JPN-90-055, Forwards Comments on Installation of Hardened Wetwell Vent at Plant,Per 891027 Response to Generic Ltr 89-16.Concludes That Hardened Vent Not Cost Beneficial & Consideration of Mods Be Deferred Until Individual Plant Evaluation Complete

Forwards Comments on Installation of Hardened Wetwell Vent at Plant,Per 891027 Response to Generic Ltr 89-16.Concludes That Hardened Vent Not Cost Beneficial & Consideration of Mods Be Deferred Until Individual Plant Evaluation Complete

ML20055H773
Person / Time
Site:
Issue date:	07/25/1990
From:	Brons J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20055H774	List: JPN-90-055, Forwards Comments on Installation of Hardened Wetwell Vent at Plant,Per 891027 Response to Generic Ltr 89-16.Concludes That Hardened Vent Not Cost Beneficial & Consideration of Mods Be Deferred Until Individual Plant Evaluation Complete
References
GL-89-16, JPN-90-055, JPN-90-55, NUDOCS 9007300048
Download: ML20055H773 (178)

Text

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• NowWrk 10601

= 914 681.G' l' h [ John C.B om E xecutive Via President Nuclear Generation July 25,1990 JPN 90055 U.S. Nuclear Regulatory Commission

. ATTN: Document Control Desk

- Mall Station P1 137 Washington, D.C. 20555

SUBJECT:

James A. FitzPatrick Nuclear Power Plant Docket No. 50333 installation of a Hardened Wetwell Vent

REFERENCES:

1. NRC Generic Letter 89-16, " Installation of a Hardened Wetwell Vent," dated September 1,1989,

2. NYPA letter, J. C. Brons to the NRC, dated October 27,1989 .

(JPN 89-070), providing the Authority's response to Generic Letter 8916.

3. NRC letter, T. E. Murley to J. C. Brons, ' Staff's Backfit Analysis for James A. FitzPatrick Nuclear Power Plant Regarding Installation of a Hardened Wetwell Vent," dated :

June 15,1990.

Dear Sir:

Generic Letter 8916 (Reference 1) requested that Mark I utilities vol'unteer to' install a hardened wetwell vent and provide a cost estimate for its installation; On October 27, 1989, the Authority responded to this Generic Letter for.the FitzPatrick (JAF) plant (Reference 2), providing an installation cost estimate of $680,000. The Authority also stated that it wanted to reserve any decision about Installing a hardened vent until JAF's IPE was completed.

On June 15,1990, the NRC staff issued its FitzPatrick backfit analysis.(Reference 3) and concluded that a hardened vent could be imposed as a backfit in accordance with 10 CFR 50.109. Reference 3 then requested that the Authority reconsider its previous decision and Inform the NRC within 30 days whether it will now commit to install a hardened vent.

The New York Power Authority (NYPA) has reviewed the potential benefits of a .

hardened wetwell vent for JAF in detall. The information in References 1 and 3 has been thoroughly considered. Portions of NYPA's plant specific IPE for JAF were accelerated out of priority to review the TW sequence in order to determine the reasonableness of voluntary action to install the hardened vent on an anticipatory basis. Although NYPA.

continues to affirm its willingness to make modifications judged cost beneficial under the severe accident management program following the IPE, all reviews to date indicate the -

value of a hardened vent at JAF to be exceedingly small. Plant specific factors leading to

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a this conclusion include the existing vent configuration, the relative slowness of the TW transient, the number of attemate successful courses of action, the low risk of vent use even if it fallect, and the extremely low probability of the TW event. NYPA continues to decline to install a hardened vent on a voluntary basis.

In the process of reviewing Reference 3, the Authority has discovered what appears - l to be a large number of flaws in the NRC's " plant specific" analysis in its support of -  !

requesting that the modification be installed on a voluntary basis. In addition,' our own plant specific analyses indicate that other accident management strategies are far more effective than the modification sought by the NRC. The attachments to this letter provide 4 I

detailed information on these points. The following is a summary of the more important points discussed in the attachments.-

. The existing wetwell vent at FitzPatrick is schedule 40 hard pipe throughout  ;

the reactor building, fully capable of withstanding the anticipated pressures. -l The vent path is " soft" only at the standby gas treatment trains which are . j outside the reactor building. Accordingly, operators are not hesitant to use the vent when called for, and the " averted costs' calculated by the NRC for - i damage in the reactor building are not applicable.  !

