ML20058M447
| ML20058M447 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 07/27/1990 |
| From: | Wallace E TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9008100131 | |
| Download: ML20058M447 (28) | |
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yd \ [ -]. " . TENNESSEE V^LLEY AUTHORITY.
< II CHATTANOOG4 TENNESSEE 37401 i
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U.S. Nuclear Regulatory Commission 1 ATTN: Document Control Desk
- Washington, D.C. -20555
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' Gentlemen:
1 In the Matter of ) Docket Nos. 50-327 i
. Tennessee Valley Authority ) 50-328 SEQUOYAH NUCLEAR PIANT (SQN) - NRC INSPECTION REPORT NOS. 50-327/90-18 AND ;
.50-328/90 As: requested in the. subject' inspection report, this letter submits TVA's response to the unresolvad items (URIs) which remain open from Inspection ;
Report;Nos. 50-327 /88-li: and 50-328/83-12. Although the responses in Enclosure 1 are titled by the 88-12 URI numbers, they address the specific. concerns raised by the subject inspection report. With .the exceptions of design' basis accident and zero period acceleration . effects, component damping values, and scismic analysis of the steel l containment vessel, TVA is in agreement with NRC on the scope of work required to address,the concerns._ For these three issues, additional inforrstion is 4 provided to support TVA's position. Summary statements of -conunitments contained :in this submittal are provided in ' Enclosure 2. Please direct questions concerning this issue to Kathy S. Whitaker at-(615).843-7748. l q Very truly yours, TENNESSEE-VALLEY AUTHORITY
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b .EG. Wallace, Manager Nuclear Licensing and l Regulatory Affairs Enclosures cc: See page 2 I l l 1
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Ms. S. C. Black, Deputy Director 1
. Project' Directorate II-4 '
U.S. Nuclear Regulatory Commission One White Flint , North 11553 Rockville Pike
'Rockville,LMaryland 20852 ,
Mr. J. N. Donohew Project. Manager
;U.S. Nuclear Regulatory Commission' -
One White Flint', North-L 11555 Rockville Pike - Rockville.. Maryland 20852 ; NRC Resident. Inspector Sequoyah Nuclear. Plant <,, 2600.Igou. Ferry Road Soddy Daisy, Tennessee 37379 1 Mr.'B. A. Wilson, Project Chief
. -i U.S. Nuclear Regulatory Commission >i Region II-101-Marietta Street,-NW, Sitite 2900 Atlanta,: Georgia 30323 ;;
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1 . p _a [4 , w ENCLOSURE'1 y Unresolved Item'(URI)- 88-12-02,-A110wable Loads for Standard Component Supports 1 U-Bolt _and Unistrut Allowables-TVA has completed the study to confirm that the Category I pipe support
, configurations installed at Sequoyah Nuclear Plant'(SQN) that use U-bolts meet the,allowableDloads tabulated in Design Standard DS-Cl-6.13.. The 185 Unit 1 installations have been evaluated and qualified;to the standard.- The:
- 323 Unit 2-installations have been evaluated, and 299.have been qualified to the standard., The 24 unqualified-U-bolt support configurations will be modified before restart from the~ Unit 2' Cycle 5 refueling outage.
A study has also been completed to confirm that the Category I pipe support 3 configurations installed at SQN that use Unistrut clamps meet the allowable loads tabulated in Design Standard DS-Cl-6.14. The 90 Unit 1 installations have'been evaluated,'and 85 have been. qualified to the standard. ior Unit'2,
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the.262. installations have been evaluated, and 237 have been qualified to the standard. The five unqualified Unit 1 Unistrut clamp support configurations will be modified before restart from.the Unit 1 Cycle 5 refueling outage. The 25 unqualified Unit 2 Unistret clamp support configurations will be modified I before restart from the Unit 2 Cycle 5 refueling outage.
-Pre-NF (Not Manufactured to the Requirements of Subsection NF of the American
- Society of Mechanical Engineers [ ASME] Code) Mechanical Snubbers '
TVA'has determined that the' current faulted load capacity of 2.0 for pre-NF snubbers will; be reduced to 1.33. For Unit 1, walkdowns must be completed Eduring the Cycle 5 refueling outage-to identify pre-NF snubbers. After a- i review to determine which snubbers do not meet the 1.33 capacity f actor, TVA will' submit a schedule for the necessary Unit 1 modifications six weeks after
. restart from.-the Unit 1 Cycle 5 refueling outage. Five pre-NF mechanical-snubbers not meeting the 1.33 capacity factor have been identified in~the initial review for Unit 2. The necessary modifications will be-completed for Unit.2 before restart from the Unit 2 Cycle 5 refueling outage.
Post-NF Mechanical Snubbers j V .The appropriate vendor-certified load capacity data sheets were used for' post-NF mechanical snubbers installed at SQN. Attachment 1 to this enclosure is. Vendor. Figure BE-416N; it was used by TVA and provides rated loads for NF
' mechanical snubbers. With the exception of.Model Number PSA-1, all allowables are-lower than'the ASME Section III allowables. The BE-416N allowable for PSA-1 is- 2,320- pounds versus the ASME Section 111 allowable of 2,300 pounds.
This difference is less than L percent, which is considered insignificant. (Figure BE-416N does not provide a faulted allowable for PSA-100, but there are no safety-related PSA-100 snubbers in Category 1 structures at SQN.) 1
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URI 88-12-03, Design Basis Accident /Zero Period Acceleration (DBA/ZPA) Effects. 'I Prior to.the restart of SQN Unit 2 TVA contracted'with Bechtel North American Power Corporation to generically address the NRC concern related to the ZPA effect on piping for the containment DBA analysis. Bechtel responded to this d concern by performing a study of five piping analysis problems of lines attached to the steel containment vessel (SCV). The attributes and bases used to select piping analyses for review were described in Reference 1.
