ML19323E295
| ML19323E295 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 05/21/1980 |
| From: | Lundvall A BALTIMORE GAS & ELECTRIC CO. |
| To: | Clark R Office of Nuclear Reactor Regulation |
| References | |
| IEB-80-04, IEB-80-4, NUDOCS 8005230356 | |
| Download: ML19323E295 (5) | |
Text
_
=
m BALTIMORE GAS AND ELECTRIC COMPANY P. O. B OX 1475 BA LTIM OR E, M ARYLAN D 21203 May 21, 1980 ARTHUR E. LUNDVALL.J A.
vice Parseccur Su mpty Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 Attn:
Mr. Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Subject:
Calvert Cliffs Nuclear Power Plant Units Nos.1 & 2, Dockets Nos. 50 ~}l7 & 50-318 Automatic Initiation of Auxiliary Feedvater System
References:
a) NRC IE Bulletin No. 80-04 dated 2/8/80, b) BG&E letter dated 1/25/80 from A. E. Lundvall, Jr.
to R. W. Reid, Chief, same subject, c) BG&E letter dated 2/12/80 from A. E. Lundvall, Jr.
to B. H. Grier, same subject, and d) NRC letter dated 3/27/80 from D. G. Eisenhut to A. E. Lundvall, Jr., Automatic Initiation of AFWS Flov Gentlemen:
Reference (a) requested that we' perform analyses to determine the potential for containment overnressure or return-to-power following a main steam line break acci' dent inside containment with continued feedvater flow. Reference (b) provided, as a result of a separate NRC request, an evaluation of the potential for such an occurrence con erning the use of auxiliary feedvater. Reference (c) transmitted this esaluation to you as our response to Reference (a).
As a result of our continuing evaluation of the auxiliary feedvater system to comply with the requirements of NUREG 06h5 and NUREG 0578, we are forwarding this supplement to Reference (c) which provides additional infor-mation, including a proposed action concerning auxiliary feedvater discharge valves, for a further assessment of the auxiliary feedvater system. This assessment was scheduled in Reference (d) for completion by June 1, 1980. We feel that this uro, nosed action and the analysis submitted previously consititute a sufficient response to the concerns addressed in Reference (a).
Very truly yours, w
/Y
_'s a
8005230 M Q
Mr. R. A. Clark May 21, 1980 cc:
J. A. Biddison, Esquire G. F. Trowbridge, Esquire Mr. E. L. Conner, Jr.
e 0
8 e
0 k
e
=
9
RESPONSE TO IE BULLETIN 80-Oh 1.
Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break inside containment included the impact of other energy sources, such as continuation of feedvater or condensate flow.
In your review, consider your ability to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable after extended operation at runout flow.
Resnonce The containment pressure response analysis described in reference (a) accounted for auxiliary feedvater ( AFW) flow at the runout flow rate of 2200 gpm.
The entire AFW flov, which was assumed to reach the ruptured steam generator, was assumed to be initiated 180 seconds after the main steam line break.
In addition, main feedvater was assumed to ramp down to 5% of full power feedvater flow in 60 seconds, and remain at that rate throughout the transient. A more realistic main feedvater flow would ramp down to zero in 20 seconds. This assumption produces a more severe reactivity transient.
The ability to detect and isolate the damaged steam generator was not considered. This provides the maximum potential water inventory to the containment for calculating,the maximum containment pressure and to veriff.that the facility can withstand a main steam line break accident without identifying the ruptured steam generator. The auxiliary feed-water (AFW) pumps cannot run without severe cavitation at the runout flow used in the analyses described in Reference (a). With two pumps operating, against maxinum steam generator pressure, a maximum combined flow rate of 960 gpm c'an be reached without cavitation to either pump.
Therefore, in order to protect the auxiliary feedvater pumps, adminis-trative controls shall be initiated on June 1, 1980 (pending NRC approval) to set and to maintain'the AFW pump discharge valve at a 57% open position. This setting vill prevent cavitation of the pump at all pressures above the stead generator isolation signal setpoint, while providing adequate flow to the steam generators in order to maintain an adequate heat sink.
The 960 gpm maximum flow rate for the AFW pumps would result in lower peak containment pressure than that in Reference (a).
2.
Review your analysis of the reactivity increase which results from a main steam line break inside or outside containment. This review should consider the reactor cooldown rate and the potential for the l
reactor to return to power with the most reactive control rod in the l
fully withdrawn position.
If your previous analysis did not consider l
all potential water sources (such as those listed in 1 above) and if l
the reactivity increase is greater than previous analysis indicated l
the report of this review should include:
i
Paga 2 The boundary conditions for the analysis e.g., the end of life a.
shutdown margin, the moderator temperature coefficient, power level and the net effect of the associated steam generator water inventory on the reactor system cooling, etc.
The most restrictive single act'ive failure in the safety injection b.
system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant
- system, The effect of extended water supply to the affected steam generator c.
on the core criticality and return to power, The hot channel factors corresponding to the most reactive rod in the d.
fully withdrawn position at the end of life, and the Minimum Departure from Nucleate Boiling Radio (MDNBR) values for the analyzed transient.
Response
The analyses described in Reference (a) were performed assuming that the most reactive control rod was in the fully withdrawn position, and water sources were as listed above. Pertinent analytical details are described below, Moderator temperature coefficient used is described by a curve of a.
reactivity insertion versus temperature, which was presented as Figure 1 of Reference (a). Both analyses in Reference (a) used a 2754 MWt power level for. the full power analyses and 1 MWt for the zero-power analyses. The end of life shutdown margin was assumed to be 6.h% Ap.
Lov steam generator level was assumed to occur at the outset of each analysis.
Each analysis in Rqference (a) assuned the operation of only one high b.
pressure and one low pressure safuty injection pump. A conservatively lov value of -1.0, 40 per 95 PPM for boron reactivity worth was also assumed.
Reference (a) presents the return-to-power analysis assuming a main c.
steam line break with 2200 gpm auxiliary feedvater flow plus 5%
main feedvater flow feeding the ruptured steam generator.
d.
Hot channel factors and a minimum DNBR ratio value are not applicable for this transient since the critical heat flux vas not approached during the transient. This statement is also made in Reference (a).
3.
If the potential for containment overpressure exists or the reactor-return-to-power response vorsens, provide a proposed corrective action and a schedule for completion of the corrective action. If the unit is operating, provide a description of any interim action that vill be taken until the proposed corrective action is completed.
t
Paca 3
Response
As discussed in Reference (a), the reactor return-to-power and containment overpressure analyses. demonstrate that the reactor does not return to power and there is no containment overpressure condition as a result of main steam line break with auxiliary feedvater flow initiated after a minimum three minute time delay. Automatic initiation of auxiliary feedwater vill be incorporated with the associated time delay after NRC approval for this modification is obtained. This approval, described in Reference (b), is expected by June 1,1980. Until that installation is complete, auxiliary feedwater vill be manually controlled, and administrative controls have been initiated to prevent operation of the AW system during the first several minutes following a reactor trip. The runout flow rate of 2200 gpm used in Reference (a) did not take into account t,ae NPSH considerations associated with AW flow.
e e
4 k
O 4
%