ML20034F617
| ML20034F617 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 03/01/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9303040054 | |
| Download: ML20034F617 (33) | |
Text
(
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GE Nuclear Energy
.a te e acw, 1 L ;u ! ' ' hv? %L (* !..W ls % 75 March 1,1993 Docket No. STN 52-001 i
Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation 1
i
Subject:
Submittal Supporting Accelerated ABWR Review Schedule - Suppression I
Pool Hypass Paths j
Dear Chet:
Enclosed are responses to Glenn Kelly's December 22,1992 memo to Jack Duncan requesting clarification on the information contained in SSAR Subsection 19E.233.
Sincerely, 5'P Jac - Fox Advanced Reactor Programs cc: Jack Duncan (GE)w/o Enclosure Norman Fletcher (DOE) l Don Knecht (GE) w/o Enclosure 1
JI B 44 9303040054 930301 PDR ADOCK 05200001 A
-g
e-fl2VA) Y
- Tchr=ry 19, 1993 CC: JD Duncan N Fletcher (DOE)
To; Glenn Kelly hQQ4 h From:
PD Knecht
Subject:
Responses to Questions
Reference:
Clarification of Submittal on LOCAs Outside of Containment, Memo: Kelly to Duncan, December 22,1992 The following responses are provided to the referenced request for clarification on the information contained in SSAR section 19E.2.3.3, " Suppression Pool Bypass Paths". A markup of this section is attached. I believe this markup satisfies your needs.
Concern 1 f))
k 1.
Table 19E.2 (a) Is this table complete in its evaluation of all possible bypass paths?
(b) If not, do we know what has not been evaluated here?
(c) Do we know the limitations?
RESPONSE
(a)
No. Table 19E.2-21 contains only those lines which were not excluded from further consideration.
g (b)
The complete listing of possible bypass paths is provided in Table 19E.2-1 along with the bases for exclusion of certain lines.
N-(c)
There are no known limitations.
2.
(a) When estimating the conditional bypass probability, explain how EQ was taken into account.
(b) Address [how] GE assured that potentially affected equipment was qualified?
(c) If equipment was not known to be qualified, how was it handled? 5 i
l
RESPONSE
9 (a)
The conditional probability of line isolation (X ) is not significantly affected by i
Equipment Qualification (EQ) concerns since actuation of the isolation valves occurs shortly after a break occurs and no active function is required after valve l
closure. Furthermore, since redundant isolation valves are not located in the same area, a diverse environment exists. Core cooling (Qo) is not affected by M
environmental concerns since the equipment in the division affected was N
conservatively assumed not to function. The environmental effects on other X
divisions are discussed with respect to the value of Qi, Second division not affected".
O (b)
The qualification of potentially affected equipment was addressed by only relying on equipment in unaffected areas.
(c)
Equipment in an affected divisional area was not relied upon in the evaluation.
i 3.
(a) For Figures 1,2, and 3 in the December 17,1992 GE draft SSAR submittal, explain how each value of X is calculated. It is unacceptable merely to state that i
the calculation is similar to other calculations in the staff's possession, although identical calculations can be referenced.
(b) Similarly, provide the calculations for Qo.
O
RESPONSE
g i\\
(a)
Figures 1,2, and 3 of the December submittal are included as Figures 19E.2-20 a, N g b, and c in the revised section 19E.2.3.3. As indicated in section 19E.2.3.3.4, the calculation of X is based on the formulas shown in Table 19E.2-21. Because the i
line failure probabilities were treated separately in the Figure 19E.2-20 trees, the
(%
values corresponding to the number of lines and the break probabilities (P13, P14 and P15) were not included in the values indicated. The formulas used in the event trees have been added to Figure 19E.2-20.
(b)
The basis for the values of Q,is provided in section 19E.2.3.3.4. The calculation of the values was accomplished by examining the ratio of core damage frequency hc to initiating event frequency in the PRA fault trees. Values with degraded divisions were obtained by recalculating the fault trees with the most limiting division (s) disabled. Only the results of this process were provided in Section O
lQ 19E.2.3.3.4.
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4.
For medium and large breaks, GE claims that because of the depressurization caused by such break sizes, the rate of loss of inventory from the break (after some unspecified time) is compensated for by available makeup sources outside of containment, such as firewater. No basis is given for this claim.
(a) how much time does an operator have to switch over to an outside source if a break occurs outside of containment?
(b) Explain how this makeup will be provided at a dry site (perhaps one with cooling towers or a spray pond).
l (c) Provide further information/ commitments to assure that makeup will be available until the plant can be brought to a safe, stable state.
RESPONSE
(a)
Emergency Procedure Guidelines specify that sources external to the containment '
be used whenever suppression pool level is not maintained. A break external to the containment would first automatically initiate HPCF (B and C) and RCIC which are both normally aligned to an external source. For large or medium breaks only HPCF would be able to provide continuous makeup. Therefore, i
switching to an external source is not necessary in response to a LOCA outside 2
containment. Other sources of external makeup which could be aligned include firewater and feedwater or condensate which can be initiated from the control room. The choice of makeup system is at the operator discretion and depends, in i
part, on the size of the break.
(.
l If the actions indicated in Section 19E.2.2.3.4 are taken, the time available for operator action is established by the decay heat levels and the inventory in the condensate storage tank (CST) or other external supply tank. As indicated in M
Section 19E.2.1.2.2.2 (4), the CST contains sufficient inventory for at least 8 i
hours of decay heat removal. Therefore, in the absence of isolation, at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> would be available to isolate the break or establish other contingency plans.
J (b)
As indicated in (a) above, makeup at a " dry site" may be drawn from the CST for at least eight hours. Long term cooling can be assured once the break has been k
i' isolated.
(c)
Actions to provide makeup to achieve a safe stable state are provided by the f
Emergency Procedure Guidelines.
)
l s J
Concern 2 1.
GE's response to concern 2 (whether GE's analysis was exhaustive in searching for and discovering potential bypass lines) is not satisfactory. Provide a judgement on bypass potential based on up-to-date P&lDs, not those from 1988.
i N l*
RESPONSE
Xk The complete listing of potential bypass lines is included in Table 19E.2-1. This listing has been verified against the most current drawings of the ABWR
.l containment isolation system (GE Drawing 107E5042).
g Concern 3 1.
It appears that the value of Qi (failure of another division) was estimated to be IE-3 if the LOCA in the secondary containment occurred near another division wall.
l (a) Please amplify on how this was determined and what was the basis for deciding which LOCAs were or were not IE-3 events.
i (b) Also please explain how the values of Qi and Q,in Figure 2 (Medium LOCA l
Outside of Containment) were determined.
