ML20040H205
| ML20040H205 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/02/1978 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Eicheldinger C WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| Shared Package | |
| ML18023A006 | List: |
| References | |
| NUDOCS 8202170441 | |
| Download: ML20040H205 (2) | |
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MAR 2 1978 Mr. C. Eiche1dinger, Manager
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Nuclear Safety Department
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Westinghouse Electric Corporation.
P. O. Box 355 c.
Pittsburgh, Pennsylvania 15230
Dear Mr. Eiche1dinger:
SUBJECT:
SAFETY EVALUATION OF WCAP-7832 The Nuclear Requiatory Comission staff has completed its review of
' J-Westinghouse Electric Corporation report WCAP-7532 entitled " Evaluation R.'
of Steam Generator Tube. Tube Sheet, and Divider Plate Under Coccined LOCA Plus SSE Condition." Our safety evaluation is enclosed.
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As a result of our review we have concluded that for the Model D and
.J Series 51 Westinghouse steam generators, the maximum stresses in new tube bundles, tube sheets and divider plates for combined seismic and b
LOCA loads (except pipe ruptures in steam generator compartrants) are 1:
less than limiting stresses in ASME Soiler and Pressure Yessel Code, Section !!!, Appendix F, and are the m fore acceptable. The exclusion of loads caused by reactor coolant pipe rupture in steam generator compart-ments cust be justified in individual applications. We have also IM-concluded that service-exoosed tubes containing defects caused by localized corrosion near the tube sheet will maintain their integrity during ccrbined seismic and LOCA loading provided the remaining wall thickness for tubes thinned by wastage is at least 72%.of the original nomir.a1 wall thickness and the length of through-wall cracks is less than the critical crack length. In Model D steam generators, the critical crack length is 0.64 inches for tubes with nominal wall thickness and 0.38 inches for tubes with 72% of the nominal wall thickness when sublected to
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mazio.nu expected pressure differential of 1435 psi. The acceptability of larger defects or other types of defects in service-excosed steam generators, such as denting in tubes, wastage and cracks at support plates, and cracks in tube U-bends must be determined by further analyses or experiments within the guidelines of Regulatory Guide 1.121, " Bases for Plugging Degraded Steam Generator Tubes.
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The use of the BLOCWM-2 coc-puter code for obtaining the primary side i
The SLODWN-2
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pressure re.sponse in the steam generator was reviewed.
code is acceptable for this use provided (1) the code is modified to j
use the correct seni~c velocity and the eorrect radial transport distance in the pressure iTessel lower plenum, and (2) a break opening tice of 1 militsecond is used.
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r Accordingly. WCAP-7832 is acceptable for referencing to support the
' I above conclusions. We do not intend to repeat our review of WCA?-7832 when it appears as a reference in a particular Itcense application t c. -
except to assure that the steam generator design.and tube defects are.'
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- In accordance with established procedure, we~ request that within three T..:
months of receiving this letter, you issue a revised version of WCAP-7832 to include this acceptance letter and additional information submitted L.
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during the review.
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' Sincerely.
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John F. Stolz, Chief
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'l.ight Water Reactors Branch No.1 E.I.:.'
Division of Project Management C~
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Enclosure:
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Accordingly. VCAP-iS32 is acceptable for referencing to support the above conclusions. We do not intend to repeat our review of WCA?-7832 wtan it appears as a reference in a particular license application t c. -
except to assure that the steam generator design.and tube defects am.',-
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In accordance with established procedure, we~ request that within three T...
months of receiving this letter, you issue a revised version of WCAP-7832 s
to include this acceptance letter and additional infor::,ation submitted during the review.
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Sincerely.
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John F. Stolz, Chief I"'
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Enclosure:
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into the progrcns yIXFM and ***ESTDYN-7 which calculste the dynamic respense (displacement histories) of the reactor coolant system. The displa:ement histery of the steam-generator is then applied to a model of the generator which includes the tube bundle, tube sheet and divider place, f rom which the internal force histories acting on these elements are obtained.
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The SSE leads are specified by an envelope of the floor response spectra at elevations in the reactor containment building corresponding to the
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upper and Icwer supports on the steam generator. The peak spectral acceleration as specified as 2.75 g's using damping of 1:;. Both hori-p.r.
zental and vertical ear;hquake =ction vere assured in the analysis.
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Following is a description of the analysis of the three components.
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Analysls of the tube bundle.
This analysis was performed by subjecting a model of the largest' 5:.
radius tube in the bundle to the loading described above, plus dead-
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veight. The analysis used the compute; progrw STASYS which perforts static and dynamic analysis of elastic bes: t'ype struc tures. The tube was assumed to have experienced thinning c,f 3 mils due to l
i eresion af ter 40 years of service (this does not include the local corrosion due to sludge deposits and local chemical hydraulic interaedion).
The largest prit.ary membrane stress intensity is based on the highest pressure dif f erence (Internal minus external), which exists before the dynamic loading is a; plied. The larges t priesty-menbrane-plus-pri..ary-bending stresa intensity was calculsco.i 'v superi posing i
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