ML20205S330
| ML20205S330 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 04/30/1986 |
| From: | CYGNA ENERGY SERVICES |
| To: | |
| Shared Package | |
| ML20205S323 | List: |
| References | |
| TASK-03-06, TASK-3-6, TASK-RR NUDOCS 8606120192 | |
| Download: ML20205S330 (200) | |
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NRC PRESENTAT ON RESOLUTION OF PENDING SEP ISSUES YANKEE NUCLEAR POWER STATION Volume 2 Jigsentggggpssp3Mwepy?: tear @ggfigyp;)RVN4%@MWfMWSMf?NN@?stitM86t@eMedWRgat##MI676eW#0FWM !
April 8-11,1986 8606120192 860606 PDR ADOCK 05000029 P PDR
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l REV. 0 VOL. 2 I
I I
I I
I NRC PRESENTATION RESOLUTION OF PENDING SEP ISSUES YANKEE NUCLEAR POWER STATION I
I I
I I
I APRIL 1986 g
I l I YNPS - RESOLUTION OF NRC ,0UESTIONS ON SEP TOPIC III-6
l YANKEE ATOMIC ELECTRIC COMPANY U.S. NUCLEAR REGULATORY COMMISSION l CYGNA ENERGY SERVICES g
I MEETING AGENDA FOR l YANKEE NUCLEAR POWER STATION RESOLUTION OF SEP TOPIC III-6 I
LOCATION: CYGNA 0FFICES - 10TH FLOOR 101 CALIFORNIA STREET I SAN FRANCISCO, CALIFORNIA 9tl111 I
l MONDAY. APRIL 7. 1986 NRC CAUCUS WITH STAFF CONSULTANTS g
I APRIL 8 TO APRIL 11. 1986 REVIEW 0F
SUMMARY
TABLE OF SEP NRC QUESTIONS REVIEW OF
SUMMARY
OF WORK MATRIX I PRESENTATION OF IN-SITU DYtJAMIC TESTING OF SMALL BORE PIPE PRESENTATION OF CRITERIA AND METHODOLOGY l
g RESPONSE TO NRC QUESTIONS SCHEDULED FOR APRIL RESOLUTION g
1 YllPS - RESOLUTION OF NRC.00ESTIONS ON SEP TOPIC III-6
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I OVERVIEW 0F PENDING SEP TOPIC III-6 ISSUES AND RESOLUTION SCHEDULE I
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' YNPS - RESOLUTION OF NRC QUESTIONS ON SEP TOPIC III-G e
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l PA6E 1 I 05/01/86 YANKEE NUCLEAR FDNER STATION -
SUMMARY
OF SEP TOPIC !!!-6 NRC QUESTIONS RESOLUT10N 90EST10N RESOLUTION REFERENCES SCHEDULE ITEM NO. ISSUE
SUMMARY
SOURCE
- I A GENERAL DUESTIONS l
Clarify saf e shutdown scope and 5/20/86 Al Hot shutdown scope should NRC he analyzed to NRC spec- criteria used.
traorjustificationfor I not performing such analyses should be provi-ded.
Clarify criteria used.
5/20/86 A2 Acceptance Criteria NRC l
should be:
(1) YCS spectra loading vs. allowable values (2) NRC spectra loading vs. SEP guidelines I or functionality.
However, in some cases YCS was used with SEP guidelines.
D. R. LeFrancois 05/20/86 A3 Issce sussary for easonry NRC malls, combine the review of IE Bulletin 80-!!
with the SEP seissic reevaluation.
Safety evaluation is cospleted. NRC SER, Resolved.
A4 Spent Fuel Pool b Chute NRC dated 2/1/83.
B EVALUATION CRITERIA NCT(0), D. R. LeFrancois YAECletter 5/20/86 B1 Soil Bearing Capatity-to NRC.
20 ksi used dif fers free NRC 10.6 ksf given in Seis- -
nic Reevaluation Cri-teria.
NCT(0), The methodologies for Presentation Resolved.
32 Effects of three earth-(1) combination of three earthquale to NRC, quale cosponents and in- NRC cosponents, and (21 generation of document structure response spec-I tra generation criteria mere not provided for in-structure response spettra are described in the Reference.
dated 2/24/86.
revien.
I B3 'Only NRC loading is spe-cified as seismic input, NCT(0) Clarify acceptance criteria used.
5/20/86 YCS is not addressed.
Desonstrate that the RSS SSI study - 6/24/06 34 The generallred assusp- NCT(0) tion that SSI is negli- is applicable for all structures, gible say not be valid and thus the SSI effects are negli-for all structures. gible.
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PA6E 2 I
05/01/96 YANKEE NUCLEAR POWER STAi!DN - SUMARY OF SEP TOPIC III-6 NRC QUESTICMS GUESTION RESOLUTION ISSUE SUMARY SOURCEe RESOLUTION REFERENCES SCHEDULE ITEM 110.
C STRUCTURES Cl VAPOR CONTAINER IVC) llCT(1), Calculate the required stresses, se Resolved, Cla Provide calculations for shear stresses in coluans NRC and tie beans.
Evaluatethesubjectbolts, se Resolved.
I C1b Provide calculations for NCT(5),
pull-out of coluen base NRC anchor bolts.
Clc See ites B1 above. NCT(2),
5/20/86 NRC Cid MS/BFD pipe ancher loads NCT(3) Re-evaluation of VC penetrations -
6/24/86 used to qualify VC penet- mill be performed when the MS/FW rations say no longer be analysis is completed.
valid because of recent I reanalysis el those lines.
et Cle Provide the evaluation NCT(4) Perforethesubjectevaluation. Resolved.
results for the clevises and turnbuckles on the diagonal tie rods.
Cif Evaluate the adequacy of NCT(6) Perfore the subject evaluations, se 5/20/06 the clearance between the VC and the ISS, and the I VC and the radioattive pipe tunnel.
I Clg Benerate ARS under NRC spectrueloads.
NCT(7) Clarify criteria used. -
5/20/86
(
I C2 REACICR SUPPORT STRUCTURE (RSS)
C?a Validatien of computer NCilB), Provide the existing documentation -
5/20/66 I code FRA should be provi-ded.
NRC to NRC for its' revism, I
I * , 4
PAGE 3 I 45/01/86- -
YAR EE NUCLEAR POWER STATIDN -
SUMMARY
OF SEP TOPIC 111-6 B C QUESTIONS 9UESTION RESOLUT104 SOURCEe RESOLUTION REFERENCES SCHEDULE ITEMNO. ISSUE
SUMMARY
ET(10), The correct value of the displace- - Resolved. l C2h Explain why easieue dis- !
placeeent of the support NRC sent under YCS is 0.0543 f t, not I coluans is greater for YCS than for NRC spettra.
0.543 f t as reported as a result of a typographical error.
I NCT(5), Evaluatethesubjectbolts. ** Resolved.
C2c Provide calculations for coluan based anchor bolts. NRC WCT(2), Saee as ites 31 above.
5/20/86 C2d See ites B1 above.
NRC ET(9) Provide the required clarification. #e Resolved.
C2e Clarify the seismic loads I spplied to the FE sodel of the steel collars.
Provide the required clarification. et Resolved.
C2f Clarify the definition of NCT(ll)
- yield' of the ring in relation to ' ultimate'.
The saae row has been entered Seisaic Analy- Resolved.
C2g Explain why there are two NCT(0) twice as a result of a typogra- sis of the RSS, identical rows in Table E-3 of the Reference. phical error. Rev. 3, dated 3/03.
C3 TURBINE BUILDING (TB)
NCT(12), Desonstrate that the results (ARS se Resolved.
C3a Turbine pedestal casping value should be 3 to 51 NRC and SAM input to Main Steas piping instead of 71. analysis) are justifiable.
Units of sodal sasses given on to Pesolved, C3h TB sodal sasses do not NCT(14),
add up to 90% of the to- NRC Table 4.1 of the TB Strututural tal sass. Analysis Report are in error.
Suasarize the cost-benefit study -
5/20/06 g C3c Code allowables are ex- NRC periorsed to choose the TB fin m creded under YCS loading.
schese.
Provide the required clarification, se Resolved.
C3d Clarify if the of fice bu- WCT(13) ilding is supported by the TB in only one or in I both horizontal direc-
tions.
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YANKEE NUCLEAR POWER STATION - SulflARY OF SEP TOPit !!!-6 NRC OVESTIONS I DUESTiJN RESOLUTION SOURCE
- RESOLUTION REFERENCES SCHEDULE ;
ITEM NO. ISSUE
SUMMARY
NCill5) 3. R. LeFrancois Suasary Design 5/20/86 C3e Evaluation of the unrein-forced easonry walls are Report for not presented for review. Block Wall llo-difications insideTB,by C.T. Main, Inc, dated 5/85.
Provide the required clarification, se Resolved.
C3f Clarif y how addition of NCTil6)
I braces will alleviate se-isait uplift.
C4 DIESEL 6ENERATOR BulLDING (DBS)
C4a.1 Clarify the vertical The subject coupling has not been 6/24/86 I
NCT(19),
coupling between the DSB NRC considered since it is assused that and the Nitrogen Tower its' effect is insignificant. This and Accueulator Tank To- assumption can be justified.
wer.
C4a.it Clarify the vertical NCT(20), The PAB provides vertical support -
6/24/86 I coupling between the DSB and the PAB (the sanner in which the coupling is NRC to a DSB girder at node point 17 (see Fig B.2 of the reference re-port).
sodeled).
[4a.iii Clarif y the direction of NCit20) Provide the required clarification. -
6/24/86 coupling between the D6B and the PAB.
C4b Provide evaluation of po- NCTl23) Desonstrate that uplift does not -
6/24/86 tential seismic uplift of exist.
the coluan base.
NCT(17) See ites B4 above.
6/24/86 3 C4c Address the effect of SSI
- g which was neglegted in the analysis.
C4d Provide the properties of NCT(18) Provide the required properties. -
6/24/86 the three beas elements shown in the vertical ao-del.
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PAGE 5 l C/01/86 TANKEE NUCLEAR PONER STAi!DN -
SUMMARY
0F SEP TOPIC 111-6 NRC SUEST10NS RESOLUTION 90ESTION SOURCEe RESOLUTION REFERENCES SCHEDULE ITEM N0. ISSUE
SUMMARY
C4e Clarify the inconsistency NCT(21) The locations of the knee brace, as - 6/24/86 in the locations of the used in the analysis model are n-kner braces between the cording to a field walkdown. Ve ify design drawing and the the accuracy of the model and t.<-
analysis model. rett if necessary.
NCT(0) Subsequent to the publication et 6/24/84 C4f Results of evaluation for YCS load case are not the reference report, the D6B sith presented for review. a revised sodification schese sas analyzed under the YCS load. Pcs-sent the results of that analysis for revies.
The correct value of the subject -
Resolved.
C4g The reported seissic NCT(22) story drif t indes of drift index is 0.0016, not 0.016 as 0.016 exceeds the allcu- reported as a result of a typogra-able story drift index of phical error.
0.005.
NCT(24) Perfore the required evaluatica and -
6/24/86 C4h Provide the evaluation results for the knee bra- present the results.
ces in the D6B.
NCT(24) The valls are designed to NRC spec- -
6/24/06 C4i Provide the evaluation results for the rein- trus leads and are included in the I forced blocknalls in the annen.
analysis models. Present the re-sults.
NCT(15) There are no SSS equipment in the -
6/24/86 C4j Provide the evaluation I results for the unrein-forced concrete sasonry DGB.
walls and the ef fect of their anticipated failure on nearby SSS equipsent.
C5 PRIMARY Aul!LIARY BUIL-DING (FAB)
NCT(12), Demonstrate that the results are 6/24/86 C5a Daa;ing value should be 3 to 51 instead of 71. NRC justifiable.
C5b Use of UBC criteria NCil25), UBC is specified as a referente co- Seistic Reeva- Pesolved.
shouldbejustified. NRC de in Section 3.1 of the Reference. luation and
Furthermore, the equation Retrofit Cri-vu=2*(fc'lH0.5 is also in the ACI teria, Rev. 2, Code, ACI 318-77. dated GIB2.
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- PAGE 6 C/01/B6 YANKEE IRJCLEAR POWER STAT 10N -
SUMMARY
OF SEP TOPlc III-6 NRC 9tESTIONS I GUESTION RESOLUTION REFERENCES SCHEDULE j I
ISSUE
SUMMARY
SOURCEe RESOLUTION ITEM NO.
l C5c Modeling of malls by NCT(27), The stiffness of the salts is ** Resolved.
equivalent coluans should NRC represented properly by the be justified. equivalent coluans used in the model.
5/20/66 C5c.i Veriitation of 8ATS NCT Cosputer Progras C5d Provide evaluation of po- NCT(33), Desonstrate that uplift forces can ** Resolved.
tential selssic uplift at NRC be resisted by existing structural coluan lines 8 and Fb. elements.
C5e Provide the criteria for NCT(0) Provide the required combination Presentation Resolved.
combining the oflects criteria. to NRC from three earthquake Docusent, cosponents. dated 2/24/B6.
C5i SSI effects say be signi- NCI(26) Shou that the effect of SSI on ver-6/24/86 ficant in the vertical tical response is insignificant.
direction analysist pro-videjustificationfor its' caission.
The radioactive pipe tunnel extends se Resolved.
C5g Provide the properties of NCT(28) the 'dussy' subdiaphrage fres Level 2 to ground and is not and explain uhy the sub- attached to Level 1. To reflect diaphrage does not appear this situation in the BATS computer in the figures illustra- sodel, the massless 'dussy' subdia-ting the sodel, phraga sentioned was used. This subdiaphrage is not connected to the level 1 sain diaphraga.
I C5h Clarify if the out-of-Pl ane degrees of freedes are also restrained by NCTI29) Out-of-plane degrees of freedos are Presentation Resolved.
not restrained by the in-plane ri- to NRC gid diaphras option of the compu- Docusent, the in plane rigid dia- ter progras BATS. dated 2/24/86.
phras assumption.
The esisting vertical sodel is a se Resolved.
C5i Clarify why the vertical NCil301 sodel shows the radioat- sisplified representation of the tive pipe tunnel to be attu.1 design. Investigate the sig-uncoupled from the bull- nificance of the coupling between ding at the roof, the pipe tunnel and the roof and
- sodify the model accordingly if the coupling is significant.
8
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PAGE7 05/01/86 YANKEE NUCLEAR PONER STATION -
SUMMARY
OF SEP TOPlc 111-6 istC QUESTIONS RESOLUTION GUESilGN SOURCEt RESOLUTION REFERENCES SCIEDULE ITEM NO. ISSUE
SUMMARY
The correct value of the subject PAB and Radio- Resolved, C5j Verify if the shear 11CT(31) stress of 130 psi report- shear stress is 13.0 psi, not 130 active Pipe ed in Table E-1 of the psi as reported as a result of a innnel Struc-Reference is accurate. typographical error, tural Analysis Report,Rev.
2, dated 1/83.
Fill-in the missing entries of the se Resolved.
C5k Cosplete Tables E-4 and NCT(32)
E-10 of the Reference by subjectTables.
providing the eissing en-tries.
NCT(15) D. R. LeFrancois Suasary Design 5/20/86 C51 Provide the evaluation I results for the unrein-forced concrete sasonry malls and the effect of Report for Bloctuall Mo-difications at PAB North Nall, I their anticipated failure on nearby SSS equipeent. Upper Pipe Chase,and Cable Spreading Race, by C. T.
Main,Inc, dated 2/86.
C5e Perfors confiraatory NRC Perfore the analysis. -
6/24/86 analysis of PA8 using NRC spectra and SEP criteria I _.
Sussarire YAEC's analysis results Fire Water 5/20/86 C6 FIRE NATER TANK I addressing the NRC concerns. Tank Seissic Analysis,by YAEC, Report I
No. YAEC-1492, dated 9/85.
1 C6a Allenable buckling stress NCil34), D. R. LeFrancois 5/20/86 of 24.1 ksi should be MRC l justified.
9 I .
PA6E 8 C/01/86 I - - - - - __ . _ _ _ _ _ _ _ _ _
l f
YANKEE NUCLEAR PONER STAi!ON -
SUMMARY
0F SEP TOPIC 111-6 NRC SUESTIONS
)
RESOLUil0N 00EST!Dil SOURCES RESOLUT!DN REFERENCES SCHEDULE ITEM NO. ISSUE
SUMMARY
NCTl37,38), Present analyses performed as
- 5/20/86 Cab,C6c Provide the following evaluations: NRC they address these issues. J D. R. LeFrancois j
- i. Task base anchor bolts li. Tank shell under nortle loads lit. Soil bearing capacity Iv. Settlesent of soit under tant.
NCil35) Present the VAEC analysis results. -
Resolved.
C6d Reconcile the difference between the design dra-I sings and the analysis model in tank mall shell sodeling.
NCil36) Present the YAEC analysis results. -
5/20/86 C6e Describe the hydrodynasic loads which result in the D. R. LeFrancois stress distribution shown I in Fig. E-1 of Rev. O Cygna analysis.
C6f Desenstrate the adequacy NCT138) Perfore the required evaluation and -
5/20/86 of free board vs. the present the results.
sarisas amplitude of wa- D. R. Lefrancois ter slashin; such that the tank roof will not be impacted during earthqu-akes.
C7 Perfore confirmatory NRC D. R. LeFrancois 6/24/86 analysis of a sasonry I sall using NRC spectra and SEP criteria.
O PIP]NS SYSTEMS
-- et Resolved, Cl Clarify the hot shutdown NRC scope under NRC spectra loading.SeeitesAl above.
I D2.1 Describe criteria used
- for pipe supports.
NRC Prepare a sussary document
- i. describing the criteria used and
- 5/20/06 D2.2 Present results of pipe . ii. sussarizing the calculations - 6/24/06 support evaluation. and results. Include a list of suppcrts installed.
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PAGE 9 05/01/86 YANKEE NUCLEAR PDuER STAil0N - SultlARY OF SEP TOPIC 111-6 NRC GUEST 10h5 RESCLUTION OLEST!DN -
RESOLUTIDel REFERENCES SCHEDULE ITEM 110. ISSUE
SUMMARY
SOURCEe E666tl.11 See Retrofit criteria locueent, 6/24/86 33 Address potential for impact loading of Appendia J.
I splifting supports.
5/20/86 H Provide codeling sethods E646(1.2) for support stiffnesses.
6/24/06 6 a6(1.D I B5 Provide correlation between the response spectra nuebers in Vol. I of t8u safety-related piping analysis report and the plot numbers in Scot 2 of Vol. 2.
For justification of deeping values 6/24/86 D6 Justify tar app opriate- E616t!.8) ness of response spectra used see itees C3a and C5a above.
I based on high desping values or complete analyses using appropriate values.
6/24/26 D7 Justify the E666(1.5)
I diesel generator building spectra.
See Retrofit Criteria Document, H Resolved.
98 Clarify the method of ES&6tl.6)
I combining stresses. Section 0.3.1.
se Resolved. ;
D9 Clarify the temperature E616(1.7)
I bases of the stress limits.
H Resolved, I DIO Esplain the differences in stress results in Tables 5.5-1 and 5.8-1 E616tl.8) of Vol. I of the piping analysis report.
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PA6E 10 C/01/86 YAREE NUCLEAR POWER STAi!0N - SUMARY OF SEP TOPIC !!!-6 MC GUESTIONS OUESTION RESOLUT10N ISSUE SUMARY SOURCEe RESOLUT10N REFERENCES SCHEDULE ITEll NO.
Ill Esplain the differ- E666(1.9) 6/24/86 entes in Equation 12 I results between Tables 5.5-1 and 5.8-1 f or probles 3 in the piping analysis report.
012 Provide results for E666(1.10) Analysis under NRC spectrue loading **' Resolved.
the NRC seismic loading is not required.
for probless 2, 3, and 23.
es Resolved.
313 Provide Section 5.11 E666(1.ll) of the safety-related piping analysis report.
914.1 Provide the Refer to Refer to lies D2 above. 5/20/B6 sethodology for the Ites D2 pipe support evaluation, above.
EG66(l.12)
D14.2 Provide the calculation Refer to Refer to Ites D2 above. 6/24/06 I package for the pipe support evaluation.
Ites D2 above.
E666(1.12)
D15 Provide evidence that E6tS(1.131 D. R. LeFrancois 5/20/06 ILE Bulletin 79-02 requirements are I satisfied.
5/20/86 D16 Clatify hot shutdoun E666(1.141 systes boundaries and the associated stress problees.
D17 Describe the procedure NRC-2(D1)I Provide the required clarification. Pipe Stress Resolved, used to determine the k alysis Re-results shown in Table port, Vol. One, 5.5-1 of the Reference. Book I of 1, Rev.1, dated 5183.
I ..
12
, = _ _ _ _ _ _ _ __
95/01/86 PAGE 11 YANKEE NUCLEAR POWER STATION -
SUMMARY
OF SEP TOPIC 11!-6 NRC QUEST 10NS SUESTION RESOLUTION ITEM NO. ISSUE
SUMMARY
SOURCEe RESDLUTIDN REFERENCES SCHEDULE B18 a. Reconcile the diffe- IItC-2(D3) Provide the required Pipe Stress Resolved.
rences between the allos- clarifications. Analysis Re-ables shown in Tables port, Vol. One, 5.5-1 and 5.8-1 of the Book i of 1, Reference for Probles 207. Rev. 1, dated
- b. Explain why the sus of 5/B3.
the component loads in Table 5.8-1 of the Refe-rence does not total the Equation 12 result for Probles 207.
D19 Explain uhy the seissic NRC-2(D41 Provide the required explanation. ** Resolved.
stress for NRC loading is less than for YCS for Probless 041A and 0418.
D20 Perfora confirsatory 6/24/B6 analysis of Reactor I Coolant Loop using NRC spectra and SEP allowables.
D21 Perfors confirsatory 6/24/86 analysis of DSSS Piping probless using NRC spectra and SEP allowables.
E MAIN STEAM AND FEEDWATER (MSIFW) PIPING AND SUPPORT STRUCTURE El Provide modeling and ana- NRC Add a section to the Retrofit MS/FW Piping 6/24/86 lysis details. Criteria Docusent to address the Analysis Re-NRC questions on modeling details. port, Rev. O, dated 7/84.
E2 Clarify why SAM analysis E6&6(2.16), SAM for probless other than Probles -
6/24/86 was periorsed only for NRC 9 are not considered because, for Prchles 9. those probless, the structures to which the piping is anchored is so-deled together with the piping.
13
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Z PA6E 12 C/01/86 VANKEE NUCLEAR POWER STAi!DN -
SUMMARY
OF SEP topic 111-6 WRC GUESTIONS DOEST 10N RESOLUTION SOURCE
- RESOLUTION REFERENCES SCHEDULE ITEN NO. ISSUE
SUMMARY
E3 Clarify how coupling bet- IRC The MS and FN lines have been so-
- 4/24/86 seen the MS and FN lines deled separately, has been considered.
E4 Analysis of MS,FN piping NRC Clarifycriteriaused. es Resolved.
and support structure un-der NRC spectra loading should be periorsed.
E5 Clarify the require- E666(2.ll Criteria are the sase. et Resolved.
eents for safe shutdoun scope piping vs. hot shutdown scope piping.
E6 Provide representative E616(2.2) 6/24/86 I models of the systes and sufficient infor-sation to verify model.
E6&6(2.3) YAEC will delete Ref. 15 froe et Resolved.
E7 Provide Reference 15 of the licensee report. a future Rev. to this report.
EB Provide a complete E616(2.4) 6/24/86 description of the modeling techniques used.
E9 Provide stiHness values E616(2.5) See El.3 6/24/86 used for anchor A-50 and the eethods used to determine thes.
E10 Provide cosplete details ES&6(2.6) 6/24/86 )
about the VC, Turbine IIRC Bldg., and RSS sass and stiffness modeling.
E616(2.7) YAEC mill subsit the reference. 5/20/84 Eli Provide Reference 4 of the report for review.
E12 Clarify hos seismic E616(2.8) The required clarification is Presentation Resolved.
anchor sovesent loads presented in the reference. to NRC, were combined with docu6ent
' other loads, dated 2/24/96.
l 14 I
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G/01/86 PA6E 13 YANKEE NUCLEAR PONER STA110N -
SUMMARY
OF SEP TOPIC 111-6 INIC OUESTIONS OUESilDil RESOLUTION ITEM 110. ISSUE SLMMARY SOURCEe RESOLUTIDM REFERENCES SCHEDULE E13 Justify why the llRC E66612.1) se Resolved.
spectra were not used in a postulated seisaic loadcase.
E!4.1 State what industry E666(2.10) se Resolved.
code and edition are used to define the stress equations used.
E14.2 Show this is in E6&Gl2.10) et Resolved.
compliante with SEP reevaluation guidelines.
E15 Clarify the sethod E666(2.ll) Modal responses are combined ee Resolved.
used to combine the according to llRC Reg. Guide 1.92.
sodal responses.
I E16 Confire that Su was E616(2.12) 6/24/86 to be included as an aliceable value in Table 3-3.
E17.1 Provide the E61612.131 6/24/86 sethodology for the pipe support evaluation.
E17.2 Provide a calculation E616(2.13) 6/24/86 I package for the pipe support evaluation.
ElB Show that all require- E646(2.14) 6/24/86 sents of I&E Bulletin 79-02 are satisfied. l I E19 Evaluate the effects of sodifications to the MS/
FW support strutture E616(2.151 6/24/86 j l
upon the stress levels j I in the attached piping.
f i
)
E20 Clarify why seismic E666(2.16) 6/24/86 anchor sotion analysis
uas performed for j probles 9 only. i )
15
PAGE14 05/01/84 TANKEE NUCLEAR PONER STAT 10N -
SUMMARY
OF SCP topic Ill-6 N C 90EST10NS 90ESTION RESOLUTION RESOLUTION REFERENCES SCHEDULE lien NO. ISSUE
SUMMARY
SOURCEe F NAJOR MECHANICAL COMPONEXTS F1 GENERAL CONCERNS Fla Perfore analyses under E616(3.11, Clarify criteria used. -
5/20/86 NRC spectra loading. NRC Flb SEP guideline stress 11- E616(3.2), Clarify criteria used.
5/20/84 sits were used under YCS Et loading while industry codes and standards should have been used.
I F2 REACTOR PRESSURE VESSEL (RPV)
F2a Analysis of RPV uplift E6tGI4.3), Perfore a refined analysis to de- -
5/20/86 under NRC spectra loading NRC sonstrate that RPV neither uplifts should be perforsed. ner slides under NRC spectra lo-ading, and thus resolve the EC concerns raised in items Fla, t2a, and F2h (beloul simultaneously.
I F2b Analysis of RPV sliding under NRC spectra loading E6tGt4.4), See ites F2a above.
NRC 5/20/86 shcald be performed.
F2c Provide sufficient detail E616(4.1) Provide the required clarification. -
5/20/86 so that reference to pro- See lies Flb above.
I per code sect: ens and al-louables can be verified.
I F2d Justify applicability of E616(4.2) the Bijlaard sethod.
Providetherequiredjustification. -
5/20/84 F3 REACTOR INTERNALS F3a Clarify what ARS is used E616(10.8 ), Provide the requested inforsation - 5/20/86 in the evaluation. NRC to NRC f or its' review.
F3b Provide the following in- NRC, Providetherequestedinforsation -
forsation: to NRC for its' review.
I '
- i. allewable stresses, ii. test data used, lii. allevables for the E616(10.3)
E616(10.6)
E616(10,5) 5/20/86 I bolts if bolts were used.
I 16 i
1
PAGE 15 C/01/86 YANKEE IRJCLEAR PONER STATION - SU M ARY OF SEP TOPlc 111-6 NRC OUESTIONS OUEST10N RESOLUTION ISSUE SumARY SOURCEe RESOLUTION REFERENCES SCHEDULE ITEM NO.
F3c Desonstrate that both YCS E6&Sil0.2) Only YCS seisaic loading has been Selseic Evalu- 5/20/B6 and NRC have been consi- considered in the analyses reported ation of YNPS dered and present results in the Reference. Justify leading Reactor Ir.ter-for both cases, considered. nals,lopell Corp., Report I e 02-0570-1204, dated B/84.
5/20/86 I F3d Clarify whether sain coo- E6&B(10.4) Provide sain coolant loop sodeling lant piping was used for support and if so, provi-details, de modeling details.