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Each of the attributes was selected to not only obtain a sample representative of the population of piping attached to the SCV, but also a representation of the most severely affected piping. The results of the study demonstrated that the sampled piping, supports, and SCV penetrations are still acceptable when the DBA/ZPA effects are considered. Based on this study, TVA concluded that a high degree of-confidence had been provided that other piping systems not included'in the sample would likewise be acceptable and no additional evaluations were required. TVA committed to incorporate DBA/ZPA effects into future piping reanalysis. During the 90-18 inspection, NRC challenged TVA's DBA/ZPA study in two areas; (1) the representativeness of the original sample and (2) the high design load-to-allowable load ratio for supports on.SCV attached piping. The first general area questioned by NRC related to the representativeness of the original five-problem study; specifically, NRC-questioned the exclusion of the hydrogen collector (HC) and containment spray (CS) piping f rom the original sample since support modifications were identified or expected from inclusion of DBA/ZPA effects. TVA.has reviewed the process used in the development of the original 1 five-pro 51em sample study. The sample selection was focused on selectinC representative piping having worst-case attributes. The HC and CS piping are the only piping systems rigidly attached to-the SCV greater than 6 inches in-diameter-and the only piping systems with extended, long axial runs. The HC ' and CS piping are thus not representative of SCV attached piping. Inclusion of the HC'and CS piping would have inappropriately skewed the conclusions of
- the sample evaluations. Also, each of these uniquely configured piping systems were identified to require full reanalysis, incorporating DBA/ZPA and revised vertical spectra effects. TVA concludes that the apparent basis for L
excluding the HC and CS from the Bechtel five-problem sample is appropriate. To more completely document this conclusion, TVA plans to amend the sample
- study report to address exclusion of the HC and CS piping.
To give additional confidence and credence to the representativeness and conclusions of the original sample, TVA contracted with Gilbert Commonwealth, Incorporated (C/C) to undertake a summary study of the 21 SCV attached piping problems formally reanalyzed since 1988. These piping problems, selected because DBA/ZPA and revised vertical spectra effects were included in their reanalysis, were performed to document design changes and to resolve conditions adverse to quality. The effects of revised vertical SCV spectra, discussed in URI 88-12-10, were considered jointly with DBA/ZPA effects since the scope of piping is the same for both issues.
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- The purpose of the sample 1 study was to conclusively determine if modificationsL :
of-pipe supports resulting from the reanalysis were caused by the inclusion of' DBA/2PA and vertical spectra effects. Modifications were first reviewed to. ; determine if conservatism in the original support evaluation could be removed and'the status of the support changed to no modification required. If this review determined a support modification was still required, the piping was
- reanalyzed without the DBA/ZPA and revised vertical spectra effects. This reanalysis was then compared to the existing analysis with DBA/ZPA and' revised vertical' spectra effects included. Pipe support load changes were assessed ;
and m'difications o reviewed to determine if load' changes beca f DBA/ZPA and vertical spectra effects caused the modification. l. summar) che scope, methodology, and results- of this review is provided in Attactw at 2 to this
' enclosure. G/C has concluded that the 21 reanalyses of SCV attached: piping since 1988 resulted in no pipe support modifications driven by DBA/ZPA and revised vertical spectra effects. This summary study substantiates the original five-problem study and provides additional evidence that the DBA/ZPA
'and revised vertical spectra effects do not have a significant potential to impact existing qualification of pipe supports on.SCV attached piping.
The-second area questioned by NRC addressed the high design-to-allowable load ratio of pipe supports adjacent to the SCV. TVA acknowledges that the current design margin of supports near the SCV are potentially low and that support margin was not a sample attribute in the original Bechtel five-problem study. TVA, however,-considers the original study results to still be valid. This-conclusion is based on the following:
- 1. The G/Cusummary study of 21 reanalyses concluded that the DBA/ZPA and revised vertical s!."etra effects did not significantly impact the existing pipe support design.-
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~ 2 .' Current pipe support design' margins are normally based on conservative evaluation. techniques. Refined support analytical methods generally provide increases in support design margin.
- 3. Additional pipe support design margin is available if the'benefical effects of the leak-before-break (LBB) technology is included in the pipe support qualification. To obtain an indication of this potential load reduction, TVA contracted with C/C to evaluate the piping analyses in the '
g 21-problem study with and without LBB DBA spectra. The results of this study indicate an average decrease in support load of 23 percent.
- 4. 'The SQN design bases criteria requires pipe stress and support evaluations
-consider'the simultaneous occurrence of a DBA and a design basis earthquake. The maximum load effects of these two extremely low probability events must be individually determined and combined absol'utely. A more reasonable approach to account for time phasing of the events and. probability of event occurrence would be to combine the effects
.of these two events by the square root of the sum of the squares (SRSS) .;
method. The G/C 21-problem study determined that using the SRSS load
'. combination-in lieu of the absolute combination resulted in an average support load ceduction of 13 percent.
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n - , i ?) y e 5.. The. standard l linear response spectrum moda1' analysis-used for evaluat!on V+ of rigorous piping _at SQN has been i shown- to be f rom 11 to _42 percent more conservative than linear elastic time-history analysistas described in-Reference,2.- !
- I
, -In Reference 3, the ef fects of nonlinearities present in pipe supports. are ' 1 combined with an evaluation ~of.the response spectra versus time-history :
methods. This' study. indicated even more pronounced pipe' support load 1 ' reductions.
.'6.--Dynamic: testing of; piping _ systems initiated byiElectric Power Research- l Institute- (EPRI) and NRC -is described in Reference 4. The authors '
conclude that: current. code rule's based on static collapse load concepts for' dynamic load-considerations are overly _ conservative. Other reports.'of 3 dynamic. testing oftpiping' systems under dynamic loadings, such as those'in ' 1 References 5 and.6, conclude that piping systems and supports are capable of withstanding neveral times their rated capacities under dynamic loadings. Considering the'significant conservatisms noted above~in the current pipe. support load development and the conclusions from the sample _ studies . pipe supports with high interaction ratios on SCV attached piping do-not represent a safety concern at SQN. q
. References
- 1. TVA letter to NRC dated March 30, 1989, "Sequoyah Nuclear Plant'(SQN)
Units'1'and 2 - NRC Inspection Report Nos. 50-327/88-12 and 50-328/88 Response to Unresolved Item (URI)- 88-12-03"
- 2. " Comparison _of-LMFBR Piping Response Obtained Using Response-Spectrum.and Time-flistory Methods" by G. M. Ilubert, presented at the ASME'PV&P Conference in Dever, Colorado in. June of 1981,
- 3. " Seismic nA'alysis of Piping with Nonlinear. Supports" by D. A. Barta, et al, ASME PV&P_ Conference, August 1980.
