ML19257B263
| ML19257B263 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 01/09/1980 |
| From: | Johnson W Maine Yankee |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0578, RTR-NUREG-578 WMY-80-4, NUDOCS 8001150432 | |
| Download: ML19257B263 (6) | |
Text
.
RIAIRE
- HARHEE, T T.
TURNPIKE ROAD (RT. 9)
$ '~
,r ENGINEERING OFFICE WESTBORO, MASSACHUSETTS 01581 617-366-9011 B.3.2.1 January 9,1980 hNY 80-4 United States Nuclear Regulatory Comission Washington, D. C.
20555 Attention: Office of Nuclear Reactor Regulation
Reference:
1)
License No. DRP-36, Docket No. 50-309 2)
Maine Yankee letter to USNRC, WMY 79-113, dated October 18, 1979 3)
Maine Yankee letter to USNRC, WMY 79-138, dated November 20, 1979 4)
USNRC letter to Maine Yankee, Automatic Initiation of Auxiliary veedwater Systems at Maine Yankee Atomic Power Station, dated December 21, 1979.
5)
YAEC-1132, " Justification for 2630 MWt Operation of the Mal. iankee Atomic Power Station," dated July 1977.
6)
Maine Yankee letter to USNRC, WMY 78-62, Proposed Change #64, dated June 26, 1978.
7)
YAEC-1202, " Maine Yankee Cycle 5 Core Performance Analysis," transmitted by Maine Yankee letter WMY 79-143, Proposed Change #73, dated December 5, 1979.
8)
YAEC-1104, Maine Yankee Plant Accident Analysis Model Using FLASH-4, dated November 1976.
Dear Sir:
SUBJECT:
Automatic Initiation of Auxiliary Feedwater Systems In response to the Short Term Recomendation 2.1.7a (NUREG-0578) Maine Yankee comitted in References 2 and 3 to automate the auxiliary feedwater system provided that the safety analysis gives assurance that no degradation of safety will occur. Maine Yankee further committed that a proposed design using control grade components would be submitted in the near future.
In your letter of December 21, 1979 (Reference 4) you indicated that some licensees have raised the issue of the applicability of current analysis of a main steam line break or main feedwater line break assuming early initiation of auxiliary feedwater.
In question is whether the change in assumptions would increase the calculated containment pressure or the likelihood of return to power. This letter is in response to your letter (Reference 4) requesting resolution of this concern, y
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8001150 N 2.
United States Nuclear Regulatory Commission January 9,1980 Page 2 Maine Yankee has been evaluating the impact of feedwater addition from the auxiliary feedwater system during a main steam or feedwater line break with respect to the increased likelihood of return to power and possible increase in containment pressure. The particular breaks in question are those breaks in the steam or feed line that will initiate a blowdown in one steam generator (affected steam generator) resulting in a primary system cooldown and/or a containment pressurization. This evaluation has been performed in conjunction with nur commitment to implement automatic initiation of the auxiliary feedwater sptem. We do not believe that these questions are applicable to a manual auxiliary feedwater system due to the consciousness of the operator in taking action to initiate the system and his instruction not to feed the affected steam generator. An automatic system is likely to be initiated too quickly for the operator to take action to limit feeding the affected steam generator.
The current steam line/feedwater line break analyses for Maine Yankee are contained in References 5, 6, and 7.
The Reference 5 analysis was performed as part of the effort to increase Maine Yankee's core power rating from 2440 MWt to 2630 MWt and contains all the pertinent assumptions, plant response and sequence of events for the design basis. Details of the analytical model used in the steam line break analysis was provided in Reference 8.
The Reference 5 analysis, which forms the basis for current operation, has assumed a significant post-trip feedwater addition to the affected steam generator through the main feedwater system. The amount of feedwater assumed is based on plant measurements and includes leakage through the main feedwater regulation valves (FWRV).
In this analysis the feedwater system has been modeled explicitly to account for the proper distribution of post-trip feedwater flow and enthalpy. For the full power case, the FWRV's are ramped closed from the 100% feedwater flow position and the FWRV bypass valves open to their 5% flow position in 10 seconds after receipt of reactor trip signal.