F3e Address any possible ad- E616(10.7) Address the subject concern. -
5/20/86 verse consequences of fuel assembly liftoff.
F4 STEAM SENERATORS (SG) AND SUPPORTS F4a Allowable stresses should E666t5.2), i. Theapplicableallowable MajorEquip- 5/20/86 be specified. NRC stresses will be identified in sent Gualifi-I the next revision of the Design cation Report, Criteria docusent.
- 11. The nuserical values of the Rev. 1, dated 5/84. 5/20/86 I allowable stresses used are given in Tables 3.7 and 3.8 of the reference report.
F4b Clarify how asin coolant E616(5.11, MCL thersal expansion will irduce Resolved.
loop (MCL) theraal expan- NRC negligible loads on the 56 because sien loads on S6 were of absence of thereal restraints on considered. the MCL, F4c Clarify the SG support E616(5.4), (1) The support frase was analyzed -
5/20/86 I frase analysis. NRC as individual coluons. However, the frasing action provided by the horizontal sesbers was taken into consideration in calculating the allowable coluen loads.
(2) Provide the required evaluation - 5/20/06 of the WlBx96 horizontal seabers.
F4t' Provide evaluation of $6 E616(5.51, The snubbers and their anchoring CYENA Calcula- 5/20/86 anchorage. NRC systes have been designed by CY6NA. tion Binders The evaluation will be sussarized E50006C/lF and and presented to the NRC. 2F.
I 17
_ = - - _ . .. - .
-- = - - . - ~. ..-
PAGE 16 95/01/D YANKEE NUCLEAR PONER STATION -
SUMMARY
OF EP TOPIC 111-6 IItC EUESTIONS GUEST 10N RESOLUT191 SOURCEe RESOLUTIDN REFERENCES SCHEDLLE ITEM NO. ISSUE
SUMMARY
F4e Justify applicability of E666(5.3) This ites will be resolved simul-
- 5/20/06 thelijlaardmethod. taneously with itees F2d and F5c.
F5 PRESSURl!ER AND SUPPORTS F5a Provide evaluation of E666(6.2), (1) The evaluation of the pressuri- Major Equip- 5/20/B6 pressuriter support frase. NRC ter support frase is described in sent Qualiff-sufficient detail le Chapter 5 of cation Report, theMajorEquipmentGualification Rev. 1, dated I Report.
(2) CY6NA has designed a modifica-tion to the pressuriter support sys-5/04.
- 5/20/B6 I tes by introducing ' SAM eliminator' bracesatElev. 1091'3.5'. This so-dification can be suesarized for NRC's review.
F5b State individual load E6&GI6.1) Individual load cases included in Major Equip- 5/20/86 the combined loading are clearly sent Qualifi-I cases included in the coabined loading, stated in the Reference. cation Report, Rev. 1, dated 5/84.
F5c Justify applicanility of E6n6(6.3) This ites mill be resolved sinut- 5/20/84 theBijlaardmethod, tanecusly with itees F2d and F4e.
F6 MAIN COOLANT PUMPS Individual load cases included in Major Equip- 5/20/86 I
F6a State the esatt sethod of E666t7.1) load case coabination. the combined loading are sent Qualifi-stated in the Reference, cation Report, Rev. I, dated 5/84.
F7 HOT SHUTCC H SYSTEM VALVES F7a 1. Shcw that the proper E616(S.1) 1. Shos that ASME Code rules as 5/20/86 ASME Code rules for the used are appropriate.
appropriate class are I used.
- 11. Show that SEP reeva- 11. SEP reevaluation guidelines are used with YCS seisaic loading. This 5/20/06 lustien guidelines are I used with NRC spectra
~ loading.
can be justified. See ites A2 above.
18
u we ,---m.w, ___
~
- , _ ~ . . _ _ . . _
PAGE 17 C/01/86 VANKEE NUCLEAR POWER STAi!DN -
SUMMARY
0F EP TOPlc ill-6 IIRC 00ESTIONS SUEST10N RESOLUTION ISSUE SUlflARY SOURCte RESOLUT10N EFERENCES SCHEDULE ITEM 110.
F7b State what loads or E666t8.2) Individual load cases included in MajorEquip- 5/20/86 coabinations are being the coebined Icading are clearly sent Qualifi-I considered. stated in the Referente. cation Report, Rev. 1, dated 5/84.
F7c Provide sasple calcula- E6n6(8.3) Saeple calculations are available -
5/20/86 tions for revies, for review.
F7d Docusent adequacy of val- E666(8.4) Subject valve has been qualified. Ilajor Equip- 5/20/86 ve HCV-205. See the Reference, sent Qualifi-cation Report, Rev. 2 (preli-ainary), dated 5/65.
F7e Docusent adequacy of val- E6&6tS.5) Subject valves have been qualified. MajorEquip- 5/20/86 vesPR-MOV-191,PR-SV-181 See the Reference, sent Gualill-andPR-SV-182, cation Report, Rev. 2 (preli-ainary), dated 5/83.
F7f Confire the validity of E61618.6) Validity of all assumptions have Major Equip- 5/20/06 all assus;tions used in been verified during the Spring 1984 sent Gualifi-the evaluations, refueling outage, and calculations cation Report, have been revised accordingly. See Rev. 2 (preli-the Reference, ainary), dated 5/85.
F7g Confire the adequacy of ES&6(8.7) Perfors the required evaluation. -
5/20/86 all HSS valves on those I systees which have re-cently been reanalyzed.
F8 FEED AND BLEED HEAT EX-EHANGERS AND SUPPORTS F8a Clarify the results to E666(9.1) The reference report states that Major Equip- 5/20/86 I desanstrate that pressure stresses are included.
the pressure stresses are included. eent Qualifi-cation Report, Rev. 1, dated 5/84.
I 19
- - - . - . - - = ~~
^
PA6EIB C/01/86 I ._ . _ _ _ _ _ . .
YAK EE NUCLEAR PONER STATION - Su m ARY OF SEP topic 111-6 B C OUESi!ONS RESOLUTION DUESTION RESOLUTION REFDENCES SCHEDULE .
ITEM NO. ISSUE SUMARY S0IRCEe
- 3. N. Hologren
- 5/20/86 F86 Clarify whether any of E646(9.21 the reconeended support modifications have been laplemented, and if so, provide detailed inforea-tion on the sodifications.
6 ELECTRICAL AND OTHER MECHANICAL COMPONENTS E666(II.21, 6/24/86 61 Define the acuptance criteria and allowable NRC stresses used to deteral-ne the adequacy of elet-trical cabinets / equipment.
E6& Sill.31, 6/24/86 62 Provide information on the qualification crite- EC ria/sethodology for I MOV-557.
6/24/86 63 Provide information which E8& Sill.ll I will allon comparison of elevation of component with elevation of ARS utilized.
6/24/86 64 Identify the allouable E616(ll.41 stresses used for the I evaluation of the alter-nate feedsater line and MOV-557.
6/24/16 65 Esplain why the alternate E666(ll.5) Analysis under both YCS and NRC feedsater line and MOV-557 spectra loads is not necessary. See are not analyzed for toth ites A2 above, I the YCS and NRC spectra.
6/24/86 66 Clarify whether anchorage E616til.61 changes on the elettrical cabinet have been coaple-ted.
I 20
II C/01/16 YANKEE NUCLEAR POWER STATION - SLNIMARY E SEP TOPIC III-6 NRC QUESTIONS RESOLUTION 90EST10N SOURCEe RESOLUTION REFERENCES SCHEDULE ITEM NO. ISSUE
SUMMARY
N DEDICATED SAFE SHUTD0let I
SYSTEM (DSSS)
Hi IssueDSSSPipingand INtc B. W. Hologren 6/24/86 Structural Report I
e Numerals in parantheses following NCT correspond to footnote nusbers in Table 1 of the NCT Review Report. A (0) appears for itens such as C2g, C4l, and C5e because they do not appear in Table I but the issues are raised in the test.
Numerals in parantheses following E6LS correspond to E646 question numbers.
el Presentation to NRC, docusent dated 4/8/95.
I
)
1 21 !
l I
%4 ham _4 _a_*J ga.._m am-,_ e - - - 2 m + - _. -
~ ~ * * " *
- e e ng -mm- + MaA &, * .sm_,
i STRUCTURAL ISSUES l
l l
1 4
22 Yi:PS - RESOLUTI0f10F flRC OUESTI0f;S ON SEP TOPIC III-6 g
l I_--_--- _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ . _ _ _ _ . . . _ _ _ _
ITEM CIA ISSUE SUMf1ARY: PROVIDE CALCULATIONS FOR SHEAR STRESSES IN VC l COLUMNS AND BEAMS
REFERENCE:
CYGNA CALCULATION BINDER 8G0G4/1-F I
RESPONSE: THE MOST CRITICAL SHEAR STRESSES IN THE VC COLUf1NS AND BEAMS ARE SHOWN IN THE FOLLOWING SHEETS.
I l . CONCLUSION: THE SHEAR STRESSES IN THE VC COLUMNS AND BEAMS ARE LESS THAN ALLOWABLES.
I I
I I
I I
l I
I YMPS - RESOL'lTI0tl 0F fRC QUESTTC';S O'l Sr.P TOPIC III-G 23 I APRIL 8, 198G
r- . . - _ - . - _ _ - - _-- - . -
I l SHEAR STRESS IN CRITICAL VC COLllflS E .
E SHEAR STRESS (KSI)
LOAD CASE OR ELEENT NO. 33 ELEKNT NO. 47 SHEAR TORSION TOTAL SHEAR TORSION TOTAL g0AD C0lBINATION 0 0 0 0 . 0
. DEAD LOAD
-0.3 0 -0.3 0.2 -0.01 0.21
. A T 199*F (SHELL)
- 0.6 *0.01 20.06 20.5 20.52 *0.52
- p. A T=2 70*F (ENVIR)
O C.01 0.01 0 0.02 0.02
- 4. 31.5 PSI PRESSURE FLOODING 0 0.01 0.01 0 0.01 0.01
. 3-D YCS 0.6 0.07 0.7 1.1 0.05 1.2
- 7. 3-D NRC 1.1 0.15 1.3 2.2 0.10 2.3
-0.9 0.03 -0.9 0.7 0.1 0.8 IACCIDENT
- 8. 1+2 2 3+4+5LOAD CASE 1.2 0.1 1.3 1.6 0.1 1.7 ISEISMIC
- 9. = 1 + 3LOAD
+6 CASE-YCS 1.7 0.2 1.9 2.7 0.1 2.8 lSEISMICLOADCASE-NRC
- 10. = 1 + 3 + 7 _ , , _ _
ALLOWABLE SHEAR STRESS:
FOR ACCIDENT LOAD CASE, F y = 0.4 Fy = 0.4 (30) = 12 KSI
= 16 KSI lFORSEISMICLOADCASES,F y
({)0.4F y I
I I
24 YNPS - RESOLtRION OF NRC OUESTIONS ON SEP TOPIC III-6 APRIL 8, 1986
v -._ :- - .- .- _. .
I SHEAR STRESS IN CRITICAL W TIE BEAtB B
E SHEAR STRESS (KSI)
LOAD CASE OR ELEMENT NO. 49 ELEMENT NO. 104 SHEAR TORSION TOTAL SHEAR TORSION TOTAL l LOAD COMBItMTION DEAD LOAD 0.09 0 0.09 0.10 -0.001 0.10 g1.
- 2. 0 -0.001 -0.001 -0.006 -0.009 -0.02 A T=199*F (SHELL) 20.01 20.006 20.02 10.002 20.06 20.06 l3. AT=*70*F(ENVIR)
- 4. 31.5 PSI PRESSURE 0.01 0.001 0.01 -0.014 0.02 0.01
- 5. FLOODING 0.09 0.03 0.12 0.02 -0.006 0.01 G. 3-D YCS 0.58 0.16 0.74 0.64 0.15 0.79 1.14 0.31 1.45 1.27 0.29 1.56
- 7. 3-D NRC 0.20 0.04 0.24 0.10 0.06 0.16 IACCIDENTLOADCASE 8 = 1+2 2 3 + 4 + 5 0.68 0.17 0.85 0.74 0.21 0.95 ISEISMICLOADCASE-YCS 0= 1 + 3 +6 SEIS!ilC LOAD CASE-flRC 1.24 0.32 1.56 1.37 0.35 1.72 I10= 1 + 3 +7 _ . - . ..
lALLOWABLESHEARSTRESS:
y = 0.4 Fy = 0.4 (30) = 12 KSI gFORACCIDENTLOADCASE,F FOR SEIStiIC LOAD CASES, F y =({)C.4Fy = 16 KSI I
I 23 YNPS - RESOLUTION OF NRC OUESTI0flS ON SEP TOPIC III-6 APRIL 8, 1986 I
l _
I ITEM C1B ISSUE
SUMMARY
- PROVIDE CALCULATIONS FOR PULL-00T OF THE VC ,
ANCHOR BOLTS l
REFERENCE:
CYGNA CALCULATION BINDER 8G004/1-F
RESPONSE
SUMMARY
OF THE VC ANCHOR BOLT PULL-0VT CALCULATIONS:
EMBEDMENT LENGTH = 25.7" (SEE FIGURE IN NEXT SHEET)
BOLT SPACING = 18.85" I TOT AL SilE AR CONE ARE A PER BOLT = 502.1 IN 2
I PULL-00T CAPACITY PER BOLT = 6 4 / F'C A C
592.1 110,260 LBS
0.85 x 4x /3000 x l YIELD CAPACITY OF 1-1/4" 6 BOLT (F y 39,300 LBS 32 KSI)
I RUPTURE CAPACITY OF 1-1/4" 6 BOLT (F U = 50 KSI)
=
61,500 LBS g
Ut: DER D + (A T = 70 F) + YCS, REQUIRED ANCHOR BOLT TENSION = 32,300 LBS I , UNDER D + ( AT = 70 F) + NRC, REQUIRED ANCHOR BOLT TENSION = 43,100 LBS I Cot CLUSI'Jf': THE PULL-00T CAPACITY OF Tile VC ANC40c BOLTS IS ADE0VATE 26 YNPS - RECOLUTIO!J OF NRC QUESTI0f;S 'Il SEP TOPIC III-G I APRIL 8, 193G
l I I
~
I PIDESTAL ,
1 3" ,
e 24" n :
I . g 8-l\" / BOLTS,
'T \
y
[*
2'-10" IDG WI'n!
3" 6.42" I
i a 8-6"x6"x1/2" !
- /.
l PIATE WASHERS V "r '
6" 18.85" d '
.L 4 L
I "
6" ^4]
A,
^1 _ .sg ,
g r a . R3 L2
[ su i
~
l
~ '
t
%^2'l q'/ />"
ANCHOR BOLT IDCATION 24 l
l/
45* ,p I SHEAR CO!E AREAS = Ay + A2 + A3 +A4 I 1 I .
VC ANCOR BOLT EVALUATION g
I 27 YNPS - RESCLUTI0rl 0F NRC OUESTIONS ON SEP TOPIC III-6 l APRIL 8. 198G
^
I ITEM CiE I ISSUE
SUMMARY
- PROVIDE Tile EVALUATION RESULTS FOR THE CLEVISES AND TURNBUCKLES ON THE VC DIAGONAL TIE RODS
REFERENCE:
CYGNA CALCULATION BINDER 86064/1-E g
RESPONSE
THE ALL0t/ARLE TENSION CAPACITY OF CLEVIS = 413 KIPS l THE ALLOWABLE TENSION CAPACITY OF TURNBUCKLE = 816 KIPS l UNDER YCS, MAXIMUM DIAGONAL TIE R00 TENSION = 117 KIPS UNDER NRC, MAXIMUM DIAGONAL TIE Roo TENSION = 228 KIPS I _ CONCLUSION: THE CLEVISES AND TURNBUCKLES ON THE VC DIAGONAL TIE R0D ARE ADEQUATE.
I I
I I
I I
I ,
, I I 28 YNPS - RESOLUTION Of tRC QUESTIONS ON CEP TOPIC III-G l APRIL 8, 1986
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Pia 1& me AL . A..t n acs
- e. *..es p ria . To et We**'9*stp l spettau t s p . . t . eit t penC E S I ort a v-CLEARANCE BETWEEN VC AND RSS AND BETWEEN VC AND RADI0 ACTIVE PIPE TUNNEL I 30 YNPS - RESOL lHION Of fEC QUESTIONS ON SEP TOPIC III-G l APRIL 8, 19D3
~.. . .
I ITEM C2c ISSUE
SUMMARY
- PROVIDE CALCULATIONS FOR THE RSS ANCHOR BOLTS I
REFERENCE:
CYGNA CALCULATION BINDER 86064/1-f
RESPONSE
SUMMARY
OF THE RSS ANCHOR BOLT CALCULATIONS:
l COLLAR ANCHOR BOLT FOR RSS EXTERIOR COLUMNS:
l 1-7/8" e WILLIAM SUPER HIGH TENSILE ROCK BOLT g EMBEDMENT LENGTH = 60" (SEE FIGURE IN NEXT SHEET)
BOLT SPACING VARIES FROM 14.5" To 21.7".
TOTAL SHEAR CONE AREA PER BOLT = 1045.8 IN2 (FOR 14.5" I SPACING) l MINIMUMPULL-00TCAPACITYPERCOLT=64dF C 'A C
= 0.85 x 4x 43000 x 1045.8 194.8 KIPS g
UflDER C + YCS. REQUIRED ANCHOR COLT TENSION = S0.2 NIPS Ur: DER D + NRC REQUIRED ANCHOR BOLT TENSION = 1C2.8 KIPS LONCLUS10'i: THE PULL-0UT CAPACITY OF THE RSS ANCHOR SCLTS IS l , ADEQUATE.
I I 31 YNPS - RESOLUTION Of fRC CUESTI0tG ON SEP TOPIC III-G APRIL 8. 1986
m I
l-7/8"p WILLIAM SUPER 11IG1 I TD4SILE NOCK BOLT l
l f:S ,s p
. i m .
G ^
l
" i 1.875" I
is" g 14. s" _
i I A \" eass nun townn l ,
~
~'s
's f ul N' COLT 3N PEDESTAL l
w /'
' 'l 60" A
2 -
I sN 8 l
_simAn corm AnrA9 l ..
g RSS COLLAR ANCil0R BOLT EVALUATION I 32 YMPS - RESOLUTION OF NRC OUESTIONS ON SEP TOPIC III-G l APRIL 8. 198G
=-- . . . . _ . -- ._
ITEM C2E ISSUE
SUMMARY
- CLARIFY THE SEISMIC LOADS APPLIED TO FINITE ELEMENT MODEL 0F STEEL COLLAR IN RSS EXTERIOR COLUMNS
REFERENCE:
CYGNA CALCULATION BINDER 80023/1-F
RESPONSE
APPLIED SEISMIC FORCES ARE AS FOLLOWS:
NODE FZ IKI FY (K) l NO. _
(VERT.) (HORIZ.) REMARKS _j 52 121.8 0 " ? " <*' to ou .
54 121.8 0 ,%- " " " " " " " " " " * '
56 121.8 0 % -
"' "^" " ' '
58 ,
121*8 0 D I
g "RCCOLA" COST;TD CLL 60 1 121.8 0 h 62 121.8 0 'z 'C '"# ' " "-"
64 121.8 0 "'"2
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66 103.9 0 % U l Ta d F TkANSFEk6 Ts F0c*1s g
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YNPS - RESOLUTION Of fRC QUESTIONS ON SEP TOPIC III-G 33 APRIL 8. 198G
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YNPS - RESOL lJTIori 0F TEC OUESTIONS ON SEP TOPIC III-G 34 APRIL 8. 1 3
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35 YMPS - RESOLUTION Of fRC CUESTI0fS Oil SEP TOPIC III-G APRIL 8. 1936
I ITEM C2F ISSUE SUM'1ARY: CLARIFY THE DEFINITION OF "YIELO" 0F THE RSS l RING FOUNDATION IN RELATION T0 " ULTIMATE" l REFEREtlCE: CYGflA COMPUTER OUTPUT 81060/71.1.F
RESPONSE
g THE DEFINITI0fl 0F " YIELD" AND " ULTIMATE" M0f1ENT CAPACITY OF THE RSS RING FOUNDATION IS SHOWN BELOW AND IN NEXT TWO SHEETS l POSITIVE fl0 MENT CAPACITIES:
l YIELD fl0l1ENT = 12C40 FT-K (SEE FIGURE IN NEXT SHEET)
ULTIMATE MOMENT = 16170 FT-K I fl0i1ENT DEVELOPED AT PLASTIC HINGE = 12632 FT-K (UNDER D + NRC)
PLASTIC HINGE ROTATION = 0.000193 RAD.
l CURVATURE DUCTILITY DEMAND = 1.1 I flEG ATIVE M0f1FNT C AP ACITIES:
I YIELO M0!1ErlT = 18100 FT-K (SEE FIGURE IN NEXT SHEET)
ULTIllATE Il0 MENT = 23740 FT-K MAXIMull M0tlEtli = 14431 FT-K
.. (fl0 YIELDIflG UNDER D + NRC)
I I
h Y'lPS - R'TDLUTI0f1 Of tRC QUESTI0flS Ofl SEP TOPIC III-G I APRIL 8, 1936 I
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(YIELD)
NEGATIVE MOMENT 16170- Q1RT_E)_
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POSITIVE D) MOMENT 10000 -
(ULTIMATE MOMENT CAPACITY IS LIMITED BY 0.004 CONCRETE STRAIN.)
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I I e 1.J 2.0 3.0 4.0 5.0 6.0 CURVATURE ( X10-4 in/in )
I ..
MOMENT - CURVATURE DIAGRAMS OF RSS RING FOUNDATION g
I 37 l
YNPS - RESOLUTI0fl 0F fRC OUESTIONS Ofl SEP TOPIC III-G l APRIL 8. 198G
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.00106 .004 .006 CONCRETE STRAIN ( in/in )
STRESS - STRAIN CURVES OF REBAR AND CONCRETE USED IN GENE l OF M0f1ENT - CURVATURE DIAGRAMS OF RSS RING FOUNDATION I 38 YNPS - RESOLUTION OF tRC OUESTIONS ON SEP TOPIC III-G I APRIL 8. 193G
i ITEM C3A ,
JSSUE
SUMMARY
- TURBINE PEDESTAL DAMPING VALUE SHOULD BE 3% TO l 5% INSTEAD OF 7%
REFERENCE:
CYGNA CALCULATION BINDER 86064/2-F g
THE TURBINE PEDESTAL MODEL WAS REVISED TO I RESPONSE:
INCLUDE THE S0IL SPRINGS. THE EFFECTIVE MODAL DAMPINGS WERE CALCULATED USING 5% CRITICAL DAMPING RATIO FOR STRUCTURAL ELEMENTS AND 75% OF THE THEORETICAL CRITICAL DAMPING RATIO FOR S0IL l SPRINGS (PER " GUIDELINES FOR SEP S0IL STRUCTURE INTERACTION REVIEUS", NRC LETTER LS05-80-12-035, g 12/15/80).
AMPLIFIED RESPONSE SPECTRA (ARS) WERE GENERATED I FOR N0DE 70 0F THE REVISED TURBINE PEDESTAL MODEL. THESE ARS ARE GENERALLY B0UNDED BY THE ENVELOPED ARS USED IN THE ANALYSIS OF THE MAIN STEAM LINES INSIDE THE TURBINE BUILDING (PROBLEM l No.009).
l CONCLUSION: THE ANALYSIS OF THE MAIN STEAM LINES USING THE EXISTING ENVELOPED ARS ARE ACCEPTABLE.
I I
I .
I I
YNPS - RESOLUTION OF tRC OUESTIONS ON SEP TOPIC III-G 39 l APRIL 8. 1986
e . 20 -
- TURBINE PEDESTAL: YCS INPUT !
I, MS NOZ:LE: N-S DIRECTION .' l
- ARS WITH SSI e****', p --- - i
. BROADENED ARS WITHOUT SSI- - -- l e
. t ENVELOPED ARS IN MS ANALYSIS ------- 8 8
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YMPS - RESOLUTION Cf tRC QUESTIONS ON SEP TOPIC III-G ao APRIL 8, 1936 4
I 1.so _ _-
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TURBINE PEDESTAL . _ _YCS INPUT.. . _ _ _ _
MS NOZZLE:
? Ahs WITH SSI E-W DIRECTION
- BROADENED ARS WITHOUT SSI- -- r-" - .
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.-- - .~ - . . - - - - . .- _ . -- .__ . - . . . _ _ . __ _ I l I 0. 9 e . . _ . . . _
- TURBINE PEDESTAL: YCS INPUT
- MS NOZZLE: VERTICAL DIRECTION 8
. .i *
- n I i
ARS WITH SSI g }
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! ENVELOPED ARS IN MS ANALYSIS ------- i i 0.90 . .