RESPONSE
The basis for the value of Qi s described in section 19E.2.3.3.4 as conservative i
(a) engineering judgement. An impact on the second division was judged to occur
'l 1
primarily due to compartment pressurization or environmental effects. Since the reactor building equipment of concern is qualified for the steam environment and J>
pressurization is largely accommodated by relief of the blowout panels, the 3
probability of consequential effect was judged to be remote. A value of IE-3 was T I
considered to be consistent with this judgement.
k4-i The values of Qi n figure 2 (now Figure 19E.2-20B) was based on engineering k
5 i
(b)
E i
judgement as discussed in the response above. The Value of Q,is indicated in a table includdwith the discussion of Q,in section 19E.2.3.3.4 for medium size -
k i
LOCAs. The development of the tree shows the loss of divisions assumed.
t i
4 2.
Please explain the assumed effect that LOCAs outside of secondary containment will have on ac or de power circuits that power divisions inside of containment.
i i
i
- i
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v y
m.
1-l 4
I
RESPONSE
9 i
s
'T As indicated on Table 19E.2-1, potential bypass paths outside of secondary q
~
containment relief into the steam tunnel portion of the turbine building. No ac or g i
a de electrical distribution centers are located in this area. The only divisional equipment are de solenoids associated with the MSIV sohnoids and ac motor k
operated containment isolation valves associated with the main steamline drains.
g l
Fuses or circuit breakers associated with these safety related components assure j
that their failure does not affect the remaining portions of the divisional electrical l
supply. Therefore no effect of this type of bypass path was considered.
j i
i l
h 4
k t
I j
.i I
i 1
i I
i 1
i i
i f
t i
I i
q
l ABWR mms Standard Plant
%s Therefore, it is necessary only to consider static loads total plant risk and therefore do not need to be on the containment.
specifically evaluated further in the PRA.
l A simple analysis was performed to determine the (1) Definition of Suppression Pool Bvpass effect of the added hydrogen mass and heat energy assocuted with 100% fuel-clad metal water reaction.
Suppression Pool Bypass is defined as the Since the design basis accident for peak containment transport of fission products through pathways which pressure is a large break LOCA, this accident was do not include the suppression pool. In such cases, chosen as the basis for the analysis.
the scrubbing action for fission product retention is lost and the potential consequences of the release in order to simplify the analysis several are higher.
conservatise assumptions were made. Since it is not possible to release the hydrogen before the first The potential for suppression pool bypass has pressure peak, only the second peak is considered.
been a subject of analysis since the early days of The hydrogen is distributed in the same manner as WASH 1400 (Reference 7). The "V" sequence which \\
the nitrogen. All of the metal water reaction heat represented a break of the low pressure line outside energy is assumed to be absorbed by the suppression of the primary containment was one of the more pool water. Finally, no credit was taken for the dominant release sequences in WASH 1400. The drywell and wetwell heat sinks.
IDCOR analysis and BMI-2104 also reviewed sequences in which the suppression pool scrubbing Consideration of 100% fuel clad metal water action was not obtained in the release pathway.
reaction results in a peak pressure of about 75 psia.
The governing service levc! C (f or steel portior.s not in order to review the importance of suppression backed by concretc)/ factored load category (for pool bypass pathways, the potential mechanisms, concrete portions including steelliner) pressure probabilities and source locations were reviewed to capability of the contaknent structure is 97 psig identify where fission products might be released which is the internal pressure required to cause the outside of the containment. The analysis has conser-i maximum stress intensity in the steel drywell head to vatively focused on the station blackout event reach general membrane yielding according to service because it leads to a higher likelihood of suppression level C limits of ASME Ill, Division 1. Subarcticle pool bypass and because it is considered one of the I
NE-3220. Therefore, the ABWR is able to withstand more probable initiating events for core damage 100% fuel clad metal water reaction as required by sequences.
The principle conclusion of the review is that, with tho exception of certain lines addressed in 19E.2JJ Suppression Pool Bypass Paths containment event trees of the PRA, suppression pool bypass pathways do not contribute significantly 19E.233.1 Introduction to risk. Consequently, the probabilistic risk assessment does not require a separate evaluation of This section reviews the potential risk of certain bypass sequences, unless the sequences develop suppression pool bypass paths and demonstrates that, during the course of an event, for example, as a with the exception of the wetwell drywell vacuum result oflow suppression pool water level. Such breakers, and certam other lines, bypass paths present cases are considered in Sect on 19D.5.7.
i no significant risk following severe accidents.
Because of this insignificance, only the vacuum Nevertheless, certain bypass lines which result breakers and the other lines require further from piping failures outside of the primary consideration in the ABWR PRA. The approach containment are included in this review in order to used in this evaluation is similar to that submitted to assess their qmncance.
l the NRC in support of the GESSAR (Reference 14) review.
(2) Mechannms for Suppression Pool Bypass The results of the evaluation is that bypass lines Alllines which originate in the reactor sessel or evaluated contribute no more than about 10% of the the primary containment are required by sections of 10CFR$0 to meet certain requirements for contain.
ment isolation. Lines which originate in the reactor 19E 2 28 Amendment
r
- ABM 22^""^5 '
Standard Plant Rev 4 vessel or the containment are required by General Design Criteria 55.and 56 to have dual barrier protection which is generally obtained by redundant isolation valv,es. Lines which are considered.
non essential in mitigating an accident are also required to automatically isolate in response to diverse isolation signals. Other lines which may be usefulin mitigating an acciden: are considered 19E.2-3.1 Amendment
AB M wtus Standard Plant R ev s exceptions to the General Design Criteria (NUREG pathways will be insignificant.
0800, Section 6.2.4} and are permitted to have remote manual isolation valves, provided that a means is The justification for this approach is as follows:
available to detect leakage or breaks in these lir.en outside of the primary containment.
Risk = Total [ Event Frequency x Consequence] (30)
A potential mechanism for suppression pool
=F xC xC (31) nbp nbp + Fbp bp bypass is the Ex-containment LOCA which results from the combined failure of a line outside of the where: F
= The total core damag-
[
UP primary containment along with the failure of its frequency of non-bypass redundant isolation valves to close. If this events combination of events occurs, the operator is made aware of the situation through leakage detection C
= The consequence of a non-nbP alarms and is instructed by plant procedures to bypass event manually isolate the lines. if possible, when the sump water levelin areas outside containment exceeds a F
He total core damage
=
bp predetermined point.
frequency by bypass events which are equivalent to a Because of these provisions the probability of nelete bypass of the suppression pool bypass occurring from the
<sion pool Ex containment LOCA" is extremely small since it requires the simultaneous failures of a piping system.
Q c.onsequence of a P
redundant and electrically separate isolation valves
_ >lete bypass event and the failure of the operator to take action.
