ML20127K909
| ML20127K909 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 01/18/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9301260268 | |
| Download: ML20127K909 (32) | |
Text
_ _ _ _ _ _
9 GE Nuclear Energy
++
- -J ';
January 18,1993 Docket No. STN 52-001 Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation
Subject:
Submittal Supporting Accele.ated ABWR Review Schedule
Dear Chet:
Attached are the responses to the memo pertaining to the clarification on LOCAs outside of containment (G. Kelly to J. Duncar., December 22,1992).
Please provide Mr. Palla with a copy of these responses.
Sincerely, Jack Fox Advanced Reactor Programs cc: Jack Duncan (GE)
Norman Fletcher (DOE)
JPM6 g7[d 9301260268 930118 PDR ADOCK 05200001 A
- [) l
January 10,1993 CC: JD Duncan N Fletcher (DOE)
To; Glenn Kelly From:
PD Knecht
Subject:
Responses to Questions
Reference:
Clarification of_ Submittal on LOCAs Outside of Containmenj, 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".
Concern 1 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
Table 19E.2-21 contains only those lines which were not excluded from further concideration. The complete listing of lines considered is provided in Table 19E.2-1 along with the bases for exclusion of certain lines.
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?
RESPONSE
(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 closure, Furthermore, since redundant isolation valves are not located in the same area, a diverse enviroment exists. Core cooling (Q) is not affected by enviromental concerns since the equipment in the division affected was conservatively assumed not to function. The environmental effects on other divisions are discussed with respect to the value of Q,, "Second division not affected".
(b)
The qualification of potentially affected equipment was addressed by only relying on equipment in unaffected areas.
l.-
l (c)
Equipment in an affected divisional area was not relied upon in the evaluation.
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.
RESPONSE
(a)
Figures 1,2, and 3 of the December submittal are included as Figures 19E.2-20 a, 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.
(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 to initiating event frequency in the PRA fault tiees. Values with degraded divisions were obtained by recalulating the fault trees with the most limiting division (s) disabled. Only the results of this process were provided in Section 19E.2.3.3.4.
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 firew ater. 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). (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. Sources of external makeup include not only firewater, but also HPCF, feedwater/ condensate and the RHR Service water Crosstie which can be initiated from the control room. The choice of makeup system is at the operator discretica and depend, in part, on the size of the break. Operator action to achieve a stable long-term response would be the time before the external supplies such as condensate storage are exhausted or until the break can be isolated. This would be expected to be on the order of _
several hours.
(b)
Makeup at a " dry site" is assured once the break has been isolated. Any further consideration of this consideration should be provided by an applicant.
(c)
Actions to provide makeup to achieve a safe stable state are provided by the Emergency Procedure Guidelines.
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.
RESPONSE
The complete listing of potential bypass lines is included in Table 19E.2-1. This listing har been verified against the most current drawings of the ABWR containment isolation system (GE Drawing 107E5042).
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. (a) Please amplify on how this was determined and what was the basis for decidir.g which LOCAs were or were not IE-3 events. (b) Also please explain how the values of Q and Q,in Figure 2 (Medium LOCA Outside of Containment) were determined.
RESPONSE
The basis for the value of Qi s described in section 19E.2.3.3.4 as conservative (a) i engineering judgement. An impact on the second division was judged to occur primarily due to compartment pressurization or environmental effects. Since the reactor building equipment of concern is qualified for the steam environment and pressurization is largely accomodated by relief of the blowout panels, the probability of consequential effect was judged to be remote. A value of IE-3 was considered to be_ consistent with this judgement.
The values of Qi n figure 2 (now Figure 19E.2-20B) was based on engineering (b) i judgement as discussed in the response above. The Value of Q,is indicated in a table include with the discussion of Q,in section 19E.2.3.3.4 for medium size LOCAs. The development of the tree shows the loss of divisions assumed. !
l 1.
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.
RESPONSE
As indicated on Table 19E.2-1, potential bypass paths outside of secondary containment relief into the steam tunnel portion of the turbine building. No ac or de electrical distribution centers are located in this area. The only divisional equipment are de solenoids associated with the MSIV solenoids and ac motor operated containment isolation valves associated with the main steamline drains.
Fuses or circuit breakers associated with these safety related components assure that their failure does not affect the remaining portions of the divisional electrical supply. Therefore no effect of this type of bypass path was considered.
l 1 l
l f
[
/t ABWR bh m amas Standard Plant m4
,)
Therefore, it is necessarv only to consider static lo total plant risk and therefore do not need to be on the containment.
specifically evaluated further in the PRA.
A sirnple analysis was performed to determine the (1) Definition of Suppression Pool Bypass effect of the added hydrogen mass and heat energy associated 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.
q conservative assumptions were made. Since it is not The potential for suppr/ession pool bypass has possible to release the hydrogen before the first pressure peak, only the second peak is considered.
been a subject of analy
- ince the early days of The hydrogen is distributed in the same manner as WASH 1400 (Referenc 6).
he "V" sequence which the nitrogen. All of the metal water reaction heat represented a break of t e 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 p:~1 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 psig.
The governing service level C (for steel portions not In order to review the importance of suppression backed by concrete)/ factored load category (for pool bypass pathways, the potential mechanisms, concrete portions including steelliner) pressure probabilities and source locations were redewed to capability of the containment 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-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 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 renew is that, with the exception of certain lines addressed in 19E.2.33 Suppression Pool Bypass Paths containment event trees of the PRA, suppression pool bypass pathways do not contribute significantly 19E.2.3.3.1 Introduction to risk. Consequently, the probabilistic risk assessment does not require a separate evaluation of This section reviews the potential risk of certain bypass sequeaces, 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 of low suppression pool water level. Such breakers, and certain other lines, bypass paths present cases are considered in Section 19D.5.7.
no significant risk following severe accidents.