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- 50 20 10 F R E Q U E N C 't ICPSI PROGRAM INSPEC CTCNA ENERCT SERV!CES I 42 YNPS - RESOLUTION OF tRC OUESTIONS ON SEP TOPIC III-6 l APRIL 8, 1986
L T T :~^::_ _ T ._ -- - - - - . - . - - . 1 I I ITEM C3B i ISSUE
SUMMARY
- TURBINE BUILDING MODAL MASSES 00 NOT ADD UP TO l 90% OF THE TOTAL MASS g
REFERENCE:
[1] " TURBINE BUILDING STRUCTURAL ANALYSIS REPORT", REV. 1. REPORT NO. EY-YR-80023-9, I CYGNA ENERGY SERVICES, DECEMBER 1982. RESPONSE: THE UNITS OF THE MODAL MASSES GIVEN ON TABLE D.1 0F THE TURBINE BUILDING REPORT [1] SHOULD BE K-SEC 2 /IN. THEREFORE THE TOTAL MODAL MASSES IN THE X AND Y DIRECTIONS, 25.6 K-l SEC2 /IN AND 27.1 K-SEC 2/IN, RESPECTIVELY, ARE LARGER THAN 90% OF THE TOTAL MASS, 330 2 2 K-SEC /FT (27.5 K-SEC /IN). CONCLUSION: TURBINE BUILDING MODAL MASSES ADD UP TO MORE THAN 90% OF THE TOTAL MASS. I I I I I YNPS - RESOLUTION OF WC OUESTIONS ON SEP TOPIC III-6 a3 l APRIL 8. 1986
-c __ _ _ __
1 I ITEM C3D REVIEW ISSUE: CLARIFY IF THE OFFICE BUILDING IS SUPPORTED BY THE TURBINE BUILDING IN ONLY ONE OR IN BOTH HORIZONTAL DIRECTIONS I
REFERENCE:
CYGNA CALCULATION BINDER 80023/4-F RESP 0tlSE: l THE OFFICE BUILDING IS RESTRAINED BY THE TURBINE BUILDING ONLY IN THE E-W DIRECTION AS SHOWN IN THE NEXT TWO SHEETS. I THE ROOF BEAMS OF THE OFFICE BUILDING ARE CONNECTED TO THE GIRDERS OF THE TURBINE BUILDING IN ELEVATION 1037'-8" ALONG COLUMN LINE C AND BETWEEN COLUMNS LINES 10 AND 13. I SINCE THERE IS NO FLOOR SLAB ATTACHED TO COLUMN LINE C IN ELEVATION 1037'-8". THE N-S STIFFNESS OF THESE GIRDERS IS Sf1ALL WHEN COMPARED WITH THE N-S STIFFNESS OF THE OFFICE BUILDING. THEREFORE. THE OFFICE BUILDING IS NOT RESTRAINED BY THE TURBINE BUILDING IN THE N-S DIRECTION. CONCLUSION: THE OFFICE BUILDING IS MODELED PROPERLY. g I I I - I I .; .; YNPS - RESOLUTION OF FRC QUESTIONS ON SEP TOPIC III-6 l APRIL 8. 1986
I l I
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.THE TUREINE BLDG lONLYINE-K
~'REO:I E =0;; ;' ' DIRECTION I
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.5 YNPS - RESOLlRION OF tRC OUESTIONS ON SEP TOPIC III-6 l APRIL 8, 1986
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46 YNPS - RESOLUTION 0F IRC QUESTIONS ON SEP TOPIC III-6 APRIL 8, 1986 g
I j I ITEM C3F I ISSUE
SUMMARY
- CLARIFY HOW ADDITION OF BRACES WILL ALLEVIATE l SEISMIC UPLIFT RESPONSE: ADDITIONAL DIAGONAL BRACES INSTALLED IN A g
DIFFERENT BAY THAN THE EXISTING BRACE CAN SIGNIFICANTLY REDUCE THE COLUMN UPLIFT FORCE AS ' I SHOWN IN THE NEXT SHEET. I I I I I I . I I 1 l -
= - .t - <
I YNPS - RESOLUTION OF fEC OUESTI0fS ON SEP TOPIC III-G 47 l APRIL 8, 1986
~
1_ _ _ -_ .__ . _ . _ __ __ l I I 3 0 20'-0" = 60'-0" _ i I 10 K i i o i [( l , DIAG. BRACE
.l
~
o N c- r== 10 K : n HA ONNECTION ( M .) 10 K 10 K I (a) FRAME WITH ONE DIAGONAL BRACE MAXIMUM COLUMN PULL-OUT FORCE = 10 kips I ; 10 K
/ \
/ N I ./ \
N, 5 K *g g ~5K e p 5K 5K 5x 5K (b) FRAME WITH TWO DIAGONAL BRACES I MAXIMUM COLUMN PULL-OUT FORCE = 5 kips l ALLEVIATION OF COLUMN UPLIFT FORCE USING ADDITIONAL DIAGONAL BRACES I 48 YNPS - RESOL lJIION OF tEC OUESTIONS ON SEP TOPIC III-6 l APRIL 8, 1986
m _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ I ITEMS C5A. C5F. AND CSI l ISSUE
SUMMARY
l CSA. DAMPING VALUE SHOULD BE 3 TO 5% INSTEAD OF 7% IN ANALYZING THE PAB. I C5F. S0IL STRUCTURE INTERACTION (SSI) EFFECTS MAY BE SIGNIFICANT IN THE VERTICAL DIRECTION ANALYSIS OF THE PAB. C51. CLARIFY WHY THE VERTICAL MODEL SHOWS THE RADI0 ACTIVE PIPE l TUNNEL TO BE UNC00 PLED FROM THE PAB AT THE ROCF. E
REFERENCE:
[1] CYGNA CALCULATION BINDER 86064/2-F [2] " GUIDELINES FOR SEP S0IL STRUCTURE INTERACTION REVIEW", NRC LETTER LS05-80-12-035 DATED I DECEMBER 15, 1980. _ RESPONSE: l THE VERTICAL PAB MODEL WAS REVISED TO INCLUDE THE S0IL SPRINGS AND COUPLING BETWEEN THE RADI0 ACTIVE PIPE TUNNEL AND THE PAB (SEE NEXT l SHEET'. THE EFFECTIVE MODAL DAl1 PINGS WERE CALCULATED USING A 5% CRITICAL I DAMPING RATIO FOR STRUCTURAL ELEMENTS AND 75% OF THE THEORETICAL CRITICAL DAMPING RATIO FOR S0IL ELEMENTS [2]. THE Ar1PLIFIED RESPONSE SPECTRA (ARS) WERE GENERATED FOR NODES 6 AND 11 0F THE REVISED VERTICAL PAB MODEL (SEE NEXT 3 SHEETS). THESE ARS ARE GENERALLY BOUNDED BY THE BROADENED ARS GENERATED FOR TILE ORIGINAL PAB MODEL. E0tlCLUSI0ti: THE BROADENED ARS FOR THE ORIGINAL PAB MODEL ARE STILL VALID. 49 YNPS - RESOLUTION OF IRC OUESTI0tlS ON SEP TOPIC III-G APRIL 8, 1986
I ! 1 I , 186" _ ;_105.6" 105.6" _ 105.6" Steel Frame (typ.) I '
.J 2
,11 s
.,10 Rad. Tunnelj -
9 2, 1 ti 2 il 3 t> 4 4>
- Rigid- Ns n
c-s 2nd Story
# *~
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fl13
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- Columns W i 7 6
N :
$ 1st Story ;.
.m
" RC Walls -
m I a 14 e 7 , h - Soil Spring r$ '16 17 , , 15 8 Node Number (typ.) i l l ' REVISED PAB VERTICAL MODEL WITH SOIL SPRINGS l t 50 YiJPS - RESOLUTI0f10F tRC QUESTIONS ON SEP TOPIC III-G APRIL 8. 1986
l 1 1 l SPECTRA 2.0 Tue. as nris isos 08:(8:39 ;
. 3, PAB YCS INPUT ARS WITil SSI FIRST FLOOR (NODE 6) BROADENED ARS WITilOUT SSI --- -
VERTICAL DIRECTION 4.30 -;
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m
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i t ;!< s, 3, n i s..., i, i.,, e.io ;. ; ., i
>. i. ,
i q ~
,if c.0a 5. ti . 10. 20. 3J. %J. 10 C. 10J.
L.I 0.2 0.3 0.1 G. 5 0. t. 4.0 2.0 3.0 4 FREQUENC1 IHil CTGNR ENERG1 SERVlfE5
/
51 YNPS - RESOLUTION OF tRC OUESTIONS ON SEP TOPIC III-G APRIL 8. 1986
Tue. as non i m os o a: v4 SPECTRA 2.0 PAB YCS INPUT ARS WITH SSI RADIOACTIVE TUNNEL (NODE 11) BROADENED ARS WITilOUT SSI - - - I t. 80 VERTICAL DIRECTION j e o 70 ! ! 8 i i G 0. 60 e t
! I 5
i !
$ . i O
+
l i
- i E u.%0 l ,i i -- - s i i* >l ti "u DAMPING RATIO ,
';i i{i;.
g 9. % ('~1 , 2g 8 I - -. j n' m 5% , f. ,1
- .-~~ ' .
I 0.20 -- j , ,t j . . . - j -si' {
- v /\iw 0 la l
- 0. 0u 3.0 4. 5. b. lo. 20 30. sa. W. 6. tJU.
L4 0.2 0.3 0.4 0. 5 0. t. l.0 2.0 FRE00 ENC 1 tilli l l C1GNH F N L B C.1 SERvlit5 I . I 52 YNPS - RESOLUTIQU (T NRC QUESTIONS ON SEP TOPIC III-G APRIL 8, 1986 I
= _ __ : - - . .-. - ... L g l ITEM CSC l ISSUE
SUMMARY
- MODELING OF WALLS IN THE PAB BY EQUIVALENT COLUMNS SHOULD BE JUSTIFIED.
I
REFERENCE:
CYGNA CALCULATION BINDER 80023/3-F RESPONSE: THE STIFFNESS OF THE WALLS IS EQUAL TO THE TOTAL STIFFNESS OF THE EQUIVALENT COLUMNS USED IN THE I PAB !10 DEL (SEE NEXT 2 SHEETS). ADDITIONALLY, THESE WALLS ARE VERY RIGID AND ANY MODELING ERROR WILL HAVE INSIGNIFICANT EFFECT ON STRUCTURAL RESPONSE. I CONCLUSION: THE STIFFNESS OF THE PAB WALLS IS MODELED PROPERLY. I I I I I ) I , I 53 YNPS - RESOLUTION OF TEC CUESTIONS ON SEP TOPIC III-6 i APRIL 8, 1986 l
]
I - I L l CROSS-SECTIONAL AREA, A = LT h THICKNESS" T 1 3 M MEHT 0F 1NERTIA, I = 77 TL SECTIONAL PROPERTIES OF WALL , i 1 l I RIGID LINK (TYP,) I'3j ['l j Y_ f 2j Ly+L2 +'3 * ' TOTAL AREA y= A gg = LyTA 2 74 3 I * 'I"A I h 2 1, 1 z 2, 2 2 3, 3 TOTAL .O S H OF INERTIA = l +1 +1y23 = I I I I THICKtESS = T I SECTIONAL PROPERTIES OF EQUIVALENT COLUMNS I 54 YNPS - RESOLtJTION OF NRC OUESTIONS ON SEP TOPIC III-G APRIL 8, 1986 I
~__. .__. .___ . _.. _l'i ___ _ _ __ __ . . _ _ _ __ _ _ - - I i I o
=
96" I _ _ 0 ,, '
~
* ' (6t) 6 4 in
_, , , , g I=26x10 SHEAR AREA =1440 in
=, .
J
- I u.A. ,
6 4
. O- ._o I=51x10 in d SHEAR AREA =2880 in 2
16"--- $ 6 I=25x10 in
. O " 3 2 J
I -- { n SHEAR AREA =1440 in WALL EQUIVALENT COLUMNS 6 6 4 I=102x10 in I=102x10 in 4 SHEAR AREA =5760 in SHEAR AREA =5760 in MODELING OF WALL AT PAB COLUMN LINE 8 ( 55 YNPS - RESOLUTION OF NRC OUESTIONS ON SEP TOPIC III-6 APRIL 8, 1986 I
= . _ __
I ITEM CSD l ISSUE
SUMMARY
- PROVIDE EVALUATION OF POTENTIAL SEISMIC UPLIFT AT COLUMN LINES 8 AND FB 0F THE PAB.
I
REFERENCE:
CYGNA CALCULATION BINDER 8606ti/1-F I RESPONSE: THE
SUMMARY
OF THE EVALUATION OF UPLIFT AT COLUMN LINES 8 ANDCE , 8 AND Fg , AND 8 AND G IS I SHOWN IN THE NEXT SHEET. THE NET UPLIFT LOADS AT THESE LOCATIONS ARE LESS THAN THE ANCHOR BOLT CAPACITY AND THE TRIBUTARY WEIGHT OF THE STRUCTURAL ELEMENTS. CONCLUSION: THERE IS N0 UPLIFT PROBLEM AT THESE LOCATIONS. I I I I I I I , I 56 YNPS - RESOLUTION OF NRC OUESTIONS ON SEP TOPIC III-G APRIL 8, 1986 I
= - -
_ _ . , --;--------------------- 7- --- --- ----- ---- - --------------- --- - ---------- y
SUMMARY
OF UPLIFT EVALUATION FOR PAB COLUMN LINE 8 G G @ 2-L4x3x5/16 15'-0" 15'-0" (EXIST) a . l l l I k" [ o r 2-L4x3x5/16 (NEW) e i 2'-0'
*
- 1'-2" E
~
9 8.84 e -- -- - w 7.45 ( J6 26 n j SRSS OF TWO HORIZ. YCS INPUT 8.84 1. 0 10. 4 4 4 4 D - VERTICAL YCS % y,47 K 1.47 2.59 (COLUMN-TO-BEAM CONNECTION) DIRECTLY CONNECTED D - VERTICAL YCS . 5 K 33.10 "K TO RC WALL (BEAM-TO-WALL CONNECTION) EVALUATION OF CONNECTION (a) COLUMN-TO-BEAM CONNECTION @ COLUMN LINES 2-7/8"/ A325 BOLTS, T a 35.2 kips per BOLT l TOTAL ALLOWABLE TENSION = 70.4 kips > (8. 8 4-1. 4 7) =7. 3 5 kips (OK) (b) COLUMN-TO-WALLCONNECTION@COLUMNLINESh&h 2-7/8"p ANCHOR BOLTS (EXIST), Ta =16 kips per BOLT AND 3-1\"/ HILTI KWIK-BOLT (NEW), T,=8.46 k & V a=9.44 k per BOLT MAXIMUM FORCE ON ANCHOR BOLT = 8.67 kips <16 kips (OK)
; MAXIMUM INTERACTION COEFFICIENT FOR HILTI BOLT = 0. 58 < 1. 0 (OK)
I I YNPS - RESOLUTION OF NRC QUESTIONS ON SEP TOPIC III-G APRIL 8, 1986 57
~ ' - := : - _ _ _ _ _ _ _ _ I ITEM C50 ISSUE
SUMMARY
- PROVIDE THE PROPERTIES OF THE " DUMMY" I SUBDIAPHRAGM AND EXPLAIN WHY THIS SUBDIAPHRAGM DOES NOT APPEAR IN THE FIGURES ILLUSTRATING THE l PAB MODEL
REFERENCE:
CYGNA CALCULATION BINDER 80023/3-F g THE COLUMNS OF THE RADI0 ACTIVE PIPE TUNNEL I RESPONSE: EXTEND FROM GROUND TO LEVEL 2 AND ARE NOT ATTACHED TO LEVEL 1 FLOOR SLAB 0F THE PAB. TO REFLECT THIS SITUATION IN THE BATS COMPUTER MODEL, THE MASSLESS " DUMMY" SUBDIAPHRAGM WAS USED FOR THE LEVEL 1 0F THE RADI0 ACTIVE PIPE TUNNEL FRAME. THIS DUMMY SUBDIAPHRAGM IS NOT CONNECTED TO THE LEVEL 1 MAIN DIAPHRAGM. CONCLUSION: THE DUMMY SUBDIAPHRAGM IS MODELED PROPERLY. I I I I I . I I 58 I YNPS - RESOLUTION OF tPC OUESTIONS ON SEP TOPIC III-G APRIL 8, 198G
I I . I FLEX 1BLE LINKS (TYP. ) STEEL ROOF
/ LEVEL 2 (TYP )
- sol SD2 t%IN sD4[ j sD3 DIAPHRAGM I
- STEEL COLttf4S f
RADIO. & BRACES (TYP.) TUNNEL
/ t
__/ MAIN LEVEL 1 DIAPHRAGM i ! I EQUIVALEtTT _
~
E MMY DIAPHRAGM WHICH HAS NO MASS AND IS DISCONNECTED FROM THE MAIN IAPHRAGM IN LE M 1 COLtit4S /' - l __/
- TUNNEL SUoPORTING COLLrts I ..
I I - HORIZONTAL STICK MODEL OF PAB I 59 YNPS - RESOLUTION OF tRC QUESTIONS ON SEP TOPIC III-G I APRIL 8, 1986 l
__ _ . . - _ _ _ _ . _ _ ~~.
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=
YNPS - RESOLUTION OF NRC OUESTIONS ON SEP TOPIC III-G 60 APRIL 8, 1986
n ~: -
- __:2r2 = _ _ . _- ---- = = _ __.
I I ITEM CSK I . ISSUE
SUMMARY
- COMPLETE TABLES E-4 AND E-10 0F THE REFERENCE
REFERENCE:
"PAB AND RADI0 ACTIVE PIPE TUNNEL STRUCTURAL l ANALYSIS REPORT", REV. 2, REPORT NO. EY-YR-80023-7, CYGNA. JANUARY 1983 RESPONSE: THE COMPLETED TABLES E-4 AND E-10 0F THE REFERENCE ARE SHOWN IN THE NEXT TWO SHEETS.
I I I I I I I I I I 61 I YNPS - RESOLUTION OF NRC OUESTIONS ON SEP TOPIC III-G APRIL 8, 1986
I 8A 4L* em C e Er - == W N N e e ** N == e m e.s
.C m eo ac ed g gl 9 b +3 C e ** e== cC e e e
=
e C e == e . ="e C. e to CC C C C C C C C C C C C C G .==
- e--
.D . en en gg= c @ N @ c e Ci e 80 g g at s N. CP. N.
3 ==.= N. ( f N. >=. A w N N
.O- oOed L 85 we .N== .N = me - == .N
-s
- to 6 eC e
se .== m N N e C CC C A C N ===
.- en en se og gy N. ED. r. Cn. C. N. C. w N w e e . C. N.
*e L C .e .= C e.= == - .-e .= w C er a LS e- eC to 68 2 L m.
k aC em C T - te m m ce gC C e en ** e - C. w. w C. ew o en . . e o e . . . .
** - L u so C 0 - i .,,.
X. C C C C C C C C C C C N N est to C C I W se o e Er 6C w CW C *** ** w W C 2 em Le > g ga en a.h er. . s%. e . m. . . . . . ce.
- T 3:r oc - C m C C N C C C C C
.. .w e LA .cc.
N w C M W I
- h y
y
.J w
w J 6 G UC Lw C eC.* C W e Ce mo e m cc. . . e er. . e e e . ee-e.
- > C C m C C C C C C C C
,E_ =. .C.C.
i === 2 { #f C
.2 et C
v C C" C* - == J W m. N. m. et-w i
. . . . e e . . e g w M ** C *E* C C N C C C C C N C
*- se N to C.
E@ e 2 CC >
== u - g; T T CK W 8e -
v CC Cr C
*- W a C % ge-N
- CC *CB CP er.
C C C C C C C % C au-N
- ar *C ==*
e .v.. 6
.e c e
6 C' T "e- * * * * * *
- v m C 8C .w C - C C === = - .= == - - N c >. N
.J C L === - s. C
*C C a #t C
- == Ci-W C .v- == - gg a
@N N O N F st >.
R* CC* == j s e== ura .N=* a== .N 6- a Cs n *3 * * *
- w ar.; p-f b C == CB @ 4E* Ch Ch CP at n.J er == >. ..
eC . g. 6. c se #") 8C 8C .C 8e 8e .C U 4 *EN I C 6 W C w w w C h a.J C a c .- c se e. C et et et et *C et eC et et et et oc >= C. CO N M M er *E* 4t* W N N N CC CC , eA se se em te se en eA e se se en =" 4 g
.C.
C C C C k q, . N
=
*b se J J .J 2 .J .,2 . ,J J .l .J mo . ,2 !E e i l' ,'w~~
e- :: y . . . . . . . . . . . . 'l,, r := C = J e a=* - e e C C C C C C C C C C C C p - .- e e = w lD m" ,pm
,* E v v v v v v v v v v v v =
OE., m 4 E YNPS - RESOLUTIQU OF tRC CUESTIONS ON SEP TOPIC III-G 62 APRIL 8e 19SG
* .C eO N m e A Ce - .= N *= te m e 0- - == 6.J La at g C. . . O. C. . ==. . ==. .=. C. m.
se6 C C C C C C C C C C C C m cr L
&# *==
e as-2 *ee e e e N e e e e C. W'i W j s N. N. N. N. yg ew ==. N. 3 7 Cr. N N e N N
,C-awC Le a === .N = a== .N = . -
- to k eC e
e == @ N N e C m C C A C A h
--ee g g eg N. C. C. &. C. e.
w
. e. N. e. C. N.
*-Lw C === N C *= == == == == - == c k
C ww s- C m eJ 6 C M e A e O 'C e @ CD Lf t m ewe- O* C* air' e m e e as Ce * * * * * * * * * *
-L as =J # C C ** C C C C C C C C e wap
{ gg CeCC N v w
- c.
- m y 6 C' g CW 2 *
- CC " 8D *'"
e /> C:. c, e v c. . . . . . . . . . m.
" ~ 'U C *C N C Ch C C e e C C C C -
seV e x C
===
Q g wMC eL I UC W Q LW C C N e e e m W
.,,o ==3 LC a
w
, N. .
* . . . N. N. . . . . m.
E C y -
= e>
> C C C C C C C C C C C -
e
>- W -D d wC aC C e N C O a w m. M. C. C. . . . CF.
, . . . . y C x N C N C C e e C C C C M c y
e N
.f vB _F ,' ,-
m T TE e s,1 - E C C - . . .
** e-e C A e e C C C C A C N & w =N c; .g -,y tw
. Cr. Cr. CP. C. C. C. e. N. e. C.
==
s. c m d g
= cs Ctw C === C C .= ="" == == *= *= '"' f%: a > fs
-6C C C *****
.- CO 6 g i d - === Cr WN N C N F u>
C
, em C * == *j % == c .N= == = .N E= i
* * * *
- aC>
y ai n L c .= CP 3 f Cr at- C* Ch C* at w C= ow e - t L t wm s.. CC R & K R & U Z s E rs C b- W c w w baJ c 6 w C 6- c-C ehC et et *C et et et *C et et *C *C se > C. CO N M f*) ac e e Lf'1 N N N C C ti. m**- E v e e e e e e e e e v e ~ ;' ... ~. = c as w a> e as as as a. as as as t-V.a. t dn,.. E e c c c
, .C - -- .c- .c- .c .c- .c. .c. .c -- .c- === s .c. =
a .J J J .,J J .J f .J J .J J J f h v, ~=
- M,3 a' '
C - - - - - - - - - - - - we ** E E
=J C C C C C C C C C C C C 1.l..
u v v v v v v v v v v v s,im g u.. = 1 63 1 YNPS - RESOLUTION OF NRC OUESTIONS ON SEP TOPIC III-6 APRIL 8. 1986
_ =- _..-- . -- - - .- -- -:
,I 'I I
I
" SEISMIC REEVALUATION AND RETROFIT CRITERIA" I FOR YNPS I REVISION 3, APRIL, 1986.
I l l lI l I I YNPS - RESOLUTION OF NRC OUESTIONS ON SEP TOPIC III-G l APRIL 8. 1986 64
I l OBJECTIVES: 0 ADDRESS, , O CLARIFY I ISSUES RAISED BY NRC AND CONSULTANTS. I FOLLOWING IS A
SUMMARY
OF: o THE HIGHLIGHTS OF REVISION 3 CRITERIA, AND I 0 THE RESPONSES TO CRITERIA-RELATED ISSUES. FOR MORE DETAILS, SEE THE br,ITERIA DOCUMENT ITSELF. I I I , I . I l I . I YNPS - RESOLUTION OF NRC CUESTIONS ON SEP TOPIC III-G APRIL 8, 1986 65
l
^
OBJECTIVE: ESTABLISH THE SEISMIC EVALUATION PROCEDURES TO DEMONSTRATE THAT PLANT CAN MAINTAIN HOT SHUTDOWN FOLLOWING THE NRC SPECTRUM EARTHOUAKE. O PIPING 0 ANALYZE l SCOPE: o EQUIPMENT y {q0 EVALUATE 0 STRUCTURE i, O MODIFY IF NECESSARY FOR EACH ITEM. DEMONSTRATE ONE OF THE FOLLOWING: 0 YCS SEISMIC LOADS VS. CQDE l 0 0 NRC SEISMIC LOADS VS. SEP ALLOWABLES NRC SEISMIC LOADS VS. FUNCTIONALITY I I I I I I I I I . I YNPS - RESOLUTION OF tEC OUESTIONS ON SEP TOPIC III-C M APRIL 8. 1936 I
=
- ^ ^
T:: -.- .- ::r: . . . ~~~ r ~^T::^~^_ .. - I IfibTERIAL PROPERTIES: l 0 0 CONCRETE, STEEL, S0ILS HILTI BOLTS SEE TABLE D.1 SEE TABLE D.2 0 DAMPING SEE TABLE D.3 g ALSO INCLUDE S0IL MATERIAL AND RADIATION DAMPING WHEN CONSIDERING I SSI. O MAJOR MECHANICAL EQU; MENT - SEE TABLE D.4 I I I I I I I l l I I . I YNPS - RESOLUTION OF NRC QUESTIONS ON SEP TOPIC III-G APRIL 8, 1986 e7 I
w
~
.e-APPENDIX D: TABLE D-1 MATERIAL PROPERTIES MAfts;ALS I
altur0 sting sittL 50tt ** l Cont aE 11 Sult0fMG !.D. Stauclue At sitti ASTM All a A 305 8 t'sf . Int. Gr. 10.6 tsf (st,th W or ()
- 1. Diesel Gen. ASTM .t? (Fy e 33 kst) t' = 1.000 psi.
81dg. & Accum. (Fy e 40 est) I Tsak Enclosure
- 2. Turbine But1 ding ASTM A7 (Fy . 33 kst) A) Footings & Grade Sea *s. f' = 2500 pst.
8 ksf 10.6 ksf (with W or () 4 Pedestal 8) Precast hMS & Wall I- shteld, f'
- 2500 pst.
C) All Other Cast In-place, f'
- 3090 pst.
ASTM All 4 A 305 lat. Gr. (Fy e 40 tst )
- 0) Tu rAine Support Met I & Pedestal. f*'
- 3000 pst.
8 tsf f' = 3000 pst ASTM A15 4 A 305 I 3. Spent Fuel Pool / Spent Fuel Chute
. ASTM A7 (Fy = 33 tst) Int. Gr.
(Fy a 40 kst) 10.6 tsf (utta W or E) 20 tsf I 4. Steel Vapor Container A) Plate Material. ASTM A-100. Class A 201 Grade 8. Fy
- 32 tst.
- 8) Steel Colums A5TM A-283, Grade C I Fy = 30 tst C) Tie Rod Asse-ely AISI C-1020 4 4320 Fy = 30 k si f'
- 3000 pst (hedestaland Footings)
ASTM A15 & A 305 Int. Gr. (Fy e 40 tst) I (asgu.e saae as cotums)
- 0) Rase Plate'. ASTM A284. Grade 8. FYa27 tst E) Aacher Bolts
- t
- 20 est (Fyr32mst)
I F F,
- 10 ksi F, . 20 ksi (5.5.)
25 kst (0.5.) to asf l Al Footing and Grade
- 5. Reactoe Support ASTM A-300. Class A-201 845. f ' . 3000 pst 5tructure Grade 8. Fy = 32 tst ASTM A15 & A 305 8)Pedestafs. Cots.
Waits & All Others Int. Gr. f ' = 4000 p s t (Fy = 40 ast) I 6. Primary au s, A5Tu A7 (Fy e 33 tst) f' c 3000 pst ASTM 415 & A 305 Int. Gr. 8 ks' 10.6 tsf (with w o- ( ? f l (Fy
- 40 kst)
I Sutiding ami RaC oactive Tunael ASTM A15 & A 305 8 ksf f *
- 3000 p s t 10.6 tsf (ita a o*()
I 7. MS/f W Support A5Tu A7 (Fy = 31 tst) lat. Gr. i Structure (Fy = 40 ast) I Continued next sheet I OT3*RWi
=1- o-er Yankee Nuclear Power Station fg g g' M Seismic Reevaluation CriteriaDoc. No. DC-1; Rev. 3 6S 1111111111111111l 1111111111111 80023/81060/81061/86064
T__
^^ -
o TABLE D-1 (Continued) , l ASTM A615 8 tsr ASTM A36 (Fy
- M ksi); Milti Estk.Solts 10.6 ksf (etth w or ()
S. elodifications* 17011 Electrode; Grade 40. 60 No Met f 'c '= 2000 psi Duck Bank f
- 3000 psi Solts: A5tM A325F. or as 5 otherwise specified All Others f g . 4000 pst f' . 4000 ps1 ASTM A615 e asf
- 9. Fire Tank A) Plate Material. ASTM Grace 60 10.6 att (with w or I)
A 283, Grade C (Fy = 60 kst) I 8) Oracing System A-36 C) Anchor Bolts I
- Applicable to all structu es.