Subsection 19E.2.3.3.4 summarizes an evaluation of If the total bypass risk is to be insignificant, the the core damage frequency from ex-containment last term in equation (31) must be much less than the 1
first, or:
The plant design criteria ensure a highly reliable F
C bP gP system for containment isolation. Nevertheless, even (32) though there is diversity in the types of vahes, all F
C types have experienced failures at operating nuclear plants and certain events, such as station blackout The total bypass and non-bypass event frequencies i
event, may make the early isolation of some lines (F) noted above are the total core damage impossible. This section evaluates the significance of frequencies for these events assuming that all esents bypass paths in order to justify that no additional have the same consequence. Since this is seldom the treatment in the PRA is neccessary.
case, the bypass frequency must be defined such that the proper consequence is applied. This is accom-
)
plished through evaluation of flow split fractiom (f) as discussed below.
he total bypass frequency can be expressed as:
(3) Methodology for Evaluation of Suppression Pool fPcbpi O
F x
Bypass F
i bp ed The evaluation of suppression pool bypass The total core damage fre-pathways is based on a methodology which evaluates where:
F
=
ed the potential relative increase in offsite consequence quency l
from bypass events over those events with 6 '. = The total conditional proba-suppression pool scrubbing. Then, knowing this P
E amount of increase,if it can be shown that the bility of full suppression pool probability of bypass is sufficiently low as to offset the bypass path i, given a core increased consequence, the added risk from these damage event.
19E M9 Amendment f
l i
1
ABWR mms Standard Plant R ev A The conditional probability of full bypass can be If equation (36) is satisfied. then the total bypass risk further refined by the expression:
is insignificant.
P x f.
(34)
(4) Criteria for Exclusion of Bypass Sequences in P
cbpi bps.
i the PRA where: f. = The fraction of fission products gener-ated during a core damage event which As noted previously,if it can be shwvn ibat tne pass tbrough line i (subsectson probability of bypass is sufficiently low as to offset 19E233.3 (1) discusses this term in the increased consequence, the risk resulting from rnore detad) release through bypass pathways will be insigni6 cant.
The flow split fraction (f) is defined as To establish a threshold for this frequency, the the ratio of the Dow rate which passes consequence ratio (right side of equation 36) was out of the bypass pathway to the total evaluated using the MAAP 3B-ABWR and CRAC flow rate of aerosols generated during codes to establish the approximate order of magni.
the core melt process. The line flow tude for evaluation purposes. To establish a split reduces the consequence associ-threshold for this frequency, the consequence ratio ated with smaller lines due to inherent
{right side of equation (36)] was evaluated using the flow restrictions I's those lines as com-MAAP 3B-ABWR and CRAC codes to establish the pared with the consequence of larger approximate order of magnitude for evaluation lines. The flow split fraction accounts purposes for this consequence reduction by re-ducing the equivalent bypass probabil-For non-bypass case, the offsite dose from normal i
iry.
containment leakage following core damage was used as a basis. 'NCL*, described in Appendix 19P, 7
and P.= The conditional probsbility of bypass in is the consequence from normal containment Pt line i (Section 19E2.333 (2) discusses leakage;" Case ~r may be used as an approximation this term in more detail).
of the full suppression pool bypass consequence.
The conditional probability of bypass is The corresponding ratio based on values in Table established through a detailed evalua-19P.2-1 is 8.4E-4 which can be used in the evaluation tion of each potential bypass pathway, of pool bypass significance. Further evaluation of establishing the failure which must
'Ex-containment LOCA* suppression pool bypass occur for a bypass path to develop and paths in the PRA is not necessary if it can be shown assigning a probability to that failure.
that the total bypass probability is significantly less than this consequence ratio.
Core damage events result in essentially two types of release: releases which bypass the suppression 19E.2J.3.2 Identification and Description of pool and those that do not. With this simplification, Suppression Pool Bypass Pathnys the total non-bypass frequency can also be defined as:
Identification of the potential suppression pool F
=Fg-F (35) bypass pathways was based on information in the bE ABWR Standard Safety Analysis Report and inserting equations (33), (34) and (35) into equation supporting piping and instrument diagrams. The (32) yields:
potential pathways are shown in matrix form in T ble 19E.2-18.
C (36) bp x f. < <
nbp/Cbp Table 19E.2-1 summarizes the results of reviewing P
i i
Assuming F is much less than F which would the ABWR design for lines which are potential d
be consistentYith the basis for' containment pathways. For each line the table provides the line isdolation.
sizes, pathways and type of isolation up to the second isolation valve. The bypass lines identified in Table 19E.2-1 were derived from a systematic review of the ABWR P& ids and other drawings.
19E -7,0 Amendment
MN auvr^s Standard Plant Rn A Several lines in Table 19E.2-1 were excluded from Wk=
the vent flow rate in a single line (SRV
~
further consideration on the basis of a variety of or drywell vent) which passes to the judgements discussed in the table notes In general, si.ppression pool the exclusion was baser! on deterministic rather than
, probabiisstic arguments. For instance, the RWCU
= the number of flow paths to the suppres-n f' return line to feedwater and LPFL Loop A were sion pool included in Table 19E.2-1 and excluded from further
' analysis because the bypass path is protected by the This can be simplified into the form:
feedwater check vahes.
f' / 1 + f' (38) l f
=
~
The remai:,ing lines are considered potential sources for significant fission product release where f'=
W./nW I
k following severe accidents. Although the probability that these lines could release a significant amount of From the formula for turbulent compressible fluid flow fission products is extremly small, they are reviewed (Reference 15) further in Subsection 19E.2.3.3.3 to assess the 1891 Yd [(dP)/KV)W (39) importance of these releases.
W
=
where W = j or k (Ib/hr)
Y Expansion factor
=
d Internal diameter (in)
=
Differential pressure (psid) 19E.233.3 Evaluation of Bypass Probability (dP)
=
Resistance coefficient = f"L/D + K' K
=
Equation (36) of Section 19E2.3.3.1 establishes f"
friction factor
=
pipe length to diameter ratio, including the need for evaluation of the flow splits and failure L/D
=
probability for each line not excluded in corrections for valves, bends Table 19E.2-1. This section provides the basis for the K'
additional factors for entrance and exit
=
evaluation of each of these factors.
effects Specific volume of fluid (cf/lb)
V
=
(1) Evaluation of Bypass Flow Split Fraction (f,)
Solving for f ',
To assess the fraction of acrosol release which bypass the suppression pool a flow split fraction is f'= 1891 Y.d."[dp/KV]II,'/1891 ny d '[dp/KVlI',~,
JJ kk needed, the flow split fraction (f) is defined as the ratio of the flow rate which passes out of a bypass
= Y.d. [dp/K]II /nY d '[dp/K] I',
(40) 33 kk pathway to the total flow rate of aerosols generated during the core melt process. Two generalized bypass Quation (40) may be rearranged to show:
paths have been evaluated: 1) a path from the RPV f * = (1/n)[Y;/Y IId /d ] x[dP;/dP l which passes to the reactor building with the k j k k
remainder passing to the suppression pool through the SRVs and 2) a path from the drywell to the
[g/ ]@
(41) reactor building with the remainder passing to the suppression pool throu6 the drywell vents.