Because of this insignificance, only the vacuum Nevertheless, certain bypars 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 te iew in order to used in this evaluation is similar to that submitted assess their significance, the NRC in support of the GESSAR (Referencer a review.
(2) Mechanisms for Suppression Pool Bypass The results of the evaluation is that bypass lines Alllines which originate in the reactor vessel or evaluated contribute no more than about 10% of the the primary containment are required by sections of 10CFR50 to meet certain requirements for contain.
ment isolation. Lines which originate in the reactor 19E.2 28 l
Amendment 24 l
ABWR uuims l
Standard Plant Rev A 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 lines outside of the primary containment.
Risk = Total [ Event Frequency x Consequence] (30) xC (3I) nbp + Fbp bp A potential mechanism for suppression pool
=F xC nbP bypass is the "Ex containment LOCA* which results nbp = The totalcore damage from the combined failure of a line outside of the where: F 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
- "'* 9 " * "'" I " " " '
aware of the situation through leakage detection C
"bP 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
= The total core damage bP predetermined point.
frequency by bypass events which are equivalent to a Because of these provisions the probability of complete bypass of the suppression pool bypass occurring from the suppression pool "Ex containment LOCA* is c.itremely small since it requires the simultaneous failures of a piping system, C
= The consequence of a bP redundant and electrically separate isolation valves complete 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 y
last term in equation %ji must be much less than the the core damage frequency from ex containment first, or:
The plant design criteria ensure a highly reliable F
C bp abp system for containment isolation. Nevertheless, even (32)l though there is diversity in the types of valves, all F
ubP bP types have experienced failures at operating nuclear plants and certain events, such as station blackout The total bypass and non bypass event frequencies 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 events 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 fractions (f) as discussed below.
The total bypass frequency can be expressed as:
(3) Methodology for Evaluation of Suppression Pool I33) l fP bpl Bypass F
=F x
bp ed The evaluation of suppression pool bypass The total core damage fre-pathways is based on a methodology which evaluates where:
F
=
d the potential relative increase in offsite consequence quency from bypass events over those events with bi= The total conditional proba-suppression pool scrubbing. Then, knowing this P
P 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, givea a core increased consequence, the added risk from these damage event.
19E.2-29 Amendment L-_--
i 23A6100A5 Standard Plant The conditional probability of full bypass can be if equation (36) is satisfied, then the total bypass risk l further refined by the expression:
is insignificant, i
)
bp. x f.
(34)
(4) Criteria for Exclusion of Bypass Sequences in P
i P
=
cbp.s i
the PRA j
where:f. = The fraction of fission products gener-ated during a core damage event which As noted previously,ifit can be shown that the pass through line i (subseetion probability of bypass is sufficiently low as to offset 4
19E2.33.3 (1) discusses this term in the increased consequence, the risk resulting from j
more detail) release through bypass pathways will be insignificant.
l T1.e flow split fraction (f) is defined as To establish a threshold for this frequency, the the ratio of the flow rate which passes consequence ratio (right side of equation 36) was l
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-j the core melt process. The line flow tudc for evaluation purposes. To establish a split reduces the consequence associ.
threshold for this fr uency, the consequence ratio i
ated with smalle4 lines due to inherent (right side of equatio
) was evaluated using the 4s flow restrictions in those lines as com.
MAAP 3B ABWR and CRAC codes to establish the i
pared with the consequence oflarger 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 j
ity.
containment leakage following core damage was j
used as a basis. "NCL*, described in Appendix 19P, c c nditional probany obypass b is th consequence from normal containment and P
b i= line i (Section 19E2.333 (2) discusses leakage;" Case 7' may be used as an approximation P
i this term in more detail).
of the full suppression pool bypass consequence.
I 1
The conditional probability of bypass is The corresponding ratio based on values in Table l
established through a detailed evalua-19P.21 is 8.4E-4 which can be used in the evaluation j
tion of each potential bypass pathway, of pool bypass significance. Further evaluation of establishing tl.e failure which must
'Ex containment LOCA* suppression pool bypass l
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 l
of release: releases which bypass the suppression 19E.233.2 Identification and Description of pool and those that do cot. With this simplification, Suppression Pool Bypass Pathways i
the total non bypass frequency can also be defined as:
Identification of the potential suppression pool l
F
=F (35) bypass pathways was based on information in the
]
nbp cd Fbp
)(
ABWR Standard Safety Analysis Report and Inserting equations (33), (34) and (M) into equation supporting piping and instrument diagrams. The (32) yields:
potential pathways are shown in matrix form in Table 19E.218.
(36) abp/Cbp Table 19E.2-1 summarizes the results of reviewing bp. x f. < <
C P
i s
the ABWR desige for lines which are potential 4**.
,, f p,
.,a t,A pathways. For each line the table provides the line j
i
- f sizes, pathways and type of isolation up to the second Q
- 44. ~ ~u k - a st- '
isolation valve. The bypass lines identified in Table 4
4.,. r*
m.
W,e~ M. --
- 19E.2-1 were derived from a systematic review of the 5
i ev' M.
19 W Amendment h
4
-- - - - =.
m
- wims Standard Plant n~ 4 k=
the vent flow rate in a single line (SRV Several lines iniable.19E.21 were excluded from W
further consideration on the basis of a variety of or drywell vent) which passes to the judgements discussed in the table notes. In general, suppression pool the exclusion was based on deterministic rather than
= the number of flow paths to the suppres.
probabilistic arguments. For instance, the RWCU n
return line to feedwater and LPFL Loop A were sion pool included in Table 19E.21 and excluded from further analysis because the bypass path is protected by the This car Se sirnplified into the form:
g ua" I f = i' / 1 + f' (38) l The remaining lines are considered potential sources for significant fission product release where f'=
W./nW I
following severe accidents. Although the probability that these lines could release a ignificant amount of From the fosquis for turbulent compressible fluid now fission products is ext (Jhfly.