- 4. " Seismic- Analysis- and . Testing of' Piping Systems and Components": by
.j S. W. Taggart, et al, PVP-Volume 144.
- 5. . " Experimental. Study of Piping Stability During Strong Earthquakes" by N
.N.-Ogawa, et'al, PVP-Volume 150.
6 "Iligh Level Seismic Tests of a Piping System at the HDR Facility" by
- L. Malcher, et al, PVP-Volume 182. ;
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g In'the" referenced letter..TVA'provided the requested-additional-;1nformation F (' concerning'(1)^the stee1Leontainment' vessel t orthotropic clastic propertles.-. ' y' -(2) the. theoretical: basis for:the double differentlationitechnique..(3);the_
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4A > Eexistence of a' spectral peak.beyond 10 Hertz,qand1(4)~the maximum steel. . containment; vessel response beyond 0.9: seconds.. 1
. Reference i
TVA letter to NRC dated June. 11, 1990i "Sequoyah-Nuclear Plant'(SQN) Units l1=
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jl, y - LandJ2 NRC. Inspection Report Nos. 50-327, 50-328/88-12 -. Unresolved Item 2 (URI ) 12-04. '.' ~ 4q g, ,
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4 u , , lURI-12-05,* Essential' Raw Cooling Water.(ERCW) Pumphouse 9 P- > - . -- '. . P : As L eonunit ted '. in Ref t:rences?l- and 2, TVA will; complete. the ' limited exploration i 0 et ' program by. September 11,,1990. The -satisfaction 'of the' design- requirements for- -! _the operating basis earthquake (OBE) concurrent with water = level at' m Elevation 704' will' be confirmed as conunttted in Reference' 3. s ^
.. -h Refe onces at ,
. 1. ' TVA' letter to NRC date'd December 28,1 1988, Sequoyah' Nuclear Plant (SQN') --
['; Essential Raw Cooling Water (ERCW) Pumphouse Foundation and ERCW. Pumping. r s Station Access Cells" [I 4 ,
- 2. _ TVA letterLto NRC dated March 1,1990,I"Sequoyah Nucle'ar Plant (SQN) -
Essentlal Raw Cooling Water (ERCW)-Pumphouse Foundation and ERCW Pumping; 5 Station' Access Cells" , 1
- 3.; NRC letter to .NRC dated March 3',- 1988, Sequoyah! Nuclear Plant = Essential :i k ,
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E . i . . Compone.nt- Damping Values URI 88-12-08',- w -
;' l g _During the'90-18 inspection it'_was agreed that the remaining NRC concern for i
URI 88-12-08:was limited to mechanical components satisfying the folloving- q
. parameters.
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- 1. Required for safe shutdown as defined for the 1979-1982 SQN seismic-margins review.
- 2. Welded construction.-
1 . 3. Located in buildings listed in the Final Safety Anal'ysis Report.(FSAR) ( Table 3.7.1-1. 0 4. Non-rigid (fundamental structural frequency less than 25 Hertz). 4 f- -5. Floor / wall mounted (not line mounted).-
- 6. . Seismically analyzed after March 1982 using safe shutdown earthquake (SSE, g SQN's' design basis earthquake) damping values greater-than 2 percent of ;
critical. 4
.N-These limiting parameters were established based on tho' recognition that NRC's '
concern relates specifically to the SQN seismic-margins review for mechanical J 1 components. -This review was completed in March 1982 and is briefly discussed @, 'in Supplenent 1, Section 2.5 of the SQN Safety Evaluation ReportLNUREG-0011. ;
, The specific-concern is that>3. percent damping eas used for some,recent design '
h basis.SSE~ analyses whe.reas the original-vendor seismic qualification used
'2 percent in'accordance with the,TVA procurement requirementr, thereby
- reducing margins implied by the seismic-margins review in 1979.
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TVA used' Appendix F of the-QualityfAssurance (QA) manual to establish-vendor requirements for ' seismic qualification of equipment prior to > A September ~1, 1974. After that date, the vendor's seismic-qualification .i m requirements were established by WB-DC-40-31.2 Revision 1. This information '[
- m. was reviewed in the seismic-qualification-review team (SQRT) audit in 1976 and 1977Jand documented by References 1 and 2. This change in: vendor requirements was made to comply with TVA's design basis licensing. commitment to-apply the :!
criteria ofxInstitute of Electrical-and Electronics Engineers (IEEE) 344-1975 5
'/, to'.the. fullest extent reasonably _possible. WB-DC-40-31.2, Revision 1 :
specified equipment _ damping values consistent with IEEE 344-1975. In j addition, as documented by References 1 and 3, the SQRT required demonstration of equivalent seismic qualification to IEEE 344-1975 for equipment procured to j y, IEEE 344-1971 (e.g.' Appendix'F) requirements. TVA completed the required equivalency demonstration in June 1979. I s .TVA: pointed.out,in Reference 4, and NRC concurred in Reference 5 that SQN is s , notian A46 plant because it was audited to IEEE 344-1975' or equivalent as its seismic-design basis for mechanical and electrical equipment.--Since this was the case, use of 2 and 3 percent damping for seismic analysis of floor- or wall-mounted equipment and components is in accordance with the SQN design
, basis.
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f " i; g , % ~: -8 . { TVA.recognises the need to clarify the FSAR and issue updated seismic-design criteria to more clearly reflect the design-basis-licensing commitments.for F
. seismic' qualification of SQN mechanical and electr.lcal equipment. An updated
' design' criteria'to replace Appendix F.of the QA'manualiwill be prepared by October 1, 1990. TVA will update he FSAR to clarify the design basis
, relative to compliance with IEEE 344-1975 or equivalent for mechanical and .
E electrical equipment and components in the next annual update. l TVA conducted a seismic-margins review as requested by NRC and the Advisory Committee on Reactor Safeguards (ACRS), in 1979 to 1982. The objective of the review was to demonstrate that adequate margins exist in the SQN designs to
, . . withstand an 84'" percentile site-specific earthquake. This review did not constitute a change in the SQN design basis. .
