NSD-NRC-96-4913, Forwards W Responses to NRC RAI Re in-vessel Retention of Molten Core Debris & Response to NRC Request Made at 960624-26 W/Nrc AP600 PRA Meeting

Forwards W Responses to NRC RAI Re in-vessel Retention of Molten Core Debris & Response to NRC Request Made at 960624-26 W/Nrc AP600 PRA Meeting

ML20132D863
Person / Time
Site:	05200003
Issue date:	12/13/1996
From:	Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NSD-NRC-96-4913, NUDOCS 9612200159
Download: ML20132D863 (52)

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l Westinghouse Energy Systems Box 355 Electric Corporation Pittsburgh Pennsylvania 15230 0355 NSD-NRC-96-4913 i

DCP/NRC0683 Docket No.: STN-52-003 December 13,1996 j

Document Control Desk U.S. Nuclear Regulatory Commission Washington, D. C., 20555 i

ATTENTION:

T.R. QUAY

SUBJECT:

AP600 RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION

Dear Mr. Quay:

Enclosure I provides Westinghouse responses to NRC requests for additional information pertaining to the in-vessel retention (IVR) of molten core debris. Specifically, the responses to RAls 480.440 through 440.461 are included in this enclosure. The RAls were transmitted to Westinghouse in a NRC letter dated November 7,1996. provides the response to a NRC request made at the June 24-26,1996 Westinghouse /NRC AP600 PRA meeting. The request was to provide a limited scope sensitivity study on the baseline PRA. This meeting open item is record number 3969 in the OITS.

The responses close, from a Westinghouse perspective, the addressed questions. The NRC technical staff should review these responses. The status of these RAIs and meeting open item will be changed to " Action N" in the OITS on January 2,1997.

A listing of the NRC requests for additional information responded to in this letter is contained in Attachment A.

Please contact Cynthia L. Ilaag on (412) 374-4277 if you have any questions concerning this transmittal.

G~

Brian A. McIntyre, Manager Advanced Plant Safety and Licensing i

f Enclosures, Attachment F D. D> S /

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190080 9612200159 961213 PDR ADOCK 05200003 A

PDR

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1 NSD-NRC-%-4913 DCP/NRC0683 i

Pag; -2 j

December 13,1996 cc:

J. Sebrosky, NRC (enclosures, attachment))

J. Flack, NRC (w/o enclosures) j J. Kudrick, NRC (w/o enclosures)

N. J. Liparulo, Westinghouse (w/o enclosures)

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1 Enclosure I to Westinghouse letter NSD-NRC-96-4913 l

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December 13,1996 l

2 1

4 3024 4

l NRC REQUEST FOR ADDITIONAL INFORMATION i

Question: 480.440 Uncertainties The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

Material Properties See responses to RAls 480.448,480.449,480.450 and 480.451.

Natural Convection lleat Transfer See responses to RAls 480.442,480.443,480.444, and 480.445.

Decay Power Uncertainties See responses to RAls 480.452,480.453, and 480.454.

Metal layer IIeat Transfer See responses to RAls 480.446, and 480.447.

?

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION jg tu

n Ouestion: 480.441 Debris Bed Configuration The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Recponse:

See responses to RAls 480.455,480.456, and 480.457.

480.441-1 W Westinghouse

~

NRC REQUEST FOR ADDITIONAL INFORMATION Question: 480.442 Molten Pool Natural Convection Heat Transfer 1

The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

1

Response

It is fundamental that correlation applicability be restricted to the same geometry. A correlation can rarely be applied to another geometry, but never without clear justification that this is appropriate. Also, it is a fundamental requirement that the range of the Rayleigh numbers in the experiments from which the correlation was derived spans the range of Rayleigh numbers in the intended application.

As explained in the IVR report, the mini-ACOPO data satisfy the first requirement, however, a small extrapolation is required to meet the second requirement. The ACOPO data, as discussed in Appendix V-2, meets or exceeds the second requirement. These are half-scale data with temperature differences of order 100 K, so beyond the scaled similarity, there is physical similarity too. The m!.ACOPO and the ACOPO experiments provide a clear test of scalability as there is a linear scale factor of 4 (i.e., volume scale up by about two orders of magnitude).

The Kymalainen data are 30% higher than the Steinberner-Reinecke correlation used in the report. By contrast, the mini-ACOPO data are in excellent agreement with it. The ACOPO data approaches somewhat lower values as the Rayleigh numbers reach and exceed the prototypical range. Also, it should be noted that the 30% departure in the Kymalainen data occurs at Rayleigh number of ~6x10", while below it the data agree with Steinberner-Reinecke.

As noted in the report, the Kymalainen data are suspect in this depanure and the ACOPO data confirm this suspicion in a definitive manner.

Regarding the sensitivity study requested, the impact has been calculated for both the base quantification and the extreme parametric case. As a result of the 30% increase in the Steinberner-Reinecke correlation, the upward fluxes increase by 11% in both cases. The consequent increase of the thermal loads in the metal layer region is 16% in both cases. For the base quantification, as seen from Figure 7.9, such an increase is negligible compared to the margins. For the extreme parametric cases, such an increase would cause the CliF to be exceeded by 10 to 20%

depending in whether the 10% increase in critical heat flux in a highly localized zone is taken into account. It is emphasized that the 30% increase postulated is refuted in a robust way by the ACOPO data and that the extreme parametric case is shown only as an example of what it takes to produce failure.

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION is!! !!Pt Question: 480.443 Molten Pool Natural Convection Heat Transfer The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

As noted in Chapter 5 the IVR report, the flux shape data from BMIT RS 48/l was utilizzed (reference 26 in the IVR report). The Rayleigh number for these data is 1.2x10"' The Jahn data shown in Mayinger's paper, which were used by INEL for Figure 3 has a Rayleigh number of 4x10' and are further removed from the range ofinterest, i

Ra' ~10" to 10'*. It should be noted that in the context in which this comparison was made in the IVR report, the deviations shown in Figure 3 have a trivial impact on the results.

As far as including the Kymalainen data in Figure 5.8 of the IVR report, the COPO geometry (torospherical)is very different from that of interest here (hemispherical) and including these data is inappropriate.

l i

1 W Westinghouse

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i NRC REQUEST FOR ADDITIONAL INFORMATION Question: 480.444 Molten Pool Natural Convection Heat Transfer The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

)

Response

The single point discrepancy in only three runs indicates an instrument error at the 40 degree location.

The ACOPO data have been published in the Park City, UT PSA '96 meeting proceedings. The paper is provided as Attachment 480.444-1 to this RAI response.

The sensitivity to a flatter flux shape can be conservatively answered for the base case (i.e., thick metal layer and oxidic pool) since for it the downward thermal loads are maximized. The result of Figure 7.10 apply and i

show the margins. The effect of any local changes in the flux shape can then be seen directly in relation to J

those margins.

For example, take Figure 11 of Attachment 480.444-1. Data can be found that exceeds the correlation by approximately 40 percent. Using this in Figure 7.10, the q(0)/qcur increases from 0.4 to 0.56. That is a local i

reduction in margin from 250% to 180%.

j T Westinghouse

Procc dings, PSA'96 Park City, UT, haud=- 29-October 3,1996 Vol. III, 1363-1350 ATTACHMENT 480.444-1 THE FIRST RESULTS FROM THE ACOPO EXPERIMENT T.G. Theofanous, M. Maguire, S. Angelini and T. Salmassi Center for Risk Studies and Safety University of California, Santa Barbara Santa Barbara, CA 9310G (805) 893-4900 ABSTRACT l

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The ACOPO experiment simulates natural convec-j tion heat transfer from volumetrically heated pools, at i

a half-scale reactor lower head geometry (hemispheri-

}

!8

'/

cal). Data for Rayleigh numbers of up to 2 10, from j

i the first round of experiments, are presented in this se paper. The results are in substantial agreement with those of the mini-ACOPO proof-of-concept experiment.

Moreover, it is shown that these ACOPO results con-OnY NO firm a key component of the in-vessel retention capabil-ity for an AP600-like design, as recently established in oudic DOE /ID-10460.

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I. INTRODUCTION Ms.1. Schematic of the inwenel retention geone The putpose of this paper is to present the first ex-try (Theofanous et al.,1995). The lower head is perimental data of natural convection heat transfer for externally cooled by boiling water.

the range of Rayleigh numbers,1025 - 10, directly rel-28 evant to a severe accident management concept known i

as "in-vessel retention" (IVR). The geometry is illus-e o4 2

trated in Figure 1, and involves a volumetrically heated oxidic pool, and a lower head that is externally sub-merged in water. For an AP600-like design, the di-y"

~

3 ameter is ~4 m, the decay power ~1.3 MW/m, and j

the distribution of Rayleigh numbers, accounting con-e o.:

servatively, for uncertainties, is as illustrated in Figure i

2 (Theofanous et al.,1995). By comparison, previous 5

data were limited to ~ 10" in the UCLA experiments yo (Asfia and Dhir,1996), to ~ 710" in the mini-ACOPO j

25 experiment (Theofanous and Liu,1995), and to ~ 10 j

for the COPO experiments (Kymsisinen et al.,1996);

s,io-so=

, io =

s,io=

io=

e,io=

that is, lacking by about one order of magnitude. For Ra' larger reactors there may be a need for almost another order of magnitude, to ~ 10". Besides this practical Fig. 2. The Ra' number distribution found in calcu-need, there are also some interesting fundamental ques-lations assessing the in-vessel retention concept for tions on the behavior as Ra' -+ ac.

an AP600-like design (Theofanous et al.,1995).

l 1

4 As explained in detail before (Theofanous et al.,

1995), the present practical need is to determine (a) the energy flow split between the upper (flat) and lower y

(hemispherical) boundaries, and (b) the shape factor

- =, -

along the hemispherical boundary, so that local heat ll e

flux conditions can be determined from the area-average 7

value. Also, it was explained that the problem is com-s=

"-8'"

pletely determined from the shape, and the isothermal

, ga 8-boundary conditions (due to the presence of crusts),

and that it is properly and completely scaled by the

[

Rayleigh number (Ra'), with the Pr number having only 8

o a minor independent efIect

/

T

~

5 W"

Nu = f(Ra')

Ra' =

(1)

/

u,o--

V

+

The long-standing difficulty in reaching the range of Ra' numbers of interest, experimentally is due to the

_3,,,,.

strong dependence on the length scale, and the diffi-culty of producing uniform volumetric heating at large Fig. 3. Schematic of the mini-ACOPO experiment, enough scales and hemispherical geometr:es. For exam-including the key construction details and instru-ple, with a radiation method (such as used in the UCLA mentation.

experiment) uniformity of power deposition requires a low-coupling system (" transparent fluid"), which really II. THE EXPERIMENTAL FACILITY limits the magnitade of power depositions and the pool superheating possible. On the other hand, for direct The ACOPO experiment is a large versica of the electrical heating, power uniformity requires a paral-mini-ACOPO, the basic design of which is illustrated lel electrode configuration (as in the slice geometry of in Figure 3. The figure shows the individual cooling COPO), which rules out the hemispherical geometry of units, the insulation between them, internal fluid tem-

)

interest.

perature measurement locations, and the expansion vol-The ACOPO idea bypasses this difficulty, by us-ume needed to accommodate the fluid during the tran-ing the internal energy of the fluid, preheated to some sient, while maintaining the vessel completely full. In high initial temperature, to simulate volumetric heat-the ACOPO, construction details were much more in-ing, by suddenly cooling the boundaries and interpret-volved. and actually building the facility proved to be a ing the transient system cooldown as a sequence of major challenge. Some perspectives of the sheer size of quasi-stationary natural convection states. That is, from the project are provided in Figures 4 through 8, which the local instantaneous fluxes at the boundaries, a total will also be used to explain its key components.

heat loss rate can be obtained to define the instanta-Starting with Figure 4, we can see the test vessel neous Rayleigh numbers, which then are correlated to (shown, in the photo, prior to insulation), the pump the instutaneous Nusselt numbers. The idea is that the and venturi racks, the heat sink tank, the tempera-cooldown would be arrested, and nothing would really ture instrumentation locations, and the data acquisi-change, if at any instant in time during the cooldown, tion and experiment control system. The heat sink is a volumetric heating rate could be supplied that was a large stainless steel cylindrical tank (2 m x 3.5 m),

equal to the then heat loss rate. The mini-ACOPO ex-loaded with ice (see Fig. 8), so as to maintain a con-periments confirmed that this idea actually works. The stant water temperature at ~0 "C. There are 15 cooling present experiments provide additional, definitive evi-units,10 on the lower and 5 on the upper boundaries, dence that this is so.

that constitute the vessel wall, as shown in Figure 7.

The mini-ACOPO test section has a diameter of 0.4 Unlike the mini-ACOPO, here each cooling unit is in-m (1/8 scale) and reached Ra' numbers of 7 10" and dependently fed by a respective pump (see Fig. 4),

3 1013, using Freon 113 and water as working fluids, whose speed is controlled such as to maintain the cool-respectively. The ACOPO test section has a diameter of ing unit operating at a near-optimum for the instru-2 m (1/2 scale) and with water it reached a Ra' number mentation. The object, as discussed in the next section, 18 of up to 2 10

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Figure V.4. The ACOPO half-scale facility.

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Figure V.5. The ACOPO test vessel lid in the Figure V.7. Schernatic of the ACOPO test ves-l final stages of polishing.

sel, showing the individual cooling units and the l

vessel support.