11, they are resiewed (Referen 7)
I gf further in Subsection 19E.2.3.3.3 to assess the 1891 Yd* [(dP)/KV]V2 (39) l importance of these releases.
W
=
where W = j or k (Ib/hr)
Expansion factor Y
=
Internal diameter (in) d
=
Differential pressure (psid)
(dP)
=
Resistance coefficient = f"L/D + K' 19E2.3.3.3 Evaluation of Hypass Probability K
=
friction factor f"
=
pipe length to diameter ratio, including Equation (36) of Section 19E2.3.3.1 establishes L/D
=
the need for evaluation of the flow splits and failure corrections for valves, bends additional factors for entrance and exit probability for each line not excluded in K'
=
Table 19E.2-1. This section provides the basis for the effects evaluation of each of these factors.
V = Specific volume of fluid (cf/lb)
(1) Evaluation of Bypass Flow Split Fraction (f,)
Sching for f ',
/1891 nyk k [dp/KV]b To assess the fraction of aerosol release which I'= 1891 Y.d.2[dp/KV]
d bypass the suppression pool a flow split fraction is
,J3 needed, the flow split fraction (f) is defined as the
= Y.d.'[dp/K]1/2/nY dk k [dp/K]1/'
(40) 1I ratio of the flow rate which passes out of a bypass pathway to the total flow rate of aerosols generated Equation ()OYay be rearranged to show:
during the core melt process. Two generalized bypass paths have been evaluated: 1) a path from the RPV f' = (1/n)[Y./Y ][d./d ] x [dP./dP l",
J k 3 k I
k which passes to the reactor building with the U
remainder passicg to tb suppression pool through
[K /Kl
(#I) k j the SRVs and 2) a path from the drywell to the reactor building with the remainder passing to the The expressions in equation (41) were evaluated nu. l suppression pool through the drywell vents.
merically for the actual line configurations to arrive at the flow split fractions used. The following assumptions The flow split fraction may be defined as:
were made in this analysis:
- 1. Containment pressure following the core melt is (37) assumed to be at an average of 45 psig during the W./W) + nW f
=
I k
post core melt period. Although the containment pressure could eventually increase to a higher level, where W. = the flow rate which passes through the the average is used to assess the total amount of bypass pathway release since a release would be occurring J
throughout this period. This pressure is typical of those calculated in severe accident analyses (see i 19E.2-3 t Amendment
^
MA6t00AS Standard Plant
%4
~
l Figures 19E.2 2 through 19E.212) y pool 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 vessel (RPV) is maintained at a relatively low below:
pressure (100 psig) by the automatic depressuriza-tion system or equivalent manual operator action.
Parameter Assumed Value Basis Four ten inch safety relief vahes (ADS valves) are conservatively assumed to be open to release RPV Resistance Coefficient (K = f"L/D) p{
effluent to the suppression pool. This is consistent
~1 7 3 with the minimum instructions in the EPGs. Ten Friction Factor
.01110.018 R d A-25) 24 inch drywell vent paths are consistent with the Siz pendent)
ABWR design configuration. For conservatism the vents are assumed to be one quarter Line Diameter (D)
Various Line size (see uncovered.
Table 19E.21)
- 3. The pressure drop in the bypass path between the Other Resistances (K)
Reff( g A-30) fission product source and the release point is a (7
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 F.n Entrante effects
.5 average 45 peig drywell pressure and an assumed Exit effects '
1.0 d
SRV design which closes the SRV when a differ-ential pressure of about 50 psid exis'.s between the Expansion Factor (Y).6 to.9 R
A022) main steamline an the SRV discharge line.
(
dep.)
Depressurization of the RPV or containment Table 19E.219 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 quantity to uncover the SRV discharge or drywell fractions is coesidered to be conservative for several vents.
reasons:
The pressure from in the non bypass path be-(a)
Bypass release paths would normally be expected 4
tween the fission product source and the suppres-to be more restricted than evaluated due to sion pool release point depends on the suppres-smaller lines, more valves and pipe bends, valves sion poollevel. 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 experience about a 20 foot (6.0M) elevation fission products in the reactor building, in piping head over the SRVs during the core snelt process.
or through radioactive decay.
For drywell sources a 15 foot (4.5M) elevation i
head is experienced over the upper horizontal (c)
For drywell sources, a higher than ana zed vent. For the station ble.kout sequence, the effect differential pressure should exist between the of ECCS system operation on suppression pool drywell and wetwell This willlead to lower flows le I has been ignored.
through the bypass path.
_/
- 5. The length oflines discharging t the suppression Evaluation of Failure Probabilities (P )
l pool and through the bypass aths affects the resistance coefficient Equation
. Based on the ABWR arrangement drawings this length is estimated to be approximately 85 ft. (25 M). For the drywell sources, the path to the suppression r/ edW Amendmta 19E2 32
M 23A6tooAs Slandard Plant n, 4 s
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) T M )dUnt p o W A MSIVs is judgeo to oc somcwud ingner than' for comparable MSIVs in currently operating
>lants becaue of its reliance on operation of aN team pilot solenoid rather than an air pilot
,olenoid. Steam pilot valves have not troyed very
- eliable in operating plants since t!'e relatively
.aigh temperature tends to le.d to bindin or
@m; la the solenoid valves.
urrent operating plant MSIV failure to close pro ability (.9C is about 4E 3/ demand with a common mode failure probability of about 1E-4/ demand For this evaluation a ingb#tommon mode failure probability of IE-$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 Group Leakane Per Valve Per Line G1
<11.5 scfh
.5.
