ML20105C177
| ML20105C177 | |
| Person / Time | |
|---|---|
| Site: | 05000000 |
| Issue date: | 11/18/1983 |
| From: | MARYLAND, UNIV. OF, COLLEGE PARK, MD |
| To: | |
| Shared Package | |
| ML20102A952 | List: |
| References | |
| FOIA-84-351 NUDOCS 8502090333 | |
| Download: ML20105C177 (186) | |
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.ro*sessoe is#de r neMs 1.0 Introduction In the Accident Sequence Precursor (ASP) study it was assumed that a specific precursor event, mitigating system failure, or initiating event was applicable to all nuclear power plants.
Further, two generic sets of standard event trees for PWR and j
BWR plants were developed and used in the analysis process.
Finally, for each precursor, the conditional probability of subsequent core damage (Pscd) was calculated from the analysis of these generic trees and averaged by dividing it by the total number of reactor years.
Much concern has risen because of the generic approach taken in the ASP study, mainly due to the fact that not all precursors that occured in a specific plant can apply to every plant of the same type.
Even it an event does apply to many plants the prob-ability of subsequent core damage may vary in plants of the same type.
Because of this concern, the study presented in this report was initiated.
The objective of this study was to estimate the im-pact of using a more plant specific approach versus the' ASP generic approach.
This study started with the calculation of more plant specific evaluations for the BWR's simply because there are fewer BWR's and a lesser number of precursors that occured in them.
The next part of this study will, however, perform similar calculations for the PWR plants.
Ideally, the most appropriate approach would be to employ specific event trees for each of the BWR power plar ts, because there are no two plants that are similar in design, operation, or maintenance.
Precursors that have happened in a plant which can potentially happen to other plants should then be identified and applied on the plant specific event trees. ' Finally, the frequency of the' subsequent core damage for each percursor and for the specific B502090333 840703 PDR FOIA
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factor of as high as 2 if the generic approach is used.
Certain precursor events were, however, observed to have an over-es-timation of more than an order of magnitude.,In a few instances an underestimation of as high as an order of 4 by using the generic approach were seen.
The plant specific approach used in this study is a very straight forward one which models and estiraates a more accurate rep-resentation of the precursor analysis than the generic approach used in the ASP report.
We strongly recommend to implement this approach for further analysis of the precursors.
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plants that they apply to, should be calculated.
Summation of all of these frequencies would yield the estimate of average industry-wide frequency of core damage.
The problem with this approach is first, that the development of plant specific event trees is a prohibitive task and is out of the scope of this study.
Secondly, application of complete plant specific approach will
. severely limit the use of LER data in estimating probability of loss of safety systems and frequency of initiating events.
The
. use of c.omplete plant specific approach is also not necessary, because it was observed that there are groups of plants that re-spond closely to an initiating event or precursor.
Therefore, one can group the plants into categories with close response.
To deal with the difficulty just stated above, in this study the BWR plants were grouped into categories that respond similarly
- to an initiating event.
The methodology to categorize the plants are dicussed in detail in the next chapter.
For each category a set of event trees for Loss of Offsite Power (LOOP), Loss of Feed T
Water Events (LOFW), Loss of Coolant Accidents (LOCA) and Main Steam Line Breaks (MSLB) initiating events were developed.
The-trees developed were based on the available PRA's of a specific plant in each category.
A total of five categories were defined and one of the categories was divided into three subcategories.
In a review of all of the BWR precursors identified in the ASP study, applicability.of each category or subcategory to these individual precursor was determined.
Then, the loss of function probabilities and frequency of initiating events were calculated.
Finally, frequency of core damage for individual precursors and for each category was calculated and the total' frequency of core damage was obtained.
The results of this more plant specific calculation showed that the frequency of core damage can be over-estimated by a
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l 2.O BWR Categori2ation 2.1 Review of Procedure A three step procedure has been used to divide the BWR plants into specific categories.
In the first two steps, the major plant categories were generated.
In the third step sub-categories for specific ever.t situations were identified.
In step 1, each plant was examined to determine what systems it utilizes to perform the various generic plant functions which must be performed in response to any initiating event.
These generic functions have been identified in many probabilistic risk assessment studies and methodology documents and referred to by many different names.
In general, they can be summarized as follows:
reactor subcriticality
- - vessel water inventory short-term core heat removal containment overpressure protection long-term core heat removal containment heat removal radioactivity removal Step 1 identified for each plant, those systems that the plant has to perform each of these functions.
The initial plar.
categories were selected so that the plants whose systems are
- nominally identical were grouped.
The plants with systems of the same type and function, without accounting for the dif-ferences in the design of those systems, were thus grouped.
In-Step-2 these categories were refined by taking into account maior differences in.the design and operation of the plant systems identified in Step 1.
There was a certain amount of h
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l subjectivity in this process, and the analyst must have the knowledge and experience to be able to judge what a major design t c" difference is.
This judgement is based not so much on f:he mechanical concept of difference in design, but rather is in-tended to be based on a probabilistic concept.
A major design difference is one which would greatly affect the availability a system to perform its intended function.
A great amount of in-l sight is required to make this judgement, since all facets of a system's operation must be considered.
The effect of system differences must consider recoverability and other human inter-i actions as well as base unreliability.
There is obviously no
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set rule which can be utilized for Step 2.
By way of example l
however, such things as three pumps rather than two, or three-out-of-four as opposed to two-out-of-four operation are generally considered not maior.
