ML19332D359
| ML19332D359 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 11/28/1989 |
| From: | Scaletti D Office of Nuclear Reactor Regulation |
| To: | Marriott P GENERAL ELECTRIC CO. |
| References | |
| NUDOCS 8912010069 | |
| Download: ML19332D359 (12) | |
Text
__.
z*?
,. y,
.t c
4 7
i 13 Docket No. ' S'IN 50-605' November-28, 1989~
i Patrick W. Marriott,. Manager Licensing & Oansulting Services 2 GE Nuclear Energy
(
Guneral Electric C2tpany a
175 Curtner Avenue San Jose, California 95125
Dear. Mr. Narriott:
SUBJECT:
Jewwl' FOR ADDITICNAL INFORMATIN RE3ARDING 'IHE GENERAL EIECIRIC CCNP,W APPLICATION FtR CERTIFICATION OF 'IHE AMR DESIGN J
In our review of your application for certification of your Advanced Boiling Water Reactor Design (AMR), we have identified a need for additional information.. Our request for additional information, contained in the.
enclos:are, ailresses the severe accident review information provided in Appendix 19D of your AMR SSAR.
j In order for us to maintain the AMR review sdudule, we request that you provide your responses to this request by January 8,1990. If you have any concerns regarding this request please call me on (301)492-1104.
Sincerely,
/s/
l Dino C. Scaletti, Project Manager Standardization and Life Extension Project Directorate Division of Reactor Projects - III, IV,.
V atrl Special Projects Office of Nuclear Reactor Regulation c
H Enclosure.
' As stated DISTRIBUI' ION:
.EDockst!Filej MRubin NRC PIR PNiyogi gpo3 PDSL Reading.GKelly DScaletti MCunningham i i CMiller RBarrett EChelliah TKenyon A'Ihadani MRood PDSL
$ b
~ ett8 CMiller 11/Jg/89 1
89 11/d/89 8912010069 891128 PDR ADOCK 05000605 PDC
% i, 9,
it-4 y
j dll
^
i.
7 (ic
.T Docket No. S'IN 50-605 November 28, 1989 Patrick W. Marriott, Manager
- Licensim & consultire Services
'GE Nuclear Energy General Electric Campany 1175 Curtner Avenue
. San Jose, California 95125 L
Dear Mr. Marriott:
SUBJECT:
REQUEST FOR ADDITIONAL INPWMATICE REXTARDING 'IHE GENERAL EIECIRTC CXMPANY APPLICATION FOR CERPIFICATION OF 'IHE AS4R DESIGN
.In our review of your application for certification of your Advanced Boiling Water Reactor Design (AIMR), we have' identified a need for additional 3
information. Our request for additional information, contained-in the enclosure, add-the severe accident review information provided in Appendix'19D of your AIMR SSAR.
.In order for us to maintain the.AIMR review schedule, we request that you provide.-your responses to this request by January 8,1990.
If you have any.
concerns regarding this request please call me on (301)492-1104.
i Sincerely,
/s/
i Dino C. Scaletti, Project Manager Standardization and Life Extension Project Directorate Division of Reactor Projects - III, IV, V and Special Projects Office'of Nuclear. Reactor Regulation L
- Enclosum:
As stated
]
DISTRIRJrION:
Docket File MRubin NRC PDR.
PNiyogi PDSL Reading GKelly i
DScaletti MCunnirgham EChell]s'!
Glille' RBarrett iah TKenyon A'Ihadani MRood L
PDSL
),'-
ett8 CMiller 11//g/89 11/' 89 11/d/89
n i
k_
l Rimussr rea Acomaat,ImmoxIcm AIMR SSAR APPENDIX 19D i
1.
In most of the currently available BWR PRAs, the loss of offsite power sequence with an==ful recovery of offsite power within 30 minutes (i.e., 'IM sequence in Fig.19D.4-4) is transferred to the MSIV closure (i.e., isolation events) event tree. - Please provide the basis for transferring it to the reactor shutdown tree (i.e., Fig.19D.4-1) instead.
2.
Should not the event tree top event, Q (Feedwater), appearing in the reactor shutdown event tree (Fig.19D.4-1) be replaced by "Feedwater and PCS"? Otherwise,- a brandi shculd be wMad to the emnat sequence L
(with an end state of OK) to determine the success or failure of the top
- event, W.
Note that wdrimer problems- (hardware or others) can lead to a manual shutdown.
3.
Please provide the basis of not crediting automatic depressurization for the safety function, X, in the reactor shutdown event tree-(Fig.
