ML20071N168

Submits Documentation of Seismic Analysis to Demonstrate That Snubbers Mods in Safety Relief Valve Line 124 Should Be Considered as long-term Fixes

ML20071N168
Person / Time
Site:	Brunswick
Issue date:	09/22/1982
From:	Zimmerman S CAROLINA POWER & LIGHT CO.
To:	Van Vliet J NRC
References
NUDOCS 8209270358
Download: ML20071N168 (4)

Text

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/ Carolina Power & Light Company September 22, 1982 Mr. J. A. Van Vliet Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 DOCKET !'0% 50-325 AND 50-324 LICENSE NOS. DPR-71 AND DPR-62 SEISMIC ANALYSIS

SUMMARY

On September 10, 1982, Carolina Power & Light Company (CP&L) presented analyses to NRC to demonstrate that modification of snubbers in safety relief valve (SRV) line 124 could be considered as long term fixes. NRC concurred with this determination. My September 20, 1982 letter to you committed to submit documentation of this analysis, which is provided in the following paragraphs.

DISCUSSION The approach used to arrive at the conclusion that pipe stressea and support loads were within structural integrity limits was as follows:

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1. Perform the analysis as per the FSAR commitments.

2. Compare the results with the pertinent FSAR limits.-

3. If the FSAR limits in (2) above were exceeded, address the significance of these noncompliance cases as follows:

Reanalyze the lines for load conditions with the design basis earthquake (DBE) loading and l

Evaluate the results against the structural integrity limits (previously defined in the initial phase of the

! NRC IE Bulletins 79-07, 79-02, and 79-14 evaluation),

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4. If the structural integrity limits in (3) above were exceeded, remove applicable conservatisms.

Step 4, as applied to SRV lines 124 was discussed in detail at our September 10, 1982 meeting. The use of a " global" envelope of amplified response spectra (ARS) was judged to yeild overly conservrtive results and thus " local" envelope ARS was used to perform the evalu tion.

8209270358 820922 PDR ADOCK 05000324 P PDR l -

411 Fayettevdle Street

• P. O. Box 1551

• Rateigh. N. C. 27602

Mr. Van Vliet September 22, 1982 Since Step 4 required certain engineering judgments, further details were provided on the approach used. To fully explain the rationale used, it was necessary to first describe certain key facts.

These were:

- The snubbers under investigation are attached to the SRV line 124 which in turn is attached to main steam (MS) line 14.

The SRV and MS lines were analyzed in a single analytical model.

- The " global" ARS envelope applicable to this single model envelopes ARS's which greatly differ from each other. Namely, the reactor pressure vessel (RPV) nozzle ARS is much higher than those applicable to all the other supporting structures such as the sacrificial wall, pedestal, drywell and the reactor building.

The lowest applicable system frequency (5.6 cps) is related to the lower section of the SRV lines rather than the MS lines.

- The snubbers under investigation are located on the lower section of the SRV line and separated from the MS/RPV nozzle anchor by several seismically active restraints.

With these facts in mind, it was judged that the analysis using the " global" ARS envelope (CASE I in Table 1) would provide overly conservative results and thus two additional analyses were performed.

CASE II used a " local" ARS which enveloped the Reactor Building (RB) at El. 50' . The basis for this selection was that the first system frequency is related to the lower section of the SRV lines where the most critical snubber 1073 under investigation is located. This snubber in turn is supported from the bottom of the drywell (El. 4') which in turn is part of the RB structure at this elevation. The conservative selection of an ARS at El. 50' in the RB was used rather than at El. 4".

CASE III was performed using a more conservative " local" ARS which enveloped all the structures supporting all the system restraints (including the MS line) excluding only the ARS at the RPV nozzle.

e_.-__ . ._. _ _ _ . _ _ . . _ . _ . _ _ - - _ _ . - - . - -__m _. . _ . _ _ . -._

Mr. Van Vliet September 22, 1982 CONCLUSION As the results in Table 1 indicate, both CASE II and III containsugstantialmarginwithrespecttothestructuralintegrity.

limits (15 ) and therefore compliance with the structural integrity limits was concurred in by the NRC.

Yours very truly, sty S. R. Zimmerm

• an Manager Licensing & Permits SRZ/dk (4671) cc: Mr. D. O. Myers (NRC-BSEP)

Mr. J. P. O'Reilly (NRC-RII) l i

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i TABLE 1 l.

BRUNSWICK STEAM ELECTRIC _ PLANT SEISMIC ANALYSIS RESULTS FOR SRV LINES 124 As DESIGNED ,

SWJBBER ERE =T "f CASE ' C ASE ' CASE A@ (pass.eired) cowa zo) $jg#U I E E N AM S 4)27SE @) Siss ,

6 64 5 ' i073 124 10 -(ts') G)21565 @ ) 369 o b) IS M ) '

45:o0) 8423 0) (3 ) vect.4 sa% 6 ao

4) & 330 (d 755 1037 124 g),3 7 30 ,, s {,,,) g) ,, g3 (2) (SIM) aso34) 28544] 0 ) <sevas am 6.ao g) co70 4) 7o4 14 t 57 124 go g (is5-) @)24330 G) 7820 19o6D) 54 2 i.O' n ) vecras.som soao

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t l

i CASE I ENVGL ARS (CEE )

i CAS E E L.cCAL ARS (DBE ) kB. Et 5o*

CASE 3Ir. ENuGL ARS (dbl?) bbdELL, S Ac. SHiGLD, PEDESTAL VP To EL SS' I I i 5

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