However, such things as turbine pumps rather 1
.than motor driven pumps, or shutoff head greater than reactor operating pressure as opposed to less than reactor operating pressure are generally considered major.
Even those examples cannot be used as herd and fast rules.
At the~ conclusion of Step 2, the major plant categories were established.
These categories served to allow construction of event trees that were reasonable representations of the response to various initiators of the plants within each category and the evaluation of event sequences for most observed precursor events.
There were, however, come specific events for which -these groupings were not sufficiently unique.
Since only a small number of events require this additional detail, it is not reasonable to further break up the categories for all cases.
This would only serve to further dilute the available data base.
Step 3 is intended to develop sub-categories within each category which will be utilized only for those events which do not apply equally to all plants in a category.
This development
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is carried out by determining the plant specific applicability and response characteristics for each precursor event and each plant.
In most cases, every plant in a category will be essentially identical in its response to a particular precursor.
For tho'se few precursors for which this is not true, sub-categories are created which are used'only when evaluating sequences which
_ include that particular event.
For the evaluation of all other events, the major categories are left intact.
It is important to note that this categorization applies l
only to the deterministic aspects of event tree development.
In many' cases, data which may be used to quantify the event tree
- sequences must be applied in a different manner.
Data for specific
- systems may span more than one category, whereas data for other systems may apply only to the plants in a specific category or subcategory.
I l
2.2 Sanmary of Categories Identified Table 1 summarizes the results of the categorization phase of the work.
Twenty-six plants were considered and seven plaint categories (A1, A2, A3, B, C, D, and E) were selected based upon presence or absence of the system functions as identified in the
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table.
In figures 1 to 6, the generic event trees for LOCA are presented for each category.
Figures 7 to 11 and 12 to 17 present analogous trees for the LOFW and LOOP initiating events respectively.
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Figures 18 to 22 are loss of PCS-initiating event.
In the remainder of this section major reasons for this categoiization and a brief summary of specifications of each category or subcategory is discussed.
In the next chapter the pro-cedure in which the category generic event trees (shown in figures 1 through 17 in this chapter) are constructed from PRA plant specific event trees are discussed.
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The BWR plants were grouped according to their engineered safety system design and feedwater pump type. (See Table 1).
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Category A This category represents the group of older BWRs.
They are not a homogeneous group, but they have similarities which allow them to be evaluated as a single categroy for many of the pre-cursors analyzed.
In particular, they all have only isolation condensors as the sole means of supplying high pressure cooling when feedwater is unavailable.
Also, they all utilize separate systems for containment cooling and shutdown cooling, giving them long term cooling diversity.
For certain precursors, the differences between these plants become important.
This requires that they be evaluated in subcategories.
Subcategroy Al-These plants would be evaluated separately for transients involving loss of offsite power.
The other plants in Category A have feedwater coolant injection systems.
This provides a means of utilizing the feedwater system to provide cooling flow at high pressure when only onsite AC power is available.
The subcategory Al plants do not have the capability, and.thus have less diversity during these transients.
Subcategory A2 ~These plants would be evaluated along with Subcategory Al for precursors which involve common mode type failures in a single low pressure injection system.
Each of the plants in these two subcategories has only one low pressure safety system, the low pressure core spray.
This system also provides the containment cooling function for these plants.
The sub-category A3 plants have both a low pressure core spray and a low pressure coolant injection, a diversity which these plants do not have.
Interstingly enough, when only random failures of the low pressure systems are evaluated the unavailability of the one system
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These plants also saw the elimination of the isolation condenser, which was replaced by the reactor core isolation cooling (RCIC) system.
This afforded additional high pressure injection for very small LOCA events, but was not as simple or reliable as the isolation condenser.
Further, this also served to make additional reductions in the diversity of long term cooling.
The isolation condenser actually provided a third method of long term cooling for the early plants, since it could maintain the plant in hot shutdown for extended periods of time.
The RCIC operates like the other injection cooling systems, thus ultimate long terms cooling by the RHR system is still required.
- Thus, the category C plants reduce long term cooling diversity from three system to only one.
Category B The category B plants continued the standardization begun with category C, and they have all of the same systems.
The difference is that the category B plants replaced motor driven main feedwater pumps (which all the other plant categories have) with turbine driven main feedwater pumps.
This reduces the availability of main feedwater as a source of injection water, with the turbine pumps, any event which causes any part of the secondary cycle to i
fail will result in a total 1,oss of feedwater.
This is because the main feedwater isolation valves will close, isolating steam to the turbine.
With the motor driven pumps, this cannot occur and feed-water can continue running or be easily recovered.
Thus, the category B plants have reduced diversity for high pressure in-f jection of coolant for pressure which in older plants would re-sult in loss of the secondary cycle without failure of the feedwater system.
Category E This category represents only the Lacrosse BWR plant.
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design versus the two system design are reasonably equivalent.
Thus, for many of the precursors, it is not necessary to make the distinction.
Subcategroy A3-These plants have both the feedwater coolant injection system and the two system low pressure systems design.
This group would be evaluated along with subcategory A2 for loss of offsite power and separately for loss of single low pressure system.
Category D The category D plants are lumped together because they have a high pressure coolant injection system in addition to Category A.
This gives the plant two high pressure cooling systems when feed-1
~ water is unavailable.
They do not have a feedwater coolant injection system, but they do have the two low pressure systems.
The major difference is in the high pressure coolant injection (HPCI).
Having two high pressure systems (HPCI and isolation condenser) improves I
response to loss of feedwater events.
Also, injection cooling is now available if a consequential LOCA occurs due to a stuck open relief valve.