The initial review results were documented in Reference 6. Piping was evaluated by analysis of selected systems using Regulatory Guide (RC) 1.61 damping values and allowable pipe stresses in accordance with ASA B31.1. The-conclusions as stated in Reference 6 for piping and equipment wer : "The i seismic margins were not quantified for selected mechanical and e =ctrical equipment in safe shutdown systems but the equipment was evaluate against the revised floor response spectra and is considered qualified. Base on the r results of the reanalysis of the selected systems;and the reevaluation of the P electrical and mechanical equipment, we concluded that the. piping systema and mechanical equipment in safe shutdown systems are sufficiently conservative in design to meet current licensing criteria." The ACRS recommended a low-power license based on these results on December 11, 1979, but also recommended an expanded study to determine the seismic-design margin of all structures and equipment necessary for safe shutdown. NRC staff accepted this recommendation and required expansion of the seismic-margins review, n TVA submitted a seismic-margins-program plan by. Reference 7. .This program was completed by TVA, and NRC was notified of completion by Reference 8. The results were that the original' qualification analysis or test for the
. equipment was generally sufficiently conservative to envelope the 4 ,
84' percentile site-specific floor responsc spectra at the equipment locations. These results were presented to NRC on March 29-30, 1982, in fulfIliment of the August 11, 1980, commitment. Although precise acceptance criteria for SQN's seismic-margins review were not defined, criteria equivalent to that used for assessing'scismic margins for the systematic-evaluation program plants would have been reasonable. This . criteria is defined in NUREG/CR-0098 entitled " Development of Criteria for Seismic Review of Selected Nuclear Power Plants." It-is based on current earthquake definitions (RG 1.60 or 84'" percentile site-specific) and recommended qualification techniques for those earthquake definitions. NUREG/CR-0098 includes a table of recommended damping values that is provided as Attachment 3 to this enclosure. The values in the left column of this table are lower-bound values and those in the right column are average
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- 4. S%jn ' N !!54aridi7 g /jl'ower-bound components percent of crit 3 caland subjected to an 84
_. -The. average'-dampity' corresponding values for-vit'al values piping?would for. welded 4 @ f lM $ ;beff 3nd 3 percent of~ critical. 4 h .. . Sdditionallyi.NUREG/CR-0098 recommends acceptable ductility factors of 1~.0'to-IO A 31.3: for vita.1 equipment , components, and structur.es. This implies a steel . MJ support / anchorage 1 allowable load increase of- 25 percent or more over the?
> design-b' asis i limit-_ of 0.9. .yleid. Other f eatures - of NUREG/CR-0098 would ,also; Mi! " 4 t
yW permit more realistic prediction of' dynamic response, for example Section 7.4
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4 4 'on:the effects of inelastic action and Section-7.9 on the responsefof Lequipment and attachments.. The approaches taken by TVAlto demonstrateL." Jadequate-seismic-margins were more conservative than those recommended in' ?;
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k'W~; . lThe' largest potential effeet of equipment / component. damping is indicated by 1
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.the peak of?the_ response spectra. A comparison follows of-the peak spectral
, if ! accelerations'in-the' auxiliary and control building lwh'ere the majority.of'the pfN , '
- Category.1' equipment is located.- This comparison:is for design: basis SSE
"" Lspectra peaks at 3' percent' damping versus 84'" W u d h m e m W h .
,. 3; ' - , 1spectt;a at' 5 -and 7. percent damping. The 3 percent design-basis damping'is in.
accordance with IEEE 344-1975, and the 5 andL7 percent damping are in- ,l y. Laccordance with-NUREG/CR-0098.- The building spectra were derived for the 9) N.A , s84'"jpercent11e site-specific earthquake using RC 1.61 damping values as t part ofLthe'scismic-margins review described above. In accordanc'c with ( ' 1
, RG 1.61, 7. percent building' damping was specifled. (This damping value J; -corresponds with the NUREG/CR-0098 recommended lower-bound value1 for.
- reinforced concrete.)
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(> !TheLfollowing table compsres. North-South and East-West horizontal peak values ,
, o : Fat ltheLb'uilding base,0 Elevation 713.5 (where the-CCW heat exchanger and other-L + ; major heat exchangers are? located), 'and Elevation' 791.25, which is the top of l tho.butidingi: . Comparisons are made for ths horizontal < direction only since 1 ,
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.that'.'dampinglis not a: factor for vertical response. . -i g
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, Direction' Elevation Spectra-Peaks
, l, _ Design Basis SSE 84 " Percentile (%) Site-Specific-gm 2'c .
3% Damping- 5% Damping 7% Dampin_g- m nNorth-South 669.0 (Base) 0.50g 0.65g 0.56g. Y Q"*,( EastpWest. 669.0 (Base) 0.50g 0.65g 0.56g- ~j n m hy (NorthiSouth 7'13.5 1.34g 1.42g 1.20g- -
' ; , East-Westt 713.5 _1.69g 1.68g 1.40g y o M, .
l y'" North-South 791.25 (Top) 2.76g 2.70g 2.31g. . East'-West _791.25-(Top) 3.17g 3.06g 2.50g- l 3 [ , i'.i D
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, ~ From this comparison it is clear that qualification of the welded construction mechanical components ofLconcern for the design basis SSE, 3 percent damping, and design-basis allowable stresses effectively ensures compliance with' NUREG/CR-0098 criteria without any further evaluation. This provides verification of the seismic-margins review results in that regard. It also
-ensures sufficient' seismic margin for the TVA-designed equipment / component supports and anchorages which were not evaluated in the 1979-1982 seismic-margins review. These supports and anchorages were evaluated and verified as satisfying design-basis criteria in 1987.and 1988 as documented in s TVA's response to IDI audit deficiency D4.6-1. Equipment / component damping values of 2 and 3 percent were used for OBE and SSE, respectively, in those evaluations.
In conclusion, TVA believes that the licensing basis for SQN is compliance with IEEE 344-1975 or equivalent for the design-basis OBE and SSE. The use of damping valucs from IEEE 344-1975 in some:recent work to resolve NRC questions is not inconsistent with the 1Icensing basis. Furthermore, the fact that the damping values are larger than the original calculations does not invalidate any of the seismic-margins review work previously done. TVA has demonstrated that the licensing basis is met and that sufficient seismic margins still exist even in the revised evaluations. As a result, further work addressing the issue of seismic margin does not appear to be a prudent use of resources, and TVA considers this issue closed. References
- 1. NRC letter to TVA dated November 16, 1976, concerning the seismic team for SQN.
