ML20040H209
| ML20040H209 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/02/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18023A006 | List: |
| References | |
| NUDOCS 8202170445 | |
| Download: ML20040H209 (8) | |
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l ENCL.05URE MAR 2 1978 TOPICAL REPORT EVALUATIO!!
Report No: WCAP7832 Repert
Title:
Evaluation of Steam Generator Tube Tube Sheet and Divider Plate Under Combined I.OCA plus SSE Conditions.
Report Date: Dec. 1973 1
Supplement 1: Oct.'1974 Supplement 2: Dec. 1975 originating Organizatien: Westinghouse Electric Corp.
SU1!".ARY OF TOPICAL REPORT This report describes the structural analysis of tubes, tubesheet and divider plate of 51-Series and Model D Westinghouse steam generators
....v when subjected to cembined LOCA and SSE, loading. The objective of this gy, analysis was to demonstrate that the maximum stresses in these components
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fall within AS!'E B & T" " Code Section III allowable limits when subjected j".
l to these loading conditions.
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The hydraulic forces acting on these components consisted of two types:
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Forces on the steam-generator internals due to rupture of a main p
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coolant pipe.
The pipe break used for this analysis was a double ended coolant pipe rupture, located in the crossover leg immediately outside the steam generater coolant outlet noz:le. These forces were, derived f rom pressure-tice histories calculated by the computer program BLCDWN-2.
2.
Torces on the internals due to motion of the steam generator caused by the LOCA forces acting on the entire reacter coolant system. The progra-SATAN-V is used to calculate the presaure-time histories at varieus points in 'the braken and unbroken loeps. which are then used as input into the pregram STFS.UST f rom which the systes external force-hist: ries are calculatec. These histories are then introduced 8202170445 920212 PDR ADOCK 05000445 g
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MAR 2 197B absolutely the combined LOCA ef fects and the SSE affect. Both
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intensities were found to be below the limits for elastic system i
l analysis prescribed by the ASME Section 111 Appendix F criteri.a.
The largest ef fect was determined to be due to the variation of i
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internal pressure. The effect due to the systee motion was shown to be nuch smaller while the effect due to SSE was f ound to be of 6
a secondary nature (sbout 10% of the total ef fect). Other loads
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were also included in the analysis such as fluid centrifugal forces 3.R a.
and fluid frictien in the U-tube region. These were found to be
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negligible. Likewise, effects due to flow induced vibration were f._l '
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also found to be neg'ligible.
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Other effects which influence tube integrity were also investigated as follows:
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a.
The critical crack length for non-thinned D-Series tubes was
[? P-calculated as.64 in., based on a high ductility fracture mechanics 4
method developed at the Battele Memorial Institute, and e =aximum expected pressure differential of 1485 psi. For maxi =um uniformly thinned tubes the critical length was calculated as.38 in. when subjected to the rame pressure differential.
b.
A parametric study was performed to detereinc the nargins of unif orm tube vall thinning due to degradation which could be tolerated f thout exceeding Taulted Condition stress limits. For D-Series I'
i tubing of.75 in, ncninal diameter and.036 in. minimum thickness a decrease to a mini.us vall thickness of.006 in. vas found to de tolerable. Fcr 5*.-Series tubing of.*75 in, neninal diar.eter l
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and.050 nominal vall thickness the minimum sall thickness was f ound to be.021 in.
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The external collapse pressure for straight tubes.of nominal vall 1
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thickness was investigated analyticallly and experimentally. For D-Series tubing at 600* F and maximum allevableI 5% ovality this l
pressure was determined to exceed by 23% the maximum dif ferential pressure existing across the tubes subsequent to' blowdown.
2.