.5 G2 11.5 to 90 scfh
,4
.2 G3
>640 scfh
.1
.01 -
f; y
0 2
t9E.2-32.1 Amendment 24
ABWR 23sumas Standard Plant n, 3 The MSIV leakage. probability (P2) is assigne,a among check valves was considered for lines value of.71 to correspond to the total line leak-containing redundant series check valves. Only age probability. Flow split fractions were deter-Feedwater and the SLC paths cantain 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 valve.
leakage probabilities) was determined for use in the evaluation.
(b) When power is available, some normally closed Ib valves open during an event in response to an (c) The probability of flow passing to the in con-injection signal, even though the actual injection denser is judged to be governed by t failure of fails (a requirement for a core damage to the bypass valve to close. This ability (P3) occur).
is taken at 4E 3 from Referen e
,Once flow passes to the main condenser, he' 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 potential for room environment degradation.
l (d) The main steamline break probability (PS) was line break probability (P15).
For station blackout events, since the valves do not open, these lines do not contribute to (c) Normally open pneumatic (P6) and DC motor potential bypass risk.
operated valves (P7) have failed to close.
Causes include improper setting of torque (i) Some normally closed velves may be open at the switches leading to valve stem failure, beginning of the event. The failure probability undetected valve operator failures and improper (Pil) fo-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 reasonable for the operators failure to rates for motor operated valves is about act during the core damage event.
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 considered for lines containimg redundant series may occur due to confusion in following a proce-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 planned procedures. A value of IE-3 is judged (f) AC solenoid and motor operated valves are reasonable during a core damage event.
subject to a common mode failure (P8) if motive power is unavailable such as during a Station (k) Pipe rupture is extremely rare in stainless steel Blackout event. For station blackout events piping. However, carbon steel piping has been these valves viill 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 probability of 1.0 was conservatively assumed, studied and shown to be in the range of IE 7 events / year. The prc$ abilities of line rupture as (g) Check valves have been observed to fail in such a function of linciac (P13, P14, P15) are taken from Reference [t-afe assumed for eacis bypass Tpur line segments outside a way as to permit full reverse flow, a condition of the containmen necessary to permit suppression pool bypass for some lines. Maintenance errors associated with line. The intermediate line size (3 to 6 inches) testable check valves have also been observeci.
probability is assumed to be twice that of the The industry failure rates for check valves large line size (greater than 6 inches).
allowing complete reverse Dow (P9), based on 7000 bours of operation per opera:ing cycle,is For pipe failures in an individual bypass line, it l
about 8.4E-3 per cycle. A common cause failure was presumed that an undetected break in an Amendment 24 19E.2-33 l
l
ABWR 23aeiooxs Standard Plant no 4 unpressurized 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 oypass 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 esent 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 inventory. In the event of an unisolated RWCU line The failure probabilities used in the evaluation break, lowering the RPV level to below the should be considered conditional probabilities, given shutdown cooling suction and depressurizing the a core melt. In general the above proDabilities 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 initiatcr or occurs simultaneously with the event is inconseq.
(3) Evaluation of Bypass Probability uentialin the evaluation based on the following discussion. The approach taken in the bypass study Table 19E.2 21 summarizes the results of these 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 d.smage 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.2-20 and I core damage and no consequence of these failures the bypass fraction for the line. The table 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 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 jrI '
is that there may be a flooding situation that could have a high consequence if it leads to an eventual j
/D(
loss of suppression pool and CST inventory or flooding of other ECCS rooms. Such an event would not be consistent with this presumed independence of the assumed conditional probabilities.
if a break in the RWCU suction line were the postulated LOCA, the containment isolation valves would be expected to close, terminating the event.
NRC concerns over Motor Operated Valve (MOV) 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 fall to close, the system arrangement assures that the core is not uncovered and EPGs require depressurization which Amendment 24 19E.241 i t
ABWR 234amas Standard Plant
%4 (4) Evaluation of Results l '
pipe break frequency is provided in Appendix 19E.2333 (2)(k).
Section 19E2.33.1 (4) pro; ides a conservative justification that bypass paths lwith a total bypass X Line Isolation - The conditional probability of g
I fraction less than 8.4E-4 do not : ubstantially increase automatic isolation valves failing to close given the offsite risk. As is shown ' Table 19E.2-21, the the ex-containment LOCA. Values used and b
bypass probability is about S for all potential the manner in which probabilities were paths not addressed in the Containment event trees.
combined are shown on Table 19E.2 21.
This total is well within the goal.
P Oper. Action - The conditional probability that 1
Potential bypass through the Wetwell Drywell operator fails to act to manually isolate the 4
Vacuum Breakers di TV are included ex-containment LOCA. Such a failure to act in 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 conservative concluded that suppression pool bypass paths and assumption was made that no operator action is Ex Containment 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 need 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.2.33.4 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 O Second Division not Affected For most lines it g
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 assumed 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 Flooding 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 Vi Line Break Outside - The frequency of piping contained in the design. The analysis of an g breaks in small, medium or 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 basis for each event Compartment pressurization and environmental initiation frequency is the jine size and the total effects of high pressure LOCAs in secondary number of lines considered. The basis for the containmeat were considered in the development of Figures 19E.2 20a through c.
Amendment 24 19E.2 M
$\\$
m6imas Standard Plant Rev A Equipment in the. ABWR design is arranged For LOCAs which occur in the reactor building, with consideration of divisional separation. A the event is assumed to fail the division in which high energy line break in a division would cause the break occurs. For other LOCAs, such as the blowout panels from the division to relieve LOCAs in the turbine building, no divisional the initial pressure spike to the steam tunnel, impact is assumed.