A!51 1141 r
** See discussions in Section 5.1.4
/
r I I _ I biUEB$E Yankee Nuclear Power Statiori T { d ' f j Seismic Reevaluation Criteria tillllllillllilllIlllllitill 80023/81060/81061/86064 Doc. No. DC-1; Rev. 3 69 I ,
I APPENDIX D: TA8tt 0-2 ALLOWA8LE LDADS FDR HILTI BOLT 5 ** KWIK-BOLT AVERAGE ULTIMATE TENSILE & SHEAR LOADS
- E 4000 PSI 6000 PSI 2000 PSI Concrete Strength Tension Shear Teesion Shear Tension Shear Diameter Esbedmen 7700 13500 6600 11562 5410 11198 13500 5/8' 2 3/4" 9100 11562 9560 ;
6250 11198 13500 3 1/2" 14500 I 12000 11562 7000 11198 15437 4 1/2" 15437 20300 13378 14300 5 1/2" 7550 21000 15437 16000 15437 8025 13378 15437 6 1/2" 21000 I 7 1/2" 9000 13378 17000 10150 15437 17133 10860 18102 8155 13257 18102 I 3/4" 3 1/4" 4" 5" 9700 11700 13257 13257 13400 16500 17133 17133 18466 13700 17600 22500 18102 21009 15195 18000 I 6" 13800 23600 21009 21000 18466 1519; ' 7" 15800 23600 21009 23000 18466 16000 15195 21009 8" 18466 23600 15195 23500 9" 16000 20500 32112 16000 26879 14000 27355 32112 l' 4 1/2" 26879 23441 27355 18900 5" 15500 23441 32112 23441 26870 17600 27355 32112 6" 26879 23441 27355 23441 7" 18200 23441 36304 23441 34491 18200 27355 363c4 8" 34491 23441 27355 23441 9" 18200 23441 3630: 23441 34491 18200 27355 10" I 31200 45105 { 23000 35680 19000 36750 45105 i 1 1/4" 5 1/2" 35680 36500 36750 27100 6 1/2" 21600 42000 45105 35680 I 31100 i 23600 36750 4709c 7 1/2" 35680 44400 39843 34600 8 1/2" 25100 44400 470ea . 37800 35680 26200 39843 495 e5 9 1/2" 35680 44400 I 10 1/2" 26800 39843 40900 Curves and test data I
- Tensiott values obtained from best fit curve through mean values of test data.
contained in A. A. Hads Report No. 8794 (HILTI No. TR-111 A). tion of the Shear anchor. values are minimum n'ean values at each embedment based j
** The allowable loads should be 1/4 of the ultimate loads listed above.
I blNN Yankee Nuclear Power Station ,g Aj ,' M Seismic Reevaluation CriteriaDoc. No. DC-1; Rev. 3 I IHilitilllillllllillllllillll 80023/81060/81061/86064
-. ; . ,-s .; -;_- .- :- _ _ ; .; ___
~
APPENDIX D TABLE D-3 I RECOMMENDED DAMPING VALUES
- Type. and Condition Percentage Stress Level of Structure Critical Damping Working stress, a. Vital piping 1 to 2
~
no more than about i b. Welded steel, prestressed 2 to 3 yield point concrete, well reinforced concrete (only slight cracking)
- c. Reinforced concrete with 3 to 5 considerable cracking
- d. Bolted and/or riveted 5 to 7 steel, wood structures' with nailed or bolted .
jcints. E At or just below a. Vital' piping 2 to 3 l yield point b. Welded steel, prestressed 5 to 7 concrete (without complete l c. loss in prestress) Prestressed concrete with 7 to 10 l d. no prestress left Reinforced concrete 7 to 10 Bolted and/or riveted steel, 10 to 15 l e. wood structures, with bolted joints g 15 to 20
- f. Wood structures with nailed joints 3
- Source: NUREG/CR-0098
[m=i Yankee Nuclear Power Station ,l e rgq i y j Seismic Reevaluation Criteria EnmanninnInnnim 80023/81060/81061/86064 ooc. No. Dc-1; Rn. 3 l
. _ . . .- - ~. = : g g - - - TABLE D-4 MATERIAL SPECIFICATIONS FOR EQUIPMENT EQUIPMENT MATERIAL SPECIFICATION Reactor Pressure Vessel (RPV) Head, Vessel Shell & Bottom Carbon Steel, SA-302 Grade B Bolting Flange Carbon Steel, SA-105 Grade 2 Closure Studs Carbon Steel, SA-193 Grade B16 Vessel Support Carbon Steel, SA-212 Grade B _ Steam Generators Head, Shell & Bottom Carbon Steel, ASTM A212 Grade B Tubes Stainless Steel, ASTM A-212, Type 304 Pressurizers Heads A Shell Course Carbon Steel, SA-302 Grade B Vessel Support Carbon Steel, SA-212 Grade B Main Coolant Pumps Casing Stainless Steel, Type 304 Main Coolant Loop Iso. Valves Valve Rody Stainless Steel, SA-351, Type 304, Grade CF8 Yoke Carbon Steel, SA-216, Grade WCA (Assumed) I Ronnet Stainless Steel, Type 304 Main Coolant Pump Disch. I Check Valves Valve Body Cap Stainless Steel, SA-351 Grade CF8 Stainless Steel, Type 316 Studs ( 20 per valve) Stainless Steel, Type 316 Main Coolant Bypass Piping Iso. Valves I Valve Body Yoke Stainless Steel, ASTM A351, Grade CF8M Carbon Steel, ASTM A216, WCB Carbon Steel, ASTM A193 Grade B7 Yoke Studs Pressurizer Spray Valves Valve Body A Stem Stainless Steel, Type 316 Main Stean Non-return Valves Valve body A Ronnet Carbon Steel, ASTM A216 Grade WCR Studs Carbon Steel, ASTM A193 Grade B7 Main Steam Code Safety Valves Valve Rody Carbon Steel, ASTM A216 Grade WCR I Yankee Nuclear Power Station I g"u=& W?F4 q;c Seismic Reevaluation Criteria (fg W ,I J 80023/81060/81061/86064 Doc. No. OC-1; Rev. 3 72 1111ll111111111111111111111111 I
---3 3. _
TABLE D-4 MATERIAL SPECIFICATIONS FOR EQUIPMENT (Continued) EQUIPMENT MATERIAL SPECIFICATION Main Feed Check Valves Valve Rody & Bonnet Carbon Steel, ASTM A216 Grade WCB Bonnet Studs Carbon Steel, ASTM A193 Grade B7 Pressurizer Relief A Block Valves Stainless Steel, ASTM A351 Grade CF8M I . Valve Body A Bonnet Bonnet Studs A Eye Rolts Carbon Steel, ASTM A193 Grade B7 Loop Isolation Valves Valve Hody & Bonnet Stainiess Steel ASTM A351, Grade CF8M Yoke Carbon Steel, ASTM A216 Grade WCB Yoke Studs Carbon Steel, ASTM A193 Grade B7 Loop Isolation Check Valves Valve Body and Cover Stainless, ASTM A182 Type 316 Cover Studs Carbon Steel, ASTM A193 Grade B7 Shutdown Cooling Iso. Valves Valve Body & Bonnet Stainless Steel, ASTM A351 Grade CF8M Yoke Carbon Steel, ASTM A216 Grade WCB Yoke & Ronnet Bolts Carbon Steel, ASTM A193 Grade B7 Feed A Bleed Heat Exchanges Shell Stainless Steel, ASTM A351 Grade CF8 Tubes Stainless Steel, ASTM A213, Type 304 I I
- m. - =m Yankee Nuclear Power Station
*= hW#4M Seismic Reevaluation Criteria 73 M e)l f j 80023/81060/81061/86064 Doc. No. DC-1; Rev. 3 11lll111111ll111111111ll111111 I
^
l ILOADS: 0 DEAD LOADS - ATTACHED EQUIPMENT, ETC. O LIVE LOADS - SNOW, MOVABLE EQUIPMENT, ETC. l 0 o EARTH PRESSURE AND GROUNDWATER TABLE FLUID LOADS - HYDROSTATIC EXCEPT UNDER SEISMIC CONDITIONS SEISMIC LOADS - NRC, YCS SPECTRA I O 0 THERMAL LOADS - T AMBIENT i 70
- F 0 PRESSURE - DESIGN PRESSURES I o WIND - PER UBC I
I I I I I I I I I .. I YNPS - RESOLUTION or tRC OUESTIONS ON SEP TOPIC III-6 74 APRIL 8, 1986 I
I lLOADCOMBIf1ATIONS: l 0 FOR STRUCTURES: (EXCEPT VC SHELL) D+R0+PO + (E OR U OR M), OR D+L+R0+Po+To I o FOR EQUIPMENT: 0+Ro+Po+E o FOR VC SHELL: 0 + L + R 'o+ To + PLOCA + TLOCA * (E OR W OR M) l (R0' INCLUDES SEISMIC CONTRIBUTIONS) I I I I I I I I I I . I 75 YNPS - RESOLUTION OF URC QUESTIONS ON SEP TOPIC III-G APRIL 8, 1986 I
l - _ . lLINEARANALYSISMETHODOLOGY: l 0 BASIC TECHNIQUE - RESPONSE SPECTRUM ANALYSIS 0 SSI - INCLUDED ON A CASE-BY-CASE BASIS g 0 INTERCONNECTED BUILDINGS - COUPLING STUDIED ON A CASE-BY-CASE BASIS l 0 USE SIMPLIFIED SDOF MODELS WHERE APPROPRIATE l 0 USE SRSS METHOD (CLOSELY SPACED MODES ABSOLUTELY ADDED) TO C0f1BINE 3-D MODAL RESPONSES I o LOCAL STRESSES IN VESSELS - (SEE ITEM NOS. F2D, F4E, F5C) USE WRC BULLETIN 107 APPROACH I - OBSERVE LIMITATIONS OF AB0VE APPROACH USE WRC BULLETIN 297 APPROACH WHERE WARRANTED 0 ARS GENERATION USE /RTIFICIAL TIME HISTORIES AND STRUCTURAL MODAL ANALYSIS RESULTS SATISFY SRP SECTION 3.7 AND REG. GUIDE 1.122 I I I l .. I I YNPS - R[ SOLUTION OF IRC OUESTIONS ON SEP TOPIC III-G APRIL 8, 1986 76 I
APPENDIX C: Figur -1 STRUCTURAL ANAL SIS ILO!!CHA5f I c, .,i.
- o. .. ..n. l I
I u .. .i o........ n.,,,. ... I l u . . . , .0, ",'.. .
.u....,
I v.',",i 8. o... .. e.., I I "g;c , a...,.
^
T..'O'3,'.. o ANALY$rS? ",0 IIuOEI I INSPE C LV ** ., N o nha.., Tim. Hi.to,y An.fy .i. An.ty.is Lin.., 8,.ct,.I u.es Il An.ly.i. u.6eg ANSYS m i u.hg uCST .t ANSYS l
- 0. . ge.t.. ANSR )
BATS DR AIN-2D ggSAP Loc. tion. l R..po..e C.icul.tlon. I I t o D.fia. S t,... I I S,.ct,. l
- In Loc.ht.. u.mgfNSPEC]
Re gson i f I , O...... e S i, . . . . . I s i i r I . ...'.., u........ t s .= - . a. SERT $i Yankee Nuclear Power Station I NlJL~#QY~]A - Seismic Reevaluation 80023/81060/81061/86064 Doc.Criteria No. DC-1; Rev. 3
.,' 7 t'111111111111111111111111111 I
~ -
_J ._ _ _ _ _ _ . _ _ _ . . _ . . . . _ _ .. _. _ I hLLOWABLESTRESSES: l 0 UNDER YCS CONCRETE: ACI STEEL: AISC, PART 1 g 1 VC SHELL BUCKLING: BASED ON STUDIES ON SHELL STABILITY EXISTING BRACES: AISC, PART 2 SECTION 2.8 ! I ! 0 UNDER NRC l - BENDING + TENSION: COMPRESSION: 0.95F y 0.95FCR, FCR " EA W/0UT F.S. l - CONCRETE AND VC SHELL BUCKLING: SAME AS UNDER YCS I I I I I I I cw l ,, f(
)
I 73 YNPS - RESOLUTION OF tRC OUESTIONS ON SEP TOPIC III-G APRIL 8, 1986 I
I ALLOWABLE STRESSES (CONT'D): l l 0 FOR EQUIPMENT AND VC SHELL PENETRATIONS PER NRC SEP GUIDELINES g ALS0: USE ASME CODE PARAGRAPH NC-3521 TO QUALIFY BODY SECTIONS OF CLASS 1 AND 2 VALVES WHEN SEISHIC STRESSES ARE LOW (SEE ITEM NO. F7A.I) 0 FOR FIRE TANK STEEL f, ANCHOR BOLT STRESSES: AISC ANCHOR BOLT PULL-00T: ACI-3t19 0 FOR REACTOR INTERNALS l - CORE SUPPORT AND OTHER STEEL: APP. F ASME CODE SECTION III, I FUEL ASSEMBLIES: TEST DATA CONTROL ROD: No YIELD No IMPACT I I I I I I .. I YNPS - RESOLlJTION Of tRC OUESTIONS 00 SEP TOPIC III-G 79 APRIL 8,19P6 I
~
* * " *
- e- --.=m, , . - _ _ _ . ,_,
lALLOWABLEDEFORMATIONS: , l l 0 LIMITED BY EXI" TING CLEARANCES 1 I MODIFICATION DESIGN: I l 0 CONVENTIONAL METHODS OF LINEAR ELASTIC STRUCTURAL AllALYSIS 0 MATERIALS. LOADS. ALLOWABLES: AS ABOVE I I I I I I I I I i 80 YNPS - RESOLUTION OF NRC QUESTIONS ON SEP TOPIC III-G APRIL 8. 1986 I -_ ._ _ - _ - _ - -
I l l STRUCTURAL NONLINEAR PERFORMANCE CRITERIA: l 0 0 USED TO DEMONSTRATE FUNCTIONALITY UNDER NRC SPECTRA ACCOUNTS FOR MATERIAL NONLINEARITIES, DUCTILE MEMBERS, l 0 CONNECTIONS USES - TIME-HISTORY ANALYSIS, OR
- EoVIVALENT LINEAR APPROACH, OR I - NONLINEAR STATIC ANALYSIS I
I I I I I I I I I .. I al Yt PS - RESOLUTIOff Of trC OUESTIONS ON SEP TOPIC III-G APRIL 8, 1936 I ,
7 7_____-...._ I MASONRY WALL PERFORMANCE CRITERIA: 0 ENSURE THAT DESIGNATED WALLS DO NOT JE0PARDIZE THEtSSS DURING A YCS BASED SEISMIC EVENT 0 LOADS - DEAD LOAD I - SEISMIC: WALLS AND EQUIPMENT
- INTERSTORY DISPLACEMENTS
- WIND PRESSURE. W TQ I - TORNADO DIFFERENTIAL PRESSURE. W TP 0 WIND AND TORNADO LOADINGS BASED ON 10-4 EVENT PROBABILITY (95% CONFIDENCE) 0 LOAD COMBINATIONS g D+W D + YCS D+W'WT"T IWTO) OR (WTP) OR (WTQ + 0.5WTP I O llATERI ALS l CONCRETE BLOCK: ASTM C-90 (EXTERIOR)
ASTM C-129 (INTERIOR) MORTAR: ASTM C-270 0 DAMPING: 5% OF CRITICAL FOR MASONRY I I I YNPS - RESOLllTICN OF fJRC OUESTIONS ON SEP TOPIC III-G 82 l APRIL 8, 1986
I hASONRYWALLANALYSISANDEVALUATIONTECHNIQUES: 0 FIELD WALKDOWN 7 0 IN GENERAL, ANALYSIS BASED ON UNCRACKED SECTIONS 0 FREQUENCY CALCULATIONS 0 SEISHIC LOADS USING APPLICABLE ARS CALCULATED FREQUENCY l - STATIC EQUIV. LOADS WHERE APPROPRIATE 0 INTERSTORY DISPLACEMENTS USING IN-PLANE STRAIN ANALYSIS (ALLOWABLE STRAIN = 0.001) I O STATIC AND DYNAMIC ANALYSIS USING MCAUTO STRUDL I - STRUDL DYNAL 0 ALLOWABLE STRESS PER ACI 531-79 l - INCREASED PCR SRP SECTION 3.8.4, APP. A, FOR (D+E') AND (D+WT ) LOAD CASES HOWEVER NO INCREASE FOR MASONRY TENSION PERPENDICULAR TO l BED JOINT
- SEE TABLES 7-2 AND 7-3 0F THE DC1, REV. 3 l
I l .. I YNPS - RESOLlfTION Of IRC OUESTIONS ON SEP TOPIC III-6 83 APRIL 8, 198G
l t i i i i,
- PIPING ISSUES i
l. i l 1 84 YNPS - RESOLUTION OF NRC QUESTIONS ON SEP TOPIC III-G i, w** _w.ey- --www,,
.._ _ ~ _ . - - -.- --
I l ITEMS Di. dig. E4. AND E13: I t ISSUE
SUMMARY
- CLARIFY THE HOT SHUTDOWN SCOPE ANALYSES WITH THE NRC SPECTRA.
REFERENCE:
YAEC PRESENTATION TO THE NRC, FEBRUARY 24-26, 1986 ll RESPONSE: ALL THE PIPING IN THE HOT SHUTDOWN SCOPE WAS ANALYZED TO YCS. l THE HOT SHUTDOWN SCOPE SYSTEMS ARE LISTED Irl TABLE D1-1 0F THE PIPING RESPONSE DOCUMENT. I CONFIRMATORY ANALYSES WILL BE PERFORt1ED ON SELECTED SYSTEMS USING THE NRC SPECTRA. I I I l I I I .. I I
, es
I ITEMS D2. Diti. AND E17: I ISSUE
SUMMARY
PROVIDE METHODS AND RESULTS OF DIPE SUPPORT EVALUATIONS. I
REFERENCE:
SEISMIC REEVALUATION AND RETROFIT CRITERIA FOR YANKEE NUCLEAR POWER STATION, R0WE, I MASSACHUSETTS, 80023/81060/81061 DC-1, REVISION 3. APRIL 1986 l RESPONSE: THE CRITERIA AND GENERAL METHODOLOGY FOR THE PIPE SUPPORT EVALUATION ARE PROVIDED IN SECTION 9.0 0F THE RETROFIT CRITERIA DOCUMENT. RESULTS ARE CURRENTLY BEING COMPILED FOR ALL HOT I SHUTDOWN PIPING SYSTEM SUPPORTS AND WILL BE PROVIDED TO THE NRC. I I I I .. I I 86 I
- . __ ~; _ .-. . .
I ITEM D3: ISSUE
SUMMARY
- ADDRESS THE POTENTIAL FOR IMPACT LOADIN$ OF l SUPPORTS SUBJECT TO PIPE UPLIFT.
I
REFERENCE:
SEISMIC REEVALUATION AND RETROFIT CRITERIA FOR YANKEE NUCLEAR POWER STATION. R0WE. I MASSACHUSETTS. 80023/81060/81061 DC-1. REVISION 3. APRIL 1986 l RESPONSE: APPENDIX J OF THE RETROFIT CRITERIA DOCUMENT DET AILS THE METHODOLOGY USED TO CONSIDER ROD HANGER UPLIFT. I I I I
- I
!I I I I I ._ - - -__ - _ - _ - .
ITEM D4: I ISSUE
SUMMARY
- DESCRIBE MODELING METHODS TO ACCOUNT FOR SUPPORT STIFFNESS.
I
REFERENCE:
SEISMIC REEVALUATION AND RETROFIT CRITERIA FOR YANKEE NUCLEAR POWER STATION, R0WE, MASSACHUSETTS, 80023/81060/810G1 DC-1, REVISION 3 APRIL 1986 l RESPONSE: ALL SUPPORTS WERE MODELED AS RIGID RESTRAINTS (STIFFNESS OF 1.0E7 FOR 8" PIPING). I THE MAIN STEAM FEE 0 WATER SYSTEM OUTSIDE THE VC IS THE EXCEPTION. THIS IS DISCUSSED IN I ITEM E1. l THE FREQUENCIES OF THE LARGE BORE SUPPORTS UERE CHECKED TO ASSURE FUNDAMENTAL FREQUEf1CIES OF 33 HZ OR GREATER. THE STIFFNESSES AND DEFLECTIONS OF THE SMALL g BORE SUPPORTS WERE CHECKE0 TO ASSURE THAT THE RIGID ASSUMPTION IS VALIO. I CONCLUSION: THE MINIt10M FREQUENCY, SlIFFNESS, AND DEFLECTION RE0tlIREMENTS INSURE THAT THE SUPPORTS ARE RIGID. I I ee g
I ITEM 05: ISSUE
SUMMARY
- PROVIDE CORRELATION BETWEEN RESPONSE SP$CTRA l IDENTIFICATION IN TABLES AND PLOTS IN THE SAFETY-RELATED PIPING REPORT.
I
REFERENCE:
PIPE STRESS ANALYSIS REPORT. REPORT N0. E-Y-YR-80023-14. REVISION 1. MAY 1983 I RESP 0tJSE: A CROSS-REFERENCE MATRIX WILL BE PROVIDED TO CORRELATE THE RESPONSE SPECTRA. I I I I I I I I .. I I I
. . = . . . ._ - _2 :-_ = -- _-
I l ITEM D8: I ISSUE
SUMMARY
CLARIFY THE METHOD OF COMBINING STRESS TERMS. I
REFERENCE:
SEISMIC REEVALUATION AND RETROFIT CRITERIA FOR I YANKEE NUCLEAR POWER STATION. R0WE. MASSACHUSETTS. 80023/81060/81061 DC-1, l REVISION 3. APRIL 198G I RESPONSE: SECTION 8.3.1 0F THE RETROFIT CRITERIA DOCUMENT LISTS THE STRESS COMBINATIONS. I IN THE MAJORITY OF ANALYSES. THE FOLLOWING LOAD C0t1BINATIONS AllD ALLOWABLES ARE USED. A '+' INDICATES COMBINATION BY ABSOLUTE SUMMATION. l ANSI B31.1 E00ATION LOAD C0l1BINATION ALLOWABLE STRESS 11 0+P 1.0 SH 12 G+P+S 1.8 SH 13 T + TAM + SAM SA l 14 G + P + T + TAM + SAM SA + SH I .. I I eo I )
I I ITEM D9: ISSUE
SUMMARY
- CLARIFY THE TEMPERATURE BASES OF THE STRESS LIMITS.
I
REFERENCE:
SEISMIC REEVALUATION AND RETROFIT CRITERIA FOR YANKEE NUCLEAR POWER STATION, R0WE, MASSACHUSETTS, 80023/81060/81061 DC-1, REVISION 3. APRIL 1986 l RESPONSE: THE STRESS ALLOWABLES FOR THE ANSI B31.1 EQUATIONS ARE OBTAINED IN APPENDIX A 0F THAT CODE. THE 11ATERI AL TYPES AND DESIGN TEMPER ATURES ARE I OBTAINED FROM THE YANKEE PIPING SPECIFICATIONS YS-497 AND YS-4G52. I I I : I I l I I - -
ITEM D10: ISSUE
SUMMARY
- EXPLAIN THE DIFFERENCES IN STRESS RESULT'S IN THE PIPING ANALYSIS REPORT FOR SYSTEMS WITHOUT MODIFICATIONS OR ADDITIONAL INFORMATION.
I PIPE STRESS ANALYSIS REPORT, REPORT N0. E-Y-YR-
REFERENCE:
I 80023-14, REVISION 1. MAY 1983 I RESPONSE: REFINEMENTS IN THE MODELS OR CORRECTIONS OF MINOR t10 DEL DEVIATIONS WERE MADE. THERMAL AND SEISMIC ANCHOR t10TIONS WERE g REDEFINED. THERMAL AND THERf1AL ANCHOR t10TI0flS WERE ANALYZED TOGETHER. l CONCLUSION: THE STRESS RESULTS IN TABLE 5.8-1 REFLECT THE LATEST ANALYSES AND SUPERSEDE THOSE IN TABLE 5.5-1. g I I I .. I I I ,
_ _ . -- - - -n _- -- - - - - ITEM Dil: I ISSUE Sunr1ARY: EXPLAIN THE DIFFERENCE Ik EQUATION 12 RESULTS BETWEEN TABLES 5.5-1 AND 5.8-1 0F THE PIPING ANALYSIS REPORT FOR PROBLEM NUMBER 3. PIPE STRESS ANALYSIS REPORT. REPORT No. E-Y-YR-I
REFERENCE:
80023-14. REVISION 1. MAY 1983 I RESPONSE: THE EQUATION 12 STRESS FOR PROBLEMS 2. 3. Afl0 23 SHOULD BE IDENTICAL IN THE TWO TABLES. THE RESULTS IN TABLE 5.8-1 ARE NOT CORRECT. I THE EQUATION 12 STRESSES FOR PR00LEMS 2. 3, Aro I CONCLUSION: 23 IN TABLE 5.8-1 SHOULD BE REPLACED BY THOSE IN TABLE 5.5-1. I I I I I .. I I I
- - - = - ,
_ . = . - - . I I ITEM D12: I ISSUE SUMr1ARY: PROVIDE RESULTS FOR THE NRC SEISMIC LOADING FOR PROBLEMS 2, 3. AND 23. I
REFERENCE:
PIPE STRESS ANALYSIS REPORT, REPORT N0. E-Y-YR-80023-14. REVISION 1, MAY 1983 l RESPONSE: ALL PIPING SYSTEMS ARE ANALYZED FOR YCS LOADS. l CONFIRMATORY ANALYSES USING THE NRC SPECTRA WILL BE PERFORMED FOR SELECTED SYSTEMS. I CONCLUSION: THESE THREE PROBLEMS ARE NOT CURRENTLY IN THE I CONFIRMATORY ANALYSIS SCOPE: THEREFORE, THE REQUESTED RESULTS ARE NOT AVAILABLE. I I I I I .. I I I
l ITEM Di3: I ISSUE
SUMMARY
- PROVIDE SECTION 5.11 0F THE SAFETY RELATED I PIPING ANALYSIS REPORT.
I
REFERENCE:
PIPE STRESS ANALYSIS REPORT. REPORT No. E-Y-YR-80023-14. REVISION 1. MAY 1983 l RESP 0!!SE: THE REQUESTED SECTION IS ATTACHED TO THE PIPING RESPONSE DOCUMENT. I lI I
- I l
j l l I es .g
2 " ~ T:- ^ - _____- _. _ __ - .- ITEMS D15 AND E18: ISSUE
SUMMARY
- PROVIDE EVIDENCE THAT ALL REQUIREMENTS 9F IE BULLETIN 79-02 ARE SATISFIED.
I l
REFERENCE:
USNRC IE BULLETIN 79-02, " PIPE SUPPORT BASE PLATE DESIGNS USING CONCRETE EXPANSION ANCHOR BOLTS," MARCH 8, 1979 SEISt1IC REEVALUATION AND RETROFIT CRITERIA FOR YANKEE NUCLEAR POWER STATION, R0WE, MASSACHUSETTS, 80023/81060/81061 DC-1, l REVISION 3 APRIL 1986 I RESPONSE: THE F0LLOWING REQUIREMENTS HAVE BEEN MET: I
- PRYING ACTION HAS BEEN CONSIDERED.
- A SAFETY FACTOR OF FOUR HAS BEEN USED FOR WEDGE AND SLEEVE TYPE ANCHORS.
I
- A SAFETY FACTOR OF FIVE HAS BEEN USED FOR SHELL TYPE ANCHORS.
- CYCLIC LOADS WERE NOT CONSIDERED IN THE g EVALUATION OR DESIGN, AS PROVISIONS FOR PROPER PRELOAD AND INSTALLATION ARE ASSUMED.
- YAEC OC AND INSTALLATION PROCEDURES ARE
.. ADEQUATE TO FULFILL THE REQUIREMENTS OF IE BULLETIN 79-02.
I CONCLUSION: ALL REQUIRE!1ENTS OF IE RULLETIN 70-02 HAVE BEEN MET. n I
.= ~;;_ _ _
1 ITEM D16: I I S S U E S Ut1!!A R Y : CLARIFY THE HOT SHUTDOWN SYSTEM BOUNDARIES AND THEIR ASSOCIATED STRESS PROBLEMS. I
REFERENCE:
YAEC PRESENTATION TO THE NRC ON FEBRUARY 24-26, 1986 I PIPE STRESS AtlALYSIS REPORT, REPORT NO. E-Y-YR-l 80023-14, REVISION 1, MAY 1983 I RESPONSE: SEE ITEM D16 IN THE PIPING RESPONSE DOCUMENT FOR A LISI 0F ALL HOT SHUTDOWN SYSTEM PROLLEMS WITH A BRIEF DESCRIPTION OF THE PROBLEM BOUNDARIES AND REFERENCE CALCULATIONS AND/0R REPORTS. I I I I I I .. . I I e, g
.w . .- -__-_-____. . . - _ _ . _ _ _ _ _ _ . _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _
. 3 I -
I ITEM D17: I ISSUE
SUMMARY
- DESCRIBE THE PROCEDURE USED TO DETERMINE THE INDIVIDUAL AND COMBINED STRESS REGULTS IN THE )
SAFETY-RELATED PIPING ANALYSIS REPORT. I PIPE STRESS ANALYSIS REPORT, REPORT NO. E-Y-YR-
REFERENCE:
I 80023-14. REVISION 1. MAY 1983 RESPONSE: STRESS RESULTS FOR INDIVIDUAL LOAD CASES ARE flAXIMA OVER ALL N0 DES. STRESSES FOR ALL LOAD CASES ARE NOT ALWAYS I MAXIMIZED AT THE SAME N0DE. I STRESS RESULTS FOR EQUATIONS ARE MAXIMUM COMBINATIONS OVER ALL N0 DES. I I I I I I .. I i lI I '
I I ITEM DIS _: FOR PROBLEM #207. EXPLAIN THE DIFFEF.ENCES IN I ISSUE
SUMMARY
ALLOWABLES BETWEEN TABLES 5.5-1 AND 5.8-1. EXPLAIN THE STRESS C0tlBIf1ATION FOR EQUATION 12 l IN TABLE 5.8-1. I 7EFEREfCE: PIPE STRESS ANALYSIS REPORT, REPORT l10. E-Y-YR-80023-14, REVISI0tl 1, MAY 1983 g I RESP 0!iSE: ALLOWABLES IN TABLE 5.5-1 ARE GIVEN FOR THE MAXIllull DESIGN TEMPER ATURE Ill T H E S Y S T Eii . THE DESIGN TE!1PERATURE IS GREATLY REDUCED PAST THE ISOLATI3t1 VALVE. l EQUATION 11 STRESS IS MAXIMIZED BETWEEt1 THE MAIN COOLANT LOOP A!JD THE ISOLATION VALVE ALLOWABLE FOR 650 DEGREES IS USED. I OTHER EQUATION STRESSES ARE MAXItlIZED PAST THE I ISOLATION VALVE; ALLOWABLE FOR 225 DEGREES At1BIENT DURING A LOCA IS USED. SEE ITEM D17 FOR STRESS COMBINATION DISCUSSION. I I .. I I l ee
l l ITEM D19: I ISSUE
SUMMARY
- FOR PROBLEMS #41A AND #41B. CHECK THAT THE 1
I SEISMIC STRESS FOR YCS IS HIGHER THAN THAT FOR NRC. REFEREllCE: PIPE STRESS AtlALYSIS REPORT, REPORT NO. E-Y-YR-l 80023-14. REVISION 1. MAY 1983 I RESPCNSE: THE MAJOR SEISMIC RESPONSE IS DUE TO THE VERTICAL COMPONENT IN THE FIRST VIBRATIONAL I MODE. AT THESE FREQUENCIES, THE YCS VALUE IS HIGHER TH AN TH AT OF THE NRC SPECTRL'M. ALTHOUGH THE HORIZONTAL NRC SPECTRA ARE HIGHER THAN THOSE OF THE YCS, ONLY HIGHER FREQUENCY MODES. WHICll ARE NOT SIGNIFICANT, CONTRIBUTE TO THE TOTAL RESPONSE. THE SPECTRAL INFORMATION FOR THE FIRST VERTICAL I MODE IS GIVEN IN THE FOLLOWING TABLE: l PROCLEM f0. FRECUEI'CY YCS ACCEL NRC ACCEL l #41A 12.183 .96 G .S7 G
, #'II B " 824 .SS G 77 G g .