Tube sheet analysis 4*.~ -
The analysis of the !!cdel D tube shact under the LOCA loading was n..
performed by using the computer program A::SYS. The model used was
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a three dinensional finite element elastic model of the channel head, divider plate, tube sheet and stud barrel. The pressure loading which vas applied to this model originated f rom the hydraulic analysis described above. The maximum pressure differential was amplified by a lead f actor of 2, and applied s tatically to ahe codel. The maximum pri=ary membrane and priaary =embrant-plus-bending stresses were found to be well below tl.e Appendix F allovables for elastic system / elastic component analysis. Stresses due to steam generator movement and SSE vere also found to be negligible.
3 Divider. place analysis 3
!ased on the results of the tube sheet analysis the only significant leading cendition wh;eh was applied to the di.'ider plate was the i
internal pressure time history. The analysis was performed by using
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i a finite dif f erence, larne def ormation elastic-plactic dynamic i
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I Computer program called PETROS-III. The calculated results show that the maximum primary-membrane stress intensity was ' lover than the value prescribed by Appendix T for invinstic component analysis.
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h, MAR I III EDet;RY CF pEGULATORY EV/.1.UATIO!!
i The review was conducted by surveying the references and the background information of some of the computer programs mentioned in the report. A number of short ennfirmatory calculations of tube behavior under internal pressure and bending, and of crack propagation in tubes with thin vall 9
axial. cracks were also performed. They were found to support specific
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conclusions in the repo'rt.
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The applicant was also requested to submit additional references which
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h-supported the analytical methods and results described in the report.
These were examined and found to be acceptable.
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The SLODWN-2 computer code was reviewed by the NRC during the evcluation of the Westinghouse 'ULTIFLEX* computer code. The Tcpical Report Evaluation on TULTIFLEX was issued on June 17, 1977 (Reference 1). !WLTIFLEX is an
- .T extension of the 3LODUN-2 code and includes the effects of fluid-structure
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interactions.
The Ecdelic'; of the rNP. prieary system presented i'n.'.: CAP-7832 is the same asthat reviewed in MULT! FLEX. T.he modeling of the steam generator primary side features was approved at that time.
During the review of the Multiflex code, two changes.iere required to the
- L5DWil-2andmodelingportionsoftheanalysis. These changes carry over i
to the use of the BLCC'.rl-2 cede for this application. These chances,are (1) the use of the correct tonic velocity and '2.5 tre use of the correct i
radial transport distance in the pressure vessel louer plenue.. These i
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- '.4 CAP-5703, '"TLT! FLEX, A FCRTRrWi IV Comput.er Dregra;n f:r Analyzia.9 ' t.c. ma l -
Hydrau!'c-S+.ruct'. ire Sy-te, Dynamics."
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MAR 2 1978 charges sbculd be included for fature analyses, for cor.pleteness. These changeswillnothaveasignificantef'ectenthepressdreresponseinthe steam generator.
In addition to these changes, the break model,is limited to either a full, complete double-ended guillotine failure or asimple slot
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rupture. A break ocening time of onemilisecond is required for a licensing calculation.
RECULATORY POSITIOil
e find this report and subsequent supplements acceptsble as a reference
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e to support conclusions that tubes thinned by localized vastage or contain-ing smaller-than-critical-length thru-vall cracks vill withstand LOCA plus SSE loads, provided the vall thickness of the tube is at least 72% of its original nominal thickness. This value is conservative because of assumptions used in the analysis. Use of lessar vall thick-u-
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nesses should be justified by more refined analysis, which should be perfcrmed within the guidelines of R. C.1.121, or experiment. These tubes may also centain longitudinal thes-vall cracks cf a length which should not be exceeded to satisfy safety require =ents under these loading conditions.
For Model D tubes critical crack' lengths are.64 inches for healthy tubes and.38 inches for uniformly thinned tubes, when subjected to maximum expected pressure different;al of 1485 psi.
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The repert does not discuss various specific forms of degradation enecuntered j
since its sub ittal, such as dencing, vastage and cracks in the support s
plates, and cracks in the U-bends.
It is, theref ore, deemed insufficient t
to deter-ine the saf ety of these tubes with these f orms of degradation.