Subsequent pressurization of the room could eventually cause a release of the energy into the Consideration of inventory depletion due to'the next adjacent division in a clockwise progression LOCA outside containment is addressed by EPGs l through the reactor building.
which specify that coolant makeup sources using inventory sources outside of containment be used 4
As doors from the corridor and penetrations are as the preferred source. In the ABWR dasign 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 makes the probability of further system condensate storage. Since condensate is unavailability unlikely. Where a LOCA could effectively an unlimited supply and makeup 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 0 value was judged to be negligible.
head drain (a unique situation addressed 1
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 l
analysis assumes that breaks inside reactor core uncovery due to inventory depletion is not a building equipment 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 effectively 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 Q Coolant Makeup - This farwr icpresents 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 the ex-containment LOCA. The compensate for the usage of water and flooding in i
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 MAKEUP FAILURE (0,)
the affected penetration,if necessary. Adequate BREAK SIZE instrumentation and long term makeup from Small Medium Laree firewater and condensate sources would normally Div. not Affected 2.2E-7 6.2E-7 6.1E-7 be available.
1 Div. affected 1.1E-6 8.6E-6 8.5E-6 2 Div. affected 3.6E-4 3.7E-3 3.7E 3 (3) Conclusion 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 division in the LOCA core damage frequencies are shown and are event trees and then calculating the resulting summarized below:
conditional probability.
19E.2 M i Amendment 24
ABWR 23Asio0As
. Standard Plant ne 4 L
s Core Damage Frequency (events /yr)
Non-Bvoass Bvoass Total m
}.
Small LOCAs L268 1.159 E8 7 '
~
Intermediate LOCAs 2.3510 1.2510
1.258 1.259 1.3E-8
^
j 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 j
PRA.
i Although 'the consequence from bypass events is j
greater than for non-bypass events, the total
}
frequency of bypass events concurrent with core damage is extremely small.' The core damage i
frequency of ex-containment LOCAs with j
bypass is less than 1% of the total evaluated core damage frequency. Large LOCAs can be j
excluded from further consideration on the basis i
oflow probability. Exclusion of Medium and j
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 i
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 i
do not significantly contribute to the offsite l
exposure risk.
4 j
19E.233.5 Suppression Pool Bypass Resulting from External Events
[4 The effect of external events on the Suppression Pool Bypass evaluation is discussed in Appendix 191 to determine if a significant potential for bypassing j
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.
i considered in the PRA.
j t
i 19E.244 2 Amendment 24 -
I m.
r.y
_, ~.....
...m,-,
......,,,,,,,-_,_,..m._,
ABWR umms Standard Plant ne. 4 TABLE 19E.2-1 i
POTENTIAL SUPPRESSION POOL BYPASS LINES i
)
PATHWAY BASIS FOR l
NUMBER SIZE (mm)
ISOLATION EXCLUSION 8
DESCRIPTION OF LINES FROM IQ (1 in. - 25.4 mm)
VALVES (SEE NOTES)
/ Main Steam Line Drain 1
e Feedwater 2
/ Reactor Inst. Lines 30 RPV RB 6
CK 4
i s CRD insert / Withdraw 103 RPV RB
<1 CK, MA 1
4
/HPCF Discharge 2
/ HPCF Warmup 2
/ HPCF Suction 2
/ Supp Poolinstrumentation
Mene cM 2
CK, CK
/ RCIC Steam Supply 1
/ RCIC Discharge 1
4
/ RCIC Min. Flow 1
/RCIC Suction 1
/ RCICTurbine Exhaust i
4 l
/ RCIC Turb. Exh Vac Bkr i
3
/ RCIC Vac Pump Discharge 1
i
< RHR LPFL Discharge 2
/RHR Warmup Lines 2
/hHR Wetwell Spray 2
WW RB 100 MO 2,4 j
/RHR Drywell Spray 2
4
/RHR SDC Suction 3
- bupp, ioot ducuon 3
.bP RB ou MO-ru EidR Supp. Pool Return 3
d i
d 1
i Amendment 24 19E2-84 1
J 4
-4
+..w-~,--n.--
-rw 1-- - v.ww w -
A.BWR m.ms Standard Plant w4 TABLE 19E.2-1 (Continued) l POTENTIAL SUPPRESSION POOL BYPASS LINES PATHWAY BASIS FOR NUMBER SIZE (mm)
. ISOLATION EXCLUSION (SEE NOTES) l j
DESCRIPTION OF LINES FROM IQ (1 in. = 25 4 mm)
VALVES l
/ RWCU Suction 1
/ RWCU Return 1
l s RWCU Head Spray Line 1
CK
- Post Accident Sampling 4
RPV RB 25 (MO, MO) i y RIP Motor Purge 10 RPV RB
<1 CK, CK.
1
/ RIP Cooling Water 4
, LDS Instruments 9
CK
{
< SPCU Suction 1
'SPCU Return 1
' Cont. Atmosphere Monitor 6 DW RB Zr t.o MA
-F LDS Samples
-2 DW RP 30 (SO, SO) j Drywell Sump Drains 2
/HVCW/RBCW Supply 4
DW.
RB 100 CK,MO W
)
/ HVCW/DWCW Return
./ DW Exhaust /SGTS 2
- 7-l
/Wetwell Vent to SGTS 1
-DW leesagg' Purge
-F
'WW Inerting/ Purge 2
< Instrument Air 2
1 l
/ SRV Pneumatic Supply 3
Flamability Control 1
)
- ADS /SRV Discharge 8
RPV WW 300 RV l
- ACS Crosstie 2
- &~
l
, WW/DW Vacuum Breaker 8 DW WW 500 CK I
/ Miscellaneous Leakage 1
i
, Access Tunnels 2
-NONE 6
r e
6 4
4 1
l i
Amendment 24 19E.2-85 i
,-..,,.y..
n~,
ABWR-msms il Standard Plant
%.4 l
TABLE 19E 2-1 (Continued)
POTENTIAL SUPPRESSION POOL IWPASS LINES LEGEND AND ACRONYSIS PATHWAY significantly reduced due to decay and other l
Source (From)
Termination (To) removal mechanisms.
