ML20071C172
| ML20071C172 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 02/24/1983 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE, YANKEE ATOMIC ELECTRIC CO. |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| SBN-479, NUDOCS 8303010533 | |
| Download: ML20071C172 (9) | |
Text
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SEABROOK STATION Enginsedng ONies:
1671 Worceshr Rood h""3*8h*" M*"ochwem 01701 Pubuc h of New 6 (617) - 872 - 8100 February 24, 1983 SBN-479 T.F. B7.1.2 United States Nuclear Regulatory Commission Washington, D. C. 20555 Attentioa:
Mr. George W. Knighton, Chief Licensing Branch No. 3 Division of Licensing
References:
(a) Construction Permits CPPR-135 and CPPR-136, Docket Nos. 50-443 and 50-444 (b) PSNH Letter, dated January 28, 1983, " Single Failure Assumptions:
(SRP 15.2.1, 15.2.2, 15.2.3, 15.2.4, 15.2.5, 15.2.6; RAIs 440.64, 440.69, and 440.125; Reactor Systems Branch)," J. DeVincentis to G. W. Knighton
Subject:
Open Item Response:
(SPP 15.2.1, 15.2.2, 15.2.3, 15.2.4, 15.2.5, 15.2.6; RAI 440.64, 440.69, and 440.125; Reactor Systems Branch)
Dear Sir:
In response to the open item regarding single failure assumptions for Condition II events, we have enclosed a response which supplements that submitted in Reference (b).
The enclosed response will be incorpore ed in OL Application Amendment 49.
Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY J. DeVincentis V
Project Manager ALL/fsf cc: Atomic Safety and Licensing Board Service List 8303b10533 830224
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PDR ADOCK 05000443 PDR 1000 Elm St.. P.O. Box 330. Monchester, NH 03105 Telephone (603) 669-4000. TWX 7102207595
ASLB SERVICE LIST Philip Ahrens, Esquire Assistant Attorney General Department of the Atterney General Augusta, ME 04333 Rep res en t a t ive Beve rly ' Hol lingwo r th Coastal Chamber of Commerec 209 Winnacunnet Road Hampton, NH 03842 William S. Jordan, III, Esquire Harmon & Weiss 1725 I Street, N.W.
Suite 506 Washington, DC 20006 E. Tupper Kinder, Esquire Assistant Attorney General Office of the Attorney General 208 State House Annex Concord, NH 03301 Robert A. Backus, Esquire 116 Lowell Street P.O. Box 516 Manchester,'NH 03105 Edward J. McDermott, Esquire Sanders and McDermott Professional Association 408 Lafayette Road Hampton, NH 03842 Jo Ann Shotwell, Esquire Assistant Attorney General Environmental Protection dureau' Department of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108
4693 783/02/24.11:34 The following is additional infonnation with respect to the assumptions of the worst single failure for Condition 11 events (incidents of moderate frequency).
f.
For each transf ent, its associated worst single failure within.the protection g
system assumed in the FSAR analyses is given in Table 440.64-1. 'The I
protection system is defined as those safety functions required to mittgate the consecuences of the event. This includes not o'niy the Solid State Protection System (SSPS), but also the Engineered Sa' feguards Features (ESF) and pressurizer and steam generator safety valves.
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For each event listed in Table 440.64-1, a brief discussion of the assumed
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single failure is provided below.
The purpose of these discussions is to Justify that the single failure assumed is inde~ d the west single failure.
e These failures are failures at the system level and consider the fatture of a-protective function. The cause or mechanical nature of the raflore which causes the system' failure is not discussed, since these are addressed in the FMEA's of the SSPS and ESF and in Chapters 6, 7, and 9 of the FSAR.. There-fore, further detali beyond the systems level single fai fure of Toss of one protection train is not provided.
The steam generator safety valves may be required.to prevent a pressurization-of the secondary system. Except Wiere it f s already stated in the FSAR, the l
steam generator valves are not challenged or required to mitigate the conse-quences of the event.