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l Figure V.6. The ACOPO test vessel tid being Figure V.8. Load ofice being transferred to the lowered upon the ACOPO test vessel.

ACOPO heat sink vessel prior to a run.

is to keep the walls as nearly isothermal as possible, air trapped, as bubbles, in the underside of the vessel and yet obtain a AT in each cooling unit that is large lid. The cooling circuits were then switched on, to ini-j enough to minimize measurement error.

tiate the cooldown, which was continued until measure-

)

The ACOPO test vesselis shown in Figures 5,6, and ment accuracy was lost, typically about I hour later.

7. Each cooling unit was manufactured separately by Data were recorded by a PC at a rate of 0.5 Hz, and welding together properly bent rings of square copper were reduced with an interfacing computer program us-tubing (1/2-inch on the side), so as to make an effec-ing a local smoothing routine before taking the time tively seamless internal surface. Within each cooling derivatives needed. All thermophysical properties were unit the rings could communicate, so that with a sin-evaluated at the -film" temperature, i.e., the average gle inlet and outlet, the flow would traverse through all value between the bulk and the wall. The energy bal-the rings. The whole vessel lower and upper parts were ance was well within the 10% error bounds, as shown in then built by putting together the cooling units, with Figure 9, and all data in fact were highly reproducible, special silicon rings between them for thermal insula-as shown below.

tion, on wooden supports, as shown in Figures 6 and 7.

The test vessel was well insulated on the outside, and IV. EXPERIMENTAL RESULTS special care was taken that it is not connected to any A total of five experiments have been run so far, in thermal masses that could introduce external heat flow the manner described above. A typical transient of the to the cooling units during operation.

Rayleigh number is shown in Figure 10, and a typical comparison of the heat flux shapes with the correlation III. MEASUREMENTS AND OPERATION obtained from mini-ACOPO is provided in Figure 11.

The data variation around the correlation in this figure As noted above, the key aspect of the operation is is also typical of what was found in mini-ACOPO; i.e.,

in regards to balancing measurement accuracy against the correlation represents a fair representation through the required condition for isothermal boundaries of the the middle of the data.

test vessel. This was resolved as follows. With a max-The upward heat transfer from Run 5/28/96 is com-imum fluid-to-wall temperature difference of the order pared to the Steinberner-Reineke (1978) correlation in of ~100 K, it was decided that the isothermal condi-Figure 12. The trend of the data veering off the corre-tion would be satisSed well enough if the cooling units lation for Ra' > 1013 was already slightly evident in the operated, inlet-to-outlet, within a few degrees K. This mini-ACOPO data, but it is quite clear now with the then led to a requirement for measuring this tempera-range extension by more than one order of magnitude.

ture difference with an accuracy of better than 0.1 K.

In this upper range of 10 < Ra' < 10", the data 15 For this purpose, we chose thermistors, with a quoted seem to indicate a Rayleigh number exponent near 0.2.

accuracy of t0.1 K. The bulk fluid temperatures were This is the highly turbulent regime, and there has been measured with chromel-alumel thermocouples to an ac-some question of whether it should tend asymptotically curacy of =1 K. Thermocouples, thermistors, as well to 0.2 or 0.25 (see Chapter 5, and the section on Natural as the venturis used for flow rate measurement, were Convection in Appendix U, of DOE /ID-104GO). By a calibrated in situ, using the complete data acquisition Ra' number of 10" the deviation from the Steinberner-system, and were found to perform very stably through-Reineke correlation is already significant. As shown in out this first experimental campaign. As shown from a Figure 12, the data from Run 5/28/96 can be well cor-typical energy balance in Figure 9, the overall accuracy related by is much better than 10%, which for an experiment of Nu = 1.95 Ra' "

(2) this size is deemed quite satisfactory.

which is shown in relation to all ACOPO data in Figure A run was begun by heating the vessel contents, to some high temperature near 95 C, very slowly, by

13. An essentially tight bound of 10% is observed.

The downward heat transfer data from ACOPO Run recirculating the contents through an external heater.

The water level was then adjusted to a few centimeters 5/28/96 is shown in comparison to the hiayinger (1975) below the top lid, and steam was injected into the free-and mini-ACOPO (Theofanous and Liu,1995) correla-board volume while also allowing for an exhaust, until tions in Figure 14. In the latter case, the extension the temperatures in this upper region reached 100 *C.

f the lower branch of the correlation, representing the This freeboard volume was then isolated, and immedi-water data obtained in the range 10 2< Ra < 4 10 3, ately connected to the expansion tank, thus allowing is used. It is seen that in the upper range both correla-this volume to fill, by the draining, of degassed,100 'C tions and the data come together to a close agreement.

water. This procedure ensured that there would be no The upper branch of the mini-ACOPO correlation, ob-3 tained with Freon 113, and extending from ~ 3 10 to

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1.4 Re. $/2856

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• 0.SeSRa* ***

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a

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1000

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Cr 1

N-i 1

g-

~

o.,

s00.,,4-0.6 Re. 5/28n6 0

1000 2000 3000 jois 10

10

10

w. t.)

Ra' Fig.12. Upward heat transfer from ACOPO Run 5/28/96 compared to the Steinberner-Reineke cor-Fig. 9.., he overall energy balance for Run 5/,28/96.

relation. The dotted line shows the i10% margins Q. is the total heating rate of the cooling circuits'.

on the correlation. The solld line shows the present and Q, la the total cooling rate of the vessel contents data St.

10'#

i i

R 5/2s/96 AM Emperis.eetaa R a

-o" 1000 10

i r

j F

500 e AcoPo Das nos scam 10 -

o Acoro D 1.95Ra* "

i 10 :O 1

i e

i i

i i

1000 2000 3000 10"

<0 10*

5 he (s)

R' Fig.13. Upward heat transfer from all Sve ACOPO I

Fig.10. The Rayleigh number transient in ACOPO runs. Data shown every 200 s, for clarity. The full Run 5/28/96*

points are from Run 5/28/96, and the solid line rep-resents the Rt to these data. The dotted lines show the 10% margins on the correlation.

O ACUPO Dea 0.DeRRs*#

2 7

,,,. co,o 1000-

• * * - 0.55as* 8 ACOPO Dua j#*

N

,4 4

i :.

a x

.. *.... ~

l o

O a

R., ms,,.

R.. ms, t

.e.

9 j

0 20 40 60 80 10c

g 10 10

10 s 10

l Ansi. (e-)

R.*

Fig.11. The heat aux distribution along the lower Fig.14. Downward heat transfer from ACOPO boundary in ACOPO Run 5/28/96 compared with Run 5/28/96 compared to the Mayinger correlation the correlation obtained from mini-ACOPO. Data

(- - -), and to the extension of the lower branch of shown only every 600 s (for clarity), for the duration the mini-ACOPO correlation (-). The dotted lines of the run.

show the 10% margins on the correlation.

t I

l

~

.-. =. -

I E

I AB 4J

' Rome while based on Eqs. (2) and (3), we have 1000-NUup.n = 6.5 Ra'~ "

(8)

R' = Nuan,n n

• 4 Now the heat flux ratios ofinterest can be obtained (see f

Section 5.1 of DOE /ID-10460) from.

. woroo a snim qup.n,1 + 2/R',

o xxwoo oaan.a a qup.o 1 + 2/%

O io i o's i o's and u

Qd"*"

1 + 0.5 R6 Ra' (10)

=

qdn,o 1 + 0.5 R'n Fig. 15. Downward heat transfer frorn all five i

ACOPO runs compared to correlation (3). Data The results, for Rayleigh numbers bounding the reg. ion f interest, are summarized in Tables 1 and 2. It can be shown every 200 s, for clarity. The dotted lines show the 10% margins on the correlation.

seen that in the previous results the upward flux was previously underestimated by ~10%, while the down-7 10", exhibits a somewhat steeper slope. This matter ward flux was overestimated by less than ~8%. These is under investigation. A fit to the data from Run 1 variations are negligible m the context of the analysts,

-yields and the margins to failure found in DOE /ID-10460.

Nuan = 0.3 Ra' '"

(3) and it is shown in relation to all ACOPO data in Figure Table 1. Illustration of the Variation in the Heat

15. An essentially 10% tight fit over the whole range of Flux Ratio, R', as a Result of the New Correla.

Ra' la observed.

tion Basis t

V. DISCUSSION Ra' R'

R"'

The in-vessel retention analysis for the AP600 noted above (Theofanous et al.,1995) was based on the Stein-berner-Reineke and the mini-ACOPO (upper branch) iots 1.59 1.63 i

correlations for the upward and downward heat trans-J fer, respectively. They are given by 5 1015 1.32 1.53 "uP = 0.345 Ra'""

(4) 1018 1.22 1.49 Nuan = 0.0038 Ra' '"

(5) although the Mayinger correlation Table 2. Bounding Values of the Effect of the New Correlation Basis on In-Vessel Retention Nudn = 0.55 Ra'"

(6) in an AP600-Like Design j

was also utilized in sensitivity analysis. It is interesting, therefore, to consider how the new results, obtained Ra' go,.n /q,,.o gan.n /qan.,

directly on the prototypic range of Rayleigh numbers (see Fig. 2), might affect the conclusions.

Given the agreement on the flux shape, it is suffi.

10!5 1.01 0.99 cient for this purpose to consider the average heat fluxes in the upward and downward directions. Let us denote 5 1015 1.09 0.94 their ratio by R', and with subscripts "o" and "n" the "old" and "new" results respectively. That is, from Eqs.

1018 1.12 0.92 (4) and (5), we have Nu g _ Nu

- 90.7 Ra""

(7) op.,

an.,

e

3 y-j VI. CONCLUSIONS The authors also wish to express their appreciation to Dr. C. Liu for his participation in the design of ACOPO, The first round of experiments from the ACOPO and to Messrs. Al Khamseh, Richard Becker and God-j e

l facility confirm the experimental concept, and ex-frey Nairn, for their essential contribution in the con-l tend the mini-ACOPO results, to fully cover the struction of the ACOPO test vessel.

j prototypic range of Rayleigh numbers of current interest to in-vessel retention REFERENCES 4

Some variations from the extensions of previous e

correlations are found, but they are mainly of a

1. F.S. Asfia and V.K. Dhir, "An experimental study j

detailed fundamental interest. The net impact on of natural convection in a volumetrically heated j

the assessment of in-vessel retention is less than spherical pool bounded on top with a rigid wall,"

10%.

Nuclear Engineering and Design 1996 (in press).

)

f

2. O. Kymsisinen, H. Tuomisto and T.G. Theofanous, NOMENCLATURE "In-vessel retention of corium at the Loviisa plant,"

Nuclear Engineering and Design 1996 (in press).

g acceleration of gravity H

depth of pool

3. F. Mayinger, M. Jahn, H. Reineke and U. Stein-Nu Nusselt numbers (qH)/k(Tmu - T.)

berner, " Examination of Thenno-hydraulic Pro-q average heat flux at pool boundaries cesses and Heat Transfer m a Core.Stelt," Fmal I

4 volumetric heat generate rate Report BMFT RS 48/1. Technical University, Ra' Rayleigh number, m, ternal = (gd4H /(kva)

Hannover, W. Germany,1975. As reviewed by 5

F.A. Kulacki, Ohio State University, for the US T

temperature NRC, March 31,1976.

Greek

4. U. Steinberner and H.-H. Reineke, " Turbulent j

a thermal diffusivity buoyancy convection heat transfer with internal thermal expansion coefficient heat sources," Proceedings Sixth International v

kinematic viscosity Heat Transfer Conference, Toronto, Canada, Au-l 4

j gust 1978.

I i

Subscripts dn downward (over the hemispherical boundary)

5. T.G. Theofanous and C. Liu, " Natural Convec-n new tion Experiments in a Hemisphere with Rayleigh j

15 o

old Numbers up to 10," Proceedings,1995 ANS Na-

~

up upward (over the flat boundary) tional Heat Transfer Conference, Portland, Ore-l w

wall value gon, August 5 9, 1995, 349-365.

J j

6. T.G. Theofanous, C. Liu, S. Additon, S. Angelini, i

O. Kymsisinen and T. Salmassi, "In-Vessel Coola-J bility and Retention of a Core Melt," DOE /ID-

]

ACKNOWLEDGEMENTS 10460, Vols. I and 2, July 1995.

4 Support from DOE's ARSAP program, and of the l

program's Project Manager, Mr. S. Sorrell (DOE, Idaho i

Operations Office), are gratefully acknowledged.

4 i

1 3

3 i

I 1

1

)

NRC REQUEST FOR ADDITIONAL INFORMATION h

Question: 480.445 Molten Pool Natural Convection Heat Transfer The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

Low volatized material cannot exist in a superheated oxidic pool. Any trapped quantities will be released as vapor and/or liquid masses as the pool crosses its solid to liquid transition region. During this period, the thermal loads are negligible as the decay heat is melting oxide and vaporizing trapped metal and hence the impact of the increased convection is negligible.

480.445-1 W Westilighouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question: 480.446 Molten Metal Layer Heat Transfer The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

The ABAQUS 5.4 code was used for the two-dimensional heat transfer modeling.

The vessel dimensions and the heat fluxes used in Figure 5.15 are discussed on Page 5-25 of the IVR Report.

The heat fluxes are from the extreme parametric case (Figure 7.16 of the IVR report) and the wall shape is adjusted to obtain an isothermal boundary at the specified melting point of 1600*K.

At 70* inclination, due to the inclination of the gravity vector, the effective gravitational force would be approximately 6% lower, and the effect on the heat transfer would be approximately 1%.