'2. TVA letter.to NRC dated February 7, 1977, " Seismic Qualification of Safety-Related Equipment - TVA Meeting with NRC Seismic Team September 28 through 30, 1976."
- 3. TVA letter to NRC dated June. 22, 1979, concerning the seismic-qualification data package.
- 4. TVA letter to NRC dated March 19, 1986, concerning seismic qualification of equipment at Sequoyah Nuclear Flant.
- 5. NRC letter to TVA dated October 29, 1987, "Sequoyah Nuclear Plar.t (SQN) -
Replacement Items Program (RIP) Seismic Screening Methodology."
- 6. NUREG-0011, Supplement 1 Section 2.5
- 7. TVA letter to NRC dated May 5, 1981, concerning the seismic-margin program plan.
-8. TVA letter to NRC dated March 1, 1982, concerning the seismic-margin program plan.
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- b b r i URL 88-12-09,, ERCW; Pumping Stiation Access Cel1s' a c r .
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-In Reference l',JTVA provided a response to NRC's concern regarding the'use of a friction' coefficient of 1.0 for tremie concrete-rock interfaces.
~
The 4
> ; response referred to American' Concrete Institute (ACI) Code 318 and shear-friction coefficients. For the ACI 318 shear-friction method, the-L normal force can be considered as either the internal member force on,the d' cross section or the: forces-caused by tensile deformation of the reinforcing R bars crossing the shear plane.- Although no reinforcing steel is present in the access cells or pumping-station foundations,.a normal force will exist I thatLwill provide shear resistance during a seismic event. The 1.0 value is j;
' intended for'use at a contact surface with an amplitude of. roughness of about 1
~1/4' inch. The.use of this value is considered to be conservative, since the )
actual amplitudes are nearer to one foot.
>- l Additionally, TVA has completed an evaluation' that further justifies the user !
y of a-l.0 friction coefficient for both concrete-rock and rock-rock interfaces. ] m The rock surface at the access cells is composed of interbedded shale !
'(56 percent) and limestone (44 percent). The rock surface is very-rough with a sperities close to each other because of the incliration of the interbedded ,
rock of different hardnesses. Based on mapping by divers, the angles of l inclination (1) of these asperities are generally in the range of 20.to, q
'25 degrees. ;
~
For the assessment of the concrete-rock contact surface, Reference 2-recommends a concrete-rock sliding friction coefficient tan p = 0.7, l (p = 35 degrees)'where p is defined as the angle of friction. This value
'is applicable for clean, sound rock on a relatively smooth surface for a ,
thorizontal plane, flowever the actual f rictional resistance of the l 1 concrete-rock contact surface is the combinatioa of concrete-rock sliding -j friction and the frictional resistance provided by the asperities. This , resistance is calculated by the following equation, which is found in ]
, Reference 3.
1 ' friction coefficient = tan (p + 1)
.Using a value for p of-35 degrees and a conservative-value for i of !
'10 degues (one-half.of the lower-bound value of the angle of asperities), the h tfriction coefficient for the concrete-rock contact surface is 1.0. 1 l
-In addition,. field-shear tests of concrete footings and blocks against various q types of rock have.been conducted at 44 Jam sites in various parts of the .[
world.- The results as reported in Refe'.ence 4 indicate that the friction =[ coefficient. exceeded 1.0 for concrete-cock contact surfaces. ;j For the assessment of the rock-rock friction coefficient, the evaluation i related the actual unconfined compressive strength test data (Reference 5) and L a
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G I: . W the anglesf of. inte'rnal f riction f rom published literature- (Reference 4) on - 1 rock 'similar to -tho'se at the ERCW access and pumping station cells. .TVA L conservatively ignored _the: cohesion measured.for limestone and shale. The _ - f riction coefficient for the rock-rock contact surf ace calculated by this approach _is 0.96.: !If:a cohesion value for the shale and limestone had been in 'taken;into~ account,-the' friction coefficients for the rock-rock contact surface
~
At would exceed 1.0. Based upon the TVA evaluation, a literature search, and the interpretation'of-3, '. ACIL318, it.is concluded that a f riction f actor of 1.0 is an appropriate value
~
for use in the sliding stability analysis'of the'ERCW cells. n,. ' -References r I TVA l'etter to NRC-dated December 28,1988, "Sequoyah Nuclear Plant (SQN) - 1. Essential Raw Cooling Water (ERCW) Pumphouse Foundation and ERCW Pumping Station Access Cells"
- 2. NAVFAC DM-7.2 Foundations on Eacth Structures, Department of the Navy,
. Alexandria,1VA.. 1982 3.'- R.-D. Lama and V.'S..Vutukuri, "llandbook on Mechanical Properties of Rocks;:VolumelIV:Trans Tech Publications, Aedermannsdorf, Switzerland, 1978 I
'4 R. 'D. ' Lama and V. S. Vutukuri, "llandbook on Mechanical Properties of .