Failures of these valves are not considered since they are not active fat lures.
These independent failures are not applicable.
Therefore, failure of these valves is not discussed below unless they are '
actuated as stated in the FSAR.
Finally, a loss of offsite powr is not considered as a single failure for these events.
The SRP.does not require consideration of a loss of offsite' powr for the accidents listed in Ta'o ie 440.04-1 (loss of AC powr,15.2.6, is by definitf on an exception). Furthermore, no single active failure hili cause:
'a loss of offsite powr to the. emergency buses. Therefore, consideration of this failure f s not appitcable.
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4693 ?83/02 4 4 11837 Feedwiter Temperature Reduction (15.1.1)
As stated in 15.1.1.1', this event is similar to the effect of increasing steam flow. This is 'uounded by the events in 15.1.2 and 15.1.3, as stated in 15.1.1.4.
Excessive Feedwater Flow (15.1.2)
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, As seen in Figure 15.1-1, the pressurizer pressure decreases until the time of turt> t ne trip.
The pressure spike is caused by the conservative delay between turbine trip and reactor trip; honever, the pressurizer PORV's and safety f
valvet do not open. Since they are not required to nittigate the consequences of the event, a single failure in these valves is not applicable and has no impact. Failure of an FIV to close W111 have no impact stoce the DNBR ts already increast~ng by the time the FIV closes (Tab le 15.1-1).
The engineered safeguards features are not required for thf s event. Therefore, a single failure in the E5F is.not 6pplicable and has no impact.
Therefore, the fat ture of one protection traf n a s listed in Table 440.64-1 is the limiting single active failure.
t iE_xcessive Steam Flow (15.1.3)
As stated in.15.1.3.2.b, the plant reaches a stao t lized condition. No reactor trip is required, no pressurizer relief valves are required to reduce pressure
! (F i gu re s.15.1-3, 15.1-5, 15.1-7, 15.1 -9), and no ESF actuation occurs. Since the' protection system f s not hxtuired to function for this event, a single.
faf fure does not apply and has no impact.
Inadvertent Secondary __ Dept _ essuriza_tton (16.1.4)
As stated tr 15.1;4.1, it is the fat ture (opening) of a steam dump, relief, or safety valve Wiich initiates the transient. As seen in Figure 15,1 this is a depressurization event, therefore pressure relieving functions of the pro-tection system arc not challenged nor required to mitigate the consequences of the event.
The only portion of the protection system required is the safety-
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injection portion of the ESF. A single failure in a protection train of the 2,
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4693 '83/02/24 11857 signals etch actuate S1 (15.1.4.1.a) w(Il have no. Impact due to the redun-dancy, diversity and independence of the S1 actuation signals.
The failure of one Si train (listed in Table 440.64-1) is the ilmiting single failure since It reduces SI flow, delays the injection of boron to the core, and, conse-quently allows a " closer" return to crittcality. This fs the single fatIvre i
Assumed in the FSAR as3 stated in 15.1.4.2.
For thf s event, the DNB design j
basis is met oy demonstrating no return to crf ticality (15.1.4.4).
T 1.oss of Externa l t.oad (1b.2.2) t This is bounded by the event described in 15.2.3, as stated in 15.2.2.4 and 15.2.3.1.
_Tuttin_e' Tr1p (15.2.3)
Unlike a depressurization transtent, for this analysis, the ability to main-
. tain RCS pressure below 110 percent of design per the SRP *crtterion must be explicitly addressed.
Since the DNBR increases with pressure (assuming all' other variables are held constant),.the event is analyzed with and without v
f pressure control to address both peak pressure and DNBR concerns. As stated in 15.2.3.2.a Item 7, both the'pressurtzer and steam generator safety valves may be required to operate. Assumptions relative to their operation are described under Items 4 and 5 in the FSAR.
b If the pressurizer relief / safety va1Ye5 fail to close once the pressure has been. reduced, there wt 11 be rio impact on the minimum DNBR.