The analyses were performed for the purpose of demonstrating that 2-dimensional conduction can dissipate highly localized hot spots. So a hot spot on the inside surface of the wall dissipates as heat is transferred through the wall to the outside surface which is cooled by water.

l 480 m T westinghouse i

f l

NRC REQUEST FOR ADDITIONAL INFORMATION b

Question: 480.447 Molten Metal Layer Heat Transfer The question can be found as Enclosure I in NRC letter to Westinghouse dated November 7,1996.

Response

The Globe-Dropkin correlation involves heat transfer through a bottom-heated layer and provides the Nusselt number in terms of the external Rayleigh nurnber. As explained in Chapter 5 of the IVR report, this correlation is widely supported by many different works.

The Kulacki-Emura correlation (the lower line in Figure 5 of the INEL question) was obtained in a 2-layer system with the bottom layer volumetrically heated. The Nusselt number is provided in terms of the internal Rayleigh number, which was defined to involve the heights of both layers.

Based on the above, it is not appropriate to compare the results of the Gl;' Dropkin correlation with the Kulacki-Emura correlation, or to use the Kulacki-Emura correlation for our system.

l

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l l

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1. '"7d T Westinghouse

s NRC REQUEST FOR ADDITIONAL INFORMATION J

Question: 480.448 Material Properties The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

To reach a liquidus below the value used in the IVR report (1300 C), the zirconium mass fraction would have to be in the range of 55 to 85%. For any other concentration the value used is conservative. In Appendix P and Chapter 7 of the IVR report, it is argued that the expected compositions would be in the iron-rich region (Zr mass fraction less than 50%). This is due to the unavoidable addition of the radial reflector (~40 tons of steel) to the melt.

The reflector slumps into the melt as the support plate is subsumed by the oxide pool.

The SCDAP/RELAP5 results do not provide a better basis for assessing the liquidus. For example, tnese calculations do not take into account the reflector which would be subsumed into the melt. Even without accounting for the reflector, the 50% Zr mass fraction quoted from SCDAP/RELAPS analysis does not produce a lower liquidus value than used in the IVR report. For 50% Zr mass fraction, the appropriate value is 1500 C. Also, there can be no unmixed regions to produce a lower liquidus, as the convection that drives the heat transfer to the vessel walls to produce the thermal loads also drives the mixing in the metal layer.

The response to Olander's comment is quoted out of context in the RAI. It could not be argued with confidence that the metal layer composition is away from the iron-rich eutectic (see Figure 6.1) to justify using a higher liquidus value than the conservative bound that was chosen.

By comparison to the above, uncertainties of 210% in conductivity and CHF pale. The effective conductivity value used in the report is 32 W/K/m which is properly obtained from Figure L.3 over the range 130 C to 1300*C. The CHF values used are the lower bound of a very tight correlation. The uncertainty in the CHF is below the normal 220% because of advantages in the experimental facility that allowed us to zero-in on the CHF value by successive runs.

Therefore, the vessel thickness used in the IVR report is appropriate. No further finite element analysis are required.

l i

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W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

.am W*.

~I Question: 480.449 Material Properties The question can be found as Enclosure I in NRC letter to Westinghouse dated November 7,1996.

Response

The mehing point for the wall was chosen conservatively as discussed in RAI 480.448. The oxidic melting point uncertainty is not important, because the superheat is what drives the thermal loads. Similarly, the emissivity is considered an intangible in the ROAAM analysis and was quantified conservatively at the lowest possible value.

Impact of variations in emissivity were covered by the parametric results in the report and also in responding to the expert's comments. Uncertainties in density and specific heat reflect negligibly on the top-to-bottom split of heat fluxes because they are reflected through the Ra' number dependency to the 0.117 power (see equation 5.31).

T Westinghouse

l j

NRC REQUEST FOR ADDITIONAL INFORMATION y

ci j

Ouestion: 480.450 Material Properties j

The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response-Uncertainties in the parameters associated with composition and temperature are addressed in the response to RAI 480.449. The thermal conductivity of the crust does not affect the peak heat loads. Rather, the heat load affects the thickness of the crusts. Thus, the uncertainties in the conductivity of the ceramic crust has no impact on the analysis. The same applies to gap conductance between the crust and the vessel wall. No discrepancies exist between Appendix L and Table 7.1, and therefore, no revisions are appropriate.

i W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question: 480.451 Material Properties The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

The difference in the thermal conductivity between steel and stainless steel is a solid state effect. For example, see variations with minute quantities of carbon in Figure L.4 of the IVR report. In the liquid state, for which this question applies, the thermal conductivities are the same.

)

" "I'I T Westinghouse

l NRC REQUEST FOR ADDITIONAL INFORMATION 1

Hm

+

g Question: 480.452 Decay Heat Assumptions The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

A 10% change in the decay power would produce a 10% change in the thermal loading, both downward and in the metal layer. However, the decay power curve (Figure 7.1) used in the IVR report contains both a reasonable estimate of the decay heat including the uncertainty (see table below) and a conservative treatment of the loss of volatile fission products. No sensitivity analysis is required.

ANS 1979 ANS 1979 IVR Report Time after Shutdown Best Estimate

+20 (Figure 7.1)

(seconds)

(WM)

(MW)

(MW) 200 55.9 58.0 62 500 46.8 48.5 48 1000 40.1 41.5 40 2000 33.2 34.3 33 3000 29.3 30.3 29 5000 25.0 25.9 2$

10000 20.4 21.I 21 20000 16.9 17.5 17 40000 14.0 14.5 14 480.452-1 t

T Westinghouse i

6 NRC REQUEST FOR ADDITIONAL INFORMATION f

Question: 480.453 Decay Heat Assumptions The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

A significantly sized circulating pool in the lower plenum cannot occur before 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Although the code results 4

quoted in the RAI do not appear to consider the impact of the core radial reflector and the time it takes the melt to penetrate it, the SCDAP/RELAPS calculation shows the relocation beginning at 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. There will be at least another hour to reach the fmal bounding state (as described in Appendix 0). Similarly, it will take additional time after the depletion of the water in the lower plenum (the 2.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> quoted) to reach the melt superheating needed to produce significant thermal loads.

The MAAP4 calculations quoted in the RAI were not oriented to the purpose of melt relocation timing as they also do not include the effect of the reflector on the melt progression. The estimates of the minimal times to the fmal bounding state in the IVR report (Appendix 0) are more reliable than the code results because the melt progression is carefully decomposed and includes the important effect of the reflector. See also the response to RAI 480.105.

480.453-1 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question: 480.454 Decay Heat Assumptions The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

Estimation of the decay power is one of the most robust aspects of the IVR ROAAM. There is no room for arbitrary variations. With proper consideration of the event timing (see RAls 480.105 and 480.453), conservative hand calculations of the volumetric heating in the oxide pool can produce values up to 1.4 MW/m'. To produce a value as high as 1.7 MW/m', it must be assumed that no oxidation of the zirconium cladding occurs. This is not a physically reasonable assumption. If SCDAP calculations can provide a basis for higher values, it must be assessed how these results are obtained and in which time frame they are applied. It is critical in this regard that the SCDAP calculation did not consider the core radial reflector and phenomenology and timing associated with its melt-through followed by core barrel melt-through.

480.4544 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION h

I Question: 480.455 Assumed " Bounding" End-State Condition The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

Configuration A It appears that the reDector was not modeled in the SCDAP calculation referred to in the RAl. It is physically impossible to relocate 85% of the core without melting a significant portion of the reflector. This is explained in the IVR report and in more detail in Appendix 0. If the SCDAP calculation is used to propose a configuration, then the calculation basis needs to be supplied, and more details on the results provided so the scenario can be constructed with the necessary details.

Configuration B The geometry and scenario are not sufficiently w l'-defined to perform calculations. For example, if the second metal layer is described as thin, where is the material of the reflector and core barrel? As in the response to Configuration A, a consistent scenario must be established prior to meamngful quantification of consequences.

1 I

l 480.455-1 W Westinghouse r

NRC REQUEST FOR ADDITIONAL INFORMATION Question: 480.456 Assumed " Bounding" End-State Condition The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

I

Response

The inclusion of the upper internals metal to the molten metal pool is considered to be highly unlikely (page 7-5 of the IVR report) and is quantified as such. The sources of metal mass are specifically outlined on page 7-5.

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION E!?

4

~t Question: 480.457 Assumed " Bounding" End-State Condition The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

As seen in Tur6 in the IVR Report Turland is satisfied on this point. Natural convection flows would naturally promote rapid segregation of metallic components on the top, and there are no sufficient forces or mechanisms to re-entrain in the wall boundary layers.

480.4s7a W Westinghouse

l 1

NRC REQUEST FOR ADDITIONAL INFORMATION Question: 480.458 Heat Addition Due to Chemical Interactions The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

a.

The water pool is not considered a source for oxidation. The water is completely vaporized before a significantly sized molten pool exists.

b.

The primary coolant volume of AP600 is approximately 200 m' and at fully depressurized conditions corresponds to a water volume of approximately 0.1 m'(100 kg). This mass of water could oxidize only 200 kg of Fe. Also note that the frothy oxide layer mentioned in the INEL comments would provide a self-limiting mechanism to oxidation. See also closing comments to Olander in Appendix V-1 of the IVR report.

The conditions stated in the RAI. oxidation on the surface of the metal pool and reduced heat transfer from the c.

upper layer due to an oxidic crust are mutually exclusive. The presence of an adiabatic oxidic crust at the top l

of the pool would prevent oxidation of the metal layer below by the steam above. Each of the conditions is assessed separately in the IVR report. The case with the adiabatic oxide layer at the top of the metal pool is presented in Figure 7.13. An assessment of Olander's metal layer oxidation scenario on page T-69 of the IVR report. Neither of the cases produces failure of the reactor vessel.

1 i

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480.458-1 W Westinghouse

1 NRC REQUEST FOR ADDITIONAL INFORMATION l

Question: 480.459 Radiative Boundary Condition on the Upper Surface of the Melt The question can be found as Enclosure ? in NRC letter to Westinghouse dated November 7,1996.

Response

The effect of radation heat transfer from steam or air at low pressure (density) is negligible. Aerosols were not a.

considered, because under strong natural circulation in the gas atmosphere and the well ooled side walls of the reactor vessel, any aerosols generated would be quickly depleted from the atmosphere by deposition. The j

sensitivity studies in Chapter 7 of the IVR report include a case with a perfectly insulated upper metal layer 1

boundary. The sensitivity case conservatively addresses this question.

b.

Specific aerosol behavior calculations were not performed. The sensitivity studies in Chapter 7 of the IVR report include a case with a perfectly insulated upper metal layer boundary. The sensitivity case conservatively addresses this question.

The 6, and 6, are the core barrel (2 inches) and vessel wall thicknesses (8 inches) respectively. The S, is the c.

side wall (cylindrical) vessel area (57.4 m') above the melt. These are used in both the base and extreme parametric cases.

3 Westinghouse

=

NRC REQUEST FOR ADDITIONAL INFORMATION Question: 480.460 Structural Analyses The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

Approximate Solution The approximate solution was provided for understanding of the key features of the problem. For the exact a.

solution and end-effects in the ablated area, see the finite element solutions provided in Chapter 4 of the IVR report.

b.

As noted for case l APC, the decay power is so low that it presents no significant challenge to the vessel wall.

For case 3DC, the pressure difference is so low as to produce a negligible effect. To get a general perspective on the pressures that can be accommodated see Appendix G of the IVR report. Conservatively, the AP600 PRA, revision 8 does not credit IVR success for any accident sequences pressurized above 1.0 MPa. The accident classes are evaluated for pressure and the vessel is assumed to fail, regardless of the success of external reactor vessel cooling, if the internal vessel pressure is greater than 1.0 MPa.

The point intended by the specine paragraph quoted in the RAI is that under thermal stresses, an clastic-perfectly c.

i plastic assumption is grossly conservative.

d.

A temperature-dependent yield stress for SA106B is used as explained on page 4-4. The use of SA106B is

)

conservative as it results in less core material holding the loads than would SA533Bl.

e.

The value used in the report was the minimum possible for iron-rich melts, and thus gives the minimum wall thickness. See also RAI 480.448.

f.

Under thermal stress loading, the position in the IVR report is correct. See Shewmon's comments on page T-144. The margins are large.

Finite Element Analysis g.

Conditions of the analysis are for a freely hung vessel (from the top). As stated in page 4-7 of the IVR report, the model is loaded with the hydrostatic pressure distribution - both inside and outside, but their effect is negligible compared to the thermal stresses. Inner surface temperature is obtained from bounding calculations (see Appendix Q). Mechanical properties are for SA106B steel. Change in clastic properties of the material with temperature is included, but perfect plasticity is conservatively assumed (see page 4-7).

h.

The purpose of Appendix G is to provide perspective. Pressures of 400 psi are of no interest to IVR. The AP600 PRA, revision 8 does not credit IVR for pressures beyond 1.0 MPa (150 psi).

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION u;

eu 9

i. The basis of the INEL analysis was refuted above. The 2.5 cm is the thickness according to the extreme parametric sensitivity, and it was used in the detailed structural analysis presented in the Addendum to Chapter

4. The pool dead load was included and found to be of negligible effect. The vessel dead weight is also negligible.

j. In the Addendum to Chapter 4, the maximum principle strain reaches values of 7% and 18% at the outer and inner layers respectively. This strain level presents no threat to the structural integrity of the vessel. The material properties are temperature dependent.

k.