. If venting resulted in a rupture of a standby gas treatment train, the impact is that much of the energy released would be at grounc level rather than elevated. Some fraction of the energy release wouk' continue to be elevated as long as standby gas treatment fans continued to operate. In no case would .

the reactor building or plant personnel be affected, i

. The time to failure (without action) in the TW sequence is so long that  ;

opportunity to use alternate success path strategies is great. There are many 1 strategies available to extend the time to failure. Plant specific and generic ,

accident management strategies are identified which can prevent containment  ;

failure without venting. The NRC does not appear to have considered any of these generic or plant specific actions in its analysis. j

. The plant specific containment failure frequency for the TW sequence at - 1 FitzPatrick is 3.2E 7/RY. This is over two orders of magnitude less than the NRC figure calculated from non-plant specific data.

l

. The NRC's cost. benefit equation is unstable. This can produce totally lilogical l

. results. For example, for certain values of increasing vent installation cost, the  ;

formula shows a greater benefit / cost ratio.  !

t The industry is performing p! ant specific individual plant evaluations (IPE) as the l NRC requested. When the IPEs are complete, the ability will exist to examine the entire i risk profile for each plant. To impose a hardened vent at this time is inconsistent witNhe j l purpose of IPE and the Severe Accident Management Program. Plant modifications, if ' l l

l any, should be made In the context of the whole risk profile. The NRC's present rush to )

judgement would deny us tne full use of the insights to be gained from a completed IPE.  !

Even if it were shown that TW sequences were major risk contributors, it is not a l

foregone conclusion that reducing this risk must be done through use of a hardened i vent. '

The initiative on the part of the NRC to seek voluntary installation of a hardened vent has been disruptive to the Authority's planned accomplishment of the FitzPatrick l

l l IPE. Authority resources necessary to respond to the NRC's request have been significant. The short time frame permitted to respond to Reference 3 required a 3

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. _ __ _ . _ __ _ _ ~_ .

.. .. 3 reordering of analytical prioritiesL The Authority must reserve the right to correct any data presented in this response upo.i completion of the IPE and its planned review process. l Nevertheless, the data so overwhelmingly suggests that the hardened vent is not cost ,

beneficial that the NRC should be comfortable in deferring consideration of vent .

. modifications until the IPE is complete.

The Authority encourages, and would be pleased to meet with the NRC staff Involved with the review of this document during the review process. Should you of your

- staff have any questions regarding this matter, please contact Mr. J. B. Ellmers of my staff.

Very truly yours, s

[ohn . Brons ~i IExecutive Vice President -

Nuclear Generation ..

Attachments:

1 Authority Comments on the NRC's Hardened Vent Backfit Analysis  ;

ll NSAC 143 " Questionable Techniques Used in Cost Benefit Analyses of Nuclear Safety Enhancements

• lll . Hardened Vent Cost / Benefit Sensitivity Studies i IV Preliminary JAFIPE Results ,

cc: U.S. Nuclear Regulatory Commission -

475 Allendale Road l King of Prussia, PA 19406 Office of the Resident inspector.

U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093 Mr. David E. LaBarge Project Directorate I 1 1 Division of Reactor Projects-l/ll U.S. Nuclear Regulatory Commission Mall Stop 14 B2 i Washington, D.C. 20555 l

l

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ATTACHMENT l to JPN 90-055 AUTHORITY COMMENTS ON THE NRC'S HARDENED VENT BACKFIT ANALYSIS i

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J NewYork Power Authority ')

l JAMES A. F 12 PATRICK NUCLEAR POWER PLANT 1 Docket No. 50-333 -!

DPR 59 1

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. . - TABLE of CONTENTS L

l Attachment 1 ' Authority Comments on the NRC's Hardened Vent Backfit Analysis -

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1. General Comments 1 1

1 1.1 I nt rod uction . . . . . . . . . . . ... . . . . . . . . . . . . . .. . . . . .. . ... . .. . . . . . . . . . . . . .. . ... . . . . . . . .. . . . .. . .. . . . . . . . . .. . . . . . . . . . . . . . . .l.1 1.2 Policy i ssues . . . . . . .. . . . . .. . . . . . .. . . . . . . .. . . . . .. . .... ... . .. . . . . . . . .. . . . . . . .. . . . .. .. . . . . . . . . . . . .. . . .. . . . . . . . . . . . . . . .l.1. .......

2. Legallssues i

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2.1 Piant Specific Requirement ........ .................................................................. 12 2 ? Excess Cost s . . . . .. . . . . . .. . . . .. . . . . . . . . .. . .. . . . . . . . ... .. .. . . . . . .. . . . . . . . .. . . . . . . . .. . . . . .. . . .. . . .. . . . . . . . . . .. .. . . . . . . l.3. . . -. . .

2.3Altematives................................................................................................................ l.4 -

2.4 Additional Considerations ...... ........ ... ... ..... .... .. ....... ............. .... . ......... ........... ....... .. . ... l.4 2.5 Recom mend at ion . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . ... . . . . . . . . .. . . . . . .. . . . . . . .. . . . .. . . .. . . . . . . . . . . . . . .. .. . . . . ... . . . . l.6 . .