In plants without RPCI, it is necessary to blow down and use low pressure cooling in this situation.
I Category C l
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These plants differ from I
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integrated, residual heat removal (RHR) system.
This reduces the number of components, but also elimantes the diversity enjoyed by the earlier plants with separate shutdown cooling systems.
This P ant group is more susceptible to precursors involvihg common mode l
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,--*w
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y
->-.*-www e-
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i 1)
LOFW Event Tree
. = -
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Figure 23:
ASP Study Event Tree for Loss of Feedwater
(
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Figure 24:
IREP Millstone 1 Event Tree for Loss of
(
i All Feedwater l
l In addition to modifications discussed in this section, the probability of success of Safety relief valve to reclose was assumed to be 1.0, thus forcing the event tree to describe only I
l LOFW event and not transient induced LOCA's. since LOCA's are treated separately.
t f
.2 L
~
3.~ 0 BWR Plant Specific Percursors Analysis 3.1 Initiating Events and Function Failures Applied For each precursor event, the initiator and the subsequent safety system failures were reviewed individually.
If the description of the actual occurence (as given in App. B of the ASP report) indicated that the event could occur at any plant, then the precursor was applied to all plant categories.
On the other hand, if the conditions inducing the precursor were plant specific or could apply only to a group of plants, then the precursor was restricted to the specific plants (s).
For example, a LOCA event caused by a stuck open relief valve was considered applicable to all plants, while the LOOP event caused by salt buildup on the 345kv lines and insulators at Millstone I (NSIC
-116780) was considered applicable to only plants next to the ocean.
Some of LOFW initiators that occured at plants of Category B were converted to loss of PCS when applied to the other plant categories, because the use of turbine driven feedwater pumps in Category B results in a LOFW following an MSIV closure transient.
In the case of the Browns Ferry Fire, the description of the event (NUREG/CR-2497 pg B-213) reveals that feedwater was lost because of the MSIV closure, while the feedwater system was not damaged by the fire.
In actuality, the core was cooled through condensate booster pumps after manual depressurization.
Thus if this event is applied to plants with motor-driven feedwater pumps, it would result in loss of PCS only and not loss of the feedwater system.
In a similar manner, mitigating system failures or deg-radations were categorized.
For example, a HPCI failure was not assumed an IC' failure and vice versas as is done in the ASP study.
In some instances a system's failure or degradation applicability was restricted to a subcategory or even to one
..r.,
. m e.
.s is required since the plant is of a different design, having been built.by Allis-Chalmere rather than General Electric.
It is the only Allis-Chalmere plant ever built.
e en Y
-e,.
.w.4.a.-
p.ea.
O
~_
f 1
plant.
For example, the RCIC/HPCI failure cause by a wrong reset logic connection was considered applicable only to the Browns Ferry 1 plant at which it occured (NSIC 85566).
However, if the mitigating system failure or degradation resulted as a consequence of another failure, it was credited for all plant categories in which the initiator was applicable.
As an example, consider the LOOP event with the relief valve, stuck open at Pilgrim 1.
(NSIC 4
).
Pilgrim 1 utilizes RCIC/HPCI systems which were degraded because of the stuck open relief valve.
Thus, in cat-egories B and C the RCIC/HPCI systems are assumed to be degraded in this analysis.
However, when the event is applied to categories j
A and D which utilize isolation condensurs and FWCI, the isolation condenser is considered failed and FWCI degraded, because the isolation condenser can not function with a stuck open relief valve.
Appendix A summarizes all precursors as they are applied to each applicable category.
i-3.2 Category Specific Event Trees i
This analysis used systematic event trees developed by the
" Interim Reliability Evaluation Program" (IREP) as followst In categories A and D the event-trees developed from the Millstone f
Point 1 Nuclear Power Plant were adopted, while for categories B and C the trees adopted were developed from the Browns Ferry 1 Nuclear Power Plant.
The IREP-Millstone 1 event trees used in category A were modified to make them category specific as follows:
(i) the IC and ICMP systems were merged into one event, (ii) the LPCI option was deleted for subcategories Al and A2 (iii) the FWCI option was deleted for subcategory A2 and (iv) the Containment Cooling (CC)' option was deleted after CS or MDP failure since both branches of the event tree lead to core damage.
(This option was i
I
-~~~~
sw.w.---i---e-,-v.-ew-,+we-+=w-
+,=e--m.,
e -v erw w-e - e -
g--
Se taken into consideration in the IREP-Millstone 1 study in order to account for the severity of the sequence in containment calculations).
Event trees used in Category D were obtained from the same set of trees by adding the HPCI option which is missing in the IREP-Millstone 1 systematic event trees.
For Category B, the event trees of Browns Ferry 1 were used with no modifications.
For Category C the Category B event trees were applied but they were modified to include the FW and/or PCS availability where applicable.
t Category E consists only of the La Crosse plant which is considerably different in design from all other BWR plants, and therefore available event trees and function failure probability data are difficult to find.
Thus for this specific category, the ASP study event trees and point estimated were applied.
In the rest of this section the ASP standard event trees, the original IREP trees (Millstone and Browns Ferry) and the corresponding category specific event trees are shown.
The foot-notes in each figure summarizes the' consideration made for any modifications.
?
l l
l l
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---Nw--
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a '.a saa.
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y 3
sea sea tocaen.e.e.
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ta a seaesama i
u seat s wer Figure 25:
IRED Study Browns Ferry 1 Event Tree where PCS is unavailable (T ).
y Note that the LOFW transient in the IREP-Browns Ferry 1 study is part of " transient systematic event tree where PCS is unavailable (TU)."