19D. 4-1).
4.
Does ABWR have a design feature whidi allows the operator to utilize RCIC in steam condensing mode to transfer reactor decay heat to the ultimate heat sink? If.yes, why is no credit given to such a feature in evaluating the safety function W (containment heat renoval)'.
5.
In a=aarttially all of the event trees shown in Fig.19D.4-1 through Fig.
l 19D.4-14, failure of the W function (long-term heat removal) is assigned a probability of failing to run RHRA or RHRB or RHRC rather than failing to start ard run RHRA or RHRB or RHRC, if the preceding V furx: tion (RHR injection or cordenser) is a s===.
'Ihis would be correct if one of-the RHR punps was sumaa= fully started and run to acocmplish the mission of the V function, and then switched to a long-term heat removal mode.
Note, however, that s e a== of the V function can also be achieved, as indicated in Table 19.3-2, by using one weersser punp ard one condenser transfer punp. In sudt a case, the approadi taken in the ABWR IER will urderestimate the failure probability of W since the RHR punp has to be started and then run throughout the mission time. Also, can one low pressure RHR pump alone always acconplish the missions of both the V and the W functions for all the transients including a large IOCA?
6.
In both the non-isolation event tree (Fig.19D.4-2) and the isolatiory' loss of feedwater event tree (Fig.19D.4-3), the qpa - t sequence (with an end-state of OK) should brandt out at the top event, W, since suma== of Q (feedwater) alone does not autcanatically warrant sucx:ess of W.
'Ihe same comment also applies to the IORV event tree (Fig.
L 19D.4-11).
/ :.
i
\\
1 7.
In Table 19D.4-1 through Table 19D.4-17, the brarx21-point value of the safety function V (IPFIA or IPFIB or IPf1C available) was assigned a value of-1.27E-02, with the source of the data given as Table 19D.4-1.
No sucta data, however, can be found in Table 19D.4-1.
Also, for the loss of offsite pcWer event trees, failure of V (IPFIA or IPFIB or IPFIC or one coMensate'and one condensate transfer punp) is given a value of
- 7.37E-03.
Again, no such data can be found in the tables. Please explain how these values were calculated.
8.
For isolatiorVloss of feedwater events, av==ful RHR operation using the PCS requires reopening of.the MSIVs and the recovery of feedwater if it i
in initially lost. In Fig.
19D.4-3, whict) event tree top event takes into consideration the reopening of MSIVa? Also, will the ctiance of reopening the MSIVs be analler if there are stuck-open SRVs?
9.
In the loss of offsite power and station blackout event tree (Fig.
19D.4-4), the probability of failing all three diesel generators (7.99E-04) is used to sort out station blackout sequences (i.e., BE2, BE8, and BEO) frun the loss of offsite power sequences (i.e., 7E2, TE8, aM TEO).
Note, however, that "all DG not fail" could insan:
(1) one DG is available, (2) two DGs are available, or (3) all three DGs are availaole.. In Figs.19D.4-5 and 19D.4-6, the unavailability of Uh (iIPCF B or C with a probability of 4.52E-03) was ocmputed b=M on the assunption that two diesel generators are available. If only one DG is available at the onset of loss of offsite power, this unavailability could bam= larger.. It appaars that sczne kind of weight-averaging
- should be applied to modify this value MaM on the probabilities of having either one or two DGs when the loss of offsite power rmws.
Also, in Fig.- 19D.4-4, the failure probability of opening SRVs follcuing an A' INS event was taken to be 1.0E-06.
For A7WS events, a large number (15) of SRVs need to be opened for pressure relief, arrl, hence, the failure probability of opening the required number of SRVs can be expected to be larger.
10.
In 11 of the loss of offsite power event trees (Figs.
19D.4-5, 4-6, and 4-7),- the failure probability of HPCF (Uh) is taken to be the same irrespective of the offsite power recovery time and regardless of whether there are stuck-open SRVs. Can the heating up of suppression pool for a prolonged period of time due to stuck-open SRVs adversely affect the availability of HPCF?
11.
Please provide the basis of not considering stuck-open SRVs in the station blackout event tree (BE2, Fig.19D.4-8).
12.
In the same event tree cited above (item 11), the failure probability of W(RHRA or RHRB or RHRC) is taken to be 5.19E-04, which does not correspond to that (1.58E-03) shown in Table 19D.4-1 for the case of loss of offsite power. Are the values shown in the column under the heading of "Ioss of Offsite Power" in Table 19D.4-1 also applicable to station blackout? If not, please explain.