T'hi s is becauss the valves are not required to close until after the tf ne of reactor' trip, at which point the DN8R I1 rising and is very'htgh (see Figares 15.2-I through 15.2-8). As stated in 15.2.3.2.a Item 4, steam relief is obtained by the steam generator safety valves.
Howver, these or any other steam relief valves wuld not be required to close until after reactor trip, den both the l
RCS pressure and DNBR are past their maximum and minimum values respectively.
Therefore, fat ture to close would have no impact.
Although the ESF may be-l required to~ function to supply emergency feedheter, a fat lure in the ESF wuld have no impact since credit for emergency feeduter is not taken (15.2.3.2.a Item 6). Therefore, the Ifmiting single fatlure is one protection train l
(Tab le 440.64-1).
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4693 'R3/02/24 11t44 ii 7
Inadvertent Closure of MS!Y (15.2.4)
This is bounded by 15.2.3 as stated in the FSAR.
Loss of Condenser Vacuum (15.2.S)
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Tat s 1s bounded by 1S.2.3 a s stated f n the FSAR.
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_ Loss of AC Power (15.2.6) for this event, the ability of the protection system to provide'long term
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cooling is verified.
The loss of one emergency feedw.ter pump of the ESF.is the 1tmiting single. fallure, as stated in Table 440.64-1. A reduction of emergency feeduter capacity reduces the capab f 11ty of the emergency feedw 7,
.to provide long tem cooling.
This results in a hf gher primry side beatup and pressure.
The pressure transient of Figure 15.2-9 shows that the pres-surizer safety valves are actuated for' tnis event.
Failure of the valves to close muld have nc impact since the emergency feeduter is adequately removing the decay heat by that time (Table 15.2-1 1 tem b).
For the case dere the single active fat ture is the failure of the pressurizer PORY or E
safety valve to close, credit can be taken for complete emergency feeduter capabi 11ty.
This muld reduce the peak pressure and cause the time at dich decay heat equa.ls heat removal capabf lity to be sooner ~.
As stated in 15.2.6.1.b and c, the steam generator safety and relief valves are'.used to i
f' dissipate decay heat during long term cooling.
Since it is desirable.to have these valves open, failure to close has no impact, aspecially since the emer-i gency feeduter supp1fes sufficient heat remoYa1 Capability.
Single failures t.hich result in loss of signals etch actuate emergency feeduter
, reactor
. trip, or valve openings have no impact due to their redundancy, diversity a independence.
Therefore, the single failure itsted in Table 440.64-1 is the limiting single faf Ture.
_ Loss o_f Noma l Feeduter (15.2.7) i As for the loss of powr event, the primary concern for the loss of normal feeduter is long term cooling capabi11ty Wich is provided by the emergency feedatter system.
Therefore, as fcr the loss of AC powr, the single active i "
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1 1.,e failure causing the foss of one emergency feedw'ter pump is the limiting single failure, as stated in 15.2.7.2
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L.oss of Flow (15.3.1 and 2)
The protection for this event is discussed in Sections 15.3.1.'4 and 15.3.2.2.
A single failure in the ESF is not applicab le since the ESF are not required -
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j to mitigate the consequences of the event. As can be seen in Figures 15.3-2 and 15.3-6, the pressurizer-PORY's may open. Howver, fallure to'close will have no impact since the point of minimum DNBR 1s past and the OffBR is rising by the time the valves close (Ff gures 15.3-4 and 15.3-8).
Therefore, the f
l wrst single failure is that of one protection train, as stated in Table i
440.64-1.
RCCA Bank Withdraw 1__ fro _m S_uberitica l (15.4.,1)
Although the pressure transient is not shown for this transient, an increase in RCS pressure is expected due to the increase in heat flux and temperature.'
Howver, f f the PORV's opened and failed to close, there wuld be'no impact on the minimum DNOR since credit for the change (increase) in pressure is not taken in the DNBR analysis. l'he ESF are not. required for this accident,.
therefore, a single fatl' re in the ESF is not applicable. Therefore, a loss u
of one protection train is the 'Ifmiting single failure.