The tree becomes a LOFW event tree by assuming o
success prob. for relief valve to reclose and MSIV to close to be 1.0.
1 T
C 4-3 L
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e w
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O E
a ac sf ee a
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es (I
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e p-c0 cp (D
~
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- LD CD Figure 26:
IREP-Study Millstone 1 for LOFW Systematic event tree modified for Categories Al and A2
- Assumption of 1.0 for RC(C) success forces the event tree to describe a LOFW event.
4
--.sn~
,,-.,---.,.-,,,,--.-.--._,_-,-m
,n,-.-.---
e E
a 7
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s ep eo cp Led co I
Figure 27.:
IREP Study Millstone 1 for LOFW Systematic event tree modified to apply in Category A3 Assumption of 1.0 for RV(C) success forces the event tree to represent a LOFW event.
9
---"w'e---,
-.--_mm_._________,__,,,h_,
LOFW Event Tree for Categories B and C
?
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n U
V A.
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e-em CD 10 Ok c0
=
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, Leo Lete CD cp Figure 28:
IREP Study Browns Ferry 1 for LOFW Event Tree Modified to apply in Categories B and C
- Assumption of 1.0 for RVCC and MSIV closure success forces the event tree to represent a LOFW event.
.+
n m--
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Figure 29:
IREP Study Millsteon 1 for LOFW Systematic Event Tree Modified to apply in Category D.
- Assumption of 1.0 for RV(C) success forces the event tree to describe a LOFW event.
+
1--
_y_.
I I
g s i g
g 3 [
g 1
8 8
..............t.gg
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[
l Figure 30:
IREP Study Millstone 1 for Loss of PCS (Excl.
l Feedwater) (T ) Loss of Nomal AC Power (T )
2 4
i The ASP Study did not include PCS failure ac an initiating event in i*a event tree sequences.
The " Transient systemic event tree where PCS is unavailable (TU)" shown in figure 20 and used for LOFW event is the appropriate tree for loss of PCS event.
This tree was used in Category B as it was and modified to include the FW availability for Category C.
N e
e J
L I
-8 T
<[= F D
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=
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(
Figure 31:
IREP Study Millstone 1 for Loss of PCS Event Tree Modified to apply in Categories Al and A2
- Assumption of 1.0 for RV(C) success forces the event tree to represent a loss of PCS event.
T W
'R g
T F
p L
e N,
Id P
V W
{
j a
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e c
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ow
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8 CD l
es et cD et et o... _
CD cp
+
Lect 49 cm Figure 32:
IREP-Study Millstone 1 Event Tree for Ioss of PCS modified to apply in Category A3 Assumption of 1.0 for RV(C) success forces the event tree to represent a loss of PCS event.
i-
-e-,-
.,-,,-,w
-,,----,,,,.,,_--,,,,,,,-_,_,,,,,,-,.,,-,w,w_,_.,,,_-,,-,.__,,.,..,,,,,,-w,,,.-,
, - _ _., -,a_
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o 0 0
. od $
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=
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it co H
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~
ct
-ce Figure 33:
IREP Study Browns Ferry 1 Event Tree for Loss of PCS applied in Category B
(
- Assumption of 1.0 for RV(C) and MSIV closure success forces the event tree to represent a loss of PCS event.
-w,w,oa
w
~
~
g W
c
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~ T h h
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- t. g 9.
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9 D
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e a
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s or 411 lA
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4.ca
.a ca Figure 34:
IREP Study Browns Ferry 1 event tree for Loss of PCS modified to apply in Category C
~
w-,e-+
t T
a M
l-h.,
,.f I
L.
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s* V R
A e
e e
=
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r v
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r P
f
'S C c as I. e so
=.
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=
e0 s
ao i..'e4
. o _. _
so Figure 35:
IREP Study Millstone 1 event tree for loss of PCS modified to aoply in Category D Assumption of 1.0 for RV(C) success forces the event tree to represent a loss of PCS success
~, - - - -
.-,.--.-e-y w-
--w-w,
--v-,
3)
Loss of Offsite Power Event Trees
- r,;,
- :
tr;"
0;;;'
0;;;;;/:*r un!":' ::"' c t' "' P-
'~'
5,i:::t.;.
cf..,
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=
s i..
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9
- r..
ie a
v.,
If i,
s..
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n Figure 36:
ASP Study Event Tree Standard event tree for BWR loss of offsite~ power.
I The ASP study treated the Emergency Power System as an separate system in its functional event tree for a loss of offsite power.
On the contrary in the IREP Study, the function
{
of the Emergency Power System was considered an integral part of the success or failure of the related safety systems.
In this analysis we followed the ASP study approach.
Thus, in addition i
to modifications discussed at the beginning of this section, the IREP study event trees used for the loss of offsite oower events were modified further to include the Emergency Power System.
The IREP study event trees used for loss of offsite power are the event trees of figures 20 and 25.
J
-m.
- e'=
w*.
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q p.,.
4 p.__.
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=
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. _. _i _
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an i
.6
~
i i
i I
t 6
e t
6 t
- n i i
e t___.__..
i i
E em I
E I
f Figure 37:
IREP Study Millstone 1 event tree for LOOP modified to aoply in Category Al
e i
i l
i i
t 7
?
g L
R 'E F
a.
j.