As noted in the report, part of the reactor vessel wall material specification is that it can withstand an accidental Hooding and still remain within operation. This means that material damage during such a cooldown event is negligible.

l 1.

The RTNDT is specified as 23 F maximum at end of life.

480.460-2 W Westingh0Use

=

NRC REQUEST FOR ADDITIONAL INFORMATION Question: 480.461 Typo / Errors The question can be found as Enclosure 1 in NRC letter to Westinghouse dated November 7,1996.

Response

a. - f.

These typographical errors are noted.

g.

For our purposes we are interested in the estimates, rather than the actual TMI wall temperatures, which were affected by an additional cooling mechanism. The caption should be clarified if the report is revised.

i s

t T westinghouse

Q O to Westinghouse Letter NSD-NRC-96-4913 December 13, 1996 i

3tM4A

1 t

1 RESPONSE TO MEETING OPEN ITEM 1

l Questbn: (#3969)

It was agreed during a Westinghouse /NRC PRA meeting on June 24-26,1996 that Westinghouse would do a lirnited scope sensitivity analysis on the baseline PRA to address the staff's concerns on focused PRA. The limited scope sensitivity analysis would keep the following systems, unless the initiating event caused them to fail: main feedwater, condensate, AC power, plant control, non-lE DC power, circulating water, main steam, chilled water, turbine building closed cooling water, component cooling water, service water, and instrument air. Westinghouse would not take credit for the following systems: chemical and volume control, startup feedwater, normal residual heat removal, diverse actuation, and the diesel generators. It was also agreed that this sensitivity analysis would be done with the Revision 7 baseline at power analysis cutsets and that the results would represent a good approximation of the actual number. Westinghouse will submit the results of this analysis to the staff for review.

Response

l As requested by the NRC at the June 24-26, 1996 meeting, Westinghouse has performed the sensitivity study as defined above. The sensitivity of the AP600 core damage frequency for internal events at power to the unavailability of five standby nonsafety-rehted systems / subsystems was studied. These five systems are: CVS, SFW, RNS, DAS, l

and the diesel generators. For the sensitivity study, these nonsafety-related systems are assumed to be unavailable in response to a reactor trip or a demand for one.

The core damage frequency for this sensitivity analysis is 4.4E-06 events per year. This increase from the baseline PRA (whose core damage frequency is 1.7E-07 events per year) is primarily due to the unavailability of DAS. The contribution of initiating event categories to the core damage frequency is summarized in Table 1. According to this table, transients (with hW available), SGTR, ATWS, and loss of main feedwater a: the main contributors to plant core damage frequency. The top 200 core damage cutsets are provided in Table 2.

As reported in Chapter 52 of the AP600 PRA, the focused PRA at-power core damage frequency is 7.7E-06 events per year.

Common cause failure of I&C software, totally failing both PMS and PLS, is a dominant contributor to the core deage frequency of this sensitivity case. This event's contribution to core damage frequency is conservative. The PLS and PMS functions nre different, and it is expected that the software used for those systems will be sufficiently different that common cause failure of the software will be smaller than is currently represented in the PRA.

Another insight evident from this sensitivity study, and the focused PRA sensitivity study, is that the AP600 design can meet the NRC safety goal without the defense in depth that is provided by the nonsafety-related systems.

PRA Revision: None.

3 Westinghouse

RESPONSE TO MEETING OPEN ITEM M

TABLEI SENSITIVITY STUDY -- CONTRIBUTION OF INITIATING EVENT CATEGORIES TO CORE DAMAGE FREQUENCY Initiating IMPORTANCE NUMBER OF CONTRIBUTION IEV Event (IEV)

(% DECREASE)

CUTSETS TO CDF FREQUENCY 1 IEV-TRANS 39.47 328 1.72E-06 1.40E+00 l

2 IEV-SGTR 15.44 139 6.73E-07 5.20E-03 3 IEV-ATWS 10.65 55 4.64E-07 4.81E-01 4 IEV-LMFW 9.79 114 4.27E-07 3.35E-01 5 IEV-LMFW1 5.39 105 2.35E-07 1.92E-01 6 IEV-LCOND 4.42 61 1.92E-07 1.12E-01 7 IEV-LCCW 4.23 140 1.84E-07 1.44E-01 8 IEV-NLOCA 2.95 747 1.28E-07 7.70E-04 9 IEV-LLOCA 1.15 616 5.02E-08 1.05E-04 10 IEV-LOSP

.95 77 4.13E-08 1.20E-01 11 IEV-SI-LB

.94 188 4.11E-08 1.04E-04 12 IEV-SLB-V

.62 60 2.69E-08 1.21E-03 13 IEV-MLOCA

.61 211 2.64E-08 1.62E-04 14 IEV-LRCS

.50 23 2.19E-08 1.80E-02 15 IEV-ATW-S

.47 32 2.06E-08 2.05E-02 16 IEV-SLOCA

.38 379 1.67E-08 1.01E-04 17 IEV-CMTLB

.34 210 1.49E-08 8.94E-05 18 IEV-PRSTR

.31 132 1.35E-08 2.50E-04 l

19 IEV-POWEX

.29 108 1.28E-08 4.50E-03 20 IEV-SLB-D

.28 7

1.24E-08 5.96E-04 21 IEV-RV-RP

.23 1

1.00E-08 1.00E-08 22 IEV-RCSLKC

.19 27 8.25E-09 5.02E-05 23 IEV-SLB-U

.18 46 8.06E-09 3.72E-04 24 IEV-LCAS

.17 69 7.45E-09 3.48E-02 25 IEV-ATW-T

.03 5

1.41E-09 1.17E+00 26 IEV-ISLOC

.00 1

5.00E-11 5.00E-11 l

2 3 Westinghouse

i e.

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TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS NUMBER CtfrSET PROB PERCENT ' BASIC EVENT NAME EVE 2r? PROB.' IDENTIFIER 1

1.68E-06 38.49 INITIATING EVENT - TRANSIENT WITH MFW EVENT OCCURS 1.40E+00 IEV-TRANS COMMON CAUSE FAILURE OF PMS AND PLS SOFWARE 1.20E-06' CCX-SF"IW 2

5.46E-07 12.51 INITIATING EVENT - STEAM CENERATOR WBE RUPWRE EVENT OCCURS 5.20E-03 IEV-SGTR OPERATOR FAILS TO MANUALLY ACWATE ADS (SGTR IF PRZ SPR FAILS) 5.00E-01 ADF-MAN 01

+

' COMMON CAUSE FAILURE OF RCP BREAKERS FAIL M OPEN 4.20E-04 RPX-CB-GO COND. PROB. OF ADN-MAN 01(OPER. FAILS TO ACT. ADS) 5.00E-01 ADN-MAN 01C 3

4.02E-07 9.21 INITIATING EVENT - LOSS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IEV-IMW COMMON CAUSE FAILURE OF PMS AND PLS SOFWARE 1.20E-06 CCX-SFW 4

3.34E-07 7.65 FAILURE OF PRS RELIEF FOR LOSS OF MFW ATWS, WITH UET 3.27E-01 OTH-PRESU INITIATING EVENT - A1WS PRECURSOR WITH NO MFW OCCURS 4.81E-01 IEV-A1WS i

COMMON CAUSE FAILURE OF PMS REACTOR TRIP SYSTEM RARDWARE 7.89E-05 CCX-PMS-HARDWARE OPERATOR FAILS W MANUALLY TRIP REACIVR VIA PMS 5.20E-02 A1W-MANO3 i

COND. PROB. OF ATW-MAN 01 (OPER. FAILS TO STEP-IN CONTROL ROD 5.17E-01 ATW-MAN 01C 5

2.30E-07 5.27 INITIATING EVENT - LOSS OF MFW TO ONE SG EVENT OCCURS 1.92E-01 IEV-LMFW1 COMMON CAUSE FAILURE OF PMS AND PLS SOF1 WARE 1.20E-06 CCX-SFIW 6

1.73E-07 3.96 INITIATING EVENT - LOSS OF CCW/SW EVENT OCCURS 1.44E-01 IEV-14CW

[

COMMON CAUSE FAILURE OF PMS AND PLS SOFTWARE 1.20E-06 CCX-SF"IW 7

1.34E-07 3.07 INITIATING EVENT - LOSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND COMMON CAUSE FAILURE OF PMS AND PLS SOFTWARE 1.20E-06 CCX-SFTW 8

5.72E-08 1.31 INITIATING EVENT - STEAM CENERATOR WBE RUPTURE EVENT OCCUMS 5.20E-03 IEV-SGTR CCMMON CAUSE FAILURE OF PMS ESF CUTPUT IDGIC SOFTWARE 1.10E-05 CCX-FMXMOD1-SW 9

4.4BE-08 1.03 INITIATING EVENT - STEAM GENERATOR WBE RUPWRE EVENT OCCURS 5.20E-03 IEV-SGTR COMMON CAUSE FAILURE OF OUTPtJT DRIVERS 8.62E-06 CCX-EP-SAM 10 4.20E-08

.96 FAILURE OF PRS RELIEF FOR 10SS OF MFW AWS WITH UET 3.27E-01 OTH-PRESU INITIATING EVENT - A1WS PRECURSOR WITH NO MFW OCCURS 4.81E-01 IEV-ATWS 5

COMMON CAUSE FAILURE OF REACTOR TRIP BREAKERS 8 10E-06 RCX-RB-FA OPERAWR FAILS TO STEP IN THE CONTROL RODS 3.30E-02 AW-MAN 01 11 3.43E-08

.79 FAILURE OF PRS RELIEF FOR LOSS OF MFW ATWS. WITH UET 3.27E-01 OTH-PRESU INITIATING EVENT - ATWS PRECURSOR WITH NO MFW OCCURS 4.81E-01 1EV-A1WS COMMON CAUSE FAILURE OF REACWR TRIP BREAKERS 8.10E-06 RCX-RB-FA I

OPERATOR FAILS TO MANUALLY TRIP REACTVR VIA DAS 5.20E-02 A1W-MAN 04 COND. PROB. OF ATW-MAN 01 (OPER. FAILS TO STEP-IN CONTROL ROD 5.17E-01 ATW-MAN 01C 12 3.00E-08

.69 INITIATING EVENT - ATWS PRECURSOR WITH NO MFW OCCURS 4.81E-01 IEV-ATWS COMMON CAUSE FAILURE OF PMS AND PLS SOFTWARE 1.20E-06 CCX-SFTW OPERATOR FAILS TO MANUALLY TRIP REACTOR VIA PMS 5.20E-02 ATW-MANO3 l

13 2.59E-08

.59 INITIATING EVENT - 14SS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-140ND FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 COMMON CAUSE FAILURE OF PMS ESF OtTTPtTF IDGIC SOF"IWARE 1.10E-05 CCX-FMXMOD1-SW 14 2.50E-08

.57 INITIATING EVENT - SAFETY INJECTION LINE BREAK EVENT OCCURS 1.04E-04 IEV-SI-LB IWRST DISCHARGE LINE

• A* STRAINER PLUGGED 2.40E-04 IMA-PLUG t

15 2.31E-08

.53 INITIATING EVENT - INTERMEDIATE IDCA EVENT OCCURS 7.70E-04 IEV-NI4CA COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-AO j

i 5

6 l

. {

TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSEI'S

(

16 2.31E-08

.53 INITIATING EVE 24T - INTERMEDIATE LOCA EVENT OCCURS 7.70E-04 IEV-NLOCA COMMON CAUSE FAILURE OF GTH STAGE ADS SQUIB VALVES TO OPERATE 3.00E-05 ADX-EV-SA 17 2.16E-08

.49 ' INITIATING EVDIT - IDSS OF RSC FLOW EVENT OCCURS 1.80E-02 IEV-LRCS COMMON CAUSE FAILURE OF PMS AND PLS SOFTWARE 1.20E-06 CCX-SFIW 18 2.03E-08

.47 INITIATING EVENT - LOSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CIOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 COMPON CAUSE FAILURE OF OUTPUT DRIVERS

8. 62 E-06 CCX-EP-SAM 19 2.00E-08

.46 INITIATING EVENT - IlfrERMEDIATE IDCA EVENT OCCURS 7.70E-04 IEV-NICCA p

COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES 2.60E-05 IWX-EV-SA s

20 2.00E-08

.46 INITIATING EVENT - INTERMEDIATE IDCA EVENT OCCURS 7.70E-04 IEV-NLOCA COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWK-EV4-SA 21 1.94E-08

.44 INITIATING EVENT - LOSS OF OFFSITE POWER EVENT OCCURS 1.20E-01 IEV-LOSP FAILURE TO RECOVER OFFSITE AC POWER IN 30 MINUTES 7.00E-01 OTH-ROS FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CIDSE (SV/PORV) 2.10E-02 OTH-SLSOV1 COMMON CAUSE FAILURE OF PMS ESF OUTPUF IDGIC SOFWARE 1.10E-05 CCX-FMXMOD1 oW 22 1.74E-08

.40 INITIATING EVENT - TRANSIENT WITH MFW EVENT OCCURS 1.40E+00 IEV-TRANS COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNEM C09 TON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA 23 1.52E-08