l Rocks; Volume III Trans Tech Publications, Aedermannsdorf, Switzerland, b 1978 i q 15.;LSQNlFSAR Section 2.5
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W ? $ ' ( W t a ( F .- r-y p : f URI 88-12-10, Seismic Analysis of the SCV q Revised SCV Spectra Effecy Before the restart of SQN Unit 2. TVA performed a detailed study.of the'effect of the time-step issue on the SCV vertical spectra and the resulting impact on
~the qualification of piping systems attached to the SCV. In this study TVA
', ? ' evaluated ten piping analysis problems to assess the effects of the time-step change. . Four of the ten piping problems selected to be evaluated contained worst-case attributes (low-allowable stress margins, modal frequency in the region of spectra value increase, high contribution of-seismic load and support qualification to Interim Criterion CEB-CI-21.89) selected to maximize the 1 +..ct of the time-step issue. The remaining six analyses evaluated bad worst-case attributes (low-stress margins, high-elevation SCV rigid attachments, etc. ) for a prior SCV vertical spectra change and were i considered representative of SCV attached piping analysis for purposes'of the time-step evaluation. The ten-problem study showed either a decrease or a negligible increase in pipe stress, SCV penetration loads, support loads, and valve accelerations. TVA concluded from this study that the increase in pipe support loads because of the time-step change is insignificant and that no additional evaluation was required. In respc,nse to NRC's current concern on the justification of the effect of the time-step issue on recent (1988-1990) reanalyses of SCV attached piping, TVA has since initiated a summary study of 21 recently reanalyzed piping problems. The study, described in the response to URI 88-12-03 and further detailed in Attachment 2 to this enclosure, evaluated the impact of the revised vertical spectra and DBA/ZPA effects. This study demonstrated conclusively that none of the 21 piping problems evaluated resulted in pipe support modification driven by revised vertical spectra and DBA/ZPA. NRC also expressed concerns about the high design load-to-allowable load ratios for pipe supports on piping attached to the SCV. This issue is also addressed in the response to URI 88-12-03. Revised Reactor Coolant Loop (RCL) Spectra Effects TVA developed a three-phase program to address the effect of the revised RCL spectra on loop attached piping. The first phase of the program consisted of developing a worst-case impact ranking of loop attached piping and the selection of a sample of the most severely impact piping. The second phase of ( the p'rogram required the evaluation of this worst-case sample to determine if modifications were required by the revised spectra. If no modifications were identified, the impact of the revised spectra would be considered bounded and no further evaluations required. If modifications were identified, the third phase of the program would conduct a horizontal review of other RCL attached piping having similar worst-case attributes to determine if additional modifications would be required.
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_14 1 At the: time of the;NRC 90-18 inspection, TVA had completed only the first phase of the evaluation program.. Each of the attached piping problems was ranked in order.of. descending ~ anticipated impact. The impact was determined by factoring the current stress levels ~ by the spectral acceleration increases fer significant participating modes. Secondary worst-case-rankings were also developed for existing piping stress interaction ratios and existing pipe support margin (as indicated by identified post-restart modifications). To-ensure the study was also representative of the population of loop attached piping,! consideration was also given to selection of a variety of pipe sizes- -
.and attachments to each of the major loop components.
3 Using the worst-case rankings. TVA selected six piping-analysis problems in addition to ongoing reanalyses (because of design changes) of other RCL attutied piping to assess the revised RCL spectra impact.. Either directly or by similarity,:these piping analysis captured (1) the 14 most severely impacted stress analysis based on projected increased pipe stress; (2) six of the top'seven' piping problems ranked on the number of post-restart modifications;.(3) five of the top seven piping analysis ranked on existing pipe stress margin; (4) pipe sizes between 3/8. inch and 10 inches in diameter and (5) attachments to the reactor coolant pump, steam generator, hot leg, cold leg, and cross-over leg. To implement the second phase of the program TVA contracted with Bechtel to
- evaluate:the six additional piping analysis problems and with G/C to assess the piping reanalyses resulting f rom ongoing design changes. The Bechtel evaluation was! :ompleted in April of 1990 and was provided to NRC during the 90-18 inspection. Bechtel concluded from this study that the change in RCL spectra did not cause any of the modifications associated with the reanalysis of the six worst-case sample piping problems. The G/C review of ongoing design changes was incomplete at the time of the-90-18 inspection. G/C has since completed this review with only one RCL spectra-driven modification :
. identified. Phase three of the program, horizontal expansion in the failed sample p'apulationi has also been completed.' The horizontal sample expansion revealed'no additional support modifications were required.
t Based on the results of this review program, TVA considers the effccts of the
- revised RCL spectra to be bounded and no further evaluations required. The '
final report.on_this open issue will be completed and submitted to NRC by September 28, L990. 1
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!NRC has' requested that TVA, incorporate;the following design attributes into the applicable calculations. ;
i Address the minimum as-built gap of 1/2 inch that C/C documented.
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u . . L 2. -' Limit the ma;;imum permissible lateralipipe movement caused by ue combined-
- ~ effect.of thermal, and seismic' loads to Lhe minimum as-butit E4r j 3. Consider: the ef fect of .the relative lateral seismic movement of the: diesel I generator roof and slab'on the size of the minimum as-built gap.:
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- 4. . Consider the effect of the radial thermal growth of the 22-inch diameter exhaust-line on the size-of.the minimum as-built gap.
- 5., Confirm'that1the' latest design spectra of' record for.the diesel generator building were used-to analyze the exhaust lines.
, 16.. Confirm ~that the exhaust line piping, lugs, and supports have~been analyzed in~ accordance with the requirements of Design- Criteria Documents -
'e" SQN-DC-V-13.3 and SQN-DC-V-24.2 for TVA Class G Seismic Category 11 piping and supports. Although punching-shear design and lug design-are not specifically addressed in-the design' criteria, qualification of these support components will be addressed and documented in the calculations. I' J -
- 7. ' Evaluate the~ piping configuration for-the thermal case alone with friction.
'8.- Confirm that the axial growth of the exhaust line silencer has been j includedLin the piping analysis. ,
tall;of these eight attributes will bel incorporated into piping analysis and
= pipe support calculations by SeptemberL28,~1990.
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; Unresolved Item (URI) 88-12-03 identifies a concern with.the exclusion of' 1
design-basis accident /zero period acceleration:(DBA/ZPA) from Sequoyah Nuclear N ' Plant-(SQN) rigorous. piping analysis performed before 1988. URI 88-12 .* Part 1 identifies a concern with the time step used to generate the steel containment vessel;(SCV) vertical seismic spectra., TVA has' developed new Y : vertical' seismic spectra curves for the,SCV, which have been in use since < 1988. 'In 1989 new~1eak-before-break (LBB) DBA spectra'were developed,.which reduce conservatisms inherent in the previously used DBA spectra and DBA-ZPA i values. LThe 21' rig % vus-piping analysis problet. rigidly attached to the SCV were j formally reanalyzed during the period from 1988 to 1990. A study of the r support modifications: required by these reanalyses _has been' performed'to [
-determine if'the modifications were driven by-the inertial response spectra l cases associated with the SCV. .The 21 problems can be grouped into four-categories: two hydrogen collector lines, seven upper containment vent cooler lines, eight lower containment vent cooler lines, and four other miscellaneous d reanalyses..