__:. e _p g
w g
c P
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-CD CD CD O>
cD t
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Figure 38:
IREP Study Millstone 1 event tree for LOOP modified to apply in Category A2
-y9--
-*- 9m
,wn---
,w-w-,.,,, _,_,.-,-
,c,-
--9--e--
--ev+-w-a--9---
-iy
l 1
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V V.g Q.y O
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a
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l
-bg I
i ok ok CD o n.
- ok
._ 4
__.._i-
'sO M_
o n. t._
0%
~
ok cD
-cp t
- o k.
_. 6 CD
-CD Figure 39:
IREP Study Millstone 1 event tree for LOOP modified to aDolv in Cateogory A3 i
~
1
LeoP nrs EM P RCIc HPCI DEP CCWD C5 LPcr 5 B (.5 Toru 0; slo of 1; 9
- ?I;;;
0 0SIISIISIII%i ?Il&;)' R Esutr
.L.
}
Figure 40 :
IREP Study Browns Ferry 1 Event Tree for LOOP modified to apply in Categories B and C h
J k
_h l
E 6
g L.
H y
- g.,g g
I
.g.g
.a
-g - -- j g
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V oc w v s
=
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e p' O t.
O W.
6 CD ob 0F J~_ ~ ep c.D cp
-ok ng
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- C0 g
CD'.-
Figure 41:
IREP Study Millstone 1 event tree i.
for LOOP Inodified to ac..oly in i
l Category D l
I l
l 7
I g
'a I
i-l l.
t I
mova'c s e w a.ww._. s v
-r
% -~~, --- _,
O Small LOCA Event Trees All of the BWR small LOCA events in the ASP report are stuck open relief valve events.
In the IREP study for Millstone l' plant, the stuck open relief valve event was treated separately and the corresponding event tree was used in this analysis for the small LOCA initiators at categories A and D.
On the other hand, the IREP study for Browns Ferry 1 plant treated the stuck open relief valve event as part of the small steam line event.
The corresponding event tree was modified in this analysis by omitting the vapor suppression system availability since this system consists of a set of relief valves.
ones6.oses sa.sess ato so. e t Evacter nis t/mC &C AH5/ AJL 4 Lang Patential Seque nc e Coelant Maintained Newputte C5 Tore Se, ore be.
Accident Sebcritical Moquate ucsponse Care Care Meguate Cooling Deesse m.
I s..
2 se 3
s Ts3 4
Tes S
Yes 6
Figure 42:
ASP Event Tree for Loss of
- Coolant Accident i
l 1
4.
.. ~.-
--e
'w--
m
a-
4 T
S
&s a
S C
a I
L S
a a
B B
P R
P S
B S
Smasts 3
P C
P B
C C
S C
9 8
T T
T a
e W
P I
S C
C 5
I I
e
- a. a.
Ta a
e S
174 IdB Tg
175 Ima Tgn
!?4 82n & I T 88 5
!?? IAE Tg5
178 IAE Tgus 179 ECn L I T 888 5
les smE Tgz 101 IAE f EN
- =
5
~
182 ECn L I
. Ts tne a
103 Les TgKt 104 1AE T Kan S
185 ten L 3 T K888 S
~
F186 ten S 9 T mSF 3
L-let sen S : Tgscaro r---les ten 8 o Tgro Litt ECn S I T E88 S
190 Cn S 1 Tg4 LEB5me E EEE IEL.Eff er se Come a LT L sams <> se aminst I. se sonaTe CR e CSS RELT n e IEmmaTS <> 3 EntM) 3 e SELAr3D 34m e gnaastgIT ISBOCEB S
- SECRT (4 3 EDURS) sen - TuA som Cons asLT tac Sim? :=meCm SACa CSS n&T l
Figure 43:
IREP Study Millstone 1 Event Tree Safety Relief Valve (Inadvertent Opening) (T I S
X e Funsteen tenuse p3 E
SC8 BC8 Drem R
S S
O 8A g
C.ee Ossen Sise M Less Then S1S C.e es w 8%"A E
Seem fasse C e so w Se e,Sn es==
Caso essess n
a gas, sons:
SOR Agg Case consed I
Casees ed I
J N
Saee dnest WREOSA
- So a sea enes G
SOE X
esta esen I
Case seeded r
X Case eenses a
1 E
E Saoe seest 8CA ESA E
Case scenes u
R Case seeses SCORg a a
3 gene seest n
x C.e so as E
Cens seeses g
n 3
E
$8ee ensfl MRASgg l
SCM GSD X
X BbA Inest SCOE 3
a sera teen SS I
leMeWA tesst SSC 2
I ledBesa teelt IIISL 3105 Figure 44:
IREP Study Browns Ferry 1
(
Event Tree for small liquid-line or steam-line break (S).
i L
l l
\\
\\
i c.
- t Q~
J
.3 U
A a
P.
s g3 D
C h
U V
e P
C F
a s
v 0
e c
w f
ex ob i
c0
)
o c.
es cp ew ek
~
r
- c. 0 1
- CD cD l
cp Figure 45:
IREP Study Millstone 1 event tree for small LOCA modified to apply in
(
Categories Al and A2 i
t.
i
~ -.
e9-et, 9
9
,,.---s--,
m,.,-.-,
,___--2
m-----
~
J k
re J
e L
L i
g a
P P
S 1
h U
C C
D C
p P
C P
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I 5
C C
W i
ex 04-c.D I
D C.
es.