.ve-
.-w-
---~w..
--w--
, +
.v
- 13. In the station blackout event _ tree (BEB, Fig.19D.4-9), why does the sequence with siw=== of RCIC need to be branted out for testing the success of HPCF? According to the a n *== criteria listed in Table 19.3-2, an===ful core. cooling using a hi@t pressure system can be achieved by using either RCIC or one train of HPCF for all transients including loss of offsite power. Furthermore, both HPCF and ~ IPFL require ac power which, in this case, is not available for nearly 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Please explain Wy both HPCF an:1 IPFL are included as event tree top
' events.
14.
For IORV transients, there is no insnediate autanatic scram signal, and the operator may be required to manually scram the reactor and start the makeup system before the suppression pool tenperature exceeds the heat capacity tenperature limit. Please provide the basis of not includirg
" timely manual scram" as an event tree top event in.the IORV event tree (Fig. 39D.4-11).
15.. Please explain Wy feedwater (Q) was not credited as a viable means of core cooling in the small LOCA event tree (Fig. 19D.4-12). Note that, according to the siw = = criteria shown in Table 19.3-2, feedwater can be-used to. sinmaafully cool the core in the event of a small steam LOCA.
16.
Please explain why HPCF is given credit in the large IOCA event tree (Fig.19D.4-14) despite the high degree of depressurization caused by the
-large IDCA.
17.
Please provide justification of not considering vapor suppression in the large IOCA event tree.
18.
In u.is:d.ructing the AINS event tree '(Fig.19D.4-15), no distinction was made between ATWS events with MSIV closure (isolation) and those with bypass available (non-isolation), although the former is generally more severe and limiting.
Please explain W y the same branch-point
[
probabilities were used in quantifying the AIWS sequence frequencies 1
despite differences in the st=== criteria, such as the time available for the operator to inhibit ADS or the unavailability of normal heat removal system for containnent heat removal (see Table 19.3-3).
19.
It appears that the low core-damage frequency (9.1E-09/RY) found for AIWS sequences is mainly driven by the low initiating event frequency (9.34E-09/RY), which was obtained by taking scram failure prubability (C) to be 1.0E-08.
Please explain in detail how this scram failure probability was calculated. Fran the fault tree developed for a single control rod drive (Fig.19D.6-17a, Figure 1), the probability of failure to insert an irriividual control rod can be estimated to be roughly 3.0E-06.
No explanation, however, is given as to how this probability is used to generate the probabilities of the basic events shown in the fault tree of control red drive system (Fig.19D.6-19a, Figure 1). Also, no probability data is given for the event RPS (RPS fails to initiate scram) appearing in the fault trees for reactivity control (Fig.
19D.6-16b).
l l
l
~.
_. _. ~.,. _.
\\
c.
20.
In Table 19.3-3, the time available for the operator to initiate one train of SIC is given to be 10 minutes for both isolation and non-isolation A%IS events. Should not the time available for the former be shorter because the sqppression pool is heated up sooner?
21.
For an A' INS event whid is initiated or amnied by closure of'all-MSIVs or loss of condenser, can adequate core coolant inventory be maintained by RCIC alone (as indicated in Table 19.3-3)? For sane IMRs of current design, su s an event requires HPCI or a ocmibination of HPCI r
and RCIC.
22.
In quantifying A%1S sequence frequencies, the same branch-point value was used for W (containment heat removal) regardless of whether there are stuck-open SRVs. Was suppression pool heating due to stuck-open SRVs taken into cocount in estimating the failure probability of W?
23.
Is there any reason why the event tree top event " ADS inhibit" in the Anis event tree is placed before "Feedwater or HPCF" and "RCIC" although it appears more logically wu ct to place it after the latter top.
events?
L
- 24. Was any functional event tree or fault tree developed to - analyze the unavailability of feedwater, condensate, and mderser system? How was the unavailability of feedwater (Q), for exanple, evaluated for different transient initiators?
25.
In the event tree quantifications, the frequency of a particular accident sequence was obtained by multiplyirg together the initiating event-frequency and the bran &-point probabilities of the failed safety functions, such as U,'V, or W, appearing in the sequence description.
'Ihis approach. is pruper if the branch point probabilities were evaluated by properly accounting for the u
.a.-a.;de failures among the event tree top events by linking together the relevant fault trees. Were these fault tree linkings done in the ABWR analyses to obtain the upper-bound of minimal cut sets for safety function failures, such as W, QUV, or UVW? If not, please explain how the branch-point probabilities were calculated for the irdividual safety functions, su& as U, V, or W.