' Hydrogen Collector Line's.
The two_ hydrogen collector lines in Unit 2, Probicm Numbers N2-14-3R and 1 N2-14-4R, were reanalyzed'and-unitized in 1989.- The civil calc 61ation ! regeneration program for Unit 2 designed three load-driven modifications from I the cutput of_these reanalyses. The Unit I calculation regeneration program usedt the Unit 2 loads for support design, as the systems were identical and
.the-Unit-2 reanalyses were-state-of-the-art.
+= The. Unit 1 program' designed no modifications. -A review of the Unit I s calculations shows that the modifications in Unit 2 could have been' eliminated by neglecting the loads. generated by differential SCV movements in the static
-load cases. Neglecting these loads is justified since any two of the ;
attachment-points are only separated by a few feet. Therefore, the modifications designed for N2-14-3R and N2-14-4R were not driven by either of the inertial load cases.related to the SCV and were not required to_ achieve
' design criteria compliance. The_aupport design margin in.these lines was'-
adequate =to accomodate the revised SCV vertical spectra, missing mass effects from seismic-inertia cases, and DBA/ZPA. I Upper-Containment Vent-Cooler Lines Seven upper-containment, vent-cooler lines were reanalyzed in response to
~
- Significant Condition Report (SCR). SQN CEB -8626 'in 1988. The seven piping-analysis problems are: TN2-67-10A, N2-67-11A, N2-67-12A, N2-67-15A, N2-67-16A,
- N2- 67-37A,' and N2-6 7-38A'.
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m 7 ' i . . . 4 . The piping configurations of the seven lines-are very similar, but there are differences in the support schemes.. The three analysis problems with the largest number of support modifications from these reanalyses were chosen for study of SCV inertial effects. These piping problems are N2-ci7-11A, N2-67-12A, and N2-67-15A. The N2-67-11A analysis had a piping stress-driven modification as well as support load-driven modifications, and was therefore thought to be one of the most severe cases among the eight. N2-67-15A The modifications designed in the 1988 reanalysis did not arise from load i increases. The loads from the 1988 reanalysis are-lower overall than the j loads that existed before the analysis. The modifications were driven by j' design criteria changes and revisions to vendor-supplied component allowables. The same niodifIcations would have been required, if no reanalysis t had been= performed, by the calculation-regeneration program. Modifications of 1 this nature have been programmatically identified and addressed by the l calculation-regeneration; program and are not related to any SCV inertial-case 3 issues, j N2-67-12A The loads on supports in this analysis problem increased under the 1988 reanalysis. A piping run was made with the 1988 geometry and new LBB spectra,- and the support loads were evaluated for their ef fect on the previous configuration-of.the supports. The modification of the support at Node R1 would not have' been necessary with the loads produced by the LBB runs. The j modification at Node 141 was not load driven, the original support was adequate for the applied load. The modification was caused by a swing angle g problem. The modification at Node 144 was driven by an interference issue. i' It is concluded that adequate design margin for the loads generated by the current LBB piping run existed at the time'of the 1988 reanalysis, j N2-67-11A. 4 The two support modifications on problem N2-67-11A were driven by load increases caused by configurational changes identified by SCR SQN CEB 8626. .h Several. piping runs have been generated to determine the relative effects of h the' piping-configuration change, the DBA/ZPA effects, and the SCV vertical spectra change. These runs demonstrate that the bulk of the load changes was-
)
because of the configuration change, and the loads generated purely by these changes would have driven the modifications, independent of other issues. On ' the' support at Node 17 the load more than doubled, from 2,556 to 6,465 pounds, simply because of the revision of the geometry. The addition of missing mass effects and new vertical spectra raised the 6,465-pound load to 6,614 pounds, ; an increase of 2 percent. The support at Node 110 was changed to a three way ;
- to take load off the support at Node 17 and reduce overall stresses in the system. These two modifications were driven by configurational changes in the pipe routing and were not significantly affected by missing mass or SCV vertical seismic spectra changes.
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'The conclusion of'the study of the three upper-containment vent-cooler lines F .is that the^ support modifications were'not driven by the revised SCV vertical spectra, ' seismic missing mass, or .DBA/ZPA ef fects, but by the configuration changes identified by the SCR. .The three lines studied effectively enveloped the support configuration of all seven lines, therefore it is believed that the conclusion applies to all seven lines.
- Lower-Containment Vent-Cooler Lines Eight of the lower-containment vent-caoler lines were reanalyzed in 1988 to incorporate a design change to the isolation valves inside containment. The isolation check valves had been failing leak rate tests and were relocated and replaced with motor-operated butterfly valves. The eight analysis problems. ,
were: N2-67-1R, N2-67-3R, N2-67-5R, N2-67-7R, N2-67-9R, N2-67-10R, N2-67-12R, and N2-67-26R. The eight lines are geometrically similar, but the support configurations vary. 'Three of the eight lines have been chosen to represent the various ' support configurations, N2-67-3R, N2-67-5R. and N2-67-7R. These three analyses have been studied for the effects of the inertial cases associated with the SCV. N2-67-3R-A piping. analysis identical to the 1988 Code of Record Analysis was performed
.using the old SCV SSE vertical spectra and neglecting missing mass in both the -c seismic and DBA cases. The faulted loads for the supports modified under the 1988 reanalysis have been compared to the loads from this run, and no
/ variation of more than 7 pounds was observed. It is concluded that.the modifications' required by the 1988 reanalysis were not caused by DBA/ZPA or by the revised:SCV vertical spectra but by the relocation and change to the valve.
.N2-67-5R A piping analysis identical to the 1988 Code of Record Analysis was performed using the old SCV SSE vertical spectra and neglecting missing mass in both the seismic and DBA cases. The faulted loads for the supports modified under the 1988-reanalysis nave been compared to the loads from this run, and no variation of. more than 8 pounds was observed. It is concluded that the
- modifications required by the 1988 reanalysis were not caused by DBA/ZPA or by the revised SCV vertical spectra but by the relocation and change to the valve.