CD
.e5 Oh
~
.c0 os
~
CL CD
- C.O c0
(
i
- cD l
Figure 46:
IREP Study Millstone 1 Event Tree for small LOCA applied in Category A3 i
I' i
1, t
f 5
t w
e A
b M e*
f w
p Q
a a
ad H
m 8
- O M
u e
o E
o EO A
A m W a
w R W 0
uJ E
2 V
a m
e et ok CP ek ok 1-CD
~
~
oE Ok
~
~ GD CD CD i
E CD l
Figure 47:
IREP Study Browns Ferry 1 Event l
Tree for small LOCA modified to apply in Category B i
I l-I
?
i
.. - ~_ _ _. - _,. _
~.
h E f b
4 d
U m
u M
da e
3 4
8 4
a w
9 v
e E
O y
53 J
Q g
em cD
.m sp en so i
e s.
e s.
so
.m
.m j
-<a s>
s0 I
cD Figure 48:
IREP Study Browns Ferry 1 Event Tree for small LOCA modified to apply in Category C
.a '
-m su.
m
--,o
c N
5" L
d h
W U
L P
g 6
0 3
L W
c C
y C
0 y
S as A
E 2
c c
ev-etr.
C.D os c.O 05
<o l_
~
- c. o o,
-es to c0
_ co
(
c.o Figure 49:
IREP Study Millstone 1 Event Tree for small LOCA modified to apply to in Category D e
.. - -, + - - - - ~
+===w--
~~
=*
~w--*
w-
=
r' "W'-
F
-'t'-'
w p=
t Wt-'
-P
"-*"'r--
P 9"&s1~*r
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- - - " " - - - - * " - - - - " - - - - - ^ - - - - -
k
~
4.0 Numerical Analysis 3
4.1 Precursor Event Frequency and System Unavailability Data The category specific precursor event frequencies were es-timated according to the number of events and number of reactor years in each category.
The ASP report data was used to cal-culate failure probabilities.
If no system failures occured in a precursor belonging to a specific category while there were system failures in other categories, the 50% f i value for zero failures was used.
Frequencies for LOFW and MSLB initiators were not categorized, the former because LOFW events are not reportable in LER and therefore the ASP study did not have complete data, and the latter because there was no MSLB initiator in the ASP data for BWR's.
For both cases the ASP study point estimates were utilized for all categories.
Sufficient data was available to compute category specific unavailability for the Emergency Power System.
Likewise, isolation condenser unavailability was computed from ASP data.
Further-more, it was observed that the RCIC/HPCI unavailability estimation was fitted in the categories B and C which have these systems and i
this estimate was used.
Similarly, the ASP HPCI unavailability for LOCA estimation was used where'only HPCI unavailability is needed.
J.
All of the data for ADS unavailability estimation in the 1
ASP report corresponds to plants of Category C.
But the ASP estimate of 0.27/
is close to 0.3/
that both the Millstone-1 D
D
[
and Browns Ferry 1 IREP studies used.
Hence, the ASP estimate for ADS failure was used'for all categories.
For the rest of the l..
Safety Systems, there was no data in the ASP report.
The sources used for the corresponding unavailabilities are listed in Table 2.
L l
t a
=m
.+.
m 64
.yo-
In analyzing the precursors, no changes were made in the recovery factors of the ASP report.
It is beyond the scope of this analysis to calculate more' accurate recovery factors.
In actuality it was not desirable to change them, since one of the objectives of this analysis is to determine the impact of the plant specific calculation on the results of the ASP study.
Since the safety systems are grouped in the ASP event trees, the recovery factors were assigned to groups of systems.
In applying the recovery factors per safety system, care was taken so the end result, would be the same with the corresponding ASP recovery factor.
For example, in the ASP report for NSlc 153810 (Loss of feedwater event), the RCIC/HPCI system is failed with a recovery factor of.1 in the ASP report.
The actual occurence was:
HPCI was unavailable due to maintenance, the RCIC turbine trip was manually reset and them put into operation.
This in this analysis for this event, the recovery factor.1 was assianed to RCIC and factor of 1.0 to HPCI.
4.2 Frequency Calculations For each of the seven plant categories identified in this study, the generalized tree representing the LOCA, LOOP and LOFW events were modeled and discussed in Chapter III.
The generalized trees for the 21 cases considered in this study (7 categories x 3 event types) are presented in Table 2.
Subsequently, specific NSIC events were considered and the generalized trees and function data were modified to reflect the specific events that occured.
To cite one example illustrating this procedure, consider NSIC 106616.
To reflect this Category Al LOOP event, the category specific event tree is modified as follows:
o The initiator (the leading constant in the equation) is set to 0.5.
Since the initiating events was part of the precursor.
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Table 4 Reactor Years by Plant Category Category Reactor Years
% of Total Al 21.5 11.57 A2 21 11.31 i
A3 20.5 11.04 B
46.83 25.21 C
46.02 24.78 j
D 18.9 10.17 E
11.'0 5.92 TOTAL 185.75 i
L i
i h
o The HPCI function failure is set to 1 to represent failures, since HPCI failed in the precursor.
A total of 19 significant precursor events from the ASP study were considered,' yielding a total of more than 200 specific event trees for the 7 plant categories (Al, A2, A3, B, C, D, and E) and the 3 event types (LOCA, LOOP, and LOFW).
The specific event trees were then used to estimate the conditional probability of core damage and the total frequency of core damage (per reactor year) for BWR's only.
In addition, the trees were grouped to analyze the 7 plant categories separately and, as a final case, all trees were grouped to yield overall estimates of core damage.
Two techniques of weighting the core damage probability based upon the number of reactor years per plant category were examined.