The expressions in equation (41) were evaluated nu-h merically for the aanalline configurations to arrive at The flow split fracion may be defined as:
the flow split fractions used. The following assumptions were made in this analysis:
l f = W./W. + nW (37)
- 1. Containment pressure following the core melt is I I k
assumed to be at an average of 45 psig during the post core melt period. Although the containment where W. = the flow rate which passes through the pressure could eventually increase to a higher lescl.
bypass pathway the average is used to assess the total amount of I
release since a release would be occurring throughout this period. This pressure is typical of 19E N1 Amendment
ABM n^ e rs Standard Plant Res A those calculated in severe accident analyses (see the drywell sources, the path to the suppression pool f
Figures 19E.2-2 through 19E.2-12).
is estimated to be 5 ft. (1.5 M)
- 2. Prior to RPV melt through, the reactor pressure Other values used in the calculation are listed I sessel (RPV) is maintained at a relatisely low below-l pressure (100 psig) by the automatic depressuriza-l tion system or equivalent manual operator action.
Parameter Assumed Value Basis Four ten inch safety relief valves (ADS valves) are l
conservathcly assumed to be open to release RPV Resistance Coefficient (K= f"L/D) effluent to the suppression pool. This is consistent with the minimum instructions in the EPGs. Ten Friction Factor
.011 to.018 Ref 14 (pg A-25) 24 inch drywell vent paths are consistent with the Size dependent)
I ABWR design configuration. For conservatism the vents are assumed to be one quarter Line Diameter (D)
Various Line size (see uncovered.
Table 19E.21) r
- 3. The pressure drop in the bypass path between the Other Resistances (K)
Ref 14 (pg A-30) fission product source and the release point is a function of whether the line produces sonic or Gate valve 13 sub-sonic velocities. For RPV sources, an average Check valve 135 100 psig internal RPV pressure is assumed during Globe valve 340 the core melt process. This is based on an Entrance effects
.5 average 45 psig drywell pressure and an assumed Exit effects 1.0 SRV design which closes the SRV when a differ-ential pressure of about 50 psid exists between the Expansion Factor (Y).6 to.9 Ref 14 (pg A022) main steamline an the SRV discharge line.
(dP, K dep.)
l Depressurization of the RPV or containment Table 19E.2-19 shows sample results (f
- from equa-through the bypass path is not considered. The tion 41) for a line with two motor operated valves. In assumption is made that pressure is continuously the evaluation of individual bypass lines the actual generated during the severe accident in sufficient configuration is used. The evaluation of flow split r
quantity to uncover the SRV dacharge or drywell fractions is considered to be conservative for several i
vents.
reasons:
4 The pressure from in the non-bypass path be-(a)
Bypass release paths would normally be expected tween the fission product source and the suppres-to be mote restricted than evaluated due to sion pool release point depends on the suppres-smaller lines, more valves and pipe bends, valves sion pool level. The suppression poollevelis being partially closed or pipe breaks being assumed to be higher than normal because of the smaller than the piping diameter.
depressurization of the RPV to the Suppression pool through the SRVs. For RPV sources, the (b)
No credit is taken for additional retention of SRVs experiena about a 20 foot (6.0M) elevation fission products in the reactor building, in piping head over the SRVs during the core melt process.
or through radioactive decay.
For drywell sources a 15 foot (4.5M) elevation head is experienced over the upper horizontal (c)
For drywell sources, a higher than analyzed vent. For the station blackout sequence, the effect differential pressure should exist between the i
of ECCS system operation on suppression pool drywell and wetwell. This willlead to lower flows level has been ignored.
through the bypass path.
- 5. The length of lines discharging to the suppression pool and through the bypass paths affects the resistance coefficient Equation (39). Based on the I
ABWR arrangement drawings this length is estimated to be approximately 85 ft. (25 M). For 19E 2 32 Amendment 1
ABWR Standard Plant 22^(ims (2) Evaluation of Failure Probabilities (P )
The failure probabilities used for the detailed calculation of the bypass probabilities are summarized in Table 19E.2-20. The bases for these probabilities are provided below:
(a) Current operating plant MSIV failure to close probability is about 4E-3/ demand with a common mode failure probability (P;) of about IE-4/de-mand. For this evaluation the common mode l
failure probability of IE-4 is assumed for failure of both valves in a single line to close.
(b) Current operating plants evaluate MSIV leakage against a leakage requirement of 11.5 SCFH per valve. About 50% of the valves typically fail this local leak rate test at this level and about 10% are believed to typically exceed the 640 SCFH level allowed by ABWR proposed technical specifica-tions. The leakage probability (P2) used in this analysis was based on three leakage groups:
Probability 612ng leakage Per Valve Per Line G1
<113 sch 3
J G2 11.5 to 640 scfh 4
.2 G3
> 640 scfh
.1
.01 19E.2 321 Ameadment
j WS umms Standard Plant ms The MSIV leakage probability (P2)is assigned a among check valves was considered for lines value of.71 totorrespond to the total line leak-containing redundant series check val es. Only I
age probabuity. Flow split fractions were deter-Feedwater and the SLC paths contain more than mined for each of the groups and a weighted one check valve. For these lines a Beta factor of average flow split fraction (weighted by the line
.18 was used for the failure of the second sahe.
i leakage probabilities) was determined for use in the evaluation.
(h) When power is available, some normally closed valves open during an event in response to an f
J (c) The probability of flow passing to the main con-injection signal. even though the actual injection l
denser is judged to be governed by the failure of fails (a requirement for a core damage to the bypass valve to close. This probability (P3) occur).
l is taken at 4E-3 from Reference 16. Once flow passes to the main condenser, the condenser is The probability that ECCS valves are not closed assumed to fail (P4) via the relatively low by an operator (P10) is considered remote positive pressure rupture disks.
during a severe accident. A value of 0.5 is judged reasonable especialy considering the (d) The main steamline break probability (PS) was potential for room environment degradation.
l line break probability (P15).
For station blackout events, since the valves do l
not open, these lines do not contribute to (c) Normally open pneumatic (P6) and DC motor potential bypass risk.
l operated valves (P7) have failed to close.
