installed inl co 'ered reinforced conicrete trenchies near surface with remiovable covers to facilitate mnainitenance and inspection acce-.ss There are C-I conduits in four electrical duct banks from the CB to the RB, Tlte duct banks, are instaalied in clo(.sed concr~ete trenches co-,vered with. backfil.11.
There are no C-I tunnels in the ESBVVR desigzn.
The access, turnnel (AT). which includes wý%alkways between and access to RB, CB. Turbine Bu1ilding (TB), and Electrical Building (EB) is clas'sified Seismlic Category II (C-IT). Since C-LI structures are designed to the samne criteria as' C-I trucuresthere is no inlipact to actjacenit C-I stru~ctures, 3.7-24