N2 7R A~ piping analysis identical to the 1988 Code of Record Analysis was performed using the-old SCV.SSE vertical spectra and neglecting missing mass in both the
-seismic and DBA cases. The faulted loads for the supports modified under the 1988 reanalysis have been compared to tbo loads from this run, and no variation of more than 21 pounds was observed. This 21-pound increase was on a 6,000-pound load and is insignificant. It is concluded that the modifications required by the 1988 reanalysis were not caused by DBA/ZPA or by
.the revised SCV vertical spectra but by the relocation and change to the valve.
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P' lis'thatLthe!supportmodificationswere'notdrivenby.thrrevised-SCV)vertleal
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; changes: relatin'g to the valve' relocations.. The'three lines. studied effectivelyjenveloped the; support. configuration of all eight lines,-therefore-the conclusion should apply to all eight' lines.' ,
P
- . Miscellaneous.Other Reanalyses
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'N2-67-2A, N2-67-3A4 N2-77-1R (Unit 1), and 060010_4'-04-01.(Unit.2) have been:
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< reanalyzed since 1988 using new SCV. vertical seismic spectra and missing mass .
9 t'echniques for b'oth- seismic and DBA load ' cases, and no modifications. iti the-
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- vicinity;of thelSCV were required. The conclusion drawn is thatisufficient design margin' existed.in pipe supports to accomodate-the two SCV-related URI 1t issues'. ~
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ri. , PIPING ANALYSES WITH RICID CONNECTIONS X'O THE CCV - REANALYZED SINCE 1988 a.. __
.9 PROBLEM #- DESCRIPTION UNIT PEN # ELEV AZIMUTH PIPE SIZE NO. OF SUPPORTS N2-14-3R HYDROGEN COLLECTOR 2 -
736'-848' 250* 12" 14 l N2-14-4R HYDROGEN COLLECTOR 2 - 736'-848' 295* 12" 14 N2-67-10A UPPER ERCW CONTNMNT 1 X69 - 772' 301* i 11A 2" 7 VENT COOLERS 1 X70 774' 301* 2" l 12A " 10 s 1 X71 776' 301* 2" 15A " 10 1 X74 774' 301* 2" 13 l 16A " 1 X75 775' 301* 37A " 2" 13 1 X68 770' 301* 2" 8 38A " 1 X72 769' 301* 2" 8 N2-67-1R LOWEP ERCW CONTNMT 1 X58 697' 7* 6" 17 j 3R VENT COOLERS 1 X60 G97' 173* i SR " 6" 17 1 X62 697' 187* 6" 17 l 7R " 1 'X56 697' 353* 6" 17 9R " 2 X56 697' 353* 6" .17-10R " 2 X60 697' 173* 12R " 6" 17 2 X62 697' 187* 6" 17 26R " 2 X58 6978 7* 6" 17 N2-67-2A ERCW SUPPLY HEADER 1,2 X58 697' 7* 6" 157 X56 697' 173* 6" X60 697' 187* 6" X62 697' 353* 6" g l N2-67-3A4 ERCW DISCHARGE 2 X59 697' 8* 6" 31 HEADER X63 697' 189* 6" l N2-77-1A WASTE DISPOSAL 1 X39A 697' 280* 1" 7 0600154-04-01 DISCH FROM EXCESS: 2 X35 697' 301" 6" 18 j LETDOWN HX TO SCV
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gr },y' v ,1 .. ^ M1 c,} -'
.,s-
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+-
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i r , h* TABLE 1 RECOMMEWED CAMP!HG VALJES 1
!'~ 5 tress i.evel' Type and Condi tion.
- , -. . Percentage of Structure Crltleal Damping-Working stress, e. Vltal piping' 1 to 2' fy] * "'
- b. Welded steel, pros tressed 2 to 3 concrete, well reinforced concrete ,
(only slight cracking)
- c. Reinforced concrete with 3 to 5 considerable cracking
- d. Bolted and/or riveted steel.
5 to 7 wood structures with nelled or
- bolted joints At or jus t below . a. Vital piping yle1d point ' 1 to 3
- b. Welded steel, prestressed concrete 5 to 7 (wlthout complete loss in prestress)
- c. Prestressed concrete with I,o prostress left 7 to 1.0 ;
l
- d. Reinforced concrete 7 to 10 a., Bolted and/or riveted steel, wood. 10 to 15 structures, with bolted Joints i
- f. Wood structures with.nelled joints 15 to 20-
).
I
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re r , O!Dp ENCLOSURF, 2 l 4
- jd%
aw; ~- ,
? LIST OF COMMITMENTS- !
GT ' ' q
- n s ;
?C [ Unresolved Item (URI)-88-12-02 [
. l .1 The'24~ unqualified-U-bolt support configurations will be modified before !
j restart from!the. Unit 2 Cycle 5 refueling outage. v z =2. The five1 unqualified Unit 1 Unistrut clamp support: configurations will.be r l modified: beforeirestart f rom the Unit 1 Cycle 5. refueling outage.
- 13. LThe:25 unqualified Unit 2 Unistrut clamp support configurations.will be i '".. modified before-restart-from the Unit 2 Cycle 5 refuelingfoutage. 1 E
4.~ TVA will submit a schedule for the necessary Unit 1 modifications on. H pre-NF (not. manufactured _to the requirements of: subsection,NF of the- .[ c American Society- of Mechanical Engineers (ASME) Code) mechanical snubbers i 6 weeks after restart from the Unit 1-Cycle 5 refueling outage.
-5. 'The necessary modifications oa pre-NF mechanical snubbers will be
~
s , completed for Unit 2 before restart from the Unit 2 Cycle 5 refueling outage.'
~
URI 88-12-08 [K :1. 1An updated-design criteria to replace Appendix F of the Quality Assurance
;(QA)' manual will be. prepared by October 1, 1990.
2.1TVA will' update the Final Safety Analysis Report (FSAR) to clarify the Ldesign basis relative to: compliance with Institute of Electrical and
,Electronica Engineers,(IEEE) 344'-1975 or equivalent for' mechanical and ;
, electrical; equipment and components in the next annual update. 4 URI88-12-[0'
, -li JThe final report on the . revised reactor coolant loop spectra ef fects will
'be' complete-and submitted ~to NRC by September 28, 1990.
'URI 88-12-11
.l. All of NRC's eight requested design attributes will'be' incorporated into piping' analysis and pipe support calculations by September 28, 1990.-
l: i i i}}