RPV Reactor Pressure Vessel W W Wetwe1I i
Some lines which originate in the primary i
SP Suppression Pool
. WW Wetwell containment are designed for operating pressures ST Steam Tunnel higher than would be :xpected in the containment i
isolation Valve Types during a severe accident. These lines (with design l
pressures greater than about 100 psig) were excluded since the probability of a break under i
AO Air Operated less than normal operating pressures and
{
h10 hiotor Operated coincident with the severe accident is extremely RV Relief Valve small.
l CK Check Valve h1A hianually Actuated 5.
Some lines return to the feedwater line. These SO Solenoid Operated pathways (such as LPCF loop A and RWCU) are i
()
Common hiode Failure Potential (See excluded since they are bounded by the evaluation Section 19E2.333 (2))
- of feedwater.
Bases for Exclusion 6.
Acceptable long term leakage from the drywell to the reactor building following a design basis 1.
Closed systems such as closed cooling water accident is specified at.4%of drywell volume per systems which do not directly connect to the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. During severe accident conditions this i
RPV or containment atmosphere require two leakage could be somewhat greater due to higher j
failures to become a bypass pathway: a leak than design basis containment pressure. However, i
or break within the cooled component and a the contribution of this leakage to overall risk is
]
line break outside of containment. Very low ignored because this leakage is through numerous flow is expected out of the break or leak at tortuous passages of small diameter which provide the cooled component is likely due to the ample opportunity for plateout and plugging high degree of restriction. These pathways effects (see subsection 19E2.13.4). A discussion of l
are not considered further on the basis of this the drywell access tunnels is included in section l
very low flow rate. Similarly, extreme 19F.
restrictions in CRD seals provides the basis for excluding those lines.
Drywell S Drains - The y ell floor ~and j
equipmeAt dr 'n sumps 3 e as med to e 2.
Pathways which originate in the primary normall/open an isolated /by mot oper ed containment wt,twe11 airspace or the valves. The dischar e is ass ' ed to pas to a frain j
suppression pool are excluded because fission heade[in the reacto buil mg. Backflo i b the i
product aerosols would first be trapped in the eac r building is a med to be prevented by 1.
suppression pool and would thus not be check valves.
1 available for release through the bypass path.
. H ' C Coolin Water, Reac;pt4 ilding Co i
3.
Some lines are closed during normal plant at. Con mm nt Atm pheric ontr - Line 4
operation and would not be expected to be.
izes ow m table 9E/' I are a. sum for these opened in the short term following a plant system accident. These lines are excluded on the basis ot low frequency of use. Furthermore, 7.
Drywell purge lines are normally closed and fail j
5 should a bypass pathway develop later when % closed. The potential for inadvertent opening is the line is used, the fission product source considered remote and is addressed by term would be expected to have oeen already Emergency Procedure guidelines.
1 j
Amendment 24 19E246 i
ABWR zutuu Standard Plant ne, a Table 19E.219 Flow Split Fractions Line Size now Spht Fraction m
m RPV bune Dnwen burce 6
0.25 1.5E-05 5.4E-05 12 0$
04E45 14Ea 25 1
5.7E-44 2.0E 03 50 2
3.3E-03 1.2E42 100 4
1.8E42 6.2E-02 150 6
4.8E 02 13E-01 200 8
8.9 E-02 2.5E-01 250 10 1.4 E-01 3.6E-01 300 12 2.0E-01 4.6E41 250 14 2.t,E-01 5.4 E-01 400 16 3.2E 01 6.2E41 450 13 3.8E41 6.7E41 500 20 4.3E-01 7.2E-01 700 28 6.1E-01 8.4E41 1000 40 7.7E41 9.2E-01 Amendment 24 19E.2-65
l ABWR mums Standard Plant nev 4 Table 19E.2-20 Failure Probabilities Svmbol Descrintion Prob / Event D.as mse v closv<<.
P1 Sicespitucugaiwd ssla W31V) 1.0E-T a
P2 MSIV leakage probability 7.1E-1 b
P3 Turbine Bypass isolation 4.0E 3 c
P4 Main condenser failure 1.0 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
PS 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) 2.8E-4 i
P12 Inadvertent opening 1.0E-3 j
P13 Smallline break 2.4E-4 k
P14 Medium lina break 1.6E-5 k
P15 Large line break 8.0E-6 k
Amendment 24 19E.2.(4
ABWR 1
msimas.
Standard Plant j
ne, 4 i
Table 19E.2 21 i
Summary of Bypass Probabilities Lines from the RPV 4
l Bypass i
Flow Split Probability Bypass Bypass Figure Pathway Fraction Eauation Probability -
Fraction 10E.2 19 1
l Main Steam 6.7E-1 4*Pl*(P3*P4 + PS) 1.6Eb-1.1Eb A
i j
Main Steam Leakage 2.2E5 4'P2*(P3'P4 + PS) 1.1E-2 2.5E 7 A
f Feedwater 5.2E-1 2*P9'P15 2.4E-8 1.3E-8 B
i f
Reactor Inst. Lines 3.1E-5 30*P13*P9 6.0E 5 1.9E-9 D
j HPCF Discharge 1.1E 1 2'P9'P10*P14 1.3E-7 1.5E-8 C'
i HPCF Warmup 1.0E-3 2*P10'P11*P13 6.7E-8 6.7E-11 C
i l
SLC Injection 3.0E-3 1*P9'P13 3.6E 7 1.1E 9 B
l j
RCIC Steam Supply 6.9E-2 1*P8'P14 1.6E5
- 1. tE-6
- E 2
j 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 1.2E1 1*P8'P14
' 1.6E 5 2.0E-6 E
1 RWCU Inst Lines 3.1E 5 4'P13*P9 8.1E-6 2 5E 10 D
l l
Post Acc Sampling 1.0E-3 4*P8'P13 9.6E 4 9.9E-7 J
+
j LDS Instruments 3.1E-5 9'P13*P9 L8E 5 5.7E 10.