Causes include improper setting of torque (i) Some normally closed valves may be open at the switches leading to valve stem failure, beginning of the event. The failure probability l
undetected valve operator failures and improper (Pil) for these valves assumes they are open 4 packing materials or lubricants. GE has issued hours during a 7000 hour0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br /> operating cycle and several service information letters on valve that the operator fails to recognize the open problems and recommended actions to prevent path and close the vale. A 0.5 probability is recurrence of the failures. The industry failure judged reasonabic for the operators failure to rates for motor operated valves is about act during the core damage event.
l i
3.6E-3/ demand and 4.1E 3 for air operated
)
valves. These failure rates are not significantly (j) Some valves may be opened by the operator
]
affected by the valve environments. A common during the course of the event. Such action may
[
cause failure among air operated valves was be in compliance with written procedures or it i
considered for lines conteimg redundant series may occur due to confusion in following a proce-I valves. For these lines a Beta factor of.18 was dure. The probability that valves are inadvert-used for the failure of the second valve.
ently opened (P12) is considered a violation of plasned procedures. A value of IE-3 is judged 1
l (f) AC solenoid and motor operated valves are reasonable during a core damage event, i
subject to a common mode failure (P8)if motive power is unavailable such as during a Station (k) Pipe rupture is enremely rare in stainless steel i
Blackout event. For station blackout events piping. However, carbon steel piping has been i
i these valves will have a conditional failure observed to fail under certain conditions. The probability of 1.0. For this analysis a failure frequency of these failures has been widely j
j probability of 1.0 was conservatively assumed.
studied and shown to be in the range of IE-7 events / year. The probabilities ofline rupture as l
a function of line size (P13, P14, P15) are taken i
(g) Check valves haw been observed to failin such from Reference 14. Four line segments outside I i
a way as to permit full reverse flow, a condition necessary to permit suppression pool bypass for of the containment are assumed for each bypass j
some lines. Maintenance errors associated with line. The intermediate line size (3 to 6 inches) l l
testable check valves have also been observed.
probability is assumed to be twice that of the i
4 The industry failure rates for check valves large line size (greater than 6 inches).
l allowing complete reverse flow (P9), based on 7000 hours0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br /> of operation per operating cycle,is For pipe failures in an individual bypass line, it l
l about S.4E-3 per cycle. A common cause failure was presumed that an undetected break in an i
19E 133 l
Amendment I
l
ABWR na m as Standard Plant nes 4 unpressur <.J line could occur at any time.
both slows the break flow and terminates any long Therefore, the conditional probability of a term release from the break. Therefore if the EPG bypass path was then taken to be the same as actions are taken, no additional consequence of the the failure rate during a one year period (which event occur.
was estimated to be 7000 hours0.081 days <br />1.944 hours <br />0.0116 weeks <br />0.00266 months <br />). This approach of estimating pipe failure probability is judged to The system arrangement routes the RWCU lines be conservative.
above the core to avoid a potential siphon of the core I
inventory. In the event of an unisolated RWCU line The f ailure probabilities used in the evaluation break, lowering the RPV level to below the
! should be considered conditional probabilities, gisen shutdown cooling suction and depressurizing the
{ a core melt. In general the abose probabilities are RPV would be sufficient to terminate the break flow
' not affected by the core melt process itself and can without causing core damage. This action should be therefore be considered independent of the event possible prior to any impact on other ECCS process.
equipment. These actions are included in Section 19D.7 Whether the bypass path is the initiator or occurs simultaneously with the event is inconseq-(3) Evaluation of Bypass Probability uential in the evaluation based on the following discussion. The approach taken in the bypass study Table 19E.2 21 summarizes the results of these i
is to consider the presence of a bypass path as an evaluations. For each potential bypass pathway, it independent event from the events which caused the shows the flow split fraction based on the line size core damage in a specific sequence. This approach is and valve configuration, the equation to calculate the acceptable because for large breaks the associated bypass probability, the results of the probability systems are not in general relied upon to prevent calculations using the data from Table 19E.2 2'.) and l core damage and no consequence of these failures the bypass fraction for the line. The tab!c also have been identified which would affect the systems includes reference to the sketch (Figure 19E.2-19) preventing core damage. Therefore whether the which illustrates the potential pathways. The evalua-break is an initiator or consequential does not affect tion is based on the conservative assumption of a the final evaluation. Similarly, none of the systems station blackout event since it is believed to be the d
associated with the smaller bypass lines are dominant core damage sequence and gives the high-associated with preventing core damage. Therefore est bypass fractions.
they too are not associated with the cause of the core melt.
The ACRS has expressed concern regarding the failure of the RWCU suction in combination with failure of the isolation valves to close. The concern is that there may be a flooding situation that could have a high consequence if it leads to an eventual loss of suppression pool and CST inventory or floodmg of other ECCS rooms. Such an event would not be consistent with this presumed independence of the assumed mah3 probabilities.
i If a break in the RWCU suction line were the postulated LOCA, the containment isolation valves would be expected to close, terminating the esent.
4 NRC concerns over Motor Opcisted Valve (MOV) i closure capability are being addressed as an industry activity. In this evaluation it was assumed that the valves fail to close due to a Station Blackout event.
Furthermore, should the isolation valves fail to close, the system arrangement assures that the core is not uncovered and EPGs require depressurization which SE333 i Amendment
ABWR o
mums Standard Plant f >o
%s (4) Evaluation of Results pipe break frequency is provided in Appendix 19E.2333 (2)(k).
Section 19E233.1 (4) pros des a conservative justification that bypass paths ith a total bypass X Line Isolation - The conditional probability of 3
I fraction less than 8.4E-4 do o substantially merease automatic isolation valses failing to close gisen the offsite risk. As is shown i Table 19E.2-21. the the ex-containment LOCA. Values used and
. bypass probability is about 5 for all potential tbe manner in which probabilities were paths not addressed in the Containment esent trees.
combined are shown op.Tabfe 19E.2%og.c,
This total ts well mihin the goal.
+t )vM
/
P Oper. Action - The conditional probability that 3
Potential bypass through the Wetwell-Drywell operator fails to act to manually isolate the Vacuum Breakers and the inerting lines are included ex-containment LOCA. Such a failure to act m the containment event trees. (Section 19D.5).
could be due to a lack of instrumentation availability or mechanical failure. For most Based on the above discussion, it can be bypass paths considered, the very conservatise concluded that suppression pool bypass paths and assumption was made that no operator action is Ex-Conta;nment LOCAs not addressed by the taken. For ECCS discharge lines and warmup Containment event trees do not contribute a lines the operator is assumed to act to close an significant offsite risk and do not ueed further open valve,if needed. The basis for the value evaluation in the PRA.
chosen (P10 in Section 19E.233 (2))is based on general operator awareness of the potential for 19E.233.4 Evaluation of Ex. containment LOCA these paths to be unisolated. Although the leak Core Damage Frequency detection system is adequate to alert the operator of a break in the system, (1) Introduction instrumentation failure is not considered to provide a strong contribution to the failure To provide a separate assessment of the probability.
importance of bypass paths, a more comprehensive analysis of the frequency of core damage from 0 Second Division not Affected - For most lines it 3
LOCAs outside containment was conducted using is conservatively assumed that the LOCA affects event tree and fault tree techniques.