In the first (referred to as Method I) the core damage condi+1onal probability for.each plant category i was weighted by-RYi /RY where RYi is the number of reactor years for that plant T
category and RY is the total nrmber of reactor years for all T
plants for which the NSIC event would occur.
The reactor years har plant category (for all plants in the category) is given in k
Table 4.
The-frequency of severe core damage was estimated by.
dividing the weighted conditional probability by the total of 185.75-BWR reactor years.
In the second techninue, (referred to as Method II), the frequency of severe core damage was directly estimated by dividing the conditional probability for each plant category i by RY ; the g
number of reactor years in that category.
In this case only precursors that actually happened in each category were considered.
O e
+ +,.
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... + -
.--ww
-g>a mer%,.
6
L TAs1.E 5 METMOD I RESULTS ASP ESTIMATE ASP ESTIMAft 4185C 9 CAT Al CAT A2 CAT A3 CAT 8 CAT C CAT D CAT E TOTAL Pecd FREQUENCY DIFFERENCE 61434 4.64s10~4 3.88al0~7 3.78s10~#
3.21:10~7 2.84ml0
4.32x10
2.58s10~6 9.03 10"*
8.8s10~3 4.73m10~3 5.24
~8
~
~I
-8
~
-2
-5 63129 4.12 10 3.81n10 3.71a10 '
3.21 1G" 2.84alC 4.15 10 5.14 10 '
5.80x10
l.8 10 9.69s10 16.7
~I
~I
~I 2.49a10"I 5.28s10"I 4.94x10
l.8m10 9.69 10
4.96 66996 6.08 10 5.94s10 S.73sto"I 2.18:10 2.40 10
~3 77916 6.08:10"I 5.94s!0"I 5.73a10'I 2.18:10 '
2.lon10'I 2.49sto'I 5.28mlo'I 4.94m10 '
2.1ml0 1.13:10
-4.37
~
~
79 % 5 4.12a10~
3.81 10 3.71x10 '
l.33 10 1.01:10 3.6s10' 9.53a10" 3.41x10
6.8s10
3.66s10 1.07
~
~0
-6 85546 4.12 10' 8.17 10 7.87m!0 4.54m10~
2.84u10' 4.15s10~8 9.9tul0~I 4.67a10 3.1m10 1.67m10
-2.80
~3 85738 7.71s!0 7.53m!0~8 7.19:10 3.95a10 3.88s10 7.82m10~8 1.08s10 '
9.14 10
3.4x10 1.83m10 2.00
~8
~8
~
~3
-5 101444 3.50m10 3.42m10 3.34 10 5.63s10 1.42 10 '
3.08:10 1.25:10 9.63a10 '
O.39 2.1 10~
2.18
-5
-5
~
-5
-0
~
~
~I
~
~I
~
~3
-5 103002 6.08s20"I 5.94al0 5.73sl0'I 3.44s10-5 8.94m10 '
4.02:10 5.26:10 '
5.0s 10 2.4x10~3 1.29 10
,3,94
~#
~8
~8
~8
-8
~8
~I 105540 6.08 10 5.94ml0~8 5.73s10 2.18:10~
2.10 10 2.49s10 5.28s10 4.94m!0 1.7m10 '
9.15:10
!.85
~
106616 9.26a10~
1.77s10
!.70s10 '
3.21 10 2.84x10 8.63a10~I 2.97ml0 1.45u10' 9.3x10 '
5.00 10
-2.90 l
~
~I
~
-0
~3
-5
~I 4.83s10'I S.68ato"I 2.67a10~'
!.83s10
5.13:10~8 8.99x10
9.25s10 ' 8 1.6 10 1.15 10 1.24 185870 9.64s1 0
!!6780 1.06al0" N/A 6.30 10"I 5.34ml0~I 5.3410"I N/A N/A 4.80x10 1.6x10 8.61m10-6 4,79
~0
~3
~
~8
~0
-6 '
~3
-5 120443 9.64a10 8.83s10'I S.61m10'I 2.67m10~
I.83s19 '
5.13:10 1.99s10 9.25 10 l.6m10 1.15 10 1.24 124222 9.64s10 8.83s10"I 8.6tx10"I 2.67x10
t.83s10 '
5.13 10 1.99s10 '
9.25 10 8 1.8x10 1.45s10~
1.24
~I
~
~8
~
~3
~
~
~8
~I
~I
~
-2
~
128 % 9 1.79a10~I 1.75s10 '
!.70m!0"I 2.18 10 '
9.70ml0 1.29:10 5.28ml0 3.46a10 '
l.4m!0 9.69 10 '
2.80
~8
~8
-5
~8
~0
-2 4
128906 7.75u10 7.53s!0 7.19:10~8 1.86s10' l.83a10 3.42 10 4.45 10 4.16 10~
2.77sto 7.54x10 1.81
-5
-2 149450 1.67s10" 7.87m10~8 7.18 10 4.21sl0"I 4.14:10 4.97m10 5.84a10
2.65m10 1.38s10 1.49s10
5.62
-8
~
~8
-8
~8
~8
~3
~3
-5 149961 3.85s10 3.76a20 3.60ml0 1.85s10~
1.82x10-5 3.36m10 4.43x10 4.13:10
,7.9 10 7.43 10 1.80
-5 153810 7.71m10' 7.53a10 7.19:10 3.95ml0 '
3.88s10 '
6.73a10 9.18sl0 8.82x10 '
- 1. % s10 g,77
~8
~8
~
~
~8
~I
~
~
~
~3
~
~3
~
2.12 TOTAL 7.11:10~
4.13 10" 3.94al0~3 7.05s10 '
1.97s10 '
3.35x10 1.68sL0 '
l.25 10 2.65x10
-3 958 Upper 1.Ms10 1.02 10 1.02:10 1.35 10
4.29 10 8.45s10 3.29a10
2.08 10 Sound 2.87x10' 5.53s10 '
l.93 Total Encludlag Broome Ferry a Factor of 0.75 != m applied in thlm table.