CONCLUSION: THE STRESSES IN THE SAFETY-RELATED PIPING ANALYSIS REPORT ARE CORRECT. I
I l ITEMS E1.A.__EB. AND ELS: I ISSUE
SUMMARY
PROVIDE A COMPLETE DESCRIPTION OF THE GENERAL i MODELING TECHNIQUE USED FOR THE MAIN STEAM /FEEDWATER PIPING ANALYSIS. I MAIN STEAM /FEEDWATER PIPING ANALYSIS REPORT, I
REFERENCE:
YANKEE NUCLEAR POWER STATION, R0WE, MASSACHUSETTS, REPORT No. -Y-YR-83033-4, REVISION 0, JULY 1984 l CYGNA CALCULATION 83033-14/F, ANALYSIS OF MAIN STEAM PIPING OUTSIDE THE VAPOR CONTAINER I CYGNA CALCULATION 83033-23/F, AflALYSIS OF FEEDWATER PIPING OUTSIDE THE VAPOR CONTAINER g CYGNA CALCULATION 83033-18/F, ANALYSIS OF THE MAIN STEAM /FEEDWATER SUPPORT STRUCTURE I RESPONSE: PIPING AND SUPPORT STRUCTURE ANALYSES ARE PERFORMED SEPARATELY. I SUPPORT AND STRUCTURE STIFFNESSES AND EQUIVALENT MASSES ARE USED AS BOUNDARY CONDITIONS FOR THE ' PIPING ANALYSES. DIPE REACTIONS ARE USED TO EVALUATED THE SUPPORT STRUCTURE. I AN ITERATIVE SOLUTION IS USED UNTIL COMPATIBILITY BETWEEN THE PIPING AND SUPPORT STRUCTURE ANALYSES IS ACHIEVED.
=
I
_ ..:__.m __ _ _ _ _ . _ - _ _ . - - I . ITEMS El.B AND E9-I , I ISSUE
SUMMARY
- PROVIDE CALCULATION METHODS AND VALUES OF STIFFNESSES FOR ANCHOR A-50.
I
REFERENCE:
CYGNA CALCULATION 83033-14/F, ANALYSIS OF MAIN STEAM PIPING OUTSIDE THE VAPOR CONTAINER CYGNA CALCULATION 83033-18/F, ANALYSIS OF THE MAIN STEAM /FEEDWATER SUPPORT STRUCTURE I RESPONSE: UNIT MASSES WERE APPLIED TO THE MATHEMATICAL MODEL OF ANCHOR A-50 TO OBTAIN THE STIFFNESS VALUES FOR EACH OF THE SIX RESTRAINED DIRECTIONS. EQUIVALENT MASSES WERE CALCULATED TO PROPERLY MODEL THE FUNDAMENTAL FREQUENCY RESPONSE OF THE ANCHOR FRAI1E IN EACH RESTRAINED DIRECTION. I THE STIFFNESSES AND MASSES WERE APPLIED TO THE PIPING SYSTEM AT THE ANCHOR LOCATION. I lI 'I .. I I 1
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I
I ITEMS E1.C.I AND E10: I ISSUE SUMt1ARY: PROV'iDE DETAILS REGARDING THE MASS STIFFNESS. AND FREQUENCIES FOR THE VAPOR CONTAINER USED IN THE PIPING ANALYSES. I
REFERENCE:
CYGNA CALCULATION 83033-34/F, DEVELOPMENT OF SIMPLIFIED STIFFNESS MODEL FOR THE VAPOR CONTAINER I l RESPONSE: A SIMPLIFIED 110 DEL OF THE V APOR CONT AINER WAS DEVELOPED TO PROVIDE GLOBAL STIFFNESS AND EQUIVALENT MASS VALUES FOR FIVE DIFFERENT I ELEVATIONS ON THE VAPOR CONTAINER. SINCE THE EXITING t1AIN STEAM AND FEEDWATER PIPES ARE RELATIVELY LARGE WITH RESPECT TO THE THICKNESS OF THE VAPOR CONTAINER SHELL, LOCAL STIFFNESSES AND MASSES ARE ALSO CALCULATED T0 t10RE ACCURATELY MODEL THE BEHAVIOR OF THE PENETRATIONS. I THE TWO SETS OF STIFFNESSES AND MASSES ARE APPLIED IN SERIES TO THE PIPING MODELS AT THE I PENETRATION LOCATIONS. I .. I I I a
^-
l ITEM Ei.c.II: l I ISSUE
SUMMARY
- PROVIDE DETAILS REGARDING THE MASS, STIFFNESS, j AND FREQUENCIES FOR THE TURBINE BUILDING USED IN l THE PIPING ANALYSES.
I CYGNA CALCULATION 83033-18/F, AN AL YSIS OF THE
)
REFERENCE:
l MAIN STEAM /FEEDWATER SUPPORT STRUCTURE I RESPONSE: SIMPLIFIED STICK AND BEAM MODELS WERE USED TO REPRESENT THE TURBINE BUILDING FOR THE STIFFNESS DETERMINATION. I FOR THE FEEDWATER ANCHORS AT THE J-WALL, GLOBAL AND LOCAL STIFFNESSES AND MASSES WERE DEVELOPED. SIl1ILAR TO THOSE OF THE VAPOR CONTAINER. I I I I I I .. I I m g
~
ITEMS Ei.C.III AND Ela: I ISSUE
SUMMARY
- PROVIDE DETAILS REGARDING THE MASS, STIFFNESS, AND FREQUENCIES FOR THE SUPPORT STRUCTURE USED IN THE PIPING ANALYSES.
REFERENCE:
CYGNA CALCULATION 83033-18/F, ANALYSIS OF THE llAIN STEAM /FEEDWATER SUPPORT STRUCTURE I RESPONSE: THE MAIN STEAM /FEEDWATER SUPPORT STRUCTURE STIFFNESSES AND MASSES ARE CALCULATED FOR EACH PIPE SUPPORT ATTACHMENT LOCATION. I FOR MULTIPLE PIPES OR SUPPORTS ON ANY SUPPORT MEMBER, THE EFFECT OF ALL PIPES IS CONSIDERED IN I DEVELOPING STIFFNESSES AND MASSES FOR EACH SINGLE LOCATION. I I I I I .. I I 105
I I ITEMS E1.D AND E3: , ISSUE
SUMMARY
- EXPLAIN THE ANALYSIS METHODS AND DISCUSS THE ADEQUACY OF BREAKING THE ENTIRE SUBSYSTEM INTO l HANY SUB-PROBLEMS.
MAIN STEAM /FEEDWATER PIPING ANALYSIS REPORT, I
REFERENCE:
YANKEE NUCLEAR POWER STATION, R0WE, MASSACHUSETTS, REPORT NO. E-Y-YR-83033-4, REVISION 0. JULY 1984 l CYGNA CALCULATION 83033-14/F, ANALYSIS OF MAIN STEAM PIPING OUTSIDE THE VAPOR CONTAINER I CYGNA CALCULATION 83033-23/F, ANALYSIS OF FEEDWATER PIPING OUTSIDE THE VAPOR CONTAINER g CYGNA CALCULATION 83033-18/F, ANALYSIS OF THE MAIN STEAM /FEEDWATER SUPPORT STRUCTURE I RESPONSE: BECAUSE OF THE SIZE OF THE MAIN STEAM /FEEDWATER SYSTEM, IT WAS NECESSARY TO PERFORM SEPARATE ANALYSES ON SMALLER SUBSYSTEMS. I AN ITERATIVE SOLUTION IS EMPLOYED TO INSURE THAT THE B0UNDARY CONDITIONS PROVIDED BY THE SUPPORT I STRUCTURE AND THE PIPE REACTIONS If1 PARTED TO THE SUPPORT STRUCTURE WERE C0f1PATIBLE. TWO ITERATIONS HAVE BEEN COMPLETED ON THE MAIN STEAM /FEEDWATER PIPING AND SUPPORT STRUCTURE ANALYSIS: ANY ADDITIONAL COUPLING EFFECT ON THE SUPPORT STRUCTURE OR PIPING WILL BE FURTHER ADDRESSED BEFORE THE ANALYSES ARE FINALIZED. 106
n- . __ _ _ . _ _ _ __ ._ _ _ _ l ITEMS E2. E12. AND E20: I ISSUE
SUMMARY
- JUSTIFY WHY SEISMIC ANCHOR MOVEMEhT (SAM) t I ANALYSIS WAS PERFOR!iED FOR PROBLEM 9 BUT WAS CONSIDERED NEGLIGIBLE FOR OTHER CASES.
I CYGNA CALCULATION 83033--14/F, ANALYSIS OF MAIN
REFERENCE:
l STEAM PIPING OUTISIDE THE VAPOR CONTAINER RESPONSE: BECAUSE SUPPORT STIFFNESSES AND EQUIVALENT MASSES WERE USE0 TO MODEL THE SUPPORT POINTS OF THE MAIN STEAM PROBLEMS OUTSIDE THE VAPOR CONTAINER THE SUPPORT POINTS COULD DISPLACE: THUS, ANCHOR MOVEMENTS WERE IMPLICITLY INCLUDED IN THE SEISMIC ANALYSES. PROBLEM 9 IS NOT ATTACHED TO THE MAIN STEAM /FEEDWATER STRUCTURE: THEREFORE THE TRADITIONAL MODELING METHOD WITH RIGID SUPPORTS WAS USED, AND SEISMIC ANCHOR MOVEMENTS WERE EXPLICITLY CONSIDERED. CONCLUSION: SEIStiIC ANCHOR MOVEMENTS WERE CONSIDERED FOR ALL ANALYSES. I I .. I I m g
l ITEM ES-I ISSUE
SUMMARY
r CLARIFY THE REQUIREMENTS FOR " SAFE SHUTDOWN l PIPING" VERSUS " HOT SHUTDOWN PIPING." I
REFERENCE:
SEISMIC REEVALUATION AND RETROFIT CRITERIA FOR 1 YANKEE NUCLEAR POWER STATION R0WE. MASSACHUSETTS, 80023/81060/81061 OC-1, REVISION 3, APRIL 1986 l RESPONSE: " SAFE SHUTDOWN PIPING" AND " HOT SHUTDOWN PIPING" ARE -INTERCHANGABLE TERMS. I ALL PIPING IN THE SAFE SHUTDOWN SCOPE WERE EVALUATED TO THE SAME RETROFIT CRITERIA. I I I I I I I : I I g me
~
__ _ _ _ _ _ _ _ _. __ Z I I ITEM E6: I ISSUE
SUMMARY
PROVIDE REPRESENTATIVE MATHEMATICAL MODELS AND l DETAILS FOR MODEL VERIFICATION. l
REFERENCE:
NONE I THE REQUESTED INFORMATION WILL BE PROVIDED RESPONSE: l DURI.NG THE NRC AUDIT. P I I I I I I I I I .. I I
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I
_ _ _ _ _ _ . . _ _ _ . . . _ m ._ . _.__, .. _ . .. l ITEM E7: ] t ISSUE
SUMMARY
- PROVIDE REFERENCE 15 of THE MAIN STEAM /FEEDWATER PIPING ANALYSIS REPORT.
REFERENCE:
MAIN STEAM /FEEDWATER PIPING ANALYSIS REPORT, YANKEE NUCLEAR POWER STATION, ROWE, MASSACHUSETTS, REPORT NO. E-Y-YR-83033-4, REVISION 0, JULY 1984 RESPONSE: THE LISTED REFERENCE WILL BE DELETED FRCH THE NEXT REVISION OF THE SUBJECT REPORT. l 110
I I ITEM Ell: I ISSUE
SUMMARY
- PROVIDE REFERENCE 4 IN THE MAIN STEAM /FEEDWATER REPORT.
I
REFERENCE:
MAIN STEAM / FEE 0 WATER PIPING ANALYSIS REPORT. YANKEE NUCLEAR POWER STATION, R0WE. MASSACHUSETTS REPORT No. E-Y-YR-83033-4. REVISION 0. JULY 1984 l RESPONSE: YAEC WILL PROVIDE THE SUBJECT LETTER TO THE NRC. I I ; I I I I I I .. I I I
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440 g,- I k-112
- . . - _ - :: :: = =- - _ - -
l ITEM E12: I ISSUE
SUMMARY
EXPLAIN HOW SEISMIC ANCHOR MOVEMENT LOADS WERE COMBINED WITH OTHER LOADS.
REFERENCE:
NONE. I RESPONSE: SEE ITEMS 08 AND E2. I , f I l I 112
.._______.-.: ._ _ um _. _ . _ . .
- -
- _ = =
^
I ITEM E14.1: I ISSUE
SUMMARY
STATE WHAT INDUSTRY CODE AND REVISION ARE USED. ANSI B31.1 1973 AND 1977 I
REFERENCE:
ASME BOILER & PRESSURE VESSEL CODE. SECTION III, DIVISION 1, 1980 THROUGH WINTER 1980 ADDENDA l SEISMIC REEVALUATION AND RETROFIT CRITERIA FOR YANKEE NUCLEAR POWER STATION, R0WE, f1ASSACHUSETTS, 80023/81060/810G1 DC-1, REVISION 3, APRIL 1986 I RESPONSE: ANSI B31.1 1977 IS SPECIFIED IN THE RETROFIT CRITERIA. l ANSI B31.1 1973 IS USED FOR SOME COMPUTER ANALYSES. I I I I I .. I I m g
.- ^~~2 ~_=- --_ --- r -._ _ _ -- z; ' ~ -
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l ITEM E14.2: I ISSUE
SUMMARY
- STATE WHETHER THE INDUSTRY CODE AND REVISION USED CONFORM TO THE SEP REQUIREMENTS.
g
REFERENCE:
ANSI B31.1 1973 AND 1977 l ASME BOILER & PRESSURE VESSEL CODE, SECTION III, DIVISION 1, 1980 THROUGH WINTER 1980 ADDENDA SEISMIC REEVALUATION AND RETROFIT CRITERIA FOR YANKEE NUCLEAR POWER STATION, R0WE, g MASSACHUSETTS, 80023/810G0/81061 DC-1, REVISION 3, APRIL 1986 RESPONSE: ANSI B31.1 1977 IS SPECIFIED IN THE RETROFIT CRITERIA. I ANSI B31.1 1973 IS USED FOR SOME COMPUTER ANALYSES. ASME 1980 THROUGH WINTER 1980 ADDENDA IS I SPECIFIED IN THE SEP REQUIREMENTS. I COMPARISON OF THE TilREE CODES YIELDS THE FOLLOWING RESULTS: I
- STRESS EQUATIONS ARE IDENTICAL, EXCEPT ANSI l '
B31.1 IS SLIGHTLY CONSERVATIVE FOR PRESSURE STRESS IN COMBINATION WITH GRAVITY AND SEISMIC LOA 3.
- MATERIAL ALLOWABLES IN ANSI D31.1 ARE LESS Til AN OR EQUAL TO THOSE IN ASME 1980.
1m g
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-. - :: 1 I
- THE COLD MODULI FOR STAINLESS STEEL IN ANSI l B31.1 1973 ARE SLIGHTLY HIGHER THAN IN ANSI B31.1 1977 AND ASME 1980. BUT THE DIFFERENCE l IS NEGLIGIBLE. l THE C0EFFICIENTS OF THERMAL EXPANSION IN ANSI I B31.1 ARE CONSISTENTLY HIGHER THAN IN ASME 1980, GIVING HIGHER THERMAL STRESSES AND l LOADS. l l
- FLEXIBILITY CHARACTERISTICS AND STRESS INTENSIFICATION FACTORS IN ANSI B31.1 AND g ASME 1980 ARE IDENTICAL.
CONCLUSION: ANSI B31.1 1973 AND 1977 MEET THE REQUIREMENTS OF THE SEP PROGRAM. I I I i I ; I l I I u I
- --. ~~- .. ITEM E15: , ISSUE
SUMMARY
- CLARIFY MODAL COMBINATION METHODS USED.
REFERENCE:
USNRC REGULATORY GUIDE 1.92 l RESPONSE: 'l0DAL COMBINATION IS PERFORMED IN THE COMPUTER PP,0 GRAM ADLPIPE USING THE 10% BANDWIDTH GROUPING METHOD IN PARAGRAPH 1.2.1 0F USNRC REG. GUIDE 1.92. I I I I I l I I 116 l
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I I ITEM E16: I ISSUE
SUMMARY
- CONFIRM THAT SU IS USED AS AN ALLOWABLE.
SEISMIC REEVALUATION AND RETROFIT CRITERIA FOR I
REFERENCE:
YANKEE NUCLEAR POWER STATION, R0WE, MASSACHUSETTS, 80023/81060/810G1 DC-1, REVISION 3 APRIL 1986 I RESP 0llSE: SU IS USED AS AU ALLOWABLE FOR NON-SEISMIC PIPING TO INSURE THAT THE PIPING WILL NOT FAIL AND IMPACT THE SAFE SHUTDOWN PIPING. I I I I I I I I .. I I 11, g
=- - --- - - - . - _-
I SECTIDN D - PIPING SYSTEMS GUESTION D1 (NRC) As noted above under Section A.1, clarification regarding the hot chutdown scope analyses with the NRC spectrum is needed. I t
RESPONSE
Most of the hot shutdown piping systems were analyzed to the YCS only in the most recent analyses. The piping systems analyzed in the hot chutdown scope are those systems required to achieve hot shutdown, tnose required to maintain the primary coolant or secondary pressure boundaries, and those interact #.ng with c.ystems required to achieve hot chutdown. A list of the analyzed systems is provided in Table D1-1, including information on hot shutdown operational status (eitner required for hot shutdown or required to maintain pressure boundaries) and parties responsible for the most recent analyses for eacn system. In response to NRC concerns, confirmatory analyses will be performed to demonstrate that the YCS analyses performed for Yankee Rowe are accouace to assure that the piping systems can also meet SEP faulted criteria I under the LLL/ TERA or NRC earthouake. The NRC will select representative pipino systems with which to perform the confirmatory analyses. I l l 1 i I ! 118
-- .- -_. -. .. _ . . - - -- .-- . ~_---_
I TA~lLE D1-1 PIPING SYSTEMS IN THE HOT SHUTDOWN SCOPE I RESPONSIBLE PIPING SYSTEM DESCRIPTION CATEGORY ORGANIZATION t 1 Mnin Steam Inside VC Hot Snutdown Cygna/Impell Main Steam Outside VC - from VC Hot Shutdown Cygna penetration to anchor A-50 ; I Main Steam Outside VC - from anchor A-50 to turbine no::les Press. Bound. Cygna Boiler Feed Discharge Inside VC Hot Snutdown Cygna/Impell Boiler Feed Discharge Outside VC Interaction Cygna Pressuri:er Control a Relief Hot Shutdown Stone &Weaster Main Coolant Loops Hot Shutdown Cygna Snutdown Cooling Press. Bound. Cycna Safety Injection Hot Shutcown Cygna Main Coolant Bleed Inside VC Press. Bound. Cycna Main Coolant Bleed Outsice VC Press. Bounc. EAS Charging Linec Inside VC Hot Sn ut down Cvona Hot & Cold Leg, Pressuri:er, Feed & Press. Bounc. Cyana Bleed Heat Exchanger, Charging, Bleed, High Pressure Sample, Safety Injection, and Steam Generator Drains and Drain Header Reactor Coolant Loop Safety Valve Press. Bounc. Cyana Discharge Cnarging Relief Press. Bounc. Cvona Pressurl:er and Auxiliary Pressurl:er Press. Bounc. Cygna Spray Pressuri:er Sample & Vent Press. Bound Cycna Steam tienerat or Blowcown Hot Shutcown Cycna Cyona I Reactor Coolant Pumo, Reactor Coolant Loco Bypass, Steam Generator, and Feed & 141ued Heat Exenanger Vents Press. Bound Reactor 8 Pressure Vent Press. Bound YAEC 119
W ~ ~ 1 2 _ - 2" ._ __ _ _ _ _ _ _ _ _ TABLE D1-1, CONT' D RESPONSIBLE PIPING SYSTEM DESCRIPTION CATEGORY ORGANIZATION t Steam Generator Wide Range Level Hot Snutdown YAEC Reference and Variable Legs Steam Generator Differential Pressure Press. Bound. YAEC Flow Steam Generator Narrow Range Level Press. Bound. Cygna Pressurizer Wide Range Level Hot Shutdown Cygna Pressurizer Narrow Range Level Press. Bound. Cygna Pressuri:er Pressure Hot Shutcown Cygna Bleed Line Flow Press. Bound. Cygna Reactor Coolant Loop Pressure, Press. Dound. Cygna , Reactor Coolant Loop, and LTOP Instrumentation Tubing l t I l i I
/
120 f l
SECTION D - PIPING SYSTEMS QUESTION D2 (NRC) No discussion or evaluation of results of the pipe succorts were presented. The licensee should provide complete details regardinn rnethods used and representative calculat ions for review.
RESPONSE
The criteria and rnethodology for the evaluation of pipe supports is provided in the retrofit criteria document (DC-1). Tnis dnrument I outlinen the desinn bases and allowables for each 5,u o nor t com;.onem . The retrofit criteria doc urne nt has been recently revised to provice clarificationn and addi t i rinal detalle as recuented by the NPC. YAEC is in the process of n.iramari; ing the suponet evaluetion resultu mf roarecentative hot sh'it dnwn E v e, t ern e; tnr t he NRC' r, review. This t oimes will indicate the critical component for each c.uocort. the d e -% ) ty n I rnar e i n. actual ca and anv ether e.iun1fIcant ooints in the a nca l y s i e, rethnd. l c u l a t i .ine: can be reviewed durino the upcor ti ne NRC avol t. . The I I I l I 121 l l 1
nn ___;_._. . . _ - SECTION D - PIPING SYSTEMS QUESTION D3 (EG&G1.1) The potential for imoact loacing of succorts subject to ozoe uolift should be addressed. RESPONSE . Aooencix J of tne retrofit criteria coeuroent (DC-1) contains tne rnetnocolocy usec to consicer rod nanoer uplitt. Ynis accresses tne effect of tne increased rnornent urn of the oicino near uoliftino succorts on Doth ospe stresses and oice succort loacs.
.=
l ' 122
- _ _ _ ^ ~ ~ ~ ~ ~ " "~~~~~:
^^
=_
? - L. - SECTION D - PIPING SYSTEMS QUESTION D4 (EG8G1.2) Information snould be provided which clearly describes the raodelino ' m2thoos used to account for suoport stiffness. l RESPONSE r The supports for all large bore safety-related oioing systems were raodeled as rigid restraints, except for the main steam /feedwater cloing outside the vapor container. The comoat ibility of the riald restraint cssumotion with the supoort conf 1puration was enecked by evaluatino tne f undataent al frequency of the support under tne ceston loaos. All supports had frecuencies above 33 Hr. As an exaraple. the support stiffnesces and frecuencies of tne four newiv installed sucoorts for boiler feed d i scharte Proolern w024 insice the vapor container are presented in Taole D4-1. As snown in tne taole. tne I difference between the actual supoort stiffnesses and the stiffnesses u4eo in tne annivsis would not have a sionificant effect on tne analvsis results. The suopoets for all email bore safety-relatea oloino systems were also moaelec as rigid rest re i nt s. For all existino succorts. stiftnesces In most I were c casos,neePed against m i n t raura values set for eacn ol oe si ne. the ac';ual supoort stiffnesses far exceeded t he min 1raum volues, wit n frequencies e r c er.co i n a 33 H.. e For all new anc modifleo succorts.
<j e f l ec t i o r.9 in the restrainec direction were checkea to limit tnem to 1/10".
I I .. I 123
~' - - - - - - -,- - . . _ _ _ . __ _ _ _ _ _ TABLE D4-1 I SUPPORT STIFFNESS AND FREQUENCY BOILER FEED DISCHARGE PROBLEM #024 ANALYSIS ACTUAL FUNDAMENTAL j 7 NODE NO. STIFFNESS STIFFNESS FREQUENCY SUPPORT NO. I BCD-024-H1 14 3.0E7 5.5E5 35 Hz 3.0E7 4.5E5 40 H: BCD-024-H2 18 24 3.0E7 5.6E5 41 H: BCD-024-H3 24 3.0E7 5.6E5 36 Hz BCD-024-H4 I 124
SECTION D - PIPING SYSTEMS QUESTION D5 (EG&G1.3) Clarif1 cation should be provioed regaroing the correlation of tne resoonse soectra nurneers used in Volurne 1 of tne licensee's saf ety related oiping analysis report versus the clot nuracers usea in Book 2 of I Volurne 2 of the sarne reoort. 7
RESPONSE
In roost of the cloing analyses, cesion soectra for eacn of tnree ortnogonal alrections are oevelooed wnlen enveloo all aoolicacie scectra for the succort ooints in tne olpino sy st era. This is a st ancarc ano conservative roetnod for oer forraing resoonse soectrurn analyses. In some cases, tne raultiple level rescor.se soect rura raet noa is usec. inis raetnoa I orovides alfferent s o e ct ra 'i inout at various levels. The resconces for the rou l t i o l e level inethoc are coroolnect in accoraance witn tne stancarc tor? t h o c outlinea in r<UREG-1061. Voi urne 4. A;1 coalve.oc in tne subiect EPfety re ateo pioint resort were Derformec t.L t. I t i.3 O nv e i o t..-o soectra. I l g. 125 I
t _ _ . SECTION D - PIPING SYSTEMS QUESTION D6 (EG&G1.4) The appropriateness of the response spectra based on high values of structural damping should be justified; otherwise, pioino analyses using I spectra based on the appropriate structural dampinq values I should be completed and submitted for review. I RESPONSE This will b r- resolved tsy t he st ruct ural ana.vs2s a rvu o. I iI i l I 126 l l
I SECTION D - PIPING SYSTEMS QUESTION D7 (EG&G1.5) i l The origin and appropriateness of the diesel generator buildino spectra should be clarified and justified. I RESPONSE This will be resolved by the structural analysis aroup. I l l I ! I ! I I .. 127
.va ...
SECTION D - PIPING SYSTEMS QUESTION D8 (EG&G1.6) The method of combining stress terms should be clarified.