D i
SRV Discharge
- 6.9E-2 8*P14 1.3E-4 8.8E-6 K
4 I,4 -
1 Total M E-5 4
i l
j-These lines may be excluded for station blackout events 1
i
{
Amendmeni 24 19E.2-67 4
y--
y-w wige-
-..y-.-._
.r-
,yw,we.u-3,-ne.,cmqrkm
- W e't
.Ngm
-e ' m em deP+w*P wr+.*'cM
'M a=
cf9---'FM+'m
ABWR uuims Standard Plant no a Table 19E.2-21 Summary of Bypass Probabilities (Continued)
Llnes from the Drwell Bypass Flow Split ProbaMlity Bypass Bypass Figure Eathway Fraction Ecuation.
Probability Fraction 10E 219 I. I H Cont Atmos Monitor 8.9E-4 6*P9'P13 1.2E-5 ME9 D
r, s. ar-7 LDS Samples 1.7E-3 2*P8*P13 4.8E-4 0$e3 E
M A
Drywell Sump Drain 3.0E-2 2*P8;P13 MMM '^ ##
N.-. /* f#~IJ W ineemaqa/ Purge E-1 F6dF5i ACS Crosstie 1.1E-1 2'P12 1.5E-6 1.6E-7 H
4 WW-DW Vac Bkr 2.6E-1 8'P9 6.7E-2 1.7E-2 G
4 S
N Total excluding vacuum breaker tdr3 /, u-t aN ase1>8~F<o(-f Grand Total excluding vacuum breaker 325 andd" ~ ..;f=
Goal 8.4E-4
- Addressed on Containment Event Trees.
Amendment 24 19E.24;8
ABWR MAMMAS l,
Standard Plant nev A -
r 5
i j
A. MAIN STEAM NO LEAKAGE (P2)
NO STREAMLINE TURBINE BYPASS OR FAILURE TO ISOLATE (P1)
BREAK (PS)
ISOLATION (P3) l OK 4
l OK-l
}
BYPASS i
i BYPASS lr RPV X
X l
X 9,
l j
Figure 19E.219A SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS i
NO CHECKVALVE NO LINEBREAK i
FAILURE (P9)
(P13, P15)
}-
OK i
OK i
j i
i j
i BYPASS I
l 1
j RPV N'
N
-TURBINE BUILDING (FW) l REACTOR BUILDING (SLC) l j
Figure 19E.2-19B. SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS j
Amendment 19E2-107 4
r
~--s,
-,n,,
Comi emm cer.ny PROPRIETARY [NTORMATION MM CAu m Standard Plant 23AoiooAs no.
C. ECCS UNES NO CHECKV ALVE 7 AILUME (P9' O*ERATOe No ey,A$s tein C'03 E,5, f,v E
,,L,IN E,s,a E,A,m,,
3.
og 5
OK i
i 1
ex 8YPASS i
I I
yA,
', X mE Acrom suiLoiNo f
arv i
as-e ts.7s 1
i a
Figum 19E.2-19C SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS i
D. INSTRUMENT LINES NO CHECK 1
NO LINE sAEAK F AILum E j
IP13.P14 Pts)
(pg3 OK f
i OK sveAss i
I i
/~ -
il arv i
//
i 88 413 77 Figum 19E.2-190 SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS i
AJnsadment 8 19E.2 108
.r-,
r- -,
l o acm: seene cepany PROPRJETARY thTORMATION MM can m 33A6100As Standard Plant wA E. STATION BLACKOUT AFFECTEO LINES NO j
ISOLATION LINE SME AK i
(PSI IP13 P14 Pl$)
OK 1
d i
OK 10 j
g BYPASS I
I 4
g 5 mPv
-cx: lx 88413 78 1
4 Figure 19E.2-19E - SUPPRES$10N POOL BYPASS PATHS AND CONFIGURATIONS 1
I
(
F. CONTAINMENT ATMOSPHERIC MONITOR '
l NO CMECEVALVE NO F AILumE MANUAL V ALVE LINE BREAK iP9' CLOsuME (P13) 4 i
OK I
i l
OK d
j 4
I' OK l
1.o avrass J
f CAYWELL N
ff b
{
es4:3 79 j
Figure 19E.2-19F SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS 195 2 109 w tg i
4
, ~, -,
w-
-.m
,-,-nv..e c
w em-><
w.
ABWR u461oots Standard Plant
. nev a -
G. DRYWELL-WETWELL VAC. BKRS CHECKVALVE FAILURE (P9)
OK BYPASS I
I DRYWELL N
WETWELL g
CONTAINMENT VENT l
ASSUMED Figure 19E.2-19G SUPPRESSION POOL BYPASS PATHS AND CONFIGURA1!ONS H. ATMOSPHERIC CONTROL SYSTEM CROSSTIE NOINADVERTENT NO AO VALVE OPENING (P12)
. FAILURE (P9)
OK i
OK BYPASS-t DRYWELL X
X WETWELL y
i Figure 19E.2-19H SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS 19E.2110 Amendment
J_1 g
Caural Eseme capacy PROPRIETARY IhTORMATION ABWR am 23,'
^5 Standard Plant
- 1. DRYWELL INE ATiNG Ne C NO AIR VALVE rdt N/
E F AILU A E ON-IP6 6 '.
g fi,)
$g oK CK ORYW(LL b
t*ui
.me s<
N A
e., o.