the division in which the break occurs. This factor represents the conditional probability that Conservative and simplified event trees of the LOCA also affects the required makeup for LOCA outside containment events were developed core cooling from a second electrical division. It and included as Figures 19E.2 20a through is asrumed that such failure results from 19E.2-20c. These trees show that the total core environmental effects from flooding or damage frequency due to LOCAs outside of pressurization effects.
containment is about 13E-8 per year. The end-point for these trees is core damage with or without bypass A systematic evaluation of potential cold of the containment.
flooding due to ex-containment breaks was summarized in Appendix 19R, Probabilistic (2) Assumptions F!ooding Analysis. Flooding in the reactor building is noted to disable the system affected The following definitions and considerations were and potentially flood the Reactor Building applied in development of the trees.
corridor, but not disable other makeup equipment due to the water-tight doors V1 Line Break Outside - The frequency of piping contained in the design. The analysis of an breaks in small, medium er large breaks outside unisolated RWCU break in subsection 19R.4.5 of containment and which communicate directly shows that no cooling systems will be damaged.
with the reactor vessel. The lines are grouped by type of isolation. The be, sis for each event Compartment pressurization and environmental initiation frequency is the line size and the total effects of high pressure LOCAs in secondary number of lines considered. The basis for the containeent were considered in the development of Figures 19E.2 20a through c.
19EM Amendme nt
AB%R wrrs Standard Plant rm3 Equipment in the ABWR design is arranged For LOCAs which occur in the reactor budding, the event is assumed to fail the division in wh with consideration of divisional separation. A high energy Gne break in a division would cause the break occurs. For other LOCAs, such as the blowout panels fro:n the division to reliese LOCAs in the turbine building, no disisional the initial pressure spike to the steam tunnel.
impact is assumed.
Subsequent pressurization of the room could esentually cause a release of the energy into the Consideration of insentory depiction due to the l next adjacent division in a clockwise progression LOCA outside containment is addressed by EPGs f through the reactor building.
which specify that coolant makeup sources using '
inventory sources outside of cou;pr. ment be used As doors from the cenidor and penetrations are as the preferred source. in the ABWR design forced open. the environment of the adjacent small breaks can be accommodated by any of the divisions could be affected by the presence of high pressure coolant makeup systems (RCIC. '
steam. However, the qualification of the HPCF B and HPCF C) which are in separate equipment to 212 degrees F and 100% humidity divisions and which draw water from the i
makes the probability of further system condensate storage. Since condensate is unasailability unlikely. Where a LOCA could effectively an unlimited supply and makeup j occur in an area adjacent to a separate division, capability exists, no additional concern is -
a value of IE-3 was assumed for 0, based on necessary for the small break LOCAs outside of -
1 conservative engineering judgement, to containment.
represent the remote possibility for failure of these adjacent systems.
Medium and large breaks outside of containment can be accommodated by any of the three For line breaks in the turbine building the effect divisions in the short term following a break of the break would not impact the divisional without concern for inventory loss in the RPV.
power distribution and, for these sequences, the All penetrations, except the RPV/RWCU bottom O value was judged to be negligible.
head drain (a unique situation addressed l g
separately in Section 19.9.1 by an event specific ;
)
Although line routing are not specified, the procedure), are above the top of active fuel so that !
analysis assumes that breaks inside reactor core uncovery due to inventory depletion is not a building equipme at rooms affect the division in concern. In the longer term, the break will which the breaks occur; LOCAs outside of the depressurize the RPV which effectisely reduces secondary containment are not assumed to fail a the loss of inventory from the break to a level well division of equipment.
within the makeup capacity of other available systems which makeup from sources outside of O Coolant Makeup - This factor represents the containment, such as firewater. Due to the conditional probability of core cooling failure by reduction in loss rate through the break, all sources of cooling with consideration to those significant time is available for operators to affected by Ibe ex-containment LOCA. The compensate for the usage of water and flooding in values used are derived from an evaluation of the affected area. Furthermore, operators are the PRA fault trees and are summarized below; assumed to follow plant procedures in isolating the break or lowering RPV level to a level below COOLANT MAIGUP FAILURE (0,)
the affected penetration,if necessary. Adequate BREAK SIZE instrumentation and long term makeup from Smm!1 Medium Larce firewater and condensate sources would normally Div. not Affected 2.2E-7 6.2E-7 6.1E-7 be available.
1 Div. affected LIE-6 S.6E-6 8.5E-6 2 Div. affected 3.6E-4 3.7E-3 3.7E-3 (3) Conclusion l
l The conditional probability when one or more For each of the event trees shown in Figures '
electrical divisions are affected were derived by 19E.2-10a through c the total non-bypass and bypass disabling the most limiting didsion in the LOCA core damage frequencies are shown and are event trees and then calculating the resulting summarized below:
conditional probability.
ME D 1 Amendment n
+
gQ
'U
< f Stang-d Plant
[
L' 2'^*[^s Core Damage Frey:n (even /yr)
Non-B To-Small LOCAs 12 1.1 9 1'
Intermedaic LOCAs 23 10 1 E 10 to Large LOCAs
-10 E-13 E-10 TOTAL llE-a
. JE. 9 13E4 Ex-containment LOCA events without bypass represent a small fraction of the total core damage frequency (1.6E-7) are therefore justified as not being further evaluated in the PRA.
nithough the consequence from bypass events is greater than for non. bypass events, the total
- frequency of bypass events concurrent with core damage is extremely small. The core damage frequency of ex. containment LOCAs with bypass is less than 1% of the total evaluated core damage frequency. Large LOCAs can be excluded from further consiueration on the basis oflow probability. Exclusion of M Jium and Small bypass sequences is based on the additional consideration of the reductions in consequences of the ex containment LOCAs due to the flow splits provided by restrictions due to line sizing. This is discussed in Section 19E.2333.
In addition, since significant margin exists between the current PRA results and the safety goals, it can be concluded that the bypass events do not significantly contribute to the offsite exposure risk.
19E.2333 Suppression Pool Bypass Resulting from External Events 4
The effect of external events on the Suppression Pool Bypass evaluation is dacuued in Apoendix 191 to determine if a significant potentia! for bypassmg the suppression pool results from component failures induced by a seismic event. Only seismic events were considered to provide a significant challenge to the creation of bypass paths beyond that already considered i:t the PRA.
19El M 2 Amendment a
ABMR unno s Standard Plant
,,, s Table 19E.2.18 Potential Bypass Pathway Matrix FROM Wetwell Suppression IQ,,
RPV Drwell Airicace Pool Drywell No NA NA NA Werwell Airspace Yes Yes NA NA Reactor Building Yes Yes Yes Yes Turbine Building Yes Yes Yes Yes The above matrix shows the paths that potentially bypass the suppression pool
- Pathways which originate in the drywell and potentially release into the werwell are potential bypass paths if the containment is vented or the wetwell fails during the severe accident.