Further cre, it 19 not applicable to later e.oisls et steam generators I
which incorporste quat re-f ull t.Le support design.
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Finally, we find acceptable the methods of analysis usEd for determining the safety of the tut e sheet and divider piste when subjected to LOCA and SSE. and the conclusions derived from these analyses.
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The use of the BLOD'.:N-2 computer code, subject 'o the r.ccification and
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res'trictions as outlined above, is acceptable fo.- the evaluation of the steam generator tube, tube sheet and divider plate under combined LOCA
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plus SSE conditions. Reference 1 presents a detaile.1 evaluation of the
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BLOOW!:-2 code.
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.q References 1.
Letter from J. F. Stolz to C. Eicheldinger, : Topical Report Evaluation of Westinghouse WCAP-8700 (P) and WCAP 3709 (;1P)," June 17,1977.
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Mr. Thomas M. Anderson, Manager t'
Nuclear Safety Department
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,z Westfrighouse Electric Corporation
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P. O. Box 355 Pittsburgh, Pennsylvania ~ 15230 4_
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Dear Mr. Anderson:
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SUBJECT:
SAFETY EVALUATION OF
-79 AND WCAP-8236
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.h-The Nuclear Regulatory Commission staff has completed its review of the following Westinghouse Electric Corporation topical nports:
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.s WCAP-7950 (Non-proprietary) entitled " Fuel Assembly Safety h,.
i Analysis for Combined Seismic and Loss of Coolant Accident,"
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July 1972
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WCAP-8236 (Proprietary) and WCAP-8288 (Non-proprietar/)
entitled " Safety Analysis of the 17 x 17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident "
December 1973 7
WCAP-8236, Addendum 1 (Proprietary) and WCAP-8288, Addendum 1 (Non-proprietary) entitled " Safety Analysis
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of the 8-Grid 17 x 17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident," March 1974 s
WCAP-7950 describes the analytical methods and results of bounding I
calculations for a fuel assembly design having fuel rods in a 15 x 15 array; WCAP-8236 describes the analyses for at: assembly design having fuel rods in a 17 x 17 array and seven fuel md spacers, and WCAP-8236 Addendum 1 describes the analyses for an assembly design having fuel rods in a 17 x 17 array and eight fuel rod spacers.- Our safety evalua-tion of these reports is enclosed.
As a result of our review. we have concluded that the analytical methods described in these reports are acceptable; we have not reviewed the results of bounding calculations. Accordingly, the above listed topical reports are acceptable for reference in license applications.to support the cone clusion that analytical methods J or calculating loss of coolant accidant f
loads are acceptable. When a proprietary report is used as a reference, both the proprietary report and the non-proprietary version must bc referenced.
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NRC FORM Hs (9.?6) NRCM 0240
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_ r Mr. Thomas M. Anderson FEB 6 1979 In 'accordance Mih established procedure, it is requested that i
. Westinghouse issue mvised versions of these reports within three months of receipt of this letter to include the NRC acceptance letter, the enclosed evaluatioc, and any changes resulting from the review, including i
reouests and positions transmitted by our letters dated June 12,1974, July 25, 1974 an1 tspril 30,1975 and your responses dated September 20, 1974 and November 20,1975.
We do not intend to repeat our review of these reports when they appear as references in a particular license application except to assure that J
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the material presented in these reports is applicable to the' specific l.
l plant -involved.
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Sticuld Nuclear Regulatory.Comission criteria or regulations change, such that cur conclusions concerning these reports are invalidated, you will be notified and given an opportunity to revise and resubmit your topical reports, should you so d4 sire.
. Sincerely,
- Original signed by 5* owr.5Woir, chief i,'
- Light Water Reactors Branch No. 1
. Division of Project Management l.