.35 INITIATING EVENT - LOSS OF OFFSITE POWER EVENT OCCURS 1.20E-01 IEV-LOSP l

FAILURE TO RECOVER OFFSITE AC POWER IN 30 MrsUTES 7.00E-01 OTH-R05 L

FAIIURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 i

COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 24 1.42E-08

.33 INITIATING EVENT - ATWS PRECURSOR WITH SI SIGNAL OCCURS 2.05E-02 IEV-AW-S FAILURE OF PRS RELIEF FOR LOSS OF MFW ATWS, WITH LTr 3.27E-01 OTH-PRESU COMMON CAUSE FAILURE OF PMS REACTOR TRIP SYSTDI HARDWARE 7.89E-05 CCX-FMS-HARDhARE OPERATOR FAILS TO MANUALLY TRIP REACTOR VIA PMS 5.20E-02 ATW-MANO3 COND. PROB. OF A W-MAN 01 (OPER. FAILS 10 STEP-IN CONTROL ROD 5.17E-01 ATW-MAN 01C 25 1.33E-08

.30 INITIATING EVENT - MAIN STEAM LINE STUCK-OPEN SV OCCURS 1.21E-03 IEV-SLB-V COMMON CAUSE FAILURE OF EMS ESF OUTPUT IMGIC SOFTWARE 1.10E-OL CCX-FMXMOD1-SW 26 1.04E-08

.24 INITIATING EVENT - MAIN STEAM LINE STUCK-OPEN SV OCCURS 1.21E-03 IEV-SLB-V COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM i

6 27 1.00E-08

.23 INITIATING EVENT - REACTOR VESSEL RUPTURE EVENT OCCURS 1.00E-08 IEV-RV-aP I

28 9.24E-09

.21 INITIATING EVENT - INTERMEDIATE LOCA EVENT OCCURS 7.70E-04 IEV-NLOCA COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK 1.20E-05 IWX-FL-GP f

29 9.24E-09

.21 INITIATING EVENT - INTERMEDIATE IDCA EVENT OCCURS 7.70E-04 IEV-NIACA

[

COMMON CAUSE FAILURE OF RECIRC LINES DUE TO SUMP SCREEN PLUGGING 1.20E-05 REX-FL-GP

}

30 9.47E-09

.19 INITIATING EVENT - INTERMEDIATE LOCA EVENT OCCURS 7.70E-04 IEV-NLOCA COMMON CAUSE FAILURE OF PMS ESF OUTPUT IDGIC SOFIWARE 1.10E-05 CCX-FMXMOD1-SW f

31 7.79E-09

.18 FAILURE OF PRZ SV FOR LOSS OF MFW AWS. NO UET 2.00E-03 OTH-PRES INITIATING EVENT - A1WS PRECURSOR WITH NO MFW OCCURS 4.81E-01 IEV-ATWS COMMON CAUSE FAILURE OF REACTVR TRIP BREAKERS 8.10E-06 RCX-RB-FA

.--w--.

m.

.m.

m.m

TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS 32 7.36E-09

.17 INITIATIPF3 EVENT - TRANSIENT WITH MRf EVENT OCCURS 1.40E+00 IEV-TRANS COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONME!C 4.7BE-04 CCX-TRNSM COMMCN CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOFTWARE 1.10E-05 CCX-FMXP80D1-SW 33 6.64E-09

.15 INITIATING EVENT - INTERMEDIATE LOCA EVEFTT OCCURS 7.70E-04 IEV-NLOCA COMMON CAUSE FAILURE OF OLTTPtTT DRIVERS 8.62E-06 CCX-EP-SAM 34 6.60E-09

.15 INITIATING EVENT - STEAM GENERATOR TUBE RUPTURE EVENT OCCURS 5.20E-03 IEV-SGTR COMMON CAUSE FAILURE OF ItCP BREAKERS FAIL TO OPEN 4.20E-04 RPX-CB-GO OPERATOR FAILS E MANUALLY ACWATE ADS 3.02E-03 ADN-MAN 01 35 6.56E-09

.15 INITIATING EVDIT - STEAM LINE BREAK DOWNSTREAM OF MSIV OCCURS 5.96E-04 IEV-SLB-D COMMON CAUSE FAILURF OF PMS ESF OLTTPUT LOGIC SOFDfARE 1.10E-05 CCI-FMXMOD1-SW 36 6.50E-09

.15 INITIATING EVENT - PASSIVE RHR WBE RUPTURE EVENT OCCURS 2.50E-04 IEV-PRSTR COMMON CAUSE FAILURE CF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA 37 6.24E-09

.14 INITIATING EVEffr - STEAM GENERA'lVR TUBE RUPWRE EVENT OCCURS 5.20E-03 IEV-SGTR COMMON CAUSE FAILURE OF PMS AND PLS SOFTWARE 1.20E-06 CCI-SFTW 39 6.15E-09

.14 INITIATING EVENT - IDSS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IEV-LMRf COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCX-XMTR COMMON CAUSE FAILURE OF CMT/SLMP LEVEL HEATED RTD SENSORS 3.84E-05 CMX-VS-FA 39 5.77E-09

.13 INITIATING EVENT - TRANSIENT WITH MRf EVENT OCCURS 1.40E+00 IEV-TRANS COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCI-EP-SAM 40 5.71E-09

.13 INITIATING EVErf" - AWS PRECURSOR WITH NO MRf OCCURS 4.81E-01 IEV-AWS OPERATOR FAILS % MANUALLY TRIP REACTOR VIA PMS 5.20E-02 A'Df-MANO 3 COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCI-XKIR COMMON CAUSE FAILURE OF PZR LEVEL SENSORS 4.7eE-04 CCX-XMTR195 41 5.41E-09

.12 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSIZE COMMON CAUSE FAILURE OF PMS ESF INPUT LOGIC GROUPS (HARIMARE) 1.03E-04 CCX-INPUT-LOGIC 42 5.41E-09

.12 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE COMMON CAUSE FAILURE OF PMS ESF INPUT LOGIC GROUPS (HARDWARE) 1.03E-04 CCX-INPUT-LOGIC 43 5.40E-09

.12 INITIATING EVENT - CORE POWER EXCURSION EVENT OCCURS 4.50E-03 IEV-PCMEX COMP 80N CAUSE FAILURE OF PMS AND PLS SOFDfARE 1.20E-06 CCI-SFDi 44 5.36E-09

.12 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA COMMON CAUSE FAILL^RE OF 2 ACCLHULATOR CHECK VALVES 5.10E-05 ACX-CV-GO 45 5.14E-09

.12 INITIATING EVENT - STEAM LINE BREAK DOWNSTREAM OF MSIV OCCURS 5.96E-04 IEV-SLB-D COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 46 4.86E-09

.11 INITIATING EVENT - MEDILH LOCA EVENT OCCURS 1.62E-04 IEV-MIOCA COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-AO 47 4.86E-09

.11 INITIATING EVENT - MEDILH LOCA EVENT OCCURS 1.62E-04 IEV-MISCA COMMON CAUSE FAILURE OF STR STAGE ADS SQUIB VALVES TO OPERATE 3.00E-05 ADX-EV-SA 48 4.21E-09

.10 INITIATING EVENT - MEDIUM LOCA EVENT OCCURS 1.622-04 IEV-MLOCA COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA W Westinghouse

e m

TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS 49 4.21E-09

.10 INITIATING EVENT - MEDIUM LOCA EVENT OCCURS 1.62E-04 IEV-MIOCA COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES 2.60E-05 IWX-EV-SA 50 4.16E-09

.10 INITIATING EVENT - IDSS OF MAIN FEEDWATER EVENT OCC' RS 3.35E-01 IEV-IFJW J

COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCX-XNTR COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA 51 4.16E-09

.10 INITIATING EVENT - I4SS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IW-LMFW COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE EVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LI E 2.60E-05 IWX-EV4-SA 52 4.09E-09

.09 INITIATING EVENT - STEAM LINE UPSTRE.AM OF MSIV OCCURS 3.72E-04 IEV-SLB-U COMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOFWARE 1.10E-05 CCX-FMXMOD1-SW 53 3.95E-09

.09 FAILURE OF PRZ SV FOR LOSS OF MFW ATWS, NO UET 2.00E-03 OTH-PRES INITIATING EVENT - A W S PPECURSOR WITH NO MFW OCCURS 4.81E-01 IEV-ATWS COMMON CAUSE FAILURE OF PMS REACTOR TRIP SYSTEM HARDWARE 7.89E-05 CCX-FMS-MARDWAPE OFERATOR FAILS TO MANUALLY TRIP REACTOR VIA PMS 5.20E-02 ATW-MANO3 54 3.83E-09

.09 INITIATING EVENT - LOSS OF COMPRESSED AIR EV m T OCCURS 3.48E-02 IEV-LCAS FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV) 1.00E-02 OTH-SLSOV2 COMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOFWARE 1.10E-OS CCX-FMXMOD1-SW 55 3.68E-09

.08 INITIATING EVENT - INTERMEDIATE LOCA EVENT OCCURS 7.70E-04 IEV-NLOCA COMMON CAUSE FAILURE OF TANK LEVEL TRANSMI'ITERS (IRWST, BAT) 4.7BE-04 IWX-XNTR OPERATOR FAILS TO ACTUATE CONT SUMP RECIR. (LEVEL SIGNAL FAILS) 1.00E-02 REN-MAN 04 56 3.21E-09

.07 INITIATING EVENT - STEAM LINE UPSTREAM OF MSIV OCCURS 3.72E-04 IEV-SLB-U COMMON CAUSE FAILURE OF OUTPUT DRIVERS B.62E-06 CCX-EP-SAM 57 3.20E-09

.07 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSIZE COMMON CAUSE FAILURE OF 4 AOVS TO OPEN 6.10E-05 CCX-AV-LA Se 3.20E-09

.07 INITIATING EVENT - LARGE LOCA EVGT OCCURS 1.05E-04 IEV-LLOCA LIDCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE COMMON CAUSE FAILURE OF 4 ADVS TO OPEN 6.10E-05 CCX-AV-LA 59 3.20E-09

.07 INITIATING EVENT - TRANSIENT WITH MFW EVENT OCCURS 1.40E+00 IEV-TRANS COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF TANK LEVEL WANSMITTERS (IRWST, BAT) 4.7eE-04 IWX-XNTR OPERATOR FAILS TO ACTUATE CONr. SUMP RECIR. (LEVEL SIGNAL FAILS) 1.00E-02 REN-MAN 04 60 3.12E-09

.07 INITI ATING EVL'NT - SAFETY INJECTION LINE BREAK EVENT OCCURS 1.04E-04 IEV-SI-LB COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWK-CV-AO 61 3.12E-09

.07 INITIATING EVENT - SAFETY INJECTION LINE BREAK EVENT OCCURS 1.04E-04 IEV-SI-LB COMMON CAUSE FAILURE OF 4TH STAGE ADS SQUIB VALVES TV OPERATE 3.00E-05 ADX-EV-SA 62 3.03E-09

.07 INITIATING EVENT - SMALL LOCA EVENT OCCURS 1.01E-04 IEV-SLOCA COMMON CAUSE FAILURE OF 4TH STAGE ADS SQUIB VALVES TO OPERATE 3.00E-05 ADX-EV-SA 63 3.03E-09

.07 INITIATING EVEN* - SMALL LOCA EVENT OCCURS 1.01E-04 IEV-SLOCA COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-A0 6

W Westingh00Se

A

[

t

. TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS t

I 64 3.00E-09

.07 INITIATING EVENT - LOSS OF CCMPRESSED AIR EVENT OCCURS 3.48E-02 IEV-LCAS i

FAILURE OF A SECONDARY SIDE RELIEF VALVE TO C1DSE (SV) 1.00E-02 OTH-SLSOV2 J

COMMON CAUSE FAILURE OF OUTPLTP DRIVERS 8.62E-06 CCX-EP-SAM 4

65 2.93E-09

.07 INITIATING EVENT - STEAM GENERATOR TUBE RUPTURE EVDET OCCURS 5.20E-03 IEV-SGTR 2

COMMON CAUSE FAILURE OF RCP BREAKERS FAIL TO OPEN 4.20E-04 RPK-CB-GO OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (SLOCA/ TRANSIENT) 1.34E-03 LPM-MAN 01 i.

66 2.75E-09

.06 INITIATING EVENT - PASSIVE RHR '1TJBE RUPTURE EVENT OCCURS 2.50E-04 IEV-PRSTR COMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGF' SOFTWARE 1.10E-05 CCX-FMKMOD1-SW 67 2.70E-09

.06 INITIATING EVENT - SAFETY INJECTION LINE BREAK EVENT OCCURS 1.04E-04 IEV-SI-LB r

COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES 2.60E-05 IWX-EV SA i

68 2.68E-09

.06 INITIATING EVENT - CMT LINE BREAK EVENT OCCURS 8.94E-05 IEV-CMTLB COMMON CAUSE. FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-AO i

I 69 2.68E-09

.06 INITIATING EVENT - CNT LINE BREAK EVENT OCCURS 8.94E-05 IEV-CMTLB L

COMMON CAUSE FAILURE OF 4TH STAGE ADS SQUIB VALVES TO OPERATE 3.00E-05 ADX-EV-SA t-70 2.68E-09

.06 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA i

LLOCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSIEE i

COMMON CAUSE FAILURE OF 4 CNT CHECK VALVES TO OPEN 5.10E-05 CMX-CV-GO 71 2.68E-09

.06 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LIACA LLOCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE i

[

COMMON CAUSE FAILURE OF 4 CMT CHECK VALVES TO OPEN 5.10E-05 CMX-CV-GO 72 2.63E-09

.06 INITIATING EVENT - SMALL LOCA EVENT OCCURS 1.01E-04 IEV-SIDCA COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA 73 2.63E-09

.06 INITIATING EVENT - SMALL LOCA EVENT OCCURS 1.01E-04 IEV-SLOC4 COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VAINES 2.60E-05 IWX-EV-SA.