U Amendment
h J
Standard Plant
,,, s Table 19E.2-19 now Split Fractions Lme Sae now Sph: Fraction a
1:1 PJV Sourre Devwelt %ume o
0 25 13E-05 5 4E 05 12 0.5 94E45 3.4 E-04 1
5E-04 2.0E 03 50 2
33E43 1.2E42 100 4
1.EE-02 52E42 150 6
4EE42 1.5E 01 200 8
8.9E 02 2.5E 01 250 10 1.4 E-01 3 6E-01 300 12 2.0E 01 4 6E 01 150 14 2.6E 01 5.4E 01 400 16 3.2E 01 6.2E 01 450 18 3.8E-01 6."T41 500 20 43E-01 72E-01 700 28 6.1E 01 8 4E 01 1000 40 7.7E 01 9.2E 01 i
19E245 Amesiment F
e..
l ABWR
- usms Standard Plant a,
s Table 19E.2-20 Failure Probabilities I
Svmbel Descrirtion Prob / Event M sc. - 3 L n. t.A ~ <J-P1 5:::. p!'c gcra:;d vahe (MSIV) 1.0EJ "/
a e ~ - _ - - <, r <-. c, <
\\
P2 MSIV leakage probability 7.1E-1 b
P3 Turbine Bypass Isolation 4.0E-3 c
P4 Main condenser failure 10 c
PS MSL break outside containment 8.0E-6 d
P6 Air operated valve (NO) 4.1E-3 e
P7 DC Motor operated valve (NO) 3.6E-3 e
P8 AC Motor Operated valve (NO-SBO) 1.0 f
P9 Check Valve 8.4E-3 g
P10 Motor operated valves (NC) 5.0E-1 h
P11 Motor operated valves (NC) 2.SE-4 i
P12 Inadvertent opening 1.0E-3 j
P13 Smallline break 2.4E-4 k
P14 Medium line break 1.6E-5 k
P15 Large line breah 8.0E4 k
1 Amendment 19E.2-te
l AB%R ume Standard Plant Table 19E.2-21 Summary of Bypass Probabilities i
Lines from the RPV Bypass Flow Split Probability Bypass B pass Figure 3
Pathwav Fraction Ecuation Probability Fraction loE 2-W Main Steam 6.7E-1 4'Pl*(P3*P4 + PS) 1.6E-5 1.1E 5 A
Main Steam Leakage 2.2E5 4* P2*(P3'P4 + PS) 1.1E-2 2.5E-7 A
Feedwater 5.2E-1 2*P9'P15 2,4E-8 1.3E-8 B
Reactor Inst. Lines 3.1E-5 30*P J'P9 6.0E-5 1.9E-9 D
HPCF Discharge 1.1E 1 2*P9'P10*P14 13E-7 1.5E-8 C
HPCF Warmup 1.0E-3 2*P10*P11*P13 6.7E-8 6.7E-11 C
SLC Injection 3.0E-3 1*P9*P13 3.6E-7 1.1E-9 B
l RCIC Stcam 5upply 6.9E-2 1*P8'P14 1.6E-5 1.1E-6 E
LPFL Discharge 1.7E-1 2*P9'P10*P15 6.7E-8 1.1E-8 C
LPFL Warmup Line 1.0E-3 2*P10'P11*P13 6.7E-8 6.7E-11 C
RWCU Suction 1.2E-1 1*P8'P14 1.6E-5 2
E RWCU Inst Lines 3.1E 5 4*P13'P9 8.1E-6 2.5E 1)
D l
Post Ace Sampling 1^E-3 4*P8'P13 9.6E-4 9.9E 7 J
LDS Instruments 3.1E-5 9'P13*P9 1.8E-5 5.7E-10 D
SRV Discharge 6.9E-2 8'P14 13E-4 8.8E-6 K
Total 2.4E-5 I
These lines may be excluded for station blackout events i
J 19E24' Amendment l
ABWR a w as Standard Plant am 4 Table 19E.2-21 Summary of Bypass Probabilities (Continued)
Lines from the Drmell Bypass Flow Split Probability Bypass Bypass Figure B1hny Fraction Eauation Probability Fraction 10E.2-10 Cont Atmos Monitor 8.9E-4 6*P9'P13 1.2E-5 1.1E-9 D
LDS Samples 1.7E-3 2*PS*P13 4.8E-4 8.2E-7 E
Drywell Sump Drain 3.0E-2 2*P8'P13 4.8E-4 1.4E-5 J
DW Purge 4.6E-1 1*P6*P11 1.1E-6 53E-7 I
2*P12f[b ACS Crosstic 1.1E-1 1.5E-6 1.6E-7 H
- WW-DW Vac Bkr 2.6E-1 8'P9 6.7E-2 1.7E-2 G
Total excluding vacuum breaker 1.6E-5 and pwg L n E.
s Grand Total excluding vacuum breaker 3.0E-5
\\
=d prge M~s Goal 8.4E4 kl L tu HPkalw [*A % ?" % NY
% pewoo3 vedud' Du/ Ad
- Lerh Ac5 Cross la (wha o % uwkgIm g wilh x
s%su m c5Tr. w, '"s'+h,iw ent>sb< I,4w t b, pss ' ~ e-s),
NMV) '
- Addressed on Con nment Event Trees.
- Id Amen.iment 24 19E.2/A
eO M
- m 6i u s Standard Plant A. MAIN STEAM NO LEAKAGE (P2)
NO STREAMLINE TURBINE BYPASS OR FAILURE TO ISOLATE (P1)
BREAK (PS)
ISOLATON (P3)
OK OK F
BYPASS 5
BYPASS RPV X
X f
X g.
Figure 19E.219A SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS B. FEEDWATER OR SLC 4
NO CHECKVALVE NO LINEBREAK FAlLURE (P9)
(P13. P15)
OK OK BYPASS I
TURBINE BULDING (FW) l y
RPV N
N REACTOR BUILDING (SLC) l Figure 19E.219B SUPPRESSION POOL BYPASS PATHS AND CONTIGL%ATIONS 19E 210?