Enclosure:
l Safety Evaluation Report cc w/ enclosure:
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.Mr. Dave Rawlins i
Westinghouse Electric Corporat6on P. O. Sox 355 L
Pittsburgh, Pennsylvania 15230 j
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.O ENCLOSURE Safety Evaluation Report
'on. Licensing Topical Reports by Westinghouse Electric Co.
" Safety Analysis of the' Fuel Assembly for Combined Seismic and Loss-of-Coolant Accident" WCAP-7950,. July 1972 WCAP-8236, December 1973
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WCAP-8236, Addendum 1. Narch 1974 7.
$* 'OUQ If;**.,j g
+s Core Perfomance Branch
,1-October 1978 i
UNITED STATES NUCLEAR REGULATORY COMMIS$10N WASHINGTON. D. C. 30555 t
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Evaluation Two Westinghouse topical reports '(Refs.1, 2) on fuel assembly mechanical response to seismic and LOCA loads have been reviewed; they both use a common analytical method.
The first report discusses the 15x15 fuel assembly design (Ref.1), and the second report dir:usses 'the 17x17 fuel assembly design (Ref. 2).
Since the issuance of these two reports, Westinghouse has changed the
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design of the 17x17 fuel assembly from 7 to 8 grids to reduce the mag-
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nitude of rod bowing. An addendum to WCAP-8236 was issued to account j
fre the change (Ref. 3).
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4 NRC initially reviewed the 17x17 fuel assembly for both 7 and 8 grid designs, and a preliminary safety evaluation was issued (Ref. 4). The staff evaluation noted that there was an adequately simulated vertical test to verify the calculated response of the vertical components from the seismic and LOCA loads. However, a simulated test for the horizontal response of the fuel assembly was not performed because the predominant load came from the vertical LOCA shock load. At that time we concluded i
that the reports were acceptable for plant licensing purposes, and that
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further review of the analytical method was not essential. No separate review was performed for the 15x15 fuel assembly because the analytical method used for the 15x15 fuel assembly duplicated that used for the 17x17 fuel assembly.
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.- :u Since the preliminary safety evaluation of the subject reports was issued, a new horizontal load on the fuel assembly was recognized in connection with a postulated inlet pipe break at the reactor nozzle safe ends (as first reported by VEPCO during the North Anna plant review). The previous NRC' conclusion has.therefore been reexamined because (1) the horizontal load h'as increased substantially, and (2) no satisfactory horizontal confirmatory test was available'.
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Because of the lack of complete experimental verification, NRC's consul-tants at the Idaho National Engineering Laboratory performed an independent calculation of horizontal loads for comparison with the Westinghouse calculation. On the basis of this comparison (Ref. 5), we conclude that
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the Westinghouse analytical method is still acceptable provided a conser-vative factor (sometires also called a safety factor) is applied to cover 7
analytical uncertainties. Conservative margin is needed because (1) the system is non-linear and (2) a full set of verification studies would be very difficult and was not performed. The magnitude of the conservative safety factor and other acceptance criteria have not been included in this review and will be considered for each plant.
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4-References 1.
L.T. Gesinski, " Fuel Assembly ~ Safety Analysis for Combined Siesmic and Loss-of Coolant Accident," Westinghouse Report WCAP-7950, July 1972.
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2.
L. T. Gesinski and D. Chiang, " Safety Analysis of the 17x17 Fuel Assembly for Combined Seismic and Luss-of-Coolant Accident," Westinghouse Report WCAP-8236 (non-proprietary version WCAP-8288), December 1973.
3.
L. T. Gesinski and G. LeBastard, " Safety Analysis of the 8-Grid 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident,"
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Addendum #1 to WCAP-8236, March 1974.
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4.
Letter from D. B. Vassallo of USNRC to R. Salvatori of Westinghouse, 0'.'
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June 12, 1974.
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5.
R.L. Grubb, "PWR Fuel Assembly Mechanical Response Analysis," Idaho i.-
National Engineering Laboratory, INEL Report RE-E-77-140 and
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Amendment No. 1 RE-E-77-140, March 22, 1977.
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