RCCA Dank Withdrawl at Powr (15.4.2)
This event is primarf ly a DNB event and demonstrates the adequacy of the over -
temperature 6T and hi gh flux trips, as stated.in.15.4.2.4.
Typical transients for the RCCA bank withdrawl at powr event are provided in Figures 16.4-4 throu gh 15.4-9.
Operation of pressure relieving va1Yes wuId serve to reduce pressure and thus minimize the DNBR.
(If no pressure control ws available, the maximum pressure wuld be limited to that sich results in a high pres-surtzer. pressure trip. This is a less ifmiting pressure transient than those events discussed in 15.2.) Failure of valves to close wuld have no impact, since the point of minimum DNBR ts.past by the time the pressure oegins to fall (after trip) as seen in the transient figures.
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4d'vd ' OJ/Ud/ d4 11:00 As discussed in 15.4.2.2.b, for some cases. the steam generator safety valves are opened. The result.is to minimize the DNDR, as seen in' Figures 15.4-11 and 15.4-12. However, failure to close has. no impact since the point of
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minimum DNBR comes right af ter reactor trip.
Failures in the ESF are not applicable since the 'ESF are not required. Therefore, the worst single faf).
ure'is one protection train as stated in Table 440.64-1.
g ropped RCCA (15.4.3)
O The worst single failure for this event is the failure of one HIS channel.
, This results in fawer' dropped RCCA's being detected in order to initiate reactor trip via negative flux rate, but has no impact if no trip is generated (i.e., if credit for trip is not taken because of the failure.) As can be seen in Figures 15.4-13 through 15.4-15, the plant reaches a new equilibrium condi-tion, and no further protective action is required. Therefore, consideration of other single failures within the protection system is not applicable.
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Statically Misaligned RCCA (15.4.3)
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As stated in Table 440.64-1, no transient analysis is required. Furthemore, i
no protective functions are required and single failures have no impact.
' Single RCCA W1thdrawal (15.4.3)
As stated in 15.4.3.1, this is a Condition III event. Since this is not a li Condition II event (incident.of moderate frequency) it is not within the scope P
of the question and should be deleted.
Inactive RC_ pump Startup (15.4.4)
The pressure transient in Figure 15.4-19 shows that the pressuri2er p0RV's are not challenged for this event.
In any case, failure to close would have no l-impact, since the. point of minimum DNBR is past by the time the failure could i
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woro co<ecec4 11:04 occur (Figure 15.4-20).
Failures in the ESF are not applicable since the ESF is not required to mitigate the consequences of the event. Therefore, the limiting single failure is the failure of one protection train, as stated in 440.64-1.
Inadvertent,A,cu_ta_t_f on of the CCCS (15.5.1)
\\ s stated in 15.5.1.1, it is a failure in the ESF which initiates the event.
A As see in Figure 15.5-2, this is initially a depressurization event. The
~ pressure then rises to the PORY setpoint. The PORV's are capable of maintain-ing system pressure below 110 percent of design.
Failure of the:PORV's to
' close would have no impact on the DNBR, since it is already high and.ncver falls below the initial value (Figure 15.5.3). Therefore, the failure listed in Table 440.64-1 is the limiting single failure.
Increase in_ RCS Inventory (10.S.2)
As stated in the FSAR, this is bounded by 15.5.1.
Inadvertent _RCS Depressurization (-15.6.1)
As stated in 15.6.1.1, 'it is a single fatture resulting in the opening of a pressurizer PORV or safety valve which initiates the transient. Although ESF features might be actuated, they are' not required to mitigate the consequences of the event, since the DNBR rises after reactor trip. Therefore ESF failures are not applicable. Therefore, the worst single failure is failure of one protection train.
Failure of Small Lines (15.6.2)
No transient analysis ts involved for this event.
The protective system is not required to function, since operator action teminates this event as stated in 15.6.2.2.
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