Armendment
ABWR uxux s Standard Plant
,, s t
4 C. ECCS UNES NO CHECKVALVE t
FArLURE (P9)
OPERATOR NO OR CLOSES VALVE UNEBREAK BYP ASS (P11)
(P10)
(P13.P14.P15)
CK OK f
BYPASS BYPASS M
i i
1 RPV N
-X
[
- REACTOR BUILDING I
I Figure 19E.219C SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS 1:
D. INSTRUMENT UNES NO LINEBREAK NO CHECK FAILURE (P13. P14. P15)
(P9)
OK i
OK k
BYPASS l
1 RPV l
(ev:
[
]
I Figure 19E.219D SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS 19E 2-108 r
Amendmen:
ABWR Standard Plant
,'^$'j}s E. STATION BLACKOUT AFFECTED LINES ISOLATION NO LINEBREAK Pi B)
(P13. P14. P15)
CK OK 1.0 i
BYPASS
-D<
X RPV
-x x
1 Figure 19E.2-19E SUPPRFSSION POOL BYPASS PATHS AND CONFIGURATIONS F. CONTAINMENT ATMOSPHERIC MONITOR NO CHECKVALVE OPERATOR NO FAILURE CLOSES VALVE LINEBREAK (P9)
(P10)
(P13)
OK OK OK 1.0 BYPASS ORYWELL N
X
[
]
Figure 19E.2-19F SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS 19E 2-109 Amendment
ABWR n^etx^s Standard Plant nv A e
G DRYWELL-WETWELL VAC. SKRS CHECKVALVE FAILURE (P9)
OK BYPASS I
I N
WETWELL DRYWELL l
CONTAINMENT VENT g
ASSUMED Figure 19E.2-19G SUPPRESSION POOL BYPASS PATHS AND CONTIGLTATIONS H. ATMOSPHERK; CONTROL SYSTEM CROSSTIE NO AO VALVE 6C%MM IMON NO INADVERTENT OPENING (P12)
FAILURE (y)
OK m) t OK 1
BYPASS I
i DRYWELL X
X WETWELL I
Figure 19E.219H SUPPRESSION POOL BYPASS PATHS AND CONTIGtJRATIONS 19E2-i!0 Amendment
i Standard Plant 2*j"l b'
- l. ORYWELL PURGE NO AIR VALVE VALVE OPEN FAILURE (PS)
(P11)
OK 1
OK BYPASS i
DRYWELL X
X STACK i
I e
Figure 19E.2-19I SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS J. SAMPLE LINES OR SUMPS NO CHECK VALVE FAILURE NO STATON NO LINE (SUMPS ONLY)
BLACKOUT (P8)
BREAK (P13)
(P9)
OK OK BYPASS BYPASS RPV X
X
] SfATON I
I l
REACTOR DRYWELL X
X
'/,
BUILDING g
g SUMP Figure 19E.2-19J SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS 19E.2111 Amendment
ABWR 2 '^ ' '"^ 5 Standard Plant l
K. SRV OISCHARGE NO SRV OL ADS /SRV BREAK OPEN (P14)
OK OK 1.0 BYPASS I
11 a
j p i
//
I RPV COiTAINW.NT VENT ww ASSUMED Figure 19E.219K SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS 19C.; 3:
Amendment
YN$
nA6rnas Standard Plant Rev A LlHE BREAK
~LINE OPER.
SECOND DIV.
COOLANT OUTSIDE 1 SOLATION ACTON NOT AFFECTED MAKEUP (V )
(X )
(P )
(0 )
(O )
1 3
1 1
o REACTOR INSTRUMENT LINE (30)
RWCU INSTRUMENT LINE (4)
LDS tNSTRUMENTS (9)
OK 1.1 E 6 1.1 E-8 NON-BYPASS 1 CE 2 n.= -l 0
NON-BYPASS e.
8.4E-3
[N) 1.1 E-6 9.2E 11 BYPASS
{
HPFL WARMUP (2) l LPFL WARMUP (2)
OK l
9.6E-4 1.1 E-6 1.1E 09 BYPASS L>f
1.4E-4 uf'?*f"3 OK
.5 1.1 E-6 7.4E 14 BYPASS POST ACCIDENT SAMPLING (4)
O PON-BYPASS 9.6E-4 o
[,,.', C 1.0 OK
,N 1 1.0
/,
h 1.1E 9 BYPASS L
SLC INJECTON OK 1.1 E-6 2.6E 10 NON-BYPASS 2.4E-4 O
Ci.<s3]
OK
[. o r. f1J 1.0
)
4.0E 13 BYPASS TOTALS
^
NON-BYPASS 1.2E-8 BYPASS 1.1E 9 Figure 19E.2 20A SMALL LOCAS OUTSIDE CONTAINMENT 19E.2-H3 Auguai I
i
MN 22^ * ^s Standard Plant R e, A LINE BREAK LNE OPER.
SECOND DIV.
COOLANT OUTSIDE '
(SCUTON ACTON NOT AFFECTED MAKEUP (V )
(X )
(P )
(0 )
(Oo) 3 3
1 1
HPCF CISCH AAGE (2)
OK 8.6E-6 2.3E-10 NON BYPASS 3 2E 5 0
t
(. F n,7 l
og 4 2E-3 8.6E-6 cr,,f.oj 5.8E-13 BYPASS
.5 k bD OK i
1E-3 3.7E 3 2.4E-13 BYPASS RCIC STE.)M SUPPLY (1)
RWCU SU(. TON (1) 0 OK 3.2E-5 0
OK p.,1 OK
'~
~
1.0 i
8.6E 6 f
2.8E 11 BYPASS 1.0 L-.
OK 1E4 3.7E-3 1.2E 11 BYPASS SRV OfSCHAAGE (8) 0 NON-BYPASS 1.3E-4 0
NON-BYPASS L/*PN7 1.0 OK 1.0 e
6.2E 7 8.1E 11 BYPASS L
TOTALS NON-BYPASS 2.3E 10 BYPASS 1.2E 10 Figure 19E.2 20B MEDIUM LOCAS OUTSIDE CONTAINMENT 19El 114 Amendmen:
1 1
f 4
ABWR nA6 *^s Standard Plant Rev A 1
LINE BREAK ~ LINE OPER.
SECOND DIV.
COOLANT OUTSIDE ISOLATION ACTION NOT AFFECTED MAKEUP (V )
(X )
(Pj)
(0 )
(0 ;
1 3
3 3
MAIN STEAMLINE (4)
OK 6.1 E 7 2.0E 11 NON-BYPASS 3.2E5 O
,,,a; OK 1E 3
/ f b
l t re; 6.1E-7 2.oE-14 BYPASS 1.o W
N4 Negl BYPASS j
FEEDWATER (2)
(INCLUDING RWCU RETURN AND LPFL A DISCHARGE)
M l'
6.1 E-7 I
c.o SleE-il NON-BYPASS 64E-5 OK j,. f a:
M
/7 1.5E4 Lb f f%
6.1 E-7 j
gy
- f. '7' T SSE 14 BYPASS 1.0 OK 4
u (L 7
Negl BYPASS SW DtSCHAAGE (f)
(LOOPS B AND C)
OK 8.5E4 1.4E 10 NON-BYPASS 1.6E-5 OK e..r;r]
OK 4.2E4 f M ' %l 2.8E 13 BYPASS
.5
)
1E4 3.7E4 1.2E-13 BYPASS TOTALS l< [
NON-BYPASS
-10 BYPASS 13 19E 2. I15 Amendment