- 26. Were all the system failure pIcbabilities (except for RCIC) listed in Table 19D.4-1 obtained by quantifyiry the fault trees shown in Section 19.D.6? Were the probabilities of failirg all ECCS systems conputed by linkirg the high pressure and low pressure system fault trees? If so, which mode of the low pressure system was used? Also, were these values actually used in the event tree quantifications?
27.
Were the fault trees for the support systems, such as electric power system, service water system and instrumentation system, individually quantified? Are the results of such fault tree quantifications (in terms of minimal cut sets) available for oczparison with BNL calculations?
e
-w.
.~,mv-y.
f. *G,
- 28. What modifications to the fault tree irlput data were made to' obtain the systen failure prtbabilities cou+41ng to loss of offsite power (last column of Table 19.D.4-1)?' Was the failure of swit& gear taken into consideration when the. failure probability of tlw W function (for exanple, in Fig.
19.D.4-7) was calculated?.
-i 29.
Please briefly dammibe the possible jnpacts of omitting the developnent of systan fault tree for plant air systen on the frontline and the support systems.
- 30. 'It was noted that a very small fraction of the failure data shown in Table 19.D.6-2 through 19 D.6-7 are inconsistent with those shown in the relevant fault trees (for exanple, DIV2WX, NW14EM and HXV032CQ in Table 19 D.6-2).
Whi& values were actually used in the fault tree quantifications?-
31.
'Ibe break areas for the various IDCAs (large, medium, ard smdi) are defined to be significantly larger than those used in, for exanple, the Limerick PPA. Do the initiating event frequencies used in the event tree' quantification reflect these cnanges in the definition of break sizes?
32.
How does the RWQJ (reactor water cleanup) system work to remove decay heat? What suction lines are used? What is the heat sink? Does the non-generative heat exchanger have enough capacity to remove decay heat?
- 33. ' For RHR shutdown cooling mode, suction is taken from RW.
Where are the points of suction for the three suction-lines?
Also, where are the discharge points for the core cooling subsystem return lines?
34.
Questions on Table 19D.4-1.
(i) What modifications were made to the fault trees to obtain the failure probabilities wu=:rci-ding to large or medium 10CAs?
e (ii) Are the RCIC failure probabilities calculated by quantifying the revised fault trees in Aiiniidimuit 8?
(iii) What are the failure probabilities corraspcidirg to station blackout?
- 35. What modifications were made to the fault trees to obtain the core damage frequency couwspordirg to incorporation of (a) gas turbine generator; and (b) fire system water wiu =Sction?
36.
Following loss of offsite power, feedwater punps (motor driven) are tripped and MSIVs are likely to be closed. Are the IW pumps or the RWCU pumps connected to DG power source?
Is re-opening of MSIVs considered in calculating the probability of NHR for the W function? In other PRAs, feedwater is considered unavailable following IDOP.
r;
~..
wy
.4 -
37.
Class II sequence friquency was calculated to be 4.29E-6.
The input to L the Class II contalment event tree, however, is 2.5E-06.
Please explain the difference. Was the CDF for Class II sequences.(4.29E-10) obtained j
'by taking 0.01% of 4.29E-06?
- 38. A7WS transient scenarios vary significantly depending on whether E IV are closed or whether offsite power is available. How can a single A7WS event tree properly handle all A7WS events of different initiators?
- 39.
In the A7WS event tree, failure to initiate S145 is given a probability of 0.2 (time available for the operator = 10 min.). A typical value used for this action in nost of other BWR PRAs is 0.87 (with time available for the operator = 8 min.). Please explain the difference, e
40.
In the A7WS event tree, the probability of failing to inhibit ADS is
. taken to be 0.1.
A typical value used in other PRAE is 0.5 if high I
i pressure core injection is a-failure, and 0.00S if HPCI is a sumaa.
'Ib
- be able to make sudi a distinction, the order of the event tree top
~
events for "HPCI" and " failure to inhibit ADS" must be intertianged.
'41.
For loss of offsite power' initiators, stuck open relief valves (SORVs).
were considered in Aiieru-ht 4, but were eliminated in Amendment 8.
Please explain why.
42.
For isolation / loss of FW events, the unavailability of feedwater is taken
.to be 0.43 (= 40%(1) + 60%(0.05)). Is not the value, 0.05, too optimistic for the E IV closure initiators?
