ML19308B987
| ML19308B987 | |
| Person / Time | |
|---|---|
| Site: | Crane, Davis Besse  |
| Issue date: | 05/17/1979 |
| From: | Carlton J, Cartin L, Parks C BABCOCK & WILCOX CO. |
| To: | Luken R BABCOCK & WILCOX CO. |
| References | |
| TASK-TF, TASK-TMR 710517, NUDOCS 8001170728 | |
| Download: ML19308B987 (5) | |
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J. D. Carlton THE BABCOCK & WILCOX COMPANY B. M. Dunn R. E. Ham POWER GENERATION GROUP i
- 3. A. Karrasch To l
C. E. Parks R. C. Luken. Nuclear Service A.
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ses eas.s L. R. Cartin Plant Integration (2835)
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File No.
or Ref.
TECo T3.13 Subj.
Date Dual Level Satpoints for AF.! Control May 17, 1979 l m.
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Re:
E. A. Womack to Distribution, "TECo Supplemental Licensing Information,"
May 16, 1979.
Per Item 1 of the above reference, a revised evaluation of the need for dual auxiliary feedwater level setpoints in the wake of TMI-2 is attached.
This material basically indicates that the dual setpoints are required and defines the B&W analyses which show the adequacy of the 35-inch (indicated) setpoint for anticipated events and tha 96-inch (indicated) setpoint for s=m11 loss of coolant accidents (LOCA). Submittal of this information to the NRC as given or in similiar form is acceptable.
Discussion of additional LOCA analyses which would be required to optimize p
a single setpoint at a low level is not provided in the attached material.
l This material is provided below for your use in a cover letter.to TECo (not
' to be submitted to the NRC). Additional discussions with TECo will be required before commitments can be made in this area.
Considerations to Optimize a Single Setuoint for A W Control Relaxation of the LOCA requirement for auxiliary feedwater control at SG levels
> 120-inch (96" indicated) is not possible unless additional analyses are per-fonned. These analyses would include a feasibility study (limited scope) and a full and/or mini-spectrum for a licensing submittal.
In the licensing climate today, B&W anticipates that design changes which affect small break analyses will require extensive analysis backup. Therefore, the ultimate analysis require-ments (i.e., LOCA and other safety analysis) to document the impact of a revised SG 1evel setpoint cannot be estimated. To determine if a single setpoint (s 35-inch) is feasible, it is anticipated that approgimately 4 breaks would be required in the range of s 401 ft.~ up to N.1 ft...
These breaks are of the size where the steam generators remove a significant portion of the core decay heat and promote RCS depressurization via steam condensation when natural cir-
,culation is lost._ This analysis scope includes the re-analysis of the recent.
"Michaelson break" documented in Reference 5'of the attached material which used a 120-inch SG level. This proposed mini-spectrum using the B&W evaluation model would provide sufficient information to deduce the acceptability of low SG 1evel control during a small break and provide recommendations leading to optirJ 7ation of a lower setpoint value.
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- Cartin to Luken P0g3 Two Dual Level Sstpoints for AEW Control May 17,1979 The objective of the feasibility study, discussed above, will be to confir:n
'that a 35-inch control setpoint is adequate. This objective may not be achieved, and it may be necessary to require use of an internediate level with con-sideration of addition design modification other than a setpoint change for the " essential level" controllers. Consequently, additional B&W analysis for non-LOCA events may be required (i.e., loss of main feedwater, loss of offsite power). The extent of these efforts cannot be defined until LOCA impact studies provide some preliminary guidance.
B&W invites TEco's review of the above material, but does not reconsnend that work in this area be started at this time. This topic should be discussed further so as to identify the impact of a design change of this type (if possible) in todays licensing climate.