LL so413 s2 Figure 19E.2-191 SUPPRESSION POOL BYPAas PATHS AND CONFIGURATIONS J. SAMPLE LINES OR SUMPS F At[uR E NC ST ATION NO BLACKOUT LINE BREAK (suups oNLyn iP81 (P131 gpgi OK OK f
8YPAS5 8YPAS$
l ST ATION I
{AEACTOA-g r viLDING DRYWELL g
p 80413 83 i
i l
Figure 19E.2-19J SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS -
l i,t 3.iii L
Generslhenc Compsey PROPRIETARY 1hTORMAT10N.
g4g Standard Plant 33^5ix^s 1
Rev A K. SRV DISCHARGE 4
AD$/SAv g
OPEN
- tria, CK OK 1.0 BYPA$$.
I J
//
W s
u RPV CONTAINMENT VENT
'm-m ASSUM E D 804 t 3-84 Figure 19E.2-19K SUPPRESSION POOL BYPASS PATHS AND CONFIGURATIONS u
N' I l'9 E.2.l l 2 i
nr.,,,,
...v,.
.s
6
-ABWR uuiwas Standard Plant Ru A j.
LINE BREAK LINE OPER.
SECOND DN.
COOLANT OUTSIDE ISOLATION ACTION NOT AFFECTED MAKEUP l
(V )
(X )
(PI)
(0 )
(O )
3 3
1 o
t REACTOR INSTRUMENT LINE (30)
RWCU INSTRUMENT LINE (4)
LDS INSTRUMENTS (9)
OK l'
1.1E 6
~
1.1 E 8 NON BYPASS 1.0E 2 0
NON BYPASS 8.4E 3 OK 1.0 1.1E-6 t'
92E 11 BYPASS HPFL WARMUP (2) i LPFL WARMUP (2)
OK 9.6E-4 1.1 E 6 -
1.1E 09 BYPASS 1.4E-4 OK 3
1.1E 6
- 7.4E 14 BYPASS POST ACCIDENT SAMPLING (4}
i l
0 NON-BYPASS 9.6E4 0
~
OK 1.0 l
1.0 1
1.1 E-9 BYPASS i
SLC IWECTION OK.
1.1 E-6
--~
2.6E-10 NON BYPASS 2.4E 4 0
1.5E 3 3
1.0 1.1 E-6 4.0E 13 BYPASS 1
1 TOTALS l!
NON-BYPASS 1.2E-8 BYPASS 1.1 E-9 Figure 19E.2 20A SMALL LOCAS OUTSIDE CONTAINMENT Amendment 19E.2 Il3 i
d
,----m,,
- a ABWR 2m as l'-
Standard Plant
' LINE BREAK LINE OPER.
SECOND DIV.
COOLANT 4
OUTSIDE ISOLATION ACTION NOT AFFECTED MAKEUP (V )
(X )
(P )
(O j)
(0 )
1 3
3 0
MAIN STEAMLINE (4)
OK 6.1E 7 I
2.0E 11 NON BYPASS l
3.2 E 5 0
a OK i
1E 3 6.1 E 7 2.0E 14 BYPASS l
1.0 OK Nogl t'
Negl BYPASS I
FEEDWATER (2)
(INCLUDING RWCU RETURN AND LPFL A DISCHARGE) l OK 6.1 E 7 1
3.9E 11 NON BYPASS '
6.4E 5 OK OK j
1.5E 3 6.1 E 7 -
l>
5.9E 14 BYPASS 1.0 OK Negl i
Negl BYPASS i
SRV DISCHARGE (8).
(LOOPS B AND C)
OK 1
8.5E 6 l
1.4E 10 NON-BYPASS 1.6E-5 2
OK OK 4.2 5 3 8.5E 6 j
2.8E 13 BYPASS 5
l 1
OK 1 E-3 3.7E-3 1.2E 13' BYPASS:
-TOTALS NON BYPASS 2.0E 10 BYPASS-4.8E 13 4[
Figure 19E.2 20C LARGE LOCAS OUTSIDE CONTAINMENT y,,,,
19E2 t15 I
.,., ~,,, ~,, _,. _,
........,,.,,.._m
,_....c,,...
ll 23A6100AS Standard Plant Rev A 4
i LINE BREAK LINE OPER.
SECONDOlV.
COOLANT OUTSIDE ISOt.ATION ACTON NOT AFFECTED MAKEUP (V )
(x3)
(p 3)
(a3)
(o )
3 g
1 4
1 HPCF OlSCHARGE (2)
OK 8.6E 6 4
2.3E 10 NON BYPASS 3.2E S
]-
0 OK I'
4.2E 3 8.6E 6 i
S.8E 13 BYPASS
.5 j
1 OK 7
1E 3 3.7E 3 l
RCIC STEAM SUPPLY (1) 2.4E 13 BYPASS RWCU SUCTON (1)
I o
OK i
l 3.2E 5 0
oK OK-1.0 8.6E 6 i
2.8E-11 BYPASS i
1,o OK 1E 3 i
3.7E 3 i
1.2E 11 BYPASS i
,I SRV DISCHARGE (8)
?
O NON BYPASS 1.3E-4 0
NON BYPASS I
1.0
'OK I
~
j,o 6.2E 7 -
5
- 8.1E 11 BYPASS
~
I TOTALS-NON-BYPASS 2.3E 10 BYPASS 1.2E 10 Figure 19E.2-20B MEDIUM LOCAS OUTSIDE CONTAINMENT g9gy,39 Amendment
+
u
--_--_,4.m-,
e 9.
m y,
-y, r-vy w-q.o-ym----4.=
w9 p.
9