I l
,1
_. e
5.0 Results Table 5 summarizes the quantitative results obtained using Method I for weighting with respect to reactor years.
The re-4
[
sults for Method II are given in Table 6.
A comparison for the core damage frequency estimates by the two methods, I and II, shows that the category totals represent different types of estimates.
The category totals for Method.I represent fractional core damage contributions be added together to obtain an overall core damage frequency estimate.
The totals for Method II, however, represent an over-l-
all core damage' frequency estimate based upon the failure data for each category and thus are larger than the figures calculated in Method I.
This is mainly due to small number of reactor years associated with each category.
To estimate the upper bound for the core damage frequency 1
in Method I, the conditional core damace orobabilities were summed by category A 954 binomial confidence interval was then computed for each category using the probability sum and the "N" figure for the category as determined by the Maximus reduction, Method
(
).
The upper 95% confidence interval was then divided by 185.75 s.
to yield the upper confidence interval for core damage frequency.
An overall upper confidence interval was also determined.by further summing all of the category probability totals and determining an
{
overall N.
For Method II, the core damage frequency for each category was multiplied by the number of reactor years in the category to determine a conditional probability.
These probabilities were then used in conjunction with the N figures from the Maximus re-duction to calculate the binomial 954 upper confidence interval The upper interval fimtres were then divided by the number of reactor years in each category to return to a frequency estimate.
4 m.+.
,m.~
l i
The 95% confidence interval in Method I represent approx-imately a factor of two increase over the base core damage con-tributions.
Comparing the frequency estimates by category shows that Category B is the largest contributor to the overall core damage frequency estimate.
The event totals indicate that the EBrown's Ferry cable tray fire (NSIC 101444) is the largest con-tributing event.
Multiplying this event's frequency by 185.75 results in a conditional probability of.179.
This is approximately one-half of the.39 figure reported in the ASP study.
Generally, the analysis indicates that other precursors contributions are over-estimated by an average of a factor of two.
M TABLE 6
~
METHOD II RESULTS llSIC #
CAT Al CAT A2 CAT A3 CAT S CAT C CAT D CAT E
-5 61434 3.10:10 63129 1.46m10"
-5 6994 4.70 10 i
77916 4.54m10'
~3 i
79565 1.64x10 3
7.31x10'5 855M 85734 6.22 10~
101444 8.86x10 '
103002 5.42x10' 105540 3.42a10'I 10M16 4.62a10'5
!!5870 2.98x10-l 116780 4.88s10
120443 2.98s10~
124222 2.98x10~
3.43a10j 128569 124906 2.93x10 149450 1.34x10' 4
4 i
149968 2.91x10,
153810 6.22:10
P TOTAL 1.39 10 1.27s10
!.15:10 1.!!a10 1.46 10'3
~3
-2 E
~3
-2 952 Upper 1.62a10 1.52x10
!.18 10 1.21:10 I.79x10'
)
Sound 1
L L
[
t l
1 1
}
r
'. i '
APPENDIX A Precursor as they are applied on the Category Event Trees W-.
e.-=
=
w-*e%we e-.
.-w
j 4
1 7
Q A 9 3
L U
g W
w v
M o
g L
em J
d W
W V
d U
l I
t t
OK 4
(B 0 5 I
get iss S (9 6 E-4 4.42d g])
0 ot7 gn 5
Lo cA l o t ' '...
7 3 g-9 LD F
- t. g E - 6 cp
( psse 15133 } s.g%. 4 We 4.c RCIC =.4 T
I M?ct hitures Duriq.agwg.5 8 % % Femt ht t Q.t a. L L% C Mt asses A 1 e6 h2 1
P = 1. R 3 E -'t RESULT
% 3 L.u: mess Op.= No c.ogs a g mgs.E Ps F= ibm CD = CORE DnMAGg
+o h ssa ag+1.w.4 1.0 f*ce.: + L e.v 4 + 4 e = =
3 LOF W e#eu+.
- t ess e,u + m e
6 J
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,y 0027 p
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sa w q,ge-g cD
(
F l.3 E-6
- co
( NSic 14$50) $9oce. of lmterest.f c,r. Losg of Cre e.V mtfSt a 4 (*
Feedw-wr F flow at 03 e.s+o-C.hT EGosy h 3 P=l9E-4 N#
0 4 ppt, E s=se cS e>s We Cee G F = F=t l"re -
ep, cagg pgmn6E
- Tssuw 7 4s.*.4 lO f re.,3 +L ew & + % = +o
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REsui.r 5
S'** 53 Ok = No cogs o Amhs,E F = F. i lu re.
CD= CORE DAMA6g of l.O becer vle. eNo4 + free +c
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C ME coSM o P = S. 0% E -6 IE 8"'I s.s u..ss ev.= No cogs nam 468 g
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Result f
Op.= t4o cogs D A te.%G.E S = Lec sy l-CD = CORE D a m n o.E F = F=i tu re.
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1.1E-4 1.0 --
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Tee P = 2.772-2 (SC = 16)
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.nu.4 tw c-+. nun a t
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