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",,,,, REVIEW AND APPROVAL 0
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C. E. Parks, Advisory Engineer. Plant Design Al
/7 )t?4f f D. Carlton, Unit Manager, PP&C#
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'%c j 7f B. M. Dunn, Unic Manager, ECCS l
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DUAL LETTEL SETp0INTS FOR AUTILIARY FEEDWATER CONTROL Current operating procedures at DB-1 specify the use of dual level setpoints for Auxiliary Feedwater Control per [interia operating instructions (Reference 1) issued by BW on November 29, 1978. The intent of these instructions was to support continued power operation of the plant by providing guidance to the operator for manual control of auxiliary feedwater level low in the SG, under certain circumstances, in lieu of high level control provided automatically by the " auto-essentail" level controllers.
In brief, the basis for the dual setpoint control scheme is as follows:
1.
Low SG Level Control (35-inches indicated on startup range instrumentation)
Test data and supplemental B&W analyses provide the design basis for the Low SG Level Setpoint for all anticipated events requiring auxiliary feedwater.
The natural circulation test (TP 800.04) demonstrated that the 35-inch (indicated) level vill provide adequate loop circulation for decay heat removal. B&W analyses (Reference 2) showed further that pressurizer level is maintained on scale for reactor trips due to or followed by a loss-of-main feedwater or a loss of offsite power. Therefore, under any condition except a small RCS break, DB-1 can be safely operated with an indicated SG 1evel of 35-inches on the startup Tange level instrumantation.
2.
High (120-inch) SG Level Control (96 -inches indicated on startup range instrumentation'.
Ratention of SG 1evel control at 96-inch (indicated) was recommended by B&W for use during design basis events only where both auxiliary feedwater and HPI are required. This recommendation was made to maintain the applicability of past small break analyses provided and/or referenced in BAW-10074A. The
- analyses provided in BAW-10074A are appropriate for DB-1 with auxiliary feedwater
_ controlled at SG 1evels as low as 120-inches :(96-inches indicated).
B&W has per-formed specific ~ analyses with the 12-Oinch SG ' level; these include the following:
1.
CIT line break 2.
HPI line break 2
3.
0.01 ft Break at the Pump Discharge
\\ Items l and 2 above have not been specifically reported'to t$a NRC; however,
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Item 3 is provided in Section 6.2.5 of Reference 5 in response to the concerns raised by the Michaelson Report. All three analyses above result in no core uncovery and meet the requirements of 10 CFR 50.46. '.B&W has also made comparison cf calenlat'ad LOCA results for DB-1 and other B&W plant types and concluded that the LOCA results reported in BAW-10074A are appropriate for a 120-inch SG level cnd fulfill licensing requirements _for continued operation of DB-1.
In the interim period pending incorporation of permanent design modifications to provide the dual setpoints for the " Auto Essential" level control, operator action is required to control the SG water level at 35-inches (indicated) for anticipated avents. A safety evaluation of the consequences that could occur if the operator fcils to sanually control SG 1evel has also been performed. This study (Reference 4) demonstrates that no unreviewed safety issues or detrimental accident consequences would result. Thus reliance on operator action in the interim is acceptable until final design modification are complete.
In summary, the dual level setpoints for auxiliary feedwater control is required. RCS Tasponse under condition where auxiliary feedwater is required and controlled per Esference 1 is supported or bounded by the Chapter 15 analyses provided in the FSAR and is consistent with the su.all break analysis provided in (1) BAW-10074A which is cpplicable for SG 1evels at or above 120-inch (96" indicated) and (2) Reference ;
which was requested by the NRC as a result of the TMI-2 incident. Furthermore, the used for the dual setpoint is independent of and not affected by the system changes (anticipatory trips, etc.) being made and the additional small break analyses performed co a result of the TMI-2 incident.
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References 1.
Fred R. Faist to T. D. Murray, " Control of Steam Generator Water Level Following SFRCS Actuation," SOM #424, November 29, 1978.
" Mini:num Pressurizar Level for Various Reactor Trip Transients," by Robert Winks, 86-2725-00, December 22, 1978.
3.
R. C. Luken to C. R. Domack, "0TSG Dual Level Setpoints - B&W Response to TBW-505," BWT-1975*, February 21, 1979.
4.
L. E. Roe to R. W. Reed, Serial No. 475, dated December 22, 1978.
5.
" Evaluation of Transient Behavior of Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant," May 7, 1979.
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