43.
In' order to expedite the staff's review, please provide a copy of the MAAP code and requisite input information that was used in the ABWR evaluation.
44.
Please provide a copy of the magnetic medium containing all system level fault trees and functional level fault trees modeled for all the initiating events applicable to the ABWR.
Back-End PRA 45.
Please provide the input files for the MAAP calculations.
46.
The probability of containment failure resulting from loss of heat removal is given as 3.4e-6 in Section 19.1.2.
However, the frequency of containment structural failure resuling from loss of containment heat I m oval is given as 2.5e-7 per reactor year in Section 19D.5.12.4.
Please clarify.
+
47.
Is the failure pressure of the upper drywell (UDW) head above 500 degrees F irxiependent of the UDW tenperature? If it is a function of temperature, please provide the function. Please also provide the leak area for the high temperature failure. Is high tenperature failure considered to be P(penetration) or D (drywell head failure) in the release mode fium containment when binning the accident sequen ?
- i
)
1 j
s
- 48. What is the location arti sizes of the passive flmiars? Please describe the melting W m of the passive flooder fuse including the tanperature distribution in the fuse. What is the reliability of these f1mdars?- Are there any exanples of their use in other indusries?
49.
CE7T for Class IV accidents was not developed because of negligibly low I
occurence frequencies (Section 19D.5.11.1.). However, CLTs for the accident claaman with similar or lower frequencies (Classes IB-3 and IIIA) were developed. Please explain.
E 50.
With respect to Firwater Addition (FA)
Is it r
_ty to have a separate "FA" category for a mitigating a.
feature? It appears that "FA" included in "IV."
(e.g., Figures 19E.2-6 hibes a sequence of SBRC-FA-DO. However, this sequence is-binned as SBRC-IV-DO in CETF IB-2, Figure 19D.5-8.)
'Ihe CETs do not show any sequences with "FA."
b.
It is stated that credit is not taken for firwater (W) for preventing core damage due to reactivity concerns. Howevax, it appears that W is credited for same of the core melt arrest in the RPV. Is there any study available reganiing the reactivity in a partially av oore?
1 L
c.
How is the firewater addition or spray handled in the CE7rs? It appears that it is included sanetimes in "ARV"(e.g., Seq.3 of CETT IA) and sometimes in " ARC" (e.g., Seq.6 of CET IA-1). Would it not sinplify arx1 clarify the CETs if W is designated as a separate headirg? W appears to play a major role in reducing the release fractions by scrubbing in case of containment failure. (A suppression pool loses its scrubbing function once the vessel fails). 'Iberefore, it is inportant to know if W is available for a particular sequence.
l.
- 51. 'It is repeatedly stated that oorium cools in the IIM after vaa=1 failure l
by the water whidt was retained in the lower plenum in many of the l-accident descriptions. Why did this water not cool corium in the vessel before va==al failure? How nuch of water is available in this manner?
E Would accidents progress differently if the water cooled the core in E
vessel?
l l
52.
Questions on Figures 19E.2-2 (Accident Sequence ICLPPFIE) l-a.
In Figure C, why does the upper drywell tenperature continue to increase throughout the accident?
l b.
In Figure E, why does the drywell water level change between the PF 1
openiry and the DW head failure?
c.
In Figure B, why does the drywell pressure decrease after water boils away?
('Ihe gas tenperature does not show any correspondirg drop during this period.)
h,.
]
53.
Questions on Figures 19E.2-5 (Accident Sequence IGPPFPH)
- Figure A shows a pressure drop at about 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br />. 'Ihis was explained a.
in the. text as being due to the flow of water fran the suppression pool into the drywell (A similar phencananon was shown in Figure 19E.2-11.).
Please clarify. It appears that the W prammire should be higher than the W pressure during this period. 'Ihis pressure drop appears to delay the m head failure by about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. What impact will this have on the final release fraction?
54.
'Ibe suppression pool bypass due to stuck open W-N vacuum breakers is of concern only for maaa involving wetwell venting. Please explain the O
consequence ratio of 825 used in the equation on Page 19E.2-40. In the same equation, the fire water unavailability of 1.5% was assumed, which is considerably lower than.10% used elsewhere. Please explain.
- 55. 'Ihe CET top event " ARC" (core melt arrest in containment) can occur if any of the following coruiitions exist, RHR is available, or RHR is recovered, or FW is available, or PF operates.