{

74 2.39E-09'

.05 INITIATING EVENT - LOSS OF MW TO ONE SG EVENT OCCURS 1.92E-01 IEV-LMW1 COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWK-EV4-SA 75 2.32E-09

.05 INITIATING EVENT - CMT LINE BREAK EVENT OCCURS 8.94E-05 IEV-CMTLB COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-03 IWX-EV4-SA l

76 2.32E-09

.05 INITIATING EVENT - CMT LINE BREAK EVENT OCCURS 8.94E-05 IEV-CMTLB COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES 2.60E-05 IWX-EV-SA 77 2.22E-09

.05 INITIATING EVENT - RANSIENT WITH MN EVENT OCCURS 1.40E+00 IEV-TRANS COMMON CAUSE FAILURE OF SENSORS IN IDW PRESSURE ENVIRONMENT 4.78E-04 CCX-RNSM COMMON CAUSE FAILURE OF 4/4 STAGE 2 Tm 3 LINE MOVs TO OPEN 1.10E-03 ADX-MV-GO OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 78 2.16E-09

.05 INITIATING EVENT - PASSIVE RHR TUBE RUPIURE EVENT OCCURS 2.50E-04 IEV-PRSTR COMMON CAUSE FAILURE OF OUTPLTP DRIVERS 8.62E-06 CCX-EP-SAM t

79 2.12E-09

.05 INITIATING EVENT - LOSS OF OFFSITE POWER EVENT OCCURS 1.20E-01 IEV-LOSP FAILURE *IV RECOVER OFFSITE AC POWER IN 30 MINUTES 7.00E-01 OTH-ROS FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 COMMON CAUSE FAILURE CF PMS AND PLS SOFIWARE 1.20E-06 CCX-SFIW

[

l

6 o

e TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS 80 2.06E-09

.05 INITIATING EVENT - LOSS OF CCW/SW EVENT OCCURS 1.44E-01 IEV-ICCW COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-AO 81 2.06E-09

.05 INITIATING EVENT - LOSS CF CCW/SW EVENT OCCURS 1.44E-01 IEV-LCCW COMMON CAUSE FAILURE OF SENSOKS IN 14W PRESSURE ENVIRONMENT 4.78E-04 CCX-TPRSM COMMON CAUSE FAILURE OF 4TH STAGE ADS SQUIB VALVES TO OPERATE 3.00E-05 ADX-EV-SA 82 2.06E-09

.05 INITIATING EVENT - IDSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMEPTT 4.78E-04 CCX-XMTR COMMON CAUSE FAILURE OF CMT/ SUMP LEVEL HEATED RTD SENSORS 3.84E-05.

CMX-VS-FA 83 2.02E-09

.05 INITIATING EVENT -- LARGE LOCA EVENT OCCURS 1.CSE-04 IEV-LthCA LLOCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSIZE COMMON CAUSE FAILURE OF CMT/ SUMP LEVEL HEATED RTD SENSORS 3.84E-05 CMX-VS-FA 84 2.02E-09

.05 INITIATING EVDIT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LIDCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE COMMON CAUSE FAILURE OF CMT/ SUMP LEVEL HEATED RTD SENSORS 3.84E-05 CMX-VS-FA 85 2.01E-09

.05 INITIATING EVENT - STEAM GENERATOR TUBE RUPTURE EVENT CCCURS 5.20E-03 IEV-SGTR OPERATOR FAILS TO DIAGNOSE SGTR EVENT 1.84E-03 CIB-MAN 00 COMMON CAUSE FAILURE OF RCP BREAKERS FAIL TO OPEN 4.20E-04 RPK-CB-GO COND. PROB. UF ADN-MAN 01(OPER. FAILS TO ACT. ADS) 5.00E-01 ADN-MAN 01C 86 1.94E-09

.04 INITIATING EVENT - MEDIUM LOCA EVENT OCCURS 1.62E-04 IEV-MLOCA COMMON CAUSE FAILURE OF RECIRC LINES DUE TO SUMP SCREEN PLUGGING 1.20E-05 REX-FL-CP 87 1.94E-09

.04 INITIATING EVENT - MEDIUM LOCA EVENT OCCURS 1.62E-04 IEV-MLOCA COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK 1.20E-05 IWX-FL-GP 88 1.86E-09

.04 INITIATING EVENT - ATWS PRECURSOR WITH NO MFW OCCURS 4.81E-01 IEV-ATWS COMMON CAUSE FAILURE OF REACTOR TRIP BREAKERS 8.10E-06 RCX-RB-FA COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCX-XMTR 89 1.86E-09

.04 INITIATING EVENT - ATWS PRECURSOR WITH NO MRf OCCURS 4.81E-01 IEV-ATWS COMMON CAUSE FAILURE OF REACTOR TRIP BREAKERS 8.10E-06 RCX-RB-FA COMMON CAUSE FAILURE OF SENSOhS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM 90 1.79E-09

.04 INITIATING EVENT - ATWS PRECURSOR WITH SI SIGNAL OCCURS 2.05E-02 IEV-ATW-S FAILURE OF PRS RELIEF FOR LOSS OF MFW ATWS. WITH UET 3.27E-01 OTH=PRESU COMMON CAUSE FAILURE OF REACTOR TRIP BREAKERS 8.10E-06 RCX-RB-FA OPERATOR FAILS TO STEP IN THE CoffTROL RODS 3.30E-02 ATW-MAN 01 91 1.79E-09

.04 INITIATING EVENT - IOSS OF CCW/SW EVENT OCCURS 1.44E-01 IEV-LCCW COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA

- 92 1.79E-09

.04 INITIATING EVENT - IDSS OF CCW/SW EVENT OCCURS 1.44E-01 IEV-14CW COMMON CAUSE FAILURE OF SENSORS IN IKPRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES 2.60E-05 IWX+EV-SA 93 1.78E-09

.04 INITIATING EVENT - MEDIUM LOCA EVENT OCCURS 1.62E-04 IEV-MLOCA COMMON CAUSE FAILURE OF PMC ESF OUTPUT LOGIC SOFTWARE 1.10E-05 CCX-PMIMOD1-SW 94 1.76E-09

.04 INITIATING EVENT - 14SS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IEV-LMMt COMMON CAUSE FAILURE OF SENSORS IN I4W PRESSURE ENViru nrunr 4.78E-04 CCX-TRNSM

.n e

TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS COMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOF"! WARE 1.10E-05 CCX-FMXMOCl-SW 95 1.76E-09

.04 INITIATING EVENT - LOSS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IEV-LMW COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCX-XMTR COMMON CAUSE FAILURE OF MS ESF OUTPUT LOGIC SOFTWARE 1.10E-05 CCX-FMXMOD1-SW 96 1.58E-09

.04 INITIATING EVENT - LARGE IDCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE - UFFER END OF BREAK SIZE 5.00E-01 BSIZE-IARGE COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-AO 97 1.58E-09

.04 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSITE COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-AO 98 1.51E-09

.03 IEV-RCSLKC 5.02E-05 IEV-RCSLKC COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-AO 99 1.51E-09

.03 IEV-RCSLKC 5.02E-05 IEV-RCSLKC COMMON CAUSE FAILURE OF 4TH STAGE ADS SQUIB VALVES TO OPERATE 3.00E-05 ADX-EV-SA 100 1.46E-09

.03 INITIATING EVENT - ATWS PRECURSOR WITH SI SIGNAL OCCURS 2.05E-02 IEV-ATW-S FAILURE OF PRS RELIEF FOR LOSS OF MFW ATWS, WITH UET 3.27E-01 OTH-PRESU COMMON CAUSE FAILURE OF REACTOR TRIP BREAKERS 8.10E-06 RCX-RB-FA OPERATOR FAILS *[O MANUALLT TRIP REACTOR VIA DAS 5.20E-02 ATW-MAN 04 COND. PROB. OF ATW-MAN 01 (OPER. FAILS TO STEP-IN CONTROL ROD 5.17E-01 ATW-MAN 01C 101 1.45E-09

.03 INITIATING EVENT - MAIN STEAM LINE STUCK-OPEN SV OCCURS 1.21E-03 IEV-SLB-V COMMON CAUSE FAILURE OF FMS AND PLS SOF"IWARE 1.20E-06 CCX-SFTW 102 1.40E-09

.03 INITIATING EVENT - MEDIUM 14CA EVENT OCCURS 1.62E-04 IEV-MLOCA COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 103 1.39E-09

.03 INITIATING EVENT - LOSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCX-XMTR COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA 104 1.39E-09

.03 INITIATING EVENT - LOSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA 105 1.39E-09

.03 INITIATING EVENT - ATWS PRECURSOR WITH MFW AVAILA. OCCURS 1.17E+00 IEV-ATW-T OPERATOR FAILS TO MANUALLY TRIP REACTOR VIA PMS 5.20E-03 ATW-MANOS COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E -04 CCX-XMTR COMMON CAUSE FAILURE OF PZR LEVEL SENSORS 4.78E-04 CCX-XMTR195 106 1.38E-09

.03 INITIATING EVENT - LOSS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IEV-LMFW COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSUPE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 107 1.3BE-09

.03 INITIATING EVENT - LOSS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IEV-IMFW COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCX-EMTR COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 108 1.37E-09

.03 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSIZE-COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA

{

e

+,

~

3 TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS 109 1.37E-09

.03 INITIATING EVENT - IARGE LOCA EVENT OCCURS

'1.05E-04 IEV-LLOCA t

LLOCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01

- BSIZE C04 TON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES 2.60E-05 IWK-EV-SA e

110 1.37E-09

.03 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LIDCA LLOCA BREAK SIZE - UFPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE f

COMMON CAUSE FAILURE OF 4 IRWST IICECTION SQUIB VALVES 2.60E-05 IWX-EV-SA 111-1.37E-09

.03 INITIATING EVENT - LARGE LOCA EVEPrr OCCURS 1.05E-04 IEV-LLOCA I

LLOCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA j

112 1.35E-09

.03 INITIATING EVENT - CORE POWER EXCURSION EVENT OCCURS 4.50E-03 IEV-POWEX FAILURE OF EITHER PZR SV FAILS % RECLOSE 1.00E-02 OTH-PRSOV 6

COMMON CAUSE FAILURE OF 4TH STAGE ADS SQUIB VALVES M OPERATE 3.00E-05 ADX-EV-SA 113 1.35E-09

.03 INITIATING EVL'N. T - CORE POWER EXCURSION EVENT OCCURS 4.50E-03 IEV-POWEX FAILURE OF EITHER PZR SV FAILS TO RECLOSE 1.00E-02 OTH-PRSOV

[

COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-AO 114 1.31E-09

.03 IEV-RCSLKC 5.02E-05 IEV-RCSLKC COMMON CAUSE FAILURE OF 4 SQUIB VALVF9 1M DECIRC LINES 2.60E-05 IWX-EV4-SA i

115 1.31E-09

.03 IEV-RCSLKC 5.02E-05 IEV-RCSLKC COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES 2.60E-05 IWX-EV-SA 116 1.28E-09

.03 INITIATING EVENT - ATWS PRECURSOR WITH SI SIGNAL OCCURS 2.05E-02 IEV-AW-S COMrtON CAUSE FAILURE OF F91S AND PLS SOFWARE 1.20E-06 CCX-SFW i

OPERATOR FAILS TO MANUALLY TRIP REACTOR VIA PMS 5.20E-02 AW-MANO3 117 1.24E-09

.03 INITIATING EVENT - STEAM GENERATOR WBE RUPWRE EVENT OCCURS 5.20E-03 IEV-SGTR MAIN GEN. BKR ES 01 FAILS TO CPEN le 12]

5.00E-03 ECOMOD01 r

COMMON CAUSE FAILURE OF CLASS 1E BATTERIES 4.70E-05 CCX-BY-PN i

118 1.21E-09

.03 INITIATING EVENT - SMALL LOCA EVENT OCCURS 1.01E-04 IEV-SLOCA COMMON CAUSE FAILURE OF RECIRC LINES DUE TO SUMP SCREEN PLUGGING 1.20E-05 REX-FL-GP 119 1.21E-09

.03 INITIATING EVENT - SMALL LOCA EVENT OCCURS 1.01E-04 IEV-SLOCA COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK 1.20E-05 IWX-FL-GP o

120 1.20E-09

.03 INITIATING EVENT - PASSIVE RHR 'liJBE RUPWRE EVENT OCCURS 2.50E-04 IEV-PRSTR 1

COMMON CAUSE FAILURE OF TANK LEVEL TRANSMITTERS (IRWST, BAT) 4.7BE-04 IWK-XMTR

[

OPERATOR FAILS TO ACTUATE CONT. SUMP RECIR. (LEVEL SIGNAL FAILSI 1.00E-02 REN-MAN 04 121 1.17E-09

.03 INITIATING EVENT - CORE POWER EXCURSION EVENT OCCURS 4.50E-03 IEV-POWEX FAILURE OF EITHER PZR SV FAILS TO RECLOSE 1.00E-02 OTH-PRSOV j

COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA 122 1.17E-09

.02 INITIATING EVENT - CORE POWER EXCURSION EVENT OCCL1tS 4.50E-03 IEV-POWEX FAILURE OF EITHER PZR SV FAILS E RECLOSE 1.00E-02 OTH-PRSOV COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES 2.60E-05 IWX-EV-SA

