ML20248J801
| ML20248J801 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/06/1989 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8904170063 | |
| Download: ML20248J801 (8) | |
Text
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n-eav Wifey Power Station Shippingport, PA 15077 0004 I
L
. f2A$sEn'r$,ow to suuss April 6, 1989 U.
S. Nuclear Regulatory Commission
-Attn:
Document Control Desk Washington, DC 20555
Reference:
Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 Primary Component Support Snubber Elimination Gentlemen:
During a
meeting on February 28, 1989 between members of my staff and the NRC staff to discuss the subject request, Mr. Arnold Lee 'of the 'NRC staff requested clarification of our October 27, 1988 submittal.
Attached is additional information in the I
following areas:
1.
Discussion of Analytical Models 2.
Use of Damping 3.
Incorporation of Loads into ;he Surge Line Stratification Analysis If there are any questio.is on this matter, please contact my office.
4 Very truly yours, cah v\\
J.
D. Sieber Vice President
/
Nuclear Group l
cc:'
Mr. J.
Beall, Sr. Resident Inspector Mr. W. T. Russell, NRC Region I Administrator Mr.
P.
Tam, Sr. Project Manager 1
ij
- $gg41;@88[$$o$y{g2 P
1 DUQUESNE LIGHT COMPANY Nuclear Group Beaver Valley Power Station Unit 2 ATTACHMENT Additional Information Regarding Prin.ary Component Support Snubber Elimination Discussion of Analvtical Models The following supplements the mathematical models discussion in Section III of Attachment A
of our October 27, 1988 submittal.
All table and figure references are to the October 27, 1988 submittal.
The use of the different models is consistent with the original design basis.
The 3-loop STARDYNE model (Figure 3) was used for seismic analysis to obtain support
- loads, component interface (nozzle and foot) loads, and embedment loads.
Each loop has the same level of detail as the 1-loop STARDYNE model (Figure 4), and all embedments are connected to the building model.
With the exception of main steam and feedwater lines, no other branch lines are of sufficient stiffness or mass to influence the requested output loads.
The 1-loop STARDYNE model (Figure
- 4) was used for pipe rupture analyses of the main steam line, feedwater line,and pressurizer spray line.
A 3-loop model is not required, because the reactor pressure vessel and building do not move significantly as the main steam line and feedwater line loads are absorbed by the steam generator
- supports, and the pressurizer spray line forces are relatively small.
- Again, stiffness and mass effects of omitted branch lines are insignificant.
Separate NUPIPE models of each of the three loops (Loop A (Loop 21) shown by Figures 5 and 6) were used to obtain Class 1 stresses in the primary loop as summarized in Table 1.
All loading conditions were analyzed except for Pipe Rupture.
This is consistent with the analyses of record, as documented in the UFSAR Section 3.9, Table 3.9B-5 which requires only that pipe-to-pipe impact and jet impingement forces be included in the faulted case.
The appropriate major branch lines includi..J the surge line and bypass lines were included in the NUPIPE models.
The NUPIPE models were used primarily to evaluate pipe stresses.
Thermal displacements were used as input to the 1-loop STARDYNE model for evaluation of embedments, supports, and the equipment interface.
Figures 5
and 6
showed work sketches of Loop A (Loop 2:
Work sketches of Loops B and C (Loops 22 & 23) are attached.
For the postulated breaks cited above, the 1-loop STARDYNE model results showed that the main steam line and feedwater line loads were primarily absorbed by the steam generator supports, rather than being transmitted to the primary piping.
Jet impingement from the pressurizer spray line break (4 inch diameter) was determined to be unable to generate significant pipe stresses even in the cold leg of the affected loop.
4 Damoina Factors Utilized in' Dynamic Analysis The damping factors utilized in the various dynamic analyses supporting the large bore snubber optimization are consistent with those used in the original dasign basia.
1 1)
For the NUPIPE model of each loop, the' seismic analysis j
was based on amplified response. spectra'(ARS) and ASME l
Code Case N-411 damping values.
The impact of the j
-increased
-flexibility of the system due to the use of Code Case N-411 damping was verified as part of the as-built reconciliation of the original design basis and reverified as part of the current snubber optimization effort.
The code case damping was not utilized in conjunction with independent support motion, and was not mixed with Regulatory Guide 1.61 criteria.
Full compliance with the provisions of Regulatory Guide 1.84 Rev.
24 and the commitments made in conjunction with the j
authorization for use of the code case has been ensured.
2)
For the 3-Loop STARDYNE model, the seismic analysis was based on time-history methods and composite modal damping as per UFSAR Section 5.4.
3)
The 1-Loop STARDYNE model was used for the rupture analysis and was not used to generate seismic loads.
Pressurizer Surce Line Stratification Analysis The following is additional clarification regarding the Pressurizer Surge Line Thermal Stratification analysis (WCAP 12093).
1)
The stratification analysis described in WCAP 12093 included all applicable boundary effects of the reactor j
coolant loop piping and pressurizer.
These boundary effects were extracted from the Loop C (Loop 23) NUPIPE Model (Stress Problem 816-XC; Sketch attached).
2)
The WCAP 12093 analysis results replace the previous design analysis results for thermal expansion and fatigue on the surge
- line, i.e. NB-3600 equations 10 through 14 and usage factors.
Since no support modifications were
- made, equation 9 stress results from the previous design analysis are still applicable (deadweight, seismic and IOCA analyses).
These are not affected by thermal stratification loadings.
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