Except for FW, other features are already designated as top events of CETs (CHR,RG,PF). Is it rwcamy to have " ARC" as a separate headirg? It p
appears to be duplicative and confusirg regartling how " ARC" occured. (It is i'
confusirg since sczne of the top events are operation / availability of-systems while some of them are events caused by operation of.the same system.)
l
- 56. High tenperature failure (HrF) occurs if corium is carried to the UN and no spray is available. Does the probability 0.01 incluSe the probability of both of these ' miring? Wouldn't it be clearer if this heading is l
' replaced by "Corium in the U W" and " Spray available"? (See also Question 18.a) 57.
Questions on Class IA/IA.1 and IIIA/IIIA.1 CETs g
l:
i a.
High temperature failure probility is identical whether RHR is available or not ill these CETs. However, if RHR is available, the probability to have UN spray appears to be higher and,therefore, the probability of high tenperature failure smaller. (See the previous question)
- b. -
Why isn't the probability for " ARC Yes" 1.0 when RHR is available j
(i.e., what does the probability of 1.e-5 represent in Sequence 4 of l-CETIA?)
c.
Sequence 3 of CET IA is binned as..FSNN. Does this inply that core melt is arrested in the containment due to FW? Why not RHR?
d.
How is core melt arrested in the containment without RHR for Sequences 4 and 6? Is this due to FW?
7 h
4
?
e.-
Mat is the basis for the containment failure probability at the time of vessel failure, 0.001, or high tenperature failure probability, 0.017 mat is the sensitivity of the final consequence to uncertainty in these rumbers?
L
' 58.
Questions on Class _IB-1/IB-1.1 and IB-3/IB-3.1 Trs.
a.
How is the core melt arrested in the contairanent for Sequences 2 and 4 of these CETs? Are these probability same for IB-1 ar.d IB-3 because they are solely due to FW? b.. M y isn't the RHR recovery probability 100% for Sequences 2 and 5 for IB-17 Wy is probability of the RHR recovery failure significantly higher c.
for Segunce 7 than for Sequence 4 in IB-17 d.
Wy is the probability of RHR recovery failure 5 times higher for i
Sequence 4 of IB-3 than Sequence 4 of IB-1, while they remain the same between Sequences 7 of IB-1 and IB3?
(Incidentally, the "ROI No" branch probability for Sequence 7 of IB-3 appears to be misprinted. It should be 0.1, not 0.01.)
Sequence 7 of IB-1 is binned as PEDI while Sequence 7 of IBl.1 as e.
PSI 24. '1his inplies that the consequence of the low pressure vanel failure is more significant than that of high pressure. Please explain. ('Ibe same uestion for IB-3.)
- 59. Questions on Class IB-2 Tf.
a.
'Ihe core rh5 fregwncy for this class is not the same as that of Table 19.3-6. Please clarify which is correct.
b.
'Ibe probability of failure to depressurize the reactor is 3 times lower for Class IB-2 campared to Class IB-1/3 (0.002 vs. 0.006). Is this due to the time available before durii:ssarization? Does this probability d= pard on how nuch time is available before the' demand of this aqni,mant? (i.e., what action can be taken to improve availability of this equipnent before challenge regardless of how much time is available?)
Please provide the basis for the "ARV No" branch probability of 0.006 c.
for Sequences 4 to 7 and 0.6 for Sequence 12.
d.
Wy is the " ARC No" branch probability of Sequence 7 significantly higher for this CETF than others (0.05 vs. 0.01)? Wy isn't this branch further divided depending on the RHR recovery? ('Ihis is done for cases which have even smaller probabilities.)
Segaence 6 is binned as FSDI. 'Ihis is the only place where a sequence e.
is binned as "High" when IW scrubbing is available. Please explain.
'O.
-t f.-
my is RHR unavailability significantly lower for Sequence 11 capared to the similar sequences for other CETs(0.01 vs. 0.05 for IA)?
g.
My isn't Sequence 12 further branched like the similar sequences of-IB-3.17
= 60.
Questions on Cla - ID and IIID CETs.
~
- a.
~ Hcre is core melt arrested in RW? Is this solely due to FY/ ('Ihis branch existed in Amendment 4 which did not have FW.)
b.
Why is the probability of RHR reocwery failure significantly higher in this CET than in others?
- 61. Questions on CET II a,
'Ibe "CC No" branch fraction is significantly rMM from Amendment 4 to Airsdigit 8(0.001 frun 0.1). Besides the availability of firewater, what else tributed to this reductiori?
r
-,A I
T l-l l.
1
.