[

123 1.14E-09

.03 INITIATING EVENT - SAFETY INJECTION LINE BREAK EVENT OCCURS 1.04E-04 IEV-SI-LB COMMON CAUSE FAILURE OF PMS ESF OUTPtfr LOGIC SOFIWARE 1,10E-05 CCX-PIC0tOD1-SW

{

124 1.11E-09

.03 INITIATING EVENT - SMALL IDCA EVENT OCCURS 1.01E-04 IEV-SLOCA COMMON CAUSE FAILURE OF PMS ESF OUTPUT ' LOGIC SOFWARE 1.10E-05 CCX-PMXMOD1-SW 6

!i f

r

->w->a.

g

-+

.e+-,

eseem-u.mww.e.we up awr s.mm ap.L a

w-..-w

1 TABLE 2 f

SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS v

L i

125 1.07E-09

.02 INITIATING EVENT - CMT LINE BREAK EVENT OCCURS 8.94E-05 IEV-CMTLB COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK 1.20E-05 IWX-FL-GP 126

'1.07E-09

.02 INITIATING EVENT - CMT LINE BREAK EVENT OCCURS 8.94E-05 IEV-CNTLB COMMON CAUSE FAILURE OF RECIRC LINES DUE TO SUMP SCREEN PLUGGING 1.20E-05 REX-FL-GP 127 1.07E-09

.02 INITIATING EVENT - INTERMEDIATE LOCA EVENT OCCURS 7.70E-04 IEV-NIDCA COMMON CAUSE FAILURE OF RCP BREAKERS FAIL TO OPEN 4.20E-04 RPX-CB-GO OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (MLOCA) 3.30E-03 LPM-MANO2 128 1.04E-09

.02 INITIATING EVENT - SAFETY IMJECTION LINE BREAK EVENT OCCURS 1.04E-04 IEV-SI-LB COMMON CAUSE FAILURE OF 2 IRWST IM7EC. SQUIBS IN 1 LINE TO OPEN 1.00E-05 IWX-EV1-SA 129 1.01E-09

.02 INITIATING EVENT - IDSS OF MFW TO ONE SG EVENT OCCURS 1.92E-01 IEV-IJEFW1 COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM

+

COMMON CAUSE FAILURE OF PMS ESF OUTPUT I4GIC SOFTWARE 1.10E-05 CCX-FMXMODI-SW 130 9.86E-10

.02 INITIATING EVENT

'DtANSIENT WITH MFW EVENT OCCURS 1.40E+00 IEV-TRANS COMMON CAUSE FAILURE OF SENSORS IN I4W PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF 4/4 STAGE 2 & 3 LINE MOVs TO OPEN 1.10E-03 ADX-MV-GO OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (SLOCA/ TRANSIENT) 1.34E-03 LPM-MAN 01 i

i 131

~9.83E-10

.02 INITIATING EVENT - CMT LINE BREAK EVENT OCCURS 8.94E-05 IEV-CMTLB T

COMMON CAUSE FAILURE OF PMS ESF OLTTPUT LOGIC SOF1 WARE 1.10E-05 CCX-FMXMOD1-SW

[

l 132 9.77E-10

.02. INITIATING EVENT - INTERMEDIATE LOCA EVENT OCCURS 7.70E-04 IEV-NLOCA

[

COMMON CAUSE FAILURE OF RCP BREAKERS FAIL TO OPEN 4.20E-04 RPX-CB-GO OPERATOR FAILS TO MANUALLY ACWATE ADS 3.02E-03 ADN-MAN 01 133 9.5BE-10

.02 INITIATING EVENT + STEAM GENERATOR 1UBE RUPTURE EVENT OCCURS 5.20E-03 IEV-SGTR COMMON CAUSE FAILURE OF 4 AOVS TO OPEN 6.10E-05 CCX-AV-LA l

OPERATOR FAILS 10 MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 i

i 134 9.43E-10

.02 INITIATING EVENT - ATWS PRECURSOR WITH NO MFW OCCURS 4.81E-01 IEV-ATWS COMMON CAUSE FAILURE OF PMS REACTOR TRIP SYSTEM HARDWARE 7.89E-05 CCX-FMS-HARDWARE OPERATOR FAILS TO MANUALLY TRIP REACTOR VIA PMS 5.20E-02 ATW-MANO3 COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E CCX-XMTR 135 9.43E-10

.02 INITIATING EVENT - ATWS PRECURSOR WITH NO MFW OCCURS 4.81E-01 IEV-ATWS

. COMMON CAUSE FAILURE OF PMS REACTOR TRIP SYSTEM HARDWARE 7.89E-05 CCX-FMS-HARDWARE OPERATOR FAILS TO MANUALLY TRIP REACTOR VIA PMS 5.20E-02 ATW-MANO3 COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM 136 9.24E-10

.02 INITIA~ TING EVENT - INTERMEDIATE LOCA EVENT OCCURS 7.70E-04 IEV-NIOCA h

COMMON CAUSE FAILUME OF PMS AND PLS SOFTWARE 1.20E-06 CCX-SFTW 137 8.96E-10

.02 INITIATING EVENT - SAFETY INJECTION LINE BREAK EVENT OCCURS 1.04E-04 IEV-SI-LB COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX+EP-SAM l

138 8.86E-10

.02 INITIATING EVENT - TRANSIENT WITH MFW EVENT OCCURS 1.40E+00 IEV-TRANS FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 1.10E-02 OTH+SLSOV TRANSMITTER FAILURE 5.23E-03 CDNTF01BRI COMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOFTWARE 1.10E-05 CCX-PMXMOD1-SW

'[

i b

139 8.86E-10

.02 INITIATING EVENT - TRANSIENT WITH MFW EVENT OCCURS 1.40E+00 IEV-TRANS FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 1.1CE-02 OTH-SLSOV FAILURE OF AIR COMPRESSOR TRANSMITTER 5.23E-03 CANTP011RI

[

COMMON CAUSE FAILURE OF PMS ESF OUTPUT IDGIC SOFTWARE 1.10E-05 CCX-FMXMOD1-SW

(

l I

e r

=..,

e

'L e

TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS 140 8.71E-10

.02 INITIATING EVENT - SMALL LOCA EVENT OCCURS 1.01E-09 IEV-SLOCA COMMON CAUSE FAILURE OF OUTPt1I DRIVERS R. 62 E-0 6 CCX-EP-SAM 141 8.49E-10

.02 INITIATING EVENT - TRANSIENT WITH MFW EVENT OCCURS 1.40E+00 IEV-TRANS COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCI-TRNSM COMMON CAUSE FAILURE OF RCP BREAKERS FAIL TO OPEN 4.20E-04 RPX-CB-GO OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 142 8.41E-10

.02 FAILURE OF PRS RELIEF FOR LOSS OF MFW AIWS, WITH UET 3.27E-01 OTH-PRESU INITIATING EVENT - ATWS PRECURSOR WITH NO MFW OCCURS 4.81E-01 IEV-ATWS COMMON CAUSE FAILURE OF REACWR TRIP BREAKERS 8.10E-06 RCX-RB-FA FAILURE OF RCD CONTROL SYSTEM TO STEP IN RODS 6.60E-04 ROD-CTRL-SYS 143 8.36E-10

.02 INITIATING EVENT - LOSS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IEV-LMFW COMMON CAUSE FAILURE OF PRHR AOVs 9.60E-05 PXX-AV-LA COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA 144 8.26E-10

.02 INITIATING EVENT - LOSS OF CCW/SW EVENT OCCURS 1.44E-01 IEV-LCCW COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK 1.20E-05 IWX-FL-GP 145 8.26E-10

.02 INITIATING EVENT - LOSS OF CCW/SW EVENT OCCURS 1.44E-01 IEV-LCCW COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.7BE-04 CCX-TRNSM COMMON CAUSE FAILURE OF RECIRC LINES DUE TO SUMP SCREEN PiUGGING 1.20E-05 REX-FL-CP 146 8.01E-10

.02 INITIATING EVENT - STEAM GENERAWR TUBE RUPWRE EVENT OCCURS 5.20E-03 IEV-SGTR COMMON CAUSE FAILURE OF 4 CMT CHECK VALVES TO OPEN 5.10E-05 CMX-CV-GO OPERATOR FAILS TO MANUALLY ACWATE ADS 3.02E-03 ADN-MAN 01 147 7.91E-10

.02 INITIATING EVENT - LOSS OF MEW TO ONE SG EVENT OCCURS 1.92E-01 IEV-LMFW1 COMMON CAUSE FAILURE OF SENSORS IN LOW FRESSURE ENVIRONMENT 4.78E-04 CCX-TPESM COMMON CAUSE FAILURE OF OUTPL7T DRIVERS 8.62E-06 CCX-EP-SAM 148 7.74E-10

.02 INITIATING EVENT - MEDIUM IDCA EVENT OCCURS 1.62E-04 IEV-MLOCA COMMON CAUSE FAILURE OF TANK LEVEL TRANSMITTERS (IRWST, BAT) 4.78E-04 IWX-XMTR OPERATOR FAILS TO ACTUATE CONT. SUMP RECIR. (LEVEL SIGNAL FAILS) 1.00E-02 REN-MAN 04 149 7.71E-10

.02 INITIATING EVENT - CMT LINE BREAK EVENT OCCURS 8.94E-05 IEV-CMTLB COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 150 7.65E-10

.02 INITIATING EVENT - LOSS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IEV-LMEW COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TPRSM COMMON CAUSE FAILURE OF TANK LEVEL TRANSMITTERS (IRWST, BAT) 4.78E-04 IWX-XMTR OPERATOR FAILS TO ACTUATE CONT. SUMP RECIR. (LEVEL SIGNAL FAILS) 1.00E-02 REN-MAN 04 151 7.65E-10

.02 INITIATING EVEffT - LOSS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IEV-LMFW COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCI-XHTR COMMON CAUSE FAILURE OF TANK LEVEL TRANSMITTERS (IRWST, BAT) 4.78E-04 IWX-XMTR OPERATOR FAILS TO ACTUATE CONT. SUMP RECIR. (LEVEL SIGNAL FAILS) 1.00E-02 REN-MAN 04 152 7.57E-10

.02 INITIATING EVENT - I4SS OF CCW/SW EVENT OCCURS 1.44E-01 IEV-LCCW COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMEPTT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOFTWARE 1.10E-05 CCX-PMIMOD1-SW 153 7.30E-10

.02 INITIATING EVENT - STEAM GENERATOR TUBE RUPWRE EVENT OCCURS 5.20E-03 IEV-SGTR FAILURE OF SG PORV & 1 SG SV ON RUPWRED SG TO CLOSE 5.40E-03 OTH-SLSOV3 COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA W Westinghouse

g

.v

, e e

TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS 154 7.15E-10

.02 INITIATING EVENT - STEAM LINE BREAK DOWNSTREAM OF MSIV OCCURS 5.96E-04 IEV-SLB-D COMMON CAUSE FAILURE OF PMS AND PLS SOFTWARE 1.20E-06 CCX-SFIW 155 7.06E.02 CONSEQUENTIAL SGTR OCCURS 1.00E-02 OTH-SGTR INITIATING EVENT - IDSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 COMMON CAUSE FAILURE OF 4TH STAGE ADS SQUIB VALVES TO OPERATE 3.00E-05 ADX-EV-SA 156 7.06E-10

.02 CONSEQUENTIAL SGTR OCCURS 1.00E-02 OTH-SGTR INITIATING EVENT - LOSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 COMMON CAUSE FAILURE OF 4 IRWST IR7ECTION CHECK VALVES 3.00E-05 IWX-CV-AO 157 6.30E-10

.01 INITIATING EVENT - LARGE 14CA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE COMMON CAUSE FAILURE OF RECIRC LINES DUE TO SUMP SCREEN PLUGGING 1.20E-05 REX-FL-GP 158 6.30E-10

.01 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK 1.20E-05 IWX-FL-GP i

159 6.30E-10

.01 INITIATING EVENT - LARGE IOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSIZE COMMON CAUSE FAILURE OF RECIRC LINES DUE TO SUMP SCREEN PLUGGING 1.20E-05 REX-FL-GP j

160 6.30E-10

.01 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSIZE COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK

1. 20 E-05 IWX-FL-GP I

161 6.12E-10

.01 CONSEQUENTIAL SGTR OCCURS 1.00E-02 OTH-SGTR INITIATING EVENT - IDSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA 162 6.12E-10

.01 CONSEQUENTIAL SGTR OCCURS 1.00E-02 OTH-SGTR INITIATING EVENT - 14SS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CIDSE (SV/PORV) 2.10E-02 OTH-SLSOV1 f

i COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES 2.60E-05 IWX-EV-SA F

163 6.03E-10

.01 IEV-RCSLKC 5.02E-05 IEV-RCSLKC COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK 1.20E-05 IWX-FL-GP l

164 6.03E-10

.01 IEV-RCSLKC 5.02E-05 IEV-RCSLKC f

COMMON CAUSE FAILURE OF RECIRC LINES DUE TO SUMP SCREEN PLUGGING 1.20E-05 REX-FL-GP 165 5.93E-10

.01 INITIATING EVENT - LOSS OF CCW/SW EVENT OCCURS 1.44E-01 IEV-LCCW COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 166 5.89E-10

.01 INITIATING EVENT - LOSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND i

COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCX-XMTR I'

CCMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOFTWARE 1.10E-05 CCX-FMIMOD1-SW 167 5.89E-10

.01 INITIATING EVENT - LOSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-If0ND COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04.