Following reactor trip feeduater flows through the fully open FWRV bypass valves. The post trip feedwater flow also includes leakage through the FWRV.
The amount of feedwater flow reaching the affected steam generator in the analysis is 2320 gpm. The primary system cooldown resulting from the continuous addition of main feedwater in this manner is limited by the final saturation pressure of the affected steam geneator. For conservatism, the Reference 5 analysis took this as atmospheric pressure. For the licensing analysic presented in References 5, 6, and 7 the action of the CEA's in combination with the negative reactivity inserted by 1 HPSI (boron injection) is sufficient to keep the reactor subcritical. The plant temperature response for the full power case with main feedwater addition post-trip is shown in Figure 1 (taken from Reference 5).
This analysis has recently been repeated by considering the auto start of the auxiliary feedwater system at time zero.
The auxiliary feedwater system was initiated at time zero due to the uncertainty associated with the level response in the affected steam generator and its effect on initiating the auxiliary feedwater system. Low hydraulic resistance in the auxiliary feedwater system coupled with low pressure in the affected steam generator will cause the auxiliary feedwater pumps to go to runout and direct all the auxiliary feedwater to the ruptured steam genrator.
The amount of feedwater flow reaching the affected steam generator from both
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United States Nuclear Regulatory Commission January 9, 1980 Page 3 the main feedwater and auxiliary feedwater system is 4095 gpm. The impact on the plant cooldown from the auto start of the auxiliary feedwater system is shown in Figure 7.
The additionsi teedwater from the auxiliary feedwater system results in a steeper cool /
t rate but does not result in a greater cooldown as shown in Figure 2 to the necessity of boron injection, shown by the licensing analysis prer. sted in Reference 5, the steeper cooldown rate would result in a temporary return to power. By taking credit for the worth of the stuck rod, which is assumed in the licensing analysis, a return to power would not occur.
The containment pressure response to a steam line or feedline break inside containment was not addressed in Maine Yankee's FSAR. However, as part of the effort to increase core power to 2630 MWt, Maine Yankee performed an analysis of a spectrum of steam line breaks inside containment at various power levels. The results, reported in Reference 5, show that the design basis LOCA produces a more limiting condition with respect to containment response. As in the analysis described above, the containment pressure analysis for steam line breaks inside containment considered the effect of feedwater addition to the ruptured steam gener ator post-trip from the main feedwater system. This analysis conservatively assumed 31% full power feedwater flow (8788 gpm) to the affected steam generator. Although auxiliary feedwater was not directly considered sensitivity analysis performed as part of this study indicates that the key assumption with respect to containment pressure is how fast the intact steam generators are isolated by the action of the non-return and excess flow check valves and is rather insensitive to continuous auxiliary feedwater addition.
In the worst case the peak containment pressure is 45 psig and occurs 112 seconds into the event.
In order to prevent a return to power post steam / feed line break with automatic initiation of auxiliary feedwater, as predicted by a conservative licensing analysis, it will be necessary to limit the flow of auxiliary feedwater to the affected steam generator. The time frame of concern for return to power post steam line break is within the first five minutes of the event. Following 5 minutes, sufficient negative reactivity will be inserted from safety injection to prevent a return to power from the subsequent cooldown induced by auxiliary feedwater addition. Therefore, if the auxiliary feedwater system is not delivering flow for five minutes post steam line break, there is no concern for a return to power.
Therefore, the design for implementation of an automatic auxiliary feedwater system will include consideration to delay the automatic start of the auxiliary feedwater system by five minutes such that the existing steam line break analyses remain valid. This allows ample time to provide auxiliary feedwater for loss of feedwater events. As committed in Reference 3, details of the proposed design will be forwarded in the near future.
Proposed Technical Specifications for the auxiliary feedwater system modifications will also be made at that time, b
United States Nuclear Regulatory Comission January 9, 1980 Page 4 We trust that you will find this submittal satisf actory; however, should yew desire additional information feel free to contact us.
Respectfully submitted, MAINE YANKEE ATOMIC POWER COMPANY
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W. Pt Johnson Vice President i
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