RESPONSE
The criteria and methodology for the pipe stress evaluation is provided in the retrofit criteria document (DC-1). This document outlines the load combinations for the ANSI B31.1 equations used in the evaluation end allowables for each stress equation. The retrofit criteria document has been recently revised to provide clarifications and additional details as requested by the NRC. Basically, separate analyses are performed for the following load cases:
- 1. Gravity + Pressure (G + P)
- 2. Thermal Expansion + Thermal Anchor Movement (T + TAM)
I 3. Seismic (YCS) - Seismic analysis is performed for each of three orthogonal directions by response spectrum Modal combination analysis methods, using the YCS. for each direction is performed according to USNRC Reg Guide 1.92. The total response is achieved by the SRSS combination of the response of the three I 4. directions. Seismic Anchor Movements (SAM) In the majority of analyses, the following load combinations and allowables are used. A ' + ' indicates combination by absolute summation. ANSI B31.1 Equation Load Combination Allowable Stress 11 G+P 1.0 Sn 12 G+P+S 1.8 Sh 13 T + TAM + SAM Sa I 14 G + P + T + TAM + SAM Sa + Sh i E In a few isolated analyses, the seismic and seismic anchor movement l stresses were combined by SRSS and considered in Equation 12. l 1 i i I 128 l l l l
~~ ' ' ~ ~ ' ~ - ~ ~ ~ ~
SECTION D - PIPING CYSTEMS GUESTION D9 (EG&G1.7) The temperature bases of the stress limits .anould be clarified.
RESPONSE
The stress allowables for ANSI B31.1 Equations 11 through 14 are obtained for the applicable material from tables in the Appendix A of the same code. The tnaterial type and design temperatures, from wnten the allowables and other piping properties are derived, are tabulated in the Yankee Rowe Soecifications for Pipino, YS-497 (J. O. No. 9693) and I YS-4652 (J. O. No. 11986.01). The retrofit criteria document (DC-1) has been updated to include these references and a clarification of the derivation of allowable stresses. I 129
.,_,_-----a__-
SECTION D - PIPING SYSTEMS QUESTION Die (EG&G1.8) The differences in stress results shown in Tables 5.5-1 and 5.8-1 of tne licensee's piping analyses reoort (Volume 1) when no rnoci ficat ions or additional data are i nd icat ed should be exp1ained. t
RESPONSE
Refinements in the roodels or corrections Additionally, of rninor rnodel deviations were t n errna l an:1 rnad e, and the systems were reanalyzea. seismic anchor rnotions were redefined. Therrna l and t nerrna l anenor raat ions were analyzed tooether. 130 I
a SECT 20N D - PIPING SYSTEMS QUESTION D11 (EG&G1.9) The difforence in Equation 12 results between Tables 5.G-1 and 5.8-1 shown in the l i censee' s piping analysis report for probl era nuraber 3 (YCS results) should be explained, t
)
I I RESPONSE This issue will be addressed at the t irne of the NRC audit. l I .. 131 I
___u- __ I SECTION D - PIPING SYSTEMS QUESTION D12 (EG8G1.10) Results for the NRC meistnic loadina should be provided for probleras 2, 3 9 and 23. I
RESPONSE
VAEC is analyzino all pipina under YCS loads to ernernency allowables. , Selected systerns will De analyzed for NRC loads and corapa r ed These t .:. f athree ui t erf ' allowables to confirra the raitivalence of t he t wo roet hods. syst erns are not currently in the conf 1rrnatory analysis K o e.- : thocef.re. the reauested results are not available. I lI 1 I .. I I 132
I SECTION D - PIPING SYSTEMS QUESTION D13 (EG&G1.11) I Section 5.11 of the safety related piping analysis report should be provided for review. I
RESPONSE
The subject section of the safety related 01pino analysis reoort i t: attached for the NRC's review. I I I
- I I
I 133 l -_ - -
5.11 Stress Sumary (Hot Shutdown Systems) with Proposed Restraints After the September,1982 outage, Cygna reviewed the as-built condition of the I Yankee Rowe hot shutdown piping systems. The results are sumarized:in Tables Table 5.11-1 shows the stress levels of overstressed 5.8 1 and 5.9-1. analyses with the Cygna-proposed restraints installed. I l l l [E3 ?? Q Yankee Atomic Electric Company L PT ri O;M" I181060,80023 6 Piping Analysis, Vol. One A 11111111l111111111111111111111 134
E E I l
?
TABLE 5.11-1 i , STRESS TABLES (H0i SMUTDOWN SYSTDt5 WlIN PROPOSED RESTRAlnis) i MAIIMUM MAllMUM - COMPUTED STRESS (PSI) ALLOWA8LE STRESS (PSI) MAtlMt94 COMPUTED STRESS (PSI) i BT E00AT10M5 87 E00 Ail 0M5 ON DIFFERENT LOAD CASES C(posENis EQ.lt EQ.12 EQ.13 EQ.14 E0 11 EQ.12 EQ.13 EQ.14 SE00ENCE ANALT515 N). PRESS. GRAvlif THERMAL SEl5Mit SAM TAM ! NO. YC5-5pectra 5949 12300 11190 17139 15000 27000 22500 37500 6986 6351 4205 -- MC-5pectra 002 a830 1119 l l 5949 17576 15513 21462 15000 36000 22500 37500 6986 !!627 8578 -- YC5-5pectra 2 002 4830 til9 15000 21000 22500 37500 11136 14820 6114 - 6270 21090 17250 23520 WC-Spectra 003 4830 1440 3 6270 29818 23511 29781 15000 36000 22500 37500 1440 18136 23548 12375 -- TC5-5pectra 003 48 30 4 6564 9348 29910 34893 15000 27000 22500 37500 1821 23323 2784 10101 -- WC-Spectra
- 5 022 4744 15000 36000 22500 37500 4528 21549 -- 6564 11092 29910 34893 022 4744 1821 23323 YC5-Spectra 6
7145 11580 22395 27416 15000 27000 22500 37500 20849 4435 5907 -- MC-Spectra 023 4744 2401 7 1145 15186 22395 27416 15000 36000 22500 37500 2401 20849 8041 11784 -- 8 023 4744 l 1 1 l 1
- ;gM;t343 Yankee Atomic Electric Company 7, Q $ N i Piping AnalyS15 Vol. One s.1 U a a a' 81060, 80023 lll!!I111111111111111111111111
E SECTION D - PIPING SYSTEMS QUESTION D14 (EG&G1.12) I Detailed informatiori regat' ding the roethodole.gy used and reoresentative calculation packages for review should be provided for the piping supports connected to the safety related piping systems, t
RESPONSE
i See Ouestion D2. I I 130
w -_ - . _ _ . _ .. .. - _ . _ .
- - _ . - _ _ _ ...- ~.. . . _ _
8 SECTION D - PIPING SYSTEMS QUESTION D15 (EGAG1.13) The licensee should provide evidence tnat all reaut rernent s of I&E Bulletin 79-02 are satisfied. t
RESPONSE
IEE Bulletin 79-02 recurres tnat baseolate flexiollity ano its effects on increasing anchor bolt loaas due to orving De consicerec in the cesion of coraDonent suoDorts. Tne suoiect bulietin has four reautrements:
- Prvino action s'1all be consioerea.
- A safety factor of four shall oo useC on WeCce ano sleeve tyDe anenor colts, fe safety factor of five cnall be usec on snell tvoe ancior o, ,l t s.
- Cveiic loaCs tna11 be consicerec, as aDolicaD1e.
- OC OoCurderstation of CyC11C loaCs and installation.
I'r y i n n aCt lein nas cee rs C o rin 1 C e r eu for all Cafe GnutcoWn sVCte? C;se SuGDOrts at 'Y N P S . For G h a fi. D l e. for the lurte DorO D1DC C U C D or t '3. no rs l i near DaGe Olate arialyCet WerC perforfdeo for TV01 Cal base Diate Conf;?drations t o Cet erful ne t ne Crlt1 Cal lo&CinD C1PECtions anC CEGICn
*DaOs. Tr i e C e TVL1Cai Con!13 Mrat 10rit 01 0 allOWaDie j'oOCS WCre UCeC fCr 1: , 1 fig s C 'J O D Or t 9. T D '? e n Ys OF D*2TG Were e%31UTte: With the G 3 f C t '.
felC*:C rf C r s'C C P 1 D C C 1 r. bullrtin 79-C.
- F. C C '.*i l V CVC1IC l o E. C s &CL)16C to tne D1DC 'i l 6 D O T.T T C in tnt EOfe C n u t C C r. n C .' De St e C JP to i n c .' fi1a l and C e l sfd 1 C CCriC i t lons. 10000 C 'v C 4 1 C .1502. 3?C CerHPai1V ef 1oW ft*GCuencv arc have no effect on t se ar.Cnor D?2 t a l l NI A D i eL G OeC . f 10'd in tne VenCor C a C a l L t's if tne orCDO" Drelo3C G :~. !
I 2DCtallat130 C.Jec t ' 1 C E.t l onG G r e fise t .
%- !" c P C C'.a r e Y e.'at o- P.ti1etin 7 9 - 0 2 W .3 e- tnorefcre n *- O C;. c r.E F_ c.
thC C e 5E i t r. 210 :.' a T '. C n . I n u 5. . the tn1r0 e riC fourtn r e c u i r elet.e n i t of I E' . s e t 1 re 79 U. 4E C o .'e r e d D. 'Y lit'C ' 5 OC D ro C P Ls f i at MDWO. I 137
V -= = a_= = __.__._.___.__..._ .___ ~_ SECTION D - PIPING CYSTEMS QUESTION D16 (EG&G1.14) I The licensee should clarify hot shutdown systern boundaries and their ccsociated stress problems. .
RESPONSE
A list of. hot shutdown problerns was provided in response to question Di cbove. The associated stress probleras for each line are indicated b71ow. Large and srnal l bore probleras are included in the list below. Additional work has been done since issuance of the subject reports. The first set of references for each problera corresponds to the latest to date. Where those references Icnalysisofrecordorstressreportto di f fer frorn those used generate results of the safety related piping report, the srnall bore piping report, and the roain stearn/feedwater piping report, a second set of references are provided in parentheses. I These references correspond to the original Cygna calculations whose results are docurnented in the abovernent ioned reoorts. REFERENCE CALCULATION PIPING SYSTEM DESCRIPTION I Main Stearn Inside VC - Loop 1 Problern #001 Calculations 83033-28/F and Cygna 83033-1/F (Cyyna Calculation 83033-1/F) Probl ern #002 - I rn oel l I Main Stearn Inside VC - Loop 2 Report 09-0570-0045 (Cyana Calculation 81060-2/F) Main St earn Incide VC - Loop 3 Problern #003 - Iranell Recort 09-0570-0045 (Cyana Calculat t on 81060-3/F) P rob l ern #004 - Cygna I Main Stearn inside VC - Loop 4 Ca1culations 83033-28/F and 83023-4/F (Cyona Calculation 83033-4/F) Prob l era s v005 and Main Stearn Outside VC - f rorn VC oenetrat ion #008 - Cvgna to* anchor A Loops 1 and 4 Calculatton 83033-14/F Probl erns #006 anc* Main St earn Oututde VC - f rorn VC penet rat ion #007 - Cygna to anchor A Loops 2 and 3 Calculatton 83033-14/F
~
I
_ _ . . _ _ _ , . _ . _. m._._
- , _ .. n ... --
E REFERENCE CALCULATION PIPING SYSTEM DESCRIPTION I Main Steam Outside VC - from anchor A-50 to Problem #009 - Cygna Calculation 83033-turbine nozzles 14/F Coller Feed Discharge Inside VC - Loop 1 Problem #021 - Cygna Calculations I 83033-28/F and 81060-12/F (Cygna Calculation 81060-12/F) Boiler Feed Discharge Inside VC - Loop 2 Problem #022 - Impell Report 09-0570-0045 (Cygna I Calculations 83033-2/F and 81060-13/F) I Boiler Feed Discharge Inside VC - Loop 3 Problem #023 - Imoell Report 09-0570-0045 (Cygna Calculation 81060-14/F and 80023-26/F) Problem #024 - Cygna I Boiler Feed Discharge Inside VC - Loop 4 Calculations 83033-28/F and 81060-15/F (Cygna I Calculation 81060-15/F) Boiler Feed Discharge Outside VC - Loop 1 Problems #25A and
#25B - Cygna Calculation 83033-23/F (Cygna I Calculation 83033-23/F)
B oiler Feed Discharge Outside VC - Loop 2 Problems #26A and I #26B - Cygna Calculation 83033-23/F (Cygna Calculation 83033-l
= 23/F) i Boiler Feed Discharge Outside VC - Loop 3 Problems #27A and
#279 - Cygna I Calculation 83033-23/F (Cygna Calculation 83033-23/F)
Bo11et!, Feed Discharge Outside VC - Loop 4 Problems #28A and ( #288 - Cygna Calculation 83033-l l 23/F (Cygna [W Calculation 83033-23/F) l 139 f 1
2-
---r-___ ,_;
~ ~-_ _ _
PIPING SYSTEM DESCRIPTION REFERENCE CALCULATION Problems #41A, #41B, Pressurizer Control & Relief and #41C - Stone & Webster Report B2-1198618-1 (Cygna Calculations 81060-20/F, 81060-21/F, and 81060-22/F. Main Coolant Loop 1 Problem #101 - Cygna Calculations 83033-28/F and I 81060-23/F (Cygna Calculat.on 81060-23/F) Problem #102 - Cygna I Main Coolant Loop 2 Calculations 83033-28/F and 81060-24/F (Cygna Calculation 81060-24/F) Main Coolant Loop 3 Problem #103 - Cygna Calculations I 83033-28/F and 81060-25/F (Cygna Calculation 810E0-I Main Coolant Loop 4 C5/F) Problem #104 - Cygna Calculations 83033-28/F and 81060-26/F (Cycna Calculation 81000-26/F) Shutdown Cooling from Main Coolant Loop 4 Problems #121 and
#122 - Cycna
!g through isolation valve to VC penetration Calculations g 85037-4/F, 83033-28/F, 83033-3/F, f I and 83033-4/F (Cygna Calculations ! 83033-2.'F and 63033-4/F) l lI 140 i
T- ~
~~- ~ :X _
I REFERENCE CALCULATION PIPING SYSTEM DESCRIPTION I Safety Injection from Main Coolant Loops Problems #201 and
#207 - Cygna through isolation valves to VC penetration Calculations 85037-11/F, 85037-4/F, 83033-28/F, 83033-5/F, 81060-I 29/F (Cygna Calculations 83033-5/F and I 81060-29/F)
Small Bore Problem Mnin Coolant Bleed Inside VC from Main Coolant I Loop 1 to Feed & Bleed Heat Exchanger E-8-1 #13 - Cygna Calculation 84060-16/F I Main Coolant Bleed Inside VC from Feed & Bleed Heat Exchanger E-8-4 to VC penetration Small Bore Problem
#14 - Cygna Calculation 84060-17/F EAS Report Main Coolant Bleed Outside VC from VC I penetration through the isolation valve to the low pressure surge tank Small Bore Problems I Charging Lines Inside VC
#2, #3, and 67 -
Cygna Calculations 84060-3/F and 840E0-8/F Hot & Cold Leg and Pressuriner Drains and Small Bore Problem #7
- Cygna I Drain Header Calculation 840~0-9/F Feed & Ble'>d Heat Exchanger, Charging, and Small Bore Problera #1 Safety injection Drains - Cygna Calculation 84060
-2/F Charging and Bleed Drains and High Pressure Small B.re Probleo #6 I Sample - Cygna Calculatton 64060-6/F Cygna Calculations Steam Generator Drains
" 84000-Rt/F, 640CD-1 22/F, and 84000-I 23/F Reactor Coolant Loon Safety Valve Discharoe Cygna Calculation Loops 1, 2, and 3 84060-20/F 141 1
'm ~_ _
.r.__ C ._ _.
_ _ . .____Z nn_ .. ._ . REFERENCE CALCULATIDN PIPING SYSTEM DESCRIPTIDN I R; actor Coolant Loop Safety Valve Discharge Small Bore Problems
#2 and #3 - Cygna Loop 4 and Charging Relief Calculation 84060-I 3/F Small Bore Problems IPressurizerandAuxiliaryPressurizerSpray #2 and #3 - Cygna Calculation 84060-i 3/F Pressurizer Sample & Vent Small Bore Problem #4
- Cygna Calculation 84060-4/F Small Bore Problems Steam Generator Blowdown #9, #10, #11, and
#12 - Cygna Calculations 84060-12/F, 84060-13/F, 84060-14/F, and 84060-15/F I Reactor Coolant Pump Vents Cygna Calculation 84060-18/F I Reactor Coolant Loop Bypass Vents Cygna Calculation 84060-19/F Cygna Calculation Steam Generator Vents 84060-24/F Cygna Calculation I Feed a Bleed Heat Exchanger Vents 84060-32/F YAEC Calculation YRC-
& Pressure Vent l I Reactor 65 YAEC Calculation YRC-Steam Generator Wide Range Level Reference 406 (Cygna f
Leps Calculation 84060-25/F) Steam Generator Wide Range Level Variable YAEC Calceelation YRC-406 (Cyrna Legs Calculat ton 84060-26/F) YAEC Calculation YRC-Steam Generator Differenttal Pressure Flow 406 (Cygna Calculation 840J0-27/F) 142
rf" .__L_.~-.--..~_.~_
. - - = . . _ - - _. .-. . . _
REFERENCE CALCULATION PIPING SYSTEM DESCRIPTION I Stearn Generator Narrow Range Level Cygna Calculation 84060-28/F Pressurizer Wide Range Level Cygna Calculation 84060-29/F Pressurizer Narrow Range Level Cygna Calculation 84060-30/F Cygna Calculation Pressurizer Pressure 84060-31/F Cygna Calculation IBleedLineFlow 84060-33/F Reactor Coolant Looo Pressure, Reactor Cygna Calculation Coolant Loop, and LTOP Inst rurnent at ion 84060-38/F Tubing I I i i I .. I 143
SECTION D - PIPING SYSTEMS QUESTIDN D17 (NRC)
- 7) presents the stress results term-by-term and the Tcble 5.5-1 4Ref.
total equation results. In many cases, such as for Ecuation 12, produce the listed equation f I combining ecsult. the appropriate terms does notThe licensee should describe the proc ' the results shown in these t ables.
RESPONSE
The stress results for the individual load cases represent the maxirna for all node points with aporopriate stress intensification factors. The stress results for the ecuations represent the system maxirnurn stress All load cases are not Ictthemost highly stressed node point. nrcessarily maximized at the same node point; thus the individual ctresses do not always add up the reported cornbined ecuation stresses. 144
~~
_ _ . . . . _ _ _ _ = _ ___. __ _ SECTION D - PIPING CYSTEMS GUESTION D18 (NRC) The allowables listed for problem 207 in table 5.5-1 and 5.8-1 are shown b]Iow 11 12 13 14 15900 28620 27350 43250 5.5-1 5.8-1 15900 31365 27975 46375 Also, the sum of the pressure, gravity and seismic ' loads in Table 5.8-1 (5950 + 6960 + 5118) does not total the Ecuation 12 result (21220). The differences should be explained. I RESPONSE In table 5.5-1, the allowables are given for TP304 stainless steel at 650 degrees, which is the maximum design temperature on the entire problem. In actuality, only the piping from the main coolant 10o0 up Most of the piping is at 130 the the isolation valve is at 650 degrees. degrees during normal ooeration and at 225 degrees during a LOCA. Since the Ecuation 11 stress was maximized on the portion of piping from the main coolant loop up to the isolation valve, the allowable stress at 650 degrees was reported. For the other ecuations, the stresses were on the piping past the isolation valve. Thus, the allowables I maximized recorted were obtained from ANSI B31.1 at a temperature of 225 degrees. The stress for Ecuation 12 does notbecause eaual the thesum of the ofindividual I ctresses for the three load cases, load cases do not occur at the same node point. more detail in the response to Question D17. maxima the individual This is explained in 145
- ^
L- ~.T._ ._ .
~
_-_~ _ __. . _ _. SECTION D - PIPING SYSTEMS CUESTION D19 (NRC) For problems 041A and 041B, the seismic stress for NRC loading is less this should be checked. IthanforYCS;
RESPONSE
For problems 041A and 041B, the major seismic response is due At these to the frecuencies, vartical component in the first vibrat ional mode. the YCS value is higher than that of the NRC soectrum. Although the I horizontal NRC spectra are higher than those of the YCS, frecuency modes, which are not significant, contribute to tne total only higher response. The spectral information for the first vertical moce is given I in the following tables FREQUENCY YCS ACCEL NRC ACCEL PROBLEM NO.
.96 g .87 g
#41A 12.183
.88 g .77 g
#41B 9.824 I
i l I .. 146
._ ___i - ~ ~ _ _ __ _ - _ _ __
I SECTION E - MAIN OTEAM AND FEEDWATER (MS/FW) PIPING AND CUPPORT STRUCTURE QUESTION E1 (NRC) The detailed information about the following iterns should be provided . f;r reviews
- c. A complete description of the general modeling technique used. This information should also describe techniques used to represent items I such as fittings, supports, and in-line eauipment. ,
- b. The stiffness values used f o t' anchor A-50 and the methods used to calculate these values.
- c. Details regarding the mass, stiffness, and frecuencies for vaoor container, turbine building and support structure, which are included in the piping rnodels.
- c. Details of the analysis methods, and the adequacy of breaking the I entire subsystem into many sub-problems.
RESPONSE
- c. The rnain steam and feedwater piping and support structure is a large system of interacting piping, pioe supports, anchors, and structural steel frames, and buildings. The flexibilities of the support structure and the large plaing are of roughly the same orcer; therefore, the standarc analysis methods of assuming rigid sucoort I points is not valid.
analyzed in its entirety. Because of its size, tne system cannot ce Therefore, portions of tne pipino and sucoort structure are analyzed separately, accounting for masses ano stiffnesses of interacting piping and structures. An iterative solution is useo, where results of the seoarate analyses are used to judge whether the entire system in any one iteration The has been adequately rnodeled to produce compatible results. s t er'at i ve I process is described in detail below. In each iteration, piping analyses are first performed to obtain support reactions to evaluate the support structure. To mocel tne bouncary condit ions of the piping problems, unit load analyses are first c e r f o r rn e d by the structural analysis grouo to obtain tne stiffnesses at each existing and proposed support location and for I each restrained direction. This includes anchor points on tne vaoor container and turbine building as well as on the succort structure. For example, six stiffnesses would be calculated for a 6-wav ancnon-I support. To properly capturu the fundamental vibrational frecuency of the buildings and support structure. ecuivalent massen calculated and applied at the support locations. are I n t n i s reanner, a set of springs and masses t'epresent i ng the supporting point s are input at each restrained node. The details of the spring-mass determination as well as the specific results fot' the buildings anc other locations on the support structure can be found in Cygna Calculation 83033-18/F. The eautvalent soringu and rnasses account for the flexibility of tne I buildings and support structure. The YNPS site-specific ground 147
W. ^
~
--.--- - -. == _- . _ __ L - . _. _- .
I cpectra cra ured in tha piping cnolycio inntoed cf cmplificd : building response spectra, since the groundInmotion additionis appropriately to providing - I arnpli fied through the spring-rnass models. the appropriate inertial input to the pioing system, the spring-mass models also allow the support points to deflect under the seismic , Seismic anchor movements (SAM) are therefore included in I loading. the inertial analysis, and a separate SAM analysis is not recuired. This accounts for the noted lack of SAM analyses for rnain steam problerrn #005 through #008. Only a srnall portion of problem #003 I (from Anchor A-50 to the Turbine building) is supported by the main steam /feedwater supoort structure. There supoorts in this portion: Anchor A-50 and are two seismic a prcrosed N-S restraint I at Hanger H-60. After roodification, these two supports will be firmly connected to the reinforced concrete wall along coluta line J Therefore, a standard analysis was of the turbine building. per fortned for that problem, using the approcriate enveloped I arnpli fied resoonse spectra and seismic anchor movements. With these special boundary conditions in place, the usual analyses I are performed. Modeling of the rernaining piping components and ec ul prnent are as described in the retrofit the safety-related piping analyses. criteria Pipe stress and as results and used in succort If the pipe stresses are not I reactions are then generated. acceptable, the model is altered by relocating supoorts (with new rnass and stiffness values for the new locations) and reanalyzed and until stresses are accettable. The pipe reactions are tabulated I t ransta l t t ed to the structural analysis group for use in evaluation of the support structure. to the sucoorting The pipe reactions are statically applied structure at the support ootnts to obtain member and cornoonent for stresses. Additional st at ic analyses are per'fortnea to accountThe self-weight and inertial effects for the support structure. I t ot al stresses are cornoared to the aopropriate criteria, aceouacy of the supoort structure is ceterrnined. and the If the stresses in tnen tne the support structure are within the allowable values, I piping and structural analyses are compatible. structure provides adeOuate restraint are witnin the allowable ranges. The succort for the p1p1hg wnlle stresses If the supoort structure stressec are not within the allowable values, foodifications are proposed, and another iterat ion is perfortned. In the second iteration, the stiffnesses at the succort locations I are again obtained for the rnodified support structure. structural analyses proceed as desct'abad for the first iteration. Ploing and The sterations are repeated until a compatible solution is achievec. I Statte unit loads were soplied to tne mathematical rnodel of anenor b. A-50 to obtain the stiffness values for each of the six restraineo Equivalent enasses were calculated to properly rnodel tne funcamental frequency response of the anchor f r ata e in each d irec t i ons. restrained direction. Two analysis iterations have been perforrned to cate on tne rnain Mod s f a cat sons to st earn / feodwater pleing and support Unit steucture. load anal yses were perf orroed I anchor A-50 have been identified. 148
h-2 . __ _ . _. _ . - _-- _ . _ ._r__ _ _ I - -. . for the 1ctoct cnchor c nfigurction to datcernina tho cquivolcnt stiffnssses and roasses for each of the six restraints of the anchor. I The values are provided in Table El-1.
- c. Veggr_G9DigiDet ,
I The specific details of the determination of mass and stiffnesses for the vapor container are provided in Cygna Calculation 81060-In this calculation, a sirnplified structural model of the I 34/F. vapor. container is used to provide global stiffness and ecuivalent mass values for five different elevations on the vapor container. Since the exiting rnain stearn and feedwater pipes are relatively j large with respect to the thickness of the vapor container shell, I local stif fnesses and masses are also calculated to more accurately model the behavior of the penetrations. The two sets of stif fnesses and masses are applied in series to the piping models at the penetration locations. The rnain steam and feedwater lines exit the vapor container at roughly the same I elevation; therefore, the global stiffness and mass values are eauivalent. The local stiffness values are also identical for the two lines, but the equivalent roass is adjusted fot* the differences The global and local stiffness and mass values used I an pipe size. to roodel the vapor container penetrations for the roain stearn and feedwater piping are provided in Tables El-2 and El-3, respectively. I9tDioe_R911d1Dn for the tur'btne bui1 ding are Stif fnesses and equivalent rnasses d et erroi ned for the following boundar'y conditions:
- Main steam /feedwater support structure anchors
- New restraints for feedwater pioing attached to the J-wall
- New anchor at the J-wall penetrations for feedwater piping For eacn of the above cases, sirnplified stick and ocard rnocels are Details and calculations I used to represent the turoine building.
are contained in Cygna Calculation 83303-18/F, The developraent of the boundary conditions for the new feedwater anchor at the J-wall penetration is discussed in this response. For development of the other boundary conditions, see Cygna Calculation 83033-18/F. For the feedwater anchors at the J-wall, a sirnilar aporoach G1 to theano I vapor cont ainer st i f f ness and rnass calculation local stiffnesses and roasses are calculated is taken. coal for the structural anc shell response of the turbine building and anchor plate at the wall One anchor as provided for all four penetrat ions. I penet rat ion. St'iffnesses and roasses for the enttre anchor assernbly are provloed in Table El-4. dato_StearnLEecdwatet_S9999tt_ Structure The rnain st earn / f eedwat er support structure st if f nesses and roasses 149
u - I cro eciculctcd fcr C ch pipa cupport cttcchmnnt locction. Whthen there are multiple pipes or supports on any support member, effect of all pipes is considered in developing stiffnesses and masses for each single location. e The calculation details and actual values for the support structure I stiffnesses and masses are provided in Cygna Calculation 83033-18/F. Some typical values used for the iterat ion 2 boundary conditions for i main steam and feedwater piping problems are presented in Tables Ei-I 5 and.El-6.