CCX-TRNSM COMMON CAUSE FAILURE OF PMS ESP OUTPUT LOGIC SOF1 WARE 1.10E-05 CCX-IHXMOD1-SW y

b N

[

_an

.--ka A

3 g

3, 4,

5 Ema_.mai+-

-=--r-v----'r--'-- - - ' - - - - - ' - - - - - - - - - ' ' - - - - - - - - - - - - - - - - - - - - - - - - - --'

4==

.B 5

t 1

TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS 168 5.78E-10

.01 INITIATING EVENT - IARGE LOCA EVENT OCCURS 1.05E-04 IEV-LIDCA r

LLOCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE COMMO'i CAUSE FAILURE OF PMS ESF ACTUATION LOGIC SOFWARE 1.10E-05 CCX-PMXMOD2-SM

. 169

- 5.78E-10

.01 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA I

LIDCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE i

COMMON CAUSE FAILURE OF PMS ESF INPUT IOGIC SOFWARE 1.10E-05 CCX-IN-LOGIC-SW 170 5.7BE-10

.01 INITIATI193 EVENT - 1ARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA I

LIDCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSIZE f

{

COMMON CAUSE FAILURE OF PMS ESF OUTPUT IOGIC SOF"IWARE 1.10E-05 CCX-PMXMOD1-SW 171 5.78E-10

.01 INITIATING EVENT - IARGE IOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE COMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOFIWARE 1.10E-05 CCX-FMXMOD1-SW 172 5.78E-10

.01 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA h

LLOCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSIZE COMMON CAUSE FAILURE OF PMS ESF INPUT IDGIC SOFIWARE 1.10E-05 CCX-IN-LOGIC-SW b

173 5.78E-10

.01 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LIDCA BREAK SIZE -LOWER END OF BREAK SIZE 5.00E-01 BSIZE I

COMMON CAUSE FAILURE OF PMS ESF ACWATION 14GIC SOFIWARE 1.10E-05 CCX-PMIMOD2-SW j

174 5.53E-10

.01 IEV-RCSLKC 5.02E-05 IEV-RCSLKC COMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOP"IWARE 1.10E-05 CCX-FMXMOD1-SW

[

.[

175 5.40E-10

.01 INITIATING EVENT - CORE POWER EXCUR.iION EVENT OCCURS 4.50E-03 IEV-POWEX FAILURE OF EITHER PZR SV FAILS TO RECIDSE 1.00E-02 OTH-PRSOV I

COMMON CAUSE FAILURE OF STRAINERS IN IRWST TANK 1.20E-05 IWX-FL-GP y

176 5.40E-10

.01 INITIATING EVENT - CORE POWER EXCURSION EVENT OCCURS

' 4.50E-03 IEV-POWEX FAILURE OF EITHER PIR SV FAILS W RECIDSE 1.00E-02 OTH-PRSOV COMMON CAUSE FAILURE OF RECIRC LINES DUE W SUMP SCREEN PLUGGING 1.20E-05 REX-FL-CP 177 5.32E-10

.01 INITIATING EVENT - IOSS OF MAIM FEEDWATER EVENT OCCURS 3.35E-01 IEV-LMFW COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCX-XMTR COMMON CAUSE FAILURE OF 4/4 STAGE 2 & 3 LINE MOVs TO OPEN 1.10E-03 ADX-MV-GO h

OPERAMR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 178 5.32E-10

.01 INITIATING EVENT - LOSS OF MAIN FEEDWATER EVENT OCCURS 3.35E-01 IEV-LMFW i

COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM

[

COMMON CAUSE FAILURE OF 4/4 STAGE 2 & 3 LINE MOVs TO OPEN 1.10E-03 ADX-MV-GO

[

OPERAMR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 179 5.29E-10

.01 CONSEQUENTIAL SGTR OCCURS 1.00E-02 OTH-SGTR r

INITIATING EVENT - LOSS OF OFFSITE POWER EVENT OCCURS 1.20E-01 IEV-LOSP

[

FAILURE TO RECOVER OFFSITE AC PCWER IN 30 MINUTES 7.00E-01 OTH-ROS

~

FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 i

COMMON CAUSE FAILURE OF 4TH STAGE ADS SQUIB VALVES 'IO OPERATE 3.00E-05 ADX-EV-SA 180 5.29E-10

.01 CONSEQUENIIAL SGTR OCCURS 1.00E-02 OW-SGTR INITIATING EVENT - 10SS OF OFFSITE POWER EVENT OCCURS 1.20E-01 IEV-14SP FAILURE TO RECOVER OFFSITE AC POWER IN 30 MINUTES 7.00E-01 OTH-ROS FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-AO I

i i

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4 TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS t

181 5.01E-10

.01 INITIATING EVENT - ATWS PRECURSOR WITH SI SIGNAL OCCURS 2.05E-02 IEV-ATW-S COMMON CAUSE FAILURE CF REACTOR TRIP BREAKERS 8.10E-06 RCX-RB-FA OPERATOR FAILS TO MANUALLY ACTUATE ADS 3.02E-03 ADN-MAN 01 182 4.95E-10

.01 INITIATING EVENT - CORE POWER EXCURSION EVENT OCCURS 4.50E-03 IEV-POWEX FAILURE OF EITHER PZR SV FATLS TO RECIDSE 1.00E-02 OTH-PRSOV COMMON CAUSE FAILURE OF PMS ESF OUTPUT LOGIC SOFTWARE 1.10E-05 CCX-PMXMOD1-SW 183 4.83E-10

.01 INITIATING EVENT - SMALL LOCA EVENT OCCURS 1.01E-04 IEV-SIDCA COMMON CAUSE FAILURE OF TANK LEVEL TRANSMITTERS (IRWST, BAT) 4.78E-04 IWX-XMTR OPERATOR FAILS TO ACTUATE CONT. SUMP RECIR. (LEVEL SIGNAL FAILS) 1.00E-02 REN-MAN 04 p

184 4.61E-10

.01 INITIAT193 EVENT - LOSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-ICOND COMMON CAUSE FAILURE OF SENSORS IN HIGH PRESSURE ENVIRONMENT 4.78E-04 CCX-XMTR COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 185 4.61E-10

.01 INITIATING EVENT - LOSS OF CONDENSER EVENT OCCURS 1.12E-01 IEV-LCOND f

COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMDIT 4.78E-04 CCX-TRNSM COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 186 4.59E-10

.01 CONSEQUENTIAL SGTR OCCURS 1.00E-02 OTH-SGTR I

INITIATING EVENT - LOSS OF OFFSITE POWER EVENT OCCURS 1.20E-01 IEV-IDSP c

FAILURE 'IO RECOVER OFFSITE AC POWER IN 30 MINUTES 7.00E-01 OTM-ROS i

FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 COMMON CAUSE FAILURE OF 4 SQUIB VALVES IN RECIRC LINES 2.60E-05 IWX-EV4-SA 187 4.59E-10

.01 CONSEQUENTIAL SGTR OCCURS 1.00E-02 OTH-SGTR INITIATING EVENT - LOSS OF OFFSITE POWER EVENT OCCURS

'1.20E-01 IEV-LOSP j.

FAILURE TO RECOVER OFFSITE AC POWER IN 30 MINUTES 7.00E-01 OTH-ROS FAILURE OF A SECONDARY SIDE RELIEF VALVE TO CLOSE (SV/PORV) 2.10E-02 OTH-SLSOV1 f

, COMMON CAUSE FAILURE OF 4 IRWST INJECTION SQUIB VALVES 2.60E-05 IWX-EV-SA 188 4.53E-10

.01 INITIATING EVENT - LARGE LOCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE -IDWER END OF BREAK SIZE 5.00E-01 BSIZE COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 7

189 4.53E-10

.01 INITIATING EVENT - LARGE IDCA EVENT OCCURS 1.05E-04 IEV-LLOCA LLOCA BREAK SIZE - UPPER END OF BREAK SIZE 5.00E-01 BSIZE-LARGE i

COMMON CAUSE FAILURE OF OUTFLTP DRIVERS 8.62E-06 CCX-EP-SAM

[

190 4.46E-10

.01 INITIATING EVENT - STEAM LINE UPSTREAM OF MSIV OCCURS 3.72E-04 IEV-SLB-U

(

COMMON CAUSE FAILUPE OF PMS AND PLS SOFTWARE 1.20E-06 CCX-SFTW r

191 4.39E-10

.01 INITIATING EVENT - IDSS OF MFW TO ONE SG EVENT OCCURS 1.92E-01 IEV-LMFW1 l

COMMON CAUSE FAILURE OF SENSORS IN LOW PRESSURE ENVIRONMENT 4.78E-04 CCX-TRNSM 6

COMMON CAUSE FAILURE OF TANK LEVEL TRANSMITTERS (IRWST. BAT) 4.78E-04 IWX-XMTR OPERATOR FAILS TO ACTUATE CONT. SUMP RECIR. (LEVEL SIGNAL FAILS) 1.00E-02 REN-MAN 04

+

192 4.33E-10

.01 IEV-RCSLKC 5.02E-05 IEV-RCSLKC COMMON CAUSE FAILURE OF OUTPUT DRIVERS 8.62E-06 CCX-EP-SAM 193 4.27E-10

.01 INITIATING EVENT - CMT LINE BREAK EVENT OCCURS 8.94E-05 IEV-CMTLB r

COMMON CAUSE FAILURE OF TANK LEVEL TRANSMITTERS (IRWST. BAT) 4.7BE-04 IWX-XNTR OPERATOR FAILS TO ACTUATE CONT. SUMP RECIR. (LEVEL SIGNAL FAILS) 1.00E-02 REN-MANO4 f

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4' t

t TABLE 2 SENSITIVITY STUDY - TOP 200 CORE DAMAGE CUTSETS

(

1 194 4.26E-10

.01-FAILURE OF PRS RELIEF FDR LOSS OF MFW A WS, WITH UET 3.27E-01 OTH-PRESU INITIATING EVENT - A W S PRECURSOR WITH NO MFW OCCURS 4.81E-01 IEV-ATWS COMMON CAUSE FAILURE OF PMS REACTOR TRIP SYSTEM HARDWARE 7.89E-05

. ATW-MANO3 "CX-EMS-HA.W.E OPERATOR FAILS TO MANUALLY TRIP REACTOR VIA PMS 5.20E-02 FAILURE OF ROD CONTROL SYSTEM TO STEP IN RODS 6.60E-04 Rots-N -SYS I

195 3.88E-10

.01 INITIATING EVENT - CORE POWER EXCURSION EVENT OCCURS 4.50E-03 IEV-POtC1[

FAILURE OF EITHER P2R SV FAILS TO RECLOSE 1.00E-02 OTH-PRS *'N COMMON CAUSE FAILURE OF OUTP"T DRIVERS 8.62E-06 CCX-EP-SAM

.196 3.79E-10

.01 INITIATING EVENT - LOSS OF OFFSITE POWER EVENT OCCURS 1.20E-01 IEV-LOSP FAILURE TO RECOVER OFFSITE AC POWER IN 30 MINUTES 7.00E-01 OTH-ROS COMMON CAUSE FAILURE OF PRHR AOVs 9.60E-05 PXX-AV-LA t

COMMON CAUSE FAILURE OF CLASS 1E BATTERIES 4.70E-05 CCX-BY-PN 197 3.77E-10

.01 INITIATING EVENT - TRANSIENT WITH MFW EVENT OCCURS 1.40E+00 IEV-TRANS' COMMON CAUSE FAILURE OF SENSORS IN IEW PRESSURE ENVIRONMENT 4.78E-04 CCX-TPRSM COMMON CAUSE FAILURE OF RCP BREAKERS FAIL TO OPEN 4.20E-04 RPX-CB-GO OPERATOR FAILS TO RECOGNIZE NEED FOR RCS DEPR. (SLOCA/ TRANSIENT) 1.34E-03 LPM-MAN 01 198 3.63E-10

.01 INITIATING EVENT - MAIN STEAM LINE SWCK-OPEN SV OCCURS 1.21E-03 IEV-SLt:-V CONSEQUENTIAL SGTR OCCURS 1.00E-02 OTH-SGTR COMMON CAUSE FAILURE OF 4 IRWST INJECTION CHECK VALVES 3.00E-05 IWX-CV-AO 199 3.63E-10

.01 INITIATING EVENT - MAIN STEAM LINE SWCK-OPEN SV OCCL7tS 1.21E-03 IEV-SLB-V CONSEQUENTIAL SGTR OCCURS 1.00E-02 OTH-SGTR COMMON CAUSE FAILURE OF 4TH STAGE ADS SQUIB VALVES TO OPERATE 3.00E-05 ADX-EV-SA 200 3.32E-10

.01 INITIATING EVENT - ATWS PRECURSOR WITH SI SIGNAL OCCURS 2.05E-02 IEV-ATW-S FAILURE OF PRZ SV FOR LOSS OF MFW ATWS, NO UET 2.00E-03 OTH-PRES f

COMMON CAUSE FAILURE OF REAC'IDR TRIP BREAKERS 8.10E-06 RCX-RB-FA

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Attachmsat A to NSD-NRC-96-4913 Enclosed Responses to NRC Requests for Additional Information i

Re: IVR 480.440 480.441 480.442 480.443 480.444 480.445 l

480.446 480.447 480.448 480.449 480.450 480.451 480.452 480.453 480.454 480 4f5 480.456 480.457 480.453 480.459 480.460 480.461 Re: Baseline PRA Sensitivity Study OITS # 3969 i

)

3024A