ML20147E905

Forwards Comments & Questions Re Review of NEDC-32523P, General Evaluations of GE Boiling Water Reactor Extended Power Update

ML20147E905
Person / Time
Issue date:	03/17/1997
From:	Kim W NRC (Affiliation Not Assigned)
To:	Marquino W GENERAL ELECTRIC CO.
References
TAC-M95087, NUDOCS 9703190270
Download: ML20147E905 (4)

Text

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e flarch 17, 1997

Mr. Wayne Marquino, Project Manager Technical Services

. General Electric Nuclear Energy 175 Curtner Avenue .

San Jose, California' 95125 ;!

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SUBJECT:

STAFF COM4ENTS"AND QUESTIONS REGARDING GE LICENSING TOPICAL REPORT j NEDC-32523P,;" GENERIC EVALUATIONS OF GENERAL ELECTRIC BOILING WATER REACTOR EXTENDED;. POWER'UPRATE"'(TAC NO. M95087)

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Dear Mr. Marquin'o:

's 1 The NRC staffl has ~ reviewed the beneral Electr'ic Licensing Topical Report

NEDC-32523P, " Generic Evaluations of General Electric Boiling Water Reactor j Extended Power Uprate,." (ELTR2), dated March 1996, and its Supplement 1, dated June'1996. Based on.this review,'the staff has generated the enclosed

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comments'and questions. The. staff requests that you provide your response l

? within 90 days. ,

4 If you.have<any questions regarding this request, please call me at

301-415-1392. >

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, Sincerely, 1

Orig. signed by l'

Tae Kim, Senior Project Manager Project Directorate III-1

Division of Reactor Projects - III/IV

Office of Nuclear Reactor Regulation i

Enclosure:

As stated DISTRIBUTION: See next page DOCUMENT NAME: G:\WPDOCS\UPRATE\GE ELTR2.RAI To receive a copy of this document, Indicate in the box C= Copy w/o attachment / enclosure E= Copy with attachment / enclosure N =

N3 copy 7

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, STAFF COMENTS AND QUESTIONS i

Reference:

General Electric submittal, Licensing Topical Report No. NEDC-32523P, " Generic Evaluation of General Electric Boiling  ;

Water Reactor Extended Power Uprate" (ELTR2), dated March 1996, l and its Supplement 1, dated June 1996 l

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1. In reference to Note #1 of Table 4-2, "ASME Code Equations as Affected by Power Uprate for Class 1, 2, and 3 Piping," clarify whether the existing

, design-basis analysis of the main steam and recirculation piping systems ,

use the design pressure of 1250 psig for calculation of primary pipe i

! stress such that the increase in operating pressure of about 75 psi has no  !

impact on the existing stress calculation.  !

! 2. In reference to Section 4.8, provide a detailed discussion of the impact I of the 26.9-percent steam flow increase on the design-basis analysis of the main steam piping due to main steam isolation valve (MSIV) closure and a turbine stop valve (TSV) closure loads, i l

3. Clarify whether the percentage increase in interface loads in Table 4-3 is applicable to pipe supports, nozzles, penetrations, guides, valves and pumps, heat exchangers, and anchors of the main steam and recirculation piping systems affected by the power uprate of up to 120 percent of the current rated power.

4. In Section 4.8.3, it is stated that pipe break locations and pipe whip restraint hardware capabilities are reviewed to demonstrate acceptability.

Table 4-3 does not provide results for the pipe break jet impingement or the pipe whip restraint loads. Please provide this information.

5. Discuss the applicability of Section 4.8, "BWR/5 Piping Evaluation,"

including stress increase percentages identified in Tables 4-3 and 4-4, to BWR/3, BWR/4, and BWR/6 nuclear plants. If Section 4.8 is not applicable, provide similar information for BWR/3, BWR/4, and BWR/6 plants.

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6. In Section 3.8.4.2, it is stated that "in most uprates, the highest pressure setpoint of SRVs (safety / relief valves) is not changed." Provide examples of the highest SRV setpoints for the power uprate as compared to the existing setpoints. Also, identify plants for which the power uprate will increase the highest SRV pressure setpoints and provide their SRV setpoints.

7. Table 3-13, " Overpressure Protection Results," shows that the power uprate of up to 120 percent of the original rated power results in an increase of 50 psi in operating pressure, and in some cases the increase in pressure results in exceeding the 1375 psia ASME Code limit. Provide similar overpressure protection results for a maximum operating pressure increase of 75 psi at the 20-percent power uprate conditions mentioned in the ELTR2. Also, discuss how the results are acceptable.

ENCLOSURE

-. - . . - - - - . . .- - - . - - . . . . . . -. . - = . . . - . . . - .

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, , 8. In reference to Section 4.4, provide an evaluation of the control rod i drive mechanism with regard to the stress and fatigue usage as a result of

! the 20-percent power uprate.

9. - In reference to Section 4.5.2, provide a discussion of the stress and fatigue usage of the recirculation piping, inlet and outlet nozzles, the jet pumps, and other safety-related components for the 20-percent power 4

uprate conditions in comparison to the existing design-basis analysis.

10. In reference to Section 4.6, provide an evaluation of flow-induced vibration for BWR safety / relief valves resulting from the increase in the steam flow rate at the 20-percent power uprate.

11. There appears to be an inconsistency in the power uprate report ELTR2 Supplement 1, Volume 1, regarding the increase in the steam flow rate: a

. flow rate increase of 24 percent was cited in Section 4.6, and a flow rate j increase of 26.9 percent was cited in Table 4-1. Please clarify.

i 12. In reference to Table 3-13, it is not clear how the results in some of the

columns were obtained. Please explain how numbers in the " Rise above Low Setpt" and the " Rise above initial" columns were obtained.

13. On page 32 of Supplement 1, the second and the third paragraphs discuss vessel pressures for BWR/5 and BWR/6 that appear to be inconsistent.

Please explain.

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DISTRIBUTION:

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S. Varga/J. Zwolinski B. Sheron G. Holahan B. Boger J. Strosnider R. Wessman G. Bagchi J. Calvo 4

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