- d. As discussed in the response to question El.a., because of the size of the main steam /feedwater system, it is necessary to perform separate analyses on smaller subsystems. However, an iterative solution is employed to insure that the boundary conditions provided I by the support structure and the pipe reactions imparted to the support structure were comoatible and that all criteria reouirements The iterative solution devised for this system allows the were met.
I effects of piping / support coupling to be considered while performing analyses in manageable subsystems. Main steam croblems #005 and #008 were analyzed together, as were I main steam problems #006 and #007, to address the coupling of tnese pipes. The feedwater lines were analyzed separately, because, although the feedwater piping is also supoorted by the main I steam /feedwater support structure, the dynamic interaction between the main steam and feedwater piping is minimal. The feedwater pipino has few lateral suoDorts attached to the main steam /feedwater supoort structure. These lateral supports are generally locatecThe on I dafferent frames than those supporting the main located only exceptions are the oroooseo E-W steam pleino. at RH-129 and At restraint these locations, additional I PH-130 of feedwater Problem #026. diaconal braces the succort framewill be installed which to increase will consecuently the the reduce E-Weffect stiffnessof of proino interaction. The st if fness ef fects of multiple pipes on any I one supoort area are addressed in the static analyses to ceveloo the oice supoort boundary conditions. As st at ed previously, two iterations have been comoleted Any on tne main steam /feedwater piping and supDort structure analysis. additional coupling effects on the support structure or piping will be further addressed before the analyses are finalized. I I . l 150 l l
--...a_. g__ _- __ _
TABLE E1-1 ANCHOR A-58 MODELING ASSUMPTIONS DIRECTION STIFFNESS MASS X (E-W) 941 k/in O.152 k-sec2/in Y (Vert) 1417 R/in 11.5 lo-sec2/in Z (N-S) 1534 k/in 3.675 k-sec2/in RX 9.22x10 k-in/ rad 20.628 x-sec2-in RY 1.614x10 k-in/rac 192.471 k-sec2-in RZ 4.29x10 k-in/ rad 12.465 k-sec2-in I I I 151 I
I TABLE El-2 MAIN STEAM AND FEEDWATER VAPOR CONTAINER PENETRATIONS - GLOBAL EFFECTS DIRECTION STIFFNESS MASS X (E-W) 1065 k/in 8170 lo-seca/sn Y (Vert) 39746 k/in 8170 lb-nec2/in Z (N-S) 1059 k/in 8170 lo-sec2/1n RX 3.21x10 k-in/ rad 296000 .b-sec2-in l RY 7.83x10 k-in/ rad 296000 lo-sec2-in RZ 1.63x10 k-in/ rad 296000 10-seca-in l .. 152
t .._ --- .- . - . - - - . - _ - - I TR$LE El-3 MAIN STERM AND FEEDWATER VAPOR CONTAINER PENETRATIONS - LOCAL EFFECTS I DIRECTION STIFFNESS MS MASS FW MASS X (E-W) 1634 k/in 0.98 lb-sec2/in 0 3d'1b-sec2/in Y (Vert) 837 k/in 0.98 lo-sec2/in 0.32 lo-sec2/in Z (N-5) 2077 k/in 0.98 lb-sec2/in 0. 3d lo-sec2/2n 1 RX 16636 k-in/ rad 115.25 lo-sec2-sn 12.29 lo-sec2-in RY 20223 k-an/ rad 115.25 lo-sec2-in 12.29 lo-sec2-in ! 1 RZ 12314 k-in/ rad 115.25 10-sec2-in 12.29 lo-sec2-in l l I 153
'"~
TABLE E1-4 NEW FEEDWATER ANCHOR ON TURBINE BUILDING J-WALL I DIRECTION GLOBAL STIFFNESS LOCAL STIFFNESS MASS X (E-W) 4.95x10 k/in Infinite 79.7 k-sec2/sn 1.62x10 k/in Infinite 22.3 k-sec2/sn Y (Vert) Z (N-S) 35500 k/in 6707 k/in 85.1 k-sec2/1n 1.08x10 in-k/ rad 27500 in-k/ rad 3.45x10 k-sec2-sn RX I RY 2.37x10 sn-k/ rad 239100 in-k/ rad 8.70x10 R-sec2-in 1.39x10 in-k/vad Infantte 3.45x10 k-sec2-in RZ I ** I 154
.~ _
TABLE E1-5 MAIN STEAM PROBLEM #006 BUPPORT NO. DIRECTIDN STIFFNESS MASS New 9 2228 X 147.3 k/in 1.0 hu-sec2/in Y RH-96 Y 355 k/in 4.97 lo-sec2/in New @ 413 X 81.6 R/in 3.98 lo-sec2/in I Y 52.7 k/in 1.43 lo-sec2/in MSVD-4 2 H-90 Y 33.3 R/in 0.28 lo-sec2/in H-89 Y 10.1 k/in 0.72 lb-sec2/in H-88 Y S8.8 R/in 1.53 lo-sec2/in H-87 Y 11.7 k/in 0.32 ID-sec2/in 155 I
TRILE E1-G FEEDWATER PROBLEM #25A SUPPORT NO. DIRECT!DN STIFFNESS MASS H-137A Y 14.4 k/in 0.21 Ib-sec2/in
- 40. k/in 1.33 lo-sec2/in I RH-132 X Y 35.4 k/in 0.13 lb-sec2/in RH-131 Y 8.8 R/in 0.55 lo-sec2/in I RH-130B Y 35.4 k/in 0.13 lo-sec2/in RH-130 Y 95.3 R/in 0.31 lo-seca/in RH-129 Y 50.0 k/in 0.25 lo-sec2/in RH-124 Y 106. k/in 0.81 lo-sec2/in RH-123 Y 262. k/in 0.24 lo-sec2/in 156
__._a_ _ . . . . - SECTIDN E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND CUPPORT STRUCTURE (NRC) IGUESTIDN E2 YAEC snould Justify why seismic anchor movement (SAM) analysis was p;rformed for proolem 9 (main steam line from anchor A-50 to, the four turbine nozzles) but SAM effects are considered negligible for other ccces. CECPONSE As described in the response to cuestion E1, the modeling tecnniques ucsd for problems #005 througn WOO 8 allowed the restraint locations to deflect; therefore, SAM was implicitly included in the inertia analysis. For problem #009, a conventional analysts was performed with rigid roctraint locations and enveloped amplified response soectral tnerefore, oxplicit SAM analyses were required for that p ro b l ern. I I I 157 l
CECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE QUESTION E3 (NRC) It is not clear whether the main steam line and feedwater line were Cnalyzed as a coupled system or separately. Justification,should be provided if the two piping systems were analyzed as separat e systems; YAEC sould describe in detarl how the coupling effect was considereo.
RESPONSE
The main steam and feedwater piping systems were analyzed seoarately. The piping systems and main steau/feedwater supoort structure were also analyzed t eparat ely. However, the aterative solution of tne analyses insured that the stiffness, mass, and stress state of the succort structure was compatible with tne response and stress state of the piping systems. See the response to question E1.d. for furtner cetails. i 158 I
~~~
T l SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE IQUESTIONE4 (NRC) The " hot shutdown" scope described by the licensee included noni su l ais l o portions of reactor coolant pressure boundary and raain stearn and r.oin feedwater piping. In the subrnitted analysis reports, sorne !pi pi ng syst erns, e. u. raain stearn and raain feedwater (incliiding support structures) and cornponents (e.g. prescurirer, s,t e,.rn generator) wer e ne-. fort e" analy:ed only for YCS. The licennee u5ould orovide analyses with the NRC site specific snectra for these systerns or justifscetion, f or not per f orrn a no such analyseu. I
RESPONSE
This incue will be address ed separat.el y. 159
I SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE QUESTION ES (EG&G 2.1) Clarification should be provided regarding the requirernents for " safe shutdown scope" piping versus those used for the " hot shutdown scope" I piping. t RESPONSE I J The " safe shittdown scoDe" and the " hot shutdown ecope" a rra r-c it i va i e it I terras and refer to the piping required to achieve sate chutdown ofn.1the olant. The retrofit criteria docuinent was used as the ba t . i s f.;r plaine analyses per f orr.1ed for Yantee Rowe. The sarne rea ti rorwr.t - or d-for the rest I allowables were used for the roain st eain/ f eedwa t er pipino a v. of the safety related pipino. I 160
- --a SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE QUESTION E6 (EG&G 2.2)
Representative rnathematical rnodels of the subject systems and sufficient related information to perrnst model verification should be provided. t
RESPONSE
The requested infortnat ion will be provided durir.g the NRC audit. I . I I I I 161 I
SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE QUESTION E7 (EG&G 2.3) Reference 15 as listed in tne sub,1ect licensee reoort snoulo De provicea for review. I t
RESPONSE
The listea reference will be celetec frora tne next revision to tne suo,1ect recort. l l l 162 I
I SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE I QUESTION E8 (EG&G 2.4) A complete deser-iption of the modelino techniques used shouic be ptovided.
RESPONSE
A desctiption of the modeling ter-hnioues and iter ative analysis matnoc +-
'3 L . ur-i i i r t s used in the mai n r>t e.4m/ f ot ciwat er- pl o t rig ano suppoit ,o.
I was pt-ovided in tetponne to nuer, tion E1. 1 I .. I 163
' ^ ^ ' ' "~ ~
-^
_ . . - . . .._:.=. -- 2.-1 - - 2 :::~'-~~ r ~ SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE QUESTION E9 (EGSG 2.5) The stiffness values used for anchor A-50 and the rnethods used to determine thern should be provided. t
RESPONSE
See the response to question E1.b. for the description of the rnethod used to deterrnine the anchor st if fnesses and masses and the act ual values used. I I 164
(MS/FW) PIPING AND SUPPORT ISECTIDNE-MAINSTEAMANDFEEDWATER STRUCTURE QUEST!DN EiB (EGSG 2.6) i I Additional details regarding the mass and stiffness calculations, distribution and values used for the vapor container and support . otructure should be provided. t
RESPONSE
S2e the response to question E1.c. for details regarcing the mass and for the vapor container, turbine otif fness calculations and values used I building, and support structure. I .. I 165
--. _ . - . . . . . _ ._ _ =_.. - . - - . ,
SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE I QUESTION E11 (EG&G 2.7) Refererice 4 as listed ir the subject report should be provided for
)
review. ' I l i
RESPONSE
The subject refererice is a letter frorn YAEC to Cypria clarifyiric tne exterit of the hot shutdowri scope with respect to the ri.airi , st earn / f eedwa t er syst er.1. The letter states that the f e-ec w a t e r c a r, i e . . - outside t he vapor corit a lrier ( prob] eri's #25AB t hrouch tt d 8 A B ) are riot reovired for hot s h u t d owri. These problFMc Arr Bria } yred 00l*3 for i
- n; i r:1p a c t ori the rnalra st eatii/ f eedwat er structure eric, i ri t ur ri, e r. t h e r.la i r.
st ea rn o s pi rip reouired for hot sh ut d ow ri. YAEC will orovide the suh,iect letter to the NRC. 166 ,
_. ~ 1 l SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT l STRUCTURE QUESTION E12 (EG8G 2.8) Clarification should be provided as to exactly how se i stn i c anchor rnovernent loads were cornbined with other loads. t
RESPONSE
For all piping systerns attached to the rnain st earn / feedwa t er suoport structure, the SAM effects and inertial effects were adcresned in iw sa rne seistnic analysis, due to the boundary cond it ion roodeli ng raa t i. ,a m . These were described in response to question E1. For rna i n st ea rn c r o n t e r.- 4t 009, frorn anchor A--50 t o t he turbine nor lec, SW1 J o ids vee e a n. . ' .m c and cornbined with t herina l expansivre and therrnal anchor t. . .v ennen t lo d. In Enuation 13 of ANSI D31.1 in accordance with the req u i rernent s e s f t iin retrofit eriteria. I I I 167 I
I - SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE QUESTION E13 (EGSG 2.9) Justification should be provided as to why the NRC spectra were [not] used in a postulated seismic load case. t
RESPONSE
This issue will be addressed separately. I 1 I I i j I . I 168 I 1
_ u- - . _ , . _ . . ._ -__ _. . _ _ . _ I SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE QUESTION E14 (EG&G 2.10) It should be clearly st ated what industry code and edition are being used to define the stress equations used. It should be clearly snown that this is in compliance with SEP reevaluation guidelines. I RESPONSE I The SEP reevaluation guidelines soecify that tne ASME Boiler & Pressure Vessel Code, Section III, Division 1, 1980 through the Winter 1980 Addendurn should be used for the evaluation of safety related cloing. ANSI B31.1 1977 is specified in the retrofit criteria, DC-1, for the I cualification of all safe shudown piping, including the piping in tne main steam /feedwater '. system. come of the ADLPIPE computer analyses. Add it ional ly. B31.1 1973 was invoked in Ecuations 11 tnrougn 19 were used to cornoare stresses with allowables, as indicated in the retrofit criteria. the ANSI B31.1 1977 code, the ANSI B31.1 1973 code, and I A comparison of the ASME 1980 code, yielded the following comments: o The stress ecuations are identical except that in ANSI B31.1 1977 and 1973 Ecuation 12, design pressure is used, whereas in A5ME 1980 Ecuation 9, peak pressure is used. Since deston pressure in a pleing syst em is either greater than or ecual to the roaximum oressure, tne pressure strecs calculated in ASME 1980 Ecuation 9 is always less than the pressure stress calculated in ANSI B31.1 1977 Ecuation 12. The rost eria l allowables for reoresentative materials ( A 1,06 A. A106D. Io A106C, A155 KCF70-1, A105 A213 TP304, TP316, A312 TP316. and A376 TP316) in ANSI B31.1 1977 and 1973 are A312 1P304, A376 TP30*. A213 less than or ecual to t hose in ASME 1980.
- The cold rn o c u l i (EC) for the stainless steel raa t er i a l s listec aoove are slightly higher in ANSI B31.1 1973 than in ANbl B31.1 1977 anc I ASNE 19e0 (29.2E6 pst versus 28.3E6 osa): however, this olfference is nenligible.
lne coefficients of expansion in ANSI B31.1 1977 and 1973 are Io consistent ly nigner than those in ASME 1980. ylelc a ng hi oner tnerrna l stresses and loaos. Io Flex 1b11 a t y enaract er1st a cs and st ress int ens t ficat lon f actors are identical in the t n r ere cooes. i l I Therefore, use of AN51 B31.1 1977 and 1973 are considereo t o roeet tne reautr.ements or the SEP reevaluation guidelines. l 1 1 i l 169 ' li
SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE CUESTION E15 (EG8G 2.11) The method used to combine the roodal responses in the se i stn i c load case should be clarified. t
RESPONSE
All raain stearn/feedwat er pi ping analyses were perfe_irrned wit h the nioino analysis computer prourarn ADLPIPE. Modal responses were con.r. I ne J in the I prograrn usina the 10 /. bandwidth aroupino rnethod, in confortcence with the requirements of paragraph 1.2.1 of USNRC Regulatory Guide 1.92. l l
}
l l l-1 l 170 l
~ ~ - - - - - -
SECTIDN E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE GUESTIDN E16 (EG&G 2.12) It should be confirmed that Su was intended to be included as an allowable stress value in Table 3-3 of the subject report. t REIPONSE YAEC is performing analyses and system uogrades to assure tnat tne st ruct ures, and equipment required to achieve a hot snutcown 10yctems, cro within code allowables for YCS loading. Additional systems, ctructures, and equipment are being analyzed and mocified if they nave an impact on the identified hot shutdown systems. Of these seditional cyctems, relaxed criteria are imposed on the non-safety related linen to coeure functionality of the hot snutdown systems. Tho non-safety related lines have oeen analy:ed using the s aroe loacs and methods as used for the safety related lines. The systems are cualified for ceadweight and pressure as soecified in ANSI B31.1 Ecuation 11. However, under ceadweignt + pressure + YCS loads, the allowable is taken co Su to insure that the line will not fall structurally and imoact the not sn ut down lines. 171
- _- _ =._ -
W SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE QUESTION E17 (EGSG 2.13) Detailed infortnation regarding the raethodology used and representative I calculation packages for review should be provided for the piping supports connected to the sub,)ect lines. f
RESPONSE
All piping suoports will be evaluated /analyrec/ designed u s i n ci I recuireraentn out1ined in the retrofit eriteria d oc urae nt , as was cone for the safety relattrd pipe supports. At t h i n t i rne, the wor 4< for the c. i n st eain/ feedwat er- pipsno and support 5 has not been corord et ed. T&c iterations of pipino and support structure analysis have tmen oe r #er raed ; hnwever, final taodifications have not been identified. a t iri s u o! so . evalurtions and de. ignt- inave not been initiated. Th.' wor-k i- : -:. : 'ed I to be corcolet ed in 198f3. 172
3_ - - - - ~ ~- . - - ._ ____. _ - . _ _ _ _ _ _ SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE I QUESTION E18 (EG8G 2.14) Evidence should be provided that all requireraents of IAE Bulletin 7'3-02 are satisfied. t
RESPONSE
The retrofit criteria doeurnent was used for the taal n 1steara/ f eedwate r r11 n i c t.o e 2. piping as well as for the safety related pining. I addressed in the resoonse to avestion D15 fnr safety related o z o i n t, . I I I I .. 173 I
SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE I DUESTION E19 (EG&G 2.15) A detailed evaluation of the effects any rnodifications to the MS/FW support structure will have on the stress levels in the attached t proing should be provided. I RESPONSE As discussed in the response to Question E1, the it eret ) m s. . ' i.? I t. r. unad to analy:e t he rnain st eard/f eedwat er piping and support st ruc t urm 2isures the coropat ibility of responses of t he subsyst ernt. Additiore) i t e r e t 1._,n s are perforrned until the boundary cond t t ions- upecified by tho <? u r o. .r t struct ure with rnodi ficat ions provide adequate reut raint for tac o i t;i ng systern and unt11 the olpe reactions resulting frou Ihe r.l o i n a ana;ynes can be resisted by the s u n a.. .r t utrueture without e n esc: i r.a t c.c r I allowables. 174
SECTION E - MAIN STEAM AND FEEDWATER (MS/FW) PIPING AND SUPPORT STRUCTURE QUESTION E20 (EG8G 2.16) The licensee should clarify why seistnic anchor roovernent (SAM) analynis was per f orrned for problern 9 (roain stearn line frorn anchor A-50 t to the four t urbine nozzles) but considered negligible in other cases.
RESPONSE
l As discussed in the respons,e to cuestion E1, the seasr.ite 4ncnor j roovernent c for the rna i n nt earn / f eedwat er orping 1inen or, t h.2 suonort structure are i rn pl i c i t l y i nc-l u d ed in the inertial analycis by us t r:a t.h e rnass-ntiffness in.dels at all bou ncj a ry l o c a t i o n s,. Thet ano , n i , F. croblern 9 wac, rnore traditional, with rioid supports and exoliciti,- specified SAMn. See ouestion El for rocre detail. l I 175 lI
_ ~ . . . . s- _ _ . _ _ _ . _ . _ __. _ _ _ . . . . ._.
~
YANKEE ROWE SMALL BORE PIPING SEISMIC QUALIFICATION IN-SITU FREQUENCY TESTING t PURPOSE Determine resonant frequencies of oloing scans in the as-built configuration. Resonant frecuencies were used to cet errnine sopropriate seismic accelerations from response spectra. APPLICATION Piping spans were first analyzed with the peak seismic accelerations from the appropriate seismic resoonse I spectra. All scans that failed at the oeak accelerations were then analyzed with the ZPA. If the scans cualified to the ZPA acceleration, the benefit of determining the actual resonant frecuencies and corresponding seismic accelerations could be realized. Frequency testing was used to determine the resonant frecuencies. METHODOLOGY The test method involved three basic steps:
- 1. Cnoose accelerometer locations for three orthogonal airections,
- 2. Excite pioing scan with impulse excit at ion,
- Hammer blow or Steo relaxation
- 3. Average frecuency resoonse functions.
RESULTS Maximum frecuency banowidth of 0 to 50 H:. This candwictn was reduced on a case by case basis to provide suf ficient frequency resolution at lower frecuencies. I ~~ Lowest resonant frecuency was 2.25 H . Typical resonant frecuency range was 3 to 12 H:. Results demonstrate well soaced, uncoucled resonances. Four tyoical frecuency response function olots are attachec. 176
11 - YANKEE ROWE SMALL BORE PIPING SEISMIC QUALIFICATION IN-SITU FREQUENCY TESTING General Discussion Instrument All instrumentation was calibrated prior to testing. o Traceable to NBS o Complete analy:er o Entire instrument loop Damping Structural damoing was not estimated from test cata. o Low impedance testing does not accurately ioentify damping at high levels of excitation (SSE). o Damoing estimates from frecuency response functions recuires measured forcing function. Excitation Impulse excitation is the application of a single transient forcing function by either hammer blow or steo relaxation. I Impulse excitation produces constant, broad band frecuency input motion as a function of exciter stiffness. Rubber mallet was chosen to concentrate excitation energy I in the low frequency range. Steo relaxation was used on flexible scans ioentified by excessive "DC offset" (displacement due to nammer olow). Two advantages of steo relaxation excitation are:
- 1. Very low noise level in response measurement ,
Lower frecuencies dominate the response. I 2. Averaging Averaging of frecuency response functions is performed to reduce nonlinearities and provice tne best linear estimate of the response. I Four averages were used because of consistently linear frecuency response functions. Stable averaging was used. Staole averaging sums all frecuency response functions and divides by the number of samples. 177
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1 I ALTERNATE CRITERIA I
- PVRC DAMPING I
- ACTUAL MATERIAL STRENGTH
- COMBINATION OF SEISMIC INERTIA AND SAM I
- RESPONSE SPECTRUM PEAK SHIFTING l
- N0ZZLE FLEXIBILITY (WRC BULLETIN 297) l
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SUMMARY
OF WORK [g ~ PERFORMED TO ADDRESS SEP TOPIC III-6 [I {I l II 'lI LI - l11 YNPS - RESOLUTION OF NRC QUESTIONS ON SEP TOPIC III-6 "4 II
Pag . C4/03/CE. NRC PRESENTATION - AP'ilt.1986 YAf4KEE ATOMIC ELECTRIC COMPANY ,
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC III-6 S T RtJCTU RE, SAFE lfi SYSTEM OR DESIGN SHUTDOWN DESIGN ANALYSES CRITERIA PERFORMED STATUS REFERENCES REMARKS COMPONENT LOADD CI) SCOPE STRUCTURES ACI 318-77 Linear Anchor bolts of " Reactor Support Collars were installed at R:3 actor D+YCS (Yankee Yes Structure", Rev. sin enterior columns. Support Composate two interior columns yield at 3 EY-YR-80023-6, l. ( Structure Spectra) early stage of Cygna, March 1983. ! (RSS) loading. l Therefore, , t interior column bases are modeled as pinned. No ! (' other yielding under YCS. C Functionality [21. ColIars: A11 coluans and Structure satiafies D+NRC(NRC Site functionality criteria' Specific Linear. ring foundation ' Spectral Upper develop plastic under NRC load. ( hinges under NRC Structures ' Nonlinear loads. ( time Functionality history. achieved by Rang limiting the Foundations number of Nonlinear snelastac step-by-step excursions and static displacements. j analysis. D+T+LOCA, Yes Steel elements: Linear Reinforced " Vapor Container Structure satisfies ! Vap..c YCS- AISC 8th Ed., concrete footings Structure", Rev. functionality criteria Cvntainer D+T+YCS, D+T+NRC & Part I, NRC- yield slightly 3, EY-YR-80023-5, under NRC load. St r.ac t ure D+T+W Setsmic urder YCS. All Cygna, April 1984. (VC)
- Reevaluation footings yield Crateriat31. under NRC. Some RC elements: anchor bolts yield ACI 318-77. under NRC. All steel elements and shell remain i
elastic. ., VC Shell at D+T+t.DCA+YCS Yes SEP Guidelinet43. Linear OK Same as VC O a pe= +Pspe A r.c hor i Penetratnons Loads T .or b a r.e D+YCS Yes Steel elements: Linear OK Cygna Cales. Modifications installeds f New faxes- AISC 81061/2-F & 2 RC shear walls, 3 Diag l ti.i s l d i ng Nath Oth Ed., part I., 83050/I-F braces, 4 roof braces, , w d a f a r a t s . .r. ) diaphragm & blockwall 4 Existang elements-Seismic fixes. Reevaluatton
>d Criterlat33.
W RC elements:
- ACI 318-77.
i l
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Pags No. 2
$4/03/86 NRC PIESENTATION - APRIL 1986 -)
YANKEE ATOMIC ELECTRIC COMPANY
)
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC 111-6 STRUCTURE, SAFE SYSTEM OR DESIGN SHUTDOWN DESIGN ANALYSES SCOPE CRITERIA PERFORMED STATUS REFEREP.CES REMARKS ; j COMPONENT LOADS [1] ' 1 No. But C.T. Main Linear OK C.T. Main Report Masonry Walls D+YCS l failure Structural for Block Walls in an Turbane Turbine Bldg., { l Bldg. r'a y Design Criteria ' affect ER-2648-10-1, May SSS 1985. ; I equap-ment. No. But ACI 318-77 Linear OK " Turbine Bldg.", l Tearbane 0+YCS & interact Rev. 1, Pedestal D+NRC with EY-YR-80623-9 SSS MS Cygna, Dec. 1982. ,4 Paping. I No Steel elements: Linear Engineering Cygna Cale. DGB modifications to be Diesel D+YCS completed by 8/88. AISC 8th Ed., analyses to be 83033/15-F. Generator Bldg. (DGP) Part I. completed after twatn Rennforced resolutton of modafacations) masonrys 1979 UBC. esind and tornado and PICS Bldq issues. Yes. Steel elements: Linear Engineering "PAB and Rad. Pipe PAB modifications to be Framary D+YCS completed by 8/88. Because AISC 8th Ed., analyses to be Tunnet", Rev. 2, Aqistliary of SSS Part I. completed after EY-YR-80023-7, Boaldtr.g (PAB) paping & RC elements: resolution of Cygna, 1/83. and (with equap- ACI 318-77. mind and tornado Cygna Calc. mod fseatnons) 83033/13-F ment. issues. No. But C.T. Main Linear OK C.T. Main Report Masonry malls' Masoney Walls D+YCS failure Structural for Block Walls in construction to be at PAD N._.rth PAB, Upper Chase & completed by 11/86. Wall, tipper may Design Crateria affect Cable Spreading Pape Ch a se. s SSS Rm., ER-2648-10-2,
- r. ble equip- Feb. 1986.
Spreading Rm. ment. ACI 318-77 Linear DK " PAD and Radioactive D+YCS & Yes Radaoactive Pipe O4pe Tunnei D+NRC Tunnet", Rev. 2, , EY-YR-80023-7, Cygna, Jan., 1983. D+YCS & No Steel Elements Linear OK under YCS. Yard " Spent Fuel pool, Spent Fuel Rev.3, D+NRC YCS- AISC 8th Ed., Area Crane Support etc.", Pool, Chute, EY-YR-80023-le, Fuel Transfer (SF Pool also Part 1, NRC- Structure has analyzed for functionality [23. minor anchor Cygna, May 198's. Pat House 8 l Yard Area t hersna l load.3 RC Elements yielding under trane Support ACI 318-77. NRC. , ( i [ Structure F* i co l 1
.a g G4/03/C6 '
NRC Pr.ESENTATIDN - RPAIL 1986 f YRNMEE ATOMIC ELECTRIC COMPANY
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TDPIC III-6 { J SAFE f STRUCTURE, SYSTEM OR DESIGN SHUTDDWN DESIGN ANALYSES PERFORMED STATUS REFERENCES REMA RKS COMPONENT LOADS [Il SCDPE CRITERIA I DK " Fire Water Tank i Fire Water D+P+YCS & Yes Steel elements: Linear ' AISC 8th Ed., Seismic Analysis", Tank D+P+NRC Part I a nd YAEC-1492, Sept. API Std. 650. 1985. Archor bolts: ACI 349-76. AISC 8th Ed., Linear Possible weak Cygna Cales. Preliminary modification Yard Pape D+YCS+ Pipe Yes Part I links are Anchor 83033/18-F & 19-F design according to pipe (MS/FW) Reactions & Support D+W+Pape A-50 Beam-to-wall (both incomplete) reactions obtained in connections and first iteration was Structure Reactions completed. Design addstional ; diagonal braces according to Iteration 2 are needed. is incomplete. Engineering analysts will be ; completed after resolution of wind I
& tornado issues.
I D+YCS, No. But Functionality [21. Linear Engineering N/A Modifications to be i NRV Enclosure completed by 8/88. ' D+NRC & failure analysis will be i D+W may completed after affect resolution of ; SSS wind and tornado equip- issues. ment. I EQUIPMENT Vessel & Nozzles Linear DK " Major Equi pment Analysis of RPV stability Ceactor D+P+YCS Yes SEP Guideline [43, Qualification" under NRC is in progress. ; Pressure Rev.2, Support Ring: AISC g vessel (RPV) EY-YR-83033-2, 8th Ed., Part ! , Cygna, May 1985. YCS Yes ASME Section III Linear + DK. Functionality " Seismic Evalution e Reactor where applicable, energy is satisfied by of YNPS Reactor internals demonstrating that Internals", Impell functionality bal arce ~ elsewhere. For technique no permanent Corp. Report No. detail see for impact deformations and 02-0570-1704, Aug. , reference report. of uplifted no impacts between 1984. l fuel on components occur. ' lower core ' support I plate. w CD -J
} I' Ptqe No. 4 04/03/86 NRC PT,ESENTATION - APRIL 1986 ,
- I YANKEE ATOMIC ELECTRIC COMPANY i
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC III-6 i I f STRUCTURE, SAFE SYSTEM OR DESIGN SHUTDOWN DESIGN ANALYSES CRITERIA PERFORMED STATUS REFERENCES REMARKS COMPONENT LOADS til SCOPE i i l P+YCS Yes FunctionalityC23. Linear OK. Functionality "CRMD Pressure C RDM' s is satisfied by Housing Seismic ' showing that no Analysis", , l J f plastic Westinghouse deformations occur Electric Comp., l and CRMD's ret urve Sept. 1982. to their original position after seismic event. Yes Same as RPV Linear OK Same as RPV Steare D+P+YCS Ger. erat ors Same as RPV Linear OK (1) Same as RPV. Four lateral braces pressiartzer D+P+YCS Yes (2) Cygna Calc. installed. , (wnth support 85037/2-F. l m.deficatnons) Yes SEP GuidelineC41. Linear OK Same as RPV , Main Coolant D+P+YCS P. imps Linear OK Same as RPV Lower support Feed & Pleed D+P+YCS Yes Same as RPV
- moodificattons to be Heat installed during 87 Euchangers outage, (with sie p por t w;d a f n eat n ons)
Yes Same as RPV Linear OK Same as RPV Maan Coolant D+P+YCS Loop Isolatson Valves D+P+YCS Yes Same as RPV Linear OK Same as RPV Main Coolant Pisup Dascharge Check Valves Yes Same as RPV Linear OK Same as RPV Main Coolant D+P+YCS bypass Paping Isolatson Valve.
~
Yes Same as RPV Linear OK Same as RPV Pressesr a zer D+P+YCS Spray Valves HCV-205 Yes ASME Section III, Linear OK (1) " Seismic Pressierszer D+P+YCS NC-3521-1, Analysts", Report Spray Valve No. 8330, CONVAL, MOV-191 Level A8 1. 7. 2 Inc., 3/26/85. (2) CONVAL, Inc., Supplemental Letter, D. Graham
# to W. Lucas, 5/7/85.
CD
Pa 3 M M M G4 / 0 3 / 06 NRC PRESENTATION - APRIL 1986 YANKEE ATOMIC ELECTRIC COMPANY
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC III-6 STRUCTURE, SAFE DESIGN SHUTDOWN DESIGN ANALYSES l SYSTEM OR STATUS REFERENCES REMARKS LOADS [13 SCOPE CRITERIA PERFORMED COMPONENT
}
Same as RPV Yes Same as RPV Linear OK Mann Steane D+P+YCS s! Non-return ! f,
, q Valves Same as RPV Linear OK Same as RPV Mann Steams D+P+YCS Yes '
.i C.Ae Safety l Valves Same as RPV Linear OK Same as RPV ,
Maar. Feed D+P+YCS Yes { Check valves OK Dresser Design !' D+P+NRC Yes ASME Section III, Linear Code Safety Report SR-317-19, Valves 1977 with 1978 Rev.1 and Seismic Summer addenda. Qualificatton Report SV-133-3ND 116. Linear OK Same as RPV , Pressur t zer- D+P+YCS Yes Same as RPV I-6 Reiser ar.d !' E* loc k Valves Same as RPV Linear OK Same as RPV , S.e fet y D+P+YCS Yes
- 4. l Insectson Loop 'l
!solatarn I!
Valves Same as RPV Linear OK Same as RPV Safety D+P+YCS Yes I n ject n on Loop 1.olatson Check Velves lf Same as RPV Linear OK Same as RPV Yes Shutdown D+P+YCS .' l Cvol n r.g l Isolatson i ;, Valves OK (1) Proto-Power Seismic load applied in Valve on D+P+3g Ves (1) 1983 ASME Linear Corp. Calc. No. three directions. , Section III, ~ l c-1/2" MC 01-01, File Class 2, Level A. E* l eed 8410201. LCV-2'22 (2) Seasmic (2) " Seismic ' Reevaluation Analysis Report" Crateria[31. Rev. 1 S. O. 8094,
.r
( Proto-Power Corp. i Same as RPV f Yes Same as RPV Linear OK SSS Valves on D+P+YCS Small I*or e Papang an VC } w W w
6 E E !is Page No. i
$4/@3/86 '
NRC P~ESENTATION - APRIL 1986 YANKEE ATOMIC ELECTRIC COMPANY (
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC III-6 STRUCTURE, SAFE ANALYSES { SYSTEM OR DESIf>N SHUTDOWN DESIGN ' CRITERIA PERFORMED STATUS REFERENCES REMARKS COMPONE NT LOADS (13 SCOPE i Yes SOUG/NUREG 1030 SOUG See Note (63. SOUG Report and l Valves TV-40lA,B,C.D EDE Walkdown Report (1983). T V-2 03 TV-406 and TV-411 PIPING ANSI D31.1, 1977 Linear OK " Pipe Stress Four supports ir.st al led Maan Steani D+P+T+YCS Yes for Probs. 001 t 004. Analysas Report *, L r.es a r.s a de Additional supports for Rev.1, vr (P.oblems EY-YR-80023-14, probs. 002 & 003 are not del tv ;' * * .* Cygna, May 1983. required per Impell , Cygna Calc. report, using PVRC 83033/28-F. damping and faulted Impell Report a!!owables. 09-0570-0045. ANSI B31.1, 1977 Linear Engineering "MS/FW Piping Pipe supports ard Maan Steam D+P+T+SV Yes analyses to be Analysis Report", structures are modeled as Lines from VC Disch+YCS 8 completed after Rev.0, equiv. springs attached to Ar.chvr A-50 D+P+T+SV resolution of EY-YR-83033-4, to ground. Iteration 2 l (with Disch+W supports to be completed wand and tornado Cygna, July 1984. Iteratnon 2 by 11/88. ' m dafications) assues. Functionality [2]. Linear Engineering "MS/FW Piping Iteration a supports to Maan Steam D+P+T+YCS 8 No. But I' analyses to be Analysis Report", be completed by 11/88. Lanes be yor.d D+P+T+W a nt eract wath completed after Rev.0, Ar.t bor A-50 SSS MS resolution of EY-YR-83033-4, (with Papang. wand and tornado Cygna, July 1984. Iteration 2 l l w.daficationst issues. Functionality is I achieved by limiting the stresses to ui t insat e. ANSI 031.1, 1977 Linear OK " Pipe Stress Eleven supports boater Feed D+P+T+YCS Yes installed. Two supports Dascharge Line Analysis Report", , Rev.t. for Problene 022 to be auAade VC EY-YR-80023-14, adjusted and three (Problems 021 supports for Problem 023 to 024) Cygna, May 1983. Cygna Calc. to be completed during 87 83033/28-F. Impell outage (per Impell report Report using PVRC damping and ! I 09-0570-0045. faulted allowables). D+P+T+YCS & No. But Functionality [21. Linear Engineering "MS/FW Piping Pipe supports and b.:.sler Feed Discharge from D+P+T+W interacts analyses to be Analysis Report", structures are modeled as ' VC to Turbane with completed after Rev.0, equiv. springs attached - I Bldg (with SSS MS resolution of EY-YR-83033-4, to ground. Iteration 2 Paping. wind and tornado Cygna, July 1984. supports will be f p Iteratnon 2 completed by 11/88. o no. .d i f i ca t ions) issues. o
pg , 0 4. / 03 / 8E, NRC PCESENTATION - ApMIL 1986 YANKEE ATOMIC ELECTRIC COMPANY ,
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC III-6 l 5iRUCTURE, SAFE SYSTEM OR DES LGN SHUTDOWN DESIGN ANALYSES CRITERIA PERFORMED STATUS REFERENCES REMARKS , COMPONENT LOADb (11 SCOPE i ANSI B31.1, 1977 Linear OK " Pipe Stress One support osas installed i Mann Coolarit D+P+T+YCS Yes Analysts Report", for Loop 3. j Papar.g ' Rev.1, EY-YR-80023-14 j Cygna, May 1983. f I Cygna Calc. 83033/28-F l Yes ANSI B31.1, 1977 Linear OK Stone & Webster Four snubbers deleted. Pressur-t ree- D+P+T+NRC One spring hanger added. j Report Safety 8 Tharteen supports d Reisef Papano B2-1198618-1.
- modified.
ANSI B31.1, 1977 Linear DK " Pipe Stress Forty nine supports Safety D+P+T+YCS Yes installed for Problem Analysis Report", injectnon Rev.1, 207. Problem 201 has no . Papang EY-YR-80023-14, snodi ficat ton. (Problenes 201 Cygna, May 1983. j 3 c0 / ) Cygna Cales. - f 83033/28-F, , 85037/4-F & l 85037/11-F. ANSI B31.1, 1977 Linear OK " Pipe Stress Six supports installed. Shiitato n D+P+T+YCS No. But C* l a r.q Pa par g part of Analysis Report", f pressure Rev.1, a Pre bIeras 121 EY-YR-80023-14, g a 12-2) boundary up to Cygna, May 1983. a farst asolatton valve. f ANSI B31.1, 1977 Linear OK " Th errna l, DL & Supports to be installed Mann Coolant D+P+T+YCS No. But part of Seasmac Pape during 87 outage. i e I bleed frons VC Stress Analysis of a l t o L PST pressure bour.dar y CRBL-302-1 & up to CRBH-2502-4", EAS { Report . first l; asolation EAS-074-84-000-1. ~ valve.
"SSS Sma11 Bore RPV rocking effeets and [
SMAL L BORE PVRC damping used for all .( Papang Inside VC, PIPING Analytical analysen. All support i INSIDE VC Results", Rev.1, modiftcations for small , E-Y-84060-1, bore paping inside VC to l Cygna, Mar. 1985, be completed during 87 : eMeept as noted. outage. W e llr (
- l P,,M.
$4/@3/86 Q E E E E NRC PRESENTATION - ApRIt.1986 YANKEE ATOMIC ELECTRIC COMPANY
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC III-6 STWUCTURE, SAFE SYSTEM OR DESIGN SHUTDOWN DESIGN ANALYSES COMPONFNT LOAD $ [13 SCOPE CRITERIA PERFORMED STATUS REFERENCES REMARKS y i l D+P+T+YCS No. But ANSI B31.1, 1977 Linear OK See above. One support added. ll Feed 8 Bleed Hr. Charging, part of One support modified. ; Safety pressure { Insectnon boundary. , Draans , (SP-FB-5) I RC Loop Safety D+p+T+YCS No. But ANSI B31.1, 1977 Linear OK by hand See above. Three supports to be ; Vatwe part of calculatton. added. l-Discharge pressure (SP-RCL-17 to boundary. f' I 199 Two supports to be added. ' Lc.o p 4 Safety D+P+T+YCS No. But ANSI B31.1, 1977 Linear OK See above. Vatwe part of Dascharge 8 pressure Charyinq boundary. i 6 Re1aef l (SP-RCL-209 Pressurizer D+P+T+YCS No. But ANSI B31.1, 1977 Linear OK See above. One support added. One support modified. ' Sorav part of (SP-PZR-7) pressure boundary. Charging & D+P+T+YCS Charging ANSI B31.1, 1977 Linear OK See above. Three supports modified. Aum. yes. Five supports to be Pressurazer Auw. pres added. Spray Spray no, (SP-CH-3) but part of pressure boundary. Charging D+P+T+YCS Yes ANSI B31.1, 1977 Linear OK See above. One support added. through One support modtfied. Drainnow i (SP-CH-2) ! Pressurizer D+P+T+YCS Ne.. But ANSI B31.1, 1977 Linear OK See above. Three supports modified. Seedp l e & Veret part of (SP-PIR-2ab) pressure boundary. Hot 8 Cold D+p+T+YCS No. But ANSI B31.1, 1977 Linear OK See above. Two supports modified. Leg Draans, part of Tharteen supports to bet Pressurizer pressure added or modified. l. Deaan, Deaan boundary. Header (SP-HCL-9 to ' W 16, SP-PZR-1,
@ a SP-DRH-11 e
PJ t I l i l _ __ __________ ..__ _ _ .___ _
PagE M/@3/86 NRC PRESENTATION - AP?!L 1986 YANKEE ATOMIC ELECTRIC COMPANY
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC III-6 STRUCTURE, SAFE SYSTEM OR DESIGN SHUTDOWN DESIGN ANALYSES COMPONENT LOADS [13 SCOPE CRITERIA PERFORMED STATUS REFERENCES REMARKS l Charging 8 D+P+T+YCS No. But ANSI B31.1, 1977 Linear OK See above. Five supports modified.
- f Bleed Draan part of a r.d HP Sample pressure (bP-CH-lab) t>3u nd a ry.
St ease D+P+T+YCS Yes ANSI B31.1, 1977 Linear OK See above. Two Supports added. Ger.er a t or Four supports a.odified. Bic.wdown One support deleted. ESP-SG-9 to 12 Two supports to be added. 8 SP-BD-1) Two supports to be modified. Two supports to be deleted. .. Bleed Lanes D+P+T+YCS No. But ANSI B31.1 1977 Linear OK See above. One support added. f e c+i a nd to HX part of Three support s enodi fied. (SP-FB-Jab 8 pressure ; SP-FB-33 boundary. g I RC Pump Vents D+P+T+YCS No. But ANSI B31.1, 1977 Linear OK by hand See above. l (SP-RCL-1 to part of calculatton.
- 4) pressure bou r.d a ry.
f RC Loop Bypass D+P+T+YCS No. But ANSI B31.1, 1977 Linear OK by hand See above. , vents part of calculatton. ll ; (SP-RCL-5 to pressure
'l en boundary.
i ANSI B31.1, 1977 Linear OK by hard See above. ; Steam D+P+T+YCS No. But toer erat or part of calculation with f Dea a r.s pressure frequency g (SP-SG-1 to 4) boundary. calculation. Steam D+P+T+YCS No. But ANSI B31.1, 1977 Linear OK by hand See above. Four supports modified. ber. era t or part of calculation with { ver.t s (SP-SG-5 pressure frequency j to BB boundary, calculation. . Steam D+P+T+YCS Yes ANSI B31.1, 1977 Linear OK YAEC Calc. Seven supports added. Ger.er a t or Wade YRC-406. Six supports modified. R.er qe Level Heferer.ce Leg (SP-SG-13 to j 165 D*P+T+YCS Yes ANSI B31.1, 1977 Linear OK YAEC Calc. Twelve supports added. I Steam ! Ge r.e r a t or Wade YRC-406. i Rar.qe Level l Variable Leg g (SP-SG-17 to l e 20) l
t Pag . E E l C4/03/86 NRC PRESENTATION - AP3IL 1986 I(r YANKEE ATOMIC ELECTRIC COMPANY
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC 111-6 : STRUCTURE, SAFE ,i SYSTEM OR DESIGN SHUTDOWN DESIGN ANALYSES PERFORMED STATUS RCFERENCES REMARKS i COMPONENT LOADS [1] SCOPE CRITERIA l No. But ANSI B31.1, 1977 Linear OK YAEC Calc. Five supports added. 'l; I St eane D+P+T+YCS part of YRC-406. f' Ger.erator , . Dafferenttal pressure i ! Pressure FIcm boundary. ( SP-SG-2'I a bc 3 D+p+T+YCS No. But ANSI B31.1, 1977 Linear OK by hand See above. Steau Generrator part of calculatson. Narrce Range pressure Level boundary. f 4 (SP-SG-22) , ANSI B31.1, 1977 Linear OK by frequency See above. One support added. One Pressierizer D+P+T+YCS Yes support snodified. Wade Range test and hand calculation. Level (SP-P2R-3abc) ANSI B31.1, 1977 Linear OK by frequency See above. Two supports snodified. Pressurtzer- D+P+T+YCS No. But Narr._m Range part of test and hand Level pressesre calculatton. TSP-PlR-4I boundary. . D+P+T+YCS Yes ANSI B31.1, 1977 Linear OK by frequency See above. l Press.irarer , I P ess.re test and hand (SP-PZR-5) ca1cuIatton. ANSI B31.1, 1977 Linear OK by hand See above. One support added. Feed a Bleed D+P+T+YCS No. But calculation. l HK Vents part of pressure ,i TSP-F B-3ab) l bouridary. l ANSI B31.1, 1977 Linear OK by hand See above. One support added. fe l eed L a r.e D+P+T+YCS No. But Flow (SP-FB-41 part of calculation. pressure boundary. ANSI B31.1, 1977 Linear OK YAEC Cale. Analysis and modification Reactor and D+P+T+YCS No. But to be completed by 1987. pressure vent part of YRC-65. pressure , bou r.d a ry. RC Loop D*P+T+YCS No. But ANSI B31.1, 1977 Linear OK by generic See above. W essure part of calculatnon for pressure tubing. I nst rumer.- t a t n or. Tubang boundary. (GP-RCL-21abl t N b
E M l, 4 81 NRC PRESENTATION - APAIL 1986 YANKEE ATOMIC ELECTRIC COMPANY
SUMMARY
OF WORK FTRFORMED TO ADDRESS SEP TOPIC !!!-6 j-SAFE 6> STRUCTURE, ANALYSES SYSTEM OR DESJ GN SHUTDOWN DESIGN REFERENCES REMARKS f SCOPE CRITERIA PEHFORMED STATUS COMPONENT LOADS C13 See above. , ANSI B31.1, 1977 Linear OK by generic ' RC Loop D+P+T+YCS No. But part of calculation for Instrumen- tubing. t at n or. Tubarig pressure (SP-RCL-22) boundary. ANSI B31.1, 1977 Linear OK by generic See above. LTDP D+P+T+YCS No. But part of calculatnon for Instrumen- tubing. tatson Tubang pressure TSP-HCL-d3) boundary. Analysis to be perforened D+P+T+YCS No. But ANSI B31.1, 1977 Lir. ear in 1986. Modificattons to Bleed Papar.g vc Per.etratson part of pressure be installed during 1917 f[ to LCV-222 outage. boundary. Analysis to be perforsned Ves ANSI B31.1, 1977 Linear in 1986. Mod a f ic at ions t o Chargang D+P+T+YCS . Papang - be installed during 1987 VC Per.etratson outage. to Valve Room Analysis to be perforeed { No. But ANSI B31.1, 1977 Linear in 1986. Modifications to ' Fare Paping D+P+T+YCS TK *J5 N.,zele part of to Isolatson pressure outage. , boundary. ! Valwe Analysis to be perforeed No. But ANSI B31.1, 1977 Linear in 1986. Modificattons to Prasaary Vent D+P+T+YCS Paparq - part of be installed during 1987 ; VC Per.etratson pressure outage. j boundary. to Valve R.sc+e Analysis to be perforteed No. But ANSI B31.1, 1977 Linear in 1986. Modifications to Steam Drain D+P+T+YCS Papang - NRV D+P+W part of be installed during 1987 , Area to Auw. pressure Outage. 14oa1er Ho.:en boundary. Analysis to be performed Yes ANSI B31.1, 1977 Linear in 1986. Modificat ions to Steam D+P+T+YCS G.r.erator be installed during 1987
, outage.
Per.et ra t s on t o n iV-401A.B,C,D Analysis to be performed i D+P+T+VCS Yes ANSI B31.1, 1977 Linear in 1986. Modificattors to Eroergency bo a t er F eed D+P+W be insta!!ed during 1987 Header to Farst Check Valve e U1
~
Eid E Page No. 04/03/C6 N".'C PRESENTAT ION - AP31L 1986 Wea(EE ATOMIC ELECT"IIC COMPANY
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC 111-6 g STRUCTURE, SAFE SYSTEM OR DESIEN SHUTDOWN DESIGN SCOPE CRITERIA ANALYSES PERFORMED STATUS REFERENCES REMARKS lf COMPONENT LOAD $ [1] Analysis to be performed Yes ANSI B31.1, 1977 Linear *n 1986. Modifications to Aumalzary D+P+T+YCS Steam Papang D+P+W be installed during 19457 NRV Area to outage. TV-406 DEDICATED SAFE SHUTDOWN SYSTEM (DSSS) Complies with SOUG SOUG - NUREG 1930 SOUG OK Mechanical Yes - Criteria. , Equapment (2 DSSS Poemps 4 Diesel Ger.erator Seti YAEC Calc. Buried section from Pump l AISC 8th Ed., Linear OK House to VC. PVRC SSS Cor.dus t D+YCS & Yes - VRC-357. D+W DSSS Part ! Cygna Calc. damping used for section 85005/1-F. from VC to Pull Bon. OK Will be resolved after Raceway Yes - SOUG - NUREG 1030 completton of SOUG , DSSS walkdown. OK. Functionality Analytical ENGR Panel is rigid VCS Yes - Functionalityf23. Linear Services Report (Frequency)33Hz). SSS Instrum- is achieved by t at ion Par.el DSSS limitang the #84124, April stresses to 0.9 1985. yteld stress. # Actual smaximum stress as 876 psi.
'l i
SSS Instrumen-tatson Conan Design includes RPV rocking Yes - IEEE-344-75 Test OK. Conan R.T.D. effects. C nan R. T. D. YCS has been qualified Report IPS-Il38 an Main DSSS and Conam Seismic for a Coolant Cold translational Qualifscation Leq spectrum with ZPA Report IPS-1165, -,
= 8.0 g. 1984.
f OK. These C.C.L. Report No. Sigma Meters VCS Yes - IEEE 344-75 Test A-295-80 and DSSS instruments have a r.d Fo= boro been qualified for Foxboro Report Spec 200 a transtational Nos. DOAAD28, Electronics spectrum with 15 g OOAAB29 DOAAB35, peak. DOAAB45, DOA G58 8 OOAAA05. W O i
. q, G4/03/86 i NRC PRESENTATION - APRIL 1986 i YANMEE ATOMIC ELECTRIC COMPANY
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC !!!-6 i i STRUCTURE, SAFE I DESIGN SHUTDOWN DESIGN ANALYSES SYSTEM OR STATUS REFERENCES REMARKS ! LOADS [13 SCOPE CRITERIA PERFORMED CCMPONENT OK. These Letter from Similarity qualification t Webster YCS Yes - Mstitary Std. 90tc Test NTS/ACTON to YAEC, from Military Std. 901c & i Pressesre Ga sge DSSS 8 167 instruments have been qualified for dated 8/15/85. 167 test results to IEEE '
& Therw> meter a translational 344-75.
spectrum with 10 g h' peak. OK YAEC Cales. PVRC damping used. M he-Up Pump D+P+T+YCS Yes - ANSI B31.1, 1977 Linear Multi-level response YRC-386 and Descharge DSSS spectrum analysis used in YRC-388. Paping Cygna Cale. the PAB. New piping and 85005/1-F. supports added. OK YAEC Cales. PVRC damping used. Make-Up P.smo D+P+T+YCS Yes - ANSI B31.1, 1977 Linear New piping and supports YRC-386 and l Recirc.elatnon DSSS YRC-388. added. ' P a p i riq Cygna Calc. B5005/2-F. Linear OK YAEC Cale. PVRC deeping used. Make-tja Pump D+P+T+YCS Yes - ANSI B31.1, 1977 New piping and supports YRC-386. S..c t n or. P a p e r.g DSSS added. Cygna Cales. 85005/1-F and ; 21-F. OK Cygna Cale. New buried piping. Fuel Oil P+YCS Yes - ANSI B31.1, 1977 Linear 85005/1-F. ; Papar.g DSSS i i SOUG E 5, 63 See Note [63. SOUG Report and Mann Cor. trol D+YCS ES) Yes SOUG/NUREG 1930 I (53 EDE Walkdown D ard Report (1983). SOUG [5,63 See Note [63. SOUG Report and , fWV/CIS D+YCS [5] Yes SOUG/NUREG 1030 (53 EDE Walkdown [ C a b a r.et Report (1983). SOUG/NUREG 1930 SOUG [5,63 See Note [63. SOUG Report and MCC-1 D+YCS (51 Yes EDE Walkdown (Walt Ur.s t s) (53 Report (1983). .
~
SOUG/NUREG 1030 SOUG (5,63 See Note C63. SOUG Report and l Dr D+YCS (51 Yes EDE Walkdown D i s t r a but n or. [53 Report (1983). bisses 1 8 2 . I SOUG/NUREG 1930 SOUG (5,63 See Note [63. SOUG Report and NRV I r.st r.e- D+YCh [53 Yes mer.tatson (51 EDE Walkdown l Report (1983). j C ta a r.et !
~ -4 I
M M Page No. 14 G4/+3/t6 NRC PRESENTATION - AP2IL 1986 YANKEE ATOMIC ELECTRIC COMPANY
SUMMARY
OF WORK PERFORMED TO ADDRESS SEP TOPIC !!I-6 ST RilCTURF , SAFE SvSTEM OR DESI6N SHUTDOWN DESIGN ANALYSES C0F00NENT LOADS'(13 SCOPE CRITERIA PERFORMED STATUS REFERENCES REMARKS , C.wpor.ent D+YCS No. But SOUG/NUREG 1930 SOUG [5] See Note [63. 5000 Report and C.,.,1 a ng Anchorage (53 EDE Walkdown '. Surge Tank failure Report (1983). could anteract
.e s t h SSS .l:
equap. i f*atternes D*NRC Yes (1) IEEE 344-1975 Test / Linear OK (1) Wyle Labs 1 arwt 2 for Racks & Report No. Batteries, 43952-1. D+NRC (2) AISC 8th Ed. (2) Power and IE 79-02 for Conversson Anchorage. Products Dualification Report No. OR-21372. EMCC-5 4 6 D+NRC Yes (1) IEEE 344-1975 Test / Linear OK (1) " Seismic TRS ZPA = 3.7g (Hor) for EMCC, Qualification 2.9g (Ver) ', l D+NRC (2) AISC 8th Ed. Re por t ", Rev. 2, ; and IE 79-02 for No.SC-467 Gould , Anchorage. Inc. (2) YAEC !: Calc. YRC-150 t VD-MOV-559 D+P+YCS Yes (1) ANSI B31.1, Linear OK (1) YAEC Cale. O R - MOV *,58 1980 arid AISC 8th YRC-65 for Piping . Ed. for Piping and M.D. Support. :j a nd Valve M.D. (2) " Seismic Test ! Support. Report", ., (2) IEEE 344-1975 No. B0037, I: for Actuator. Limitorque, Inc. l l l ,i m NOTES: (1) Notatnons for loads are those used in SRP Section 3.0. [21 Functionalsty as defaned as the ability to achieve and maintain hot shutdown after a seismic event. [3] -Se n su n c Reeval uat iore and Retroftt Criteria", Cygna Energy Services, Project Nos. 80023, 81060, 81061 8 86064, Dve. No. DC-1, Rev. 3, Apral 1986. (4) Letter f rosis R. Caruso (NRC) to J. May (YAEC) dated Septeenber 20, 1982, forwarding Rev. 1 SEP Seismic Reevaluation Guidelines. h CD (51 Anchorage evaluation and upgrading performed using D+YCS. Design criteria are RISC Eighth Edition. ; I r r.1 cir. s e m a r- rmalification is beinn envered as part of the ongoing SO(IG Program as documented in NUREG 1930. I
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