View source for W3P86-2740, Forwards Addl Info Re 860624 Tech Spec Change Request NPF-38-23 to Implement Bypass for nonsafety-related High Steam Generator Level Trip & 861001 Tech Spec Change Request NPF-38-48 Re Regulating Control Element Assembly Limits

{{Adams
| number = ML20214D838
| issue date = 11/19/1986
| title = Forwards Addl Info Re 860624 Tech Spec Change Request NPF-38-23 to Implement Bypass for nonsafety-related High Steam Generator Level Trip & 861001 Tech Spec Change Request NPF-38-48 Re Regulating Control Element Assembly Limits
| author name = Cook K
| author affiliation = LOUISIANA POWER &amp; LIGHT CO.
| addressee name = Knighton G
| addressee affiliation = NRC OFFICE OF NUCLEAR REACTOR REGULATION (NRR)
| docket = 05000382
| license number = 
| contact person = 
| document report number = W3P86-2740, NUDOCS 8611240270
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE, UTILITY TO NRC
| page count = 2
}}

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                                                                                                                              +             P. O. BOX 60340 LOUISI&ANA POWER                                 LIGHT/                         317NEW BARONNESTREET ORLEANS, LOUISIANA 70160 * (504)595-3100 Eh?0$5?sNIO                                                                                                                                                                                                                     ,
November 19, 1986 L
U3P86-2740 A4.05 l                                                                                                                                                                                     QA Mr. George W. Knighton, Director PWR Project Directorate No. 7                                                                                                                                                                                                   (
Division of PWR Licensing-B Office of Nuclear Reactor Regulation Washington, D.C. 20555

==SUBJECT:==
Waterford SES Unit 3 Docket No. 50-382 Technical Specification Change Requests NPF-38-23, 48 Additional Information

==REFERENCES:==
(1) W3P86-2163, dated June 24, 1986 (2) W3P86-3331, dated October 1, 1986

==Dear Mr. Knighton:==

By the Reference (1) letter LP&L requested a Technical Specification change (NPF-38-23) to implement a bypass for the non-safety related high steam generator level trip.                       In Reference (2), changes to the Regulating CEA Insertion limits (NPF-38-48) were requested to support Cycle 2 operation.
In subsequent discussions with your staff additional information was requested.
Enclosed please find the LP&L response. Should you require additional information please contact Mike Meisner at (504) 595-2832.
Yours very truly, Y             OL K.W. Cook Nuclear Support & Licensing Manager KWC/MJM/pim cc:                     B.W. Churchill, W.M. Stevenson, R.D. Martin, J.H. Wilson, L. Kopp (NRC/MRR),
NRC Resident Inspector's Office (W3) h P
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                                                                                  "AN EQUAL OPPORTUNITY EMPLOYER''
                                                                                                                                                                                                                  '\\

e Attachment to e                                                                   W3P86-2740 Page 1 of 1 Additional Information License Amendment Requests NPF-38-23, 48 Question:
Provide a short written explanation of "tip-to-tip" operation.

===Response===
Tip-to-tip operation of CEA Regulating Groups 5 and 6 means that these groups can be inserted into the core by essentially the same amount. The intent is to be able to better control top peaked axial power shapes during low power opera-tion (particularly when returning to power following a reactor trip). The CEA groups will always be limited to the transient insertion limits shown in the proposed change to Figure 3.1-2 and, in practice, Group 5 will be slightly more withdrawn than Group 6 to avoid an out-of-sequence penalty factor and a pe'ential CPC generated reactor trip.
Question:
Provide justification that the containment pressure and temperature resulting from a Steam Line Break at 20% power or less with a higher Steam Generator inventory is bounded by the LOCA containment pressure and temperature results.

===Response===
The maximum containment temperatures and pressures resulting from a design basis accident are discussed in Section 6.2 of the Waterford-3 FSAR. Table 6.2.4 summarizes the results of both the LOCAs and the Steam Line Breaks that were analyzed in this section. As can be seen from this table (and associated Figures such as 6.2-7a and 6.2-8b), the highest containment pressure and temperature occur within 60 seconds following a break in the main steam line and blowdown of the affected steam generator ends at 200 seconds. The peak containment pressure and temperature therefore, occur well before the end of steam generator blowdown.
Thus, the additional inventory that may be present in the steam generator at 20%
power or less will not change the maximum containment pressure or temperature presented in the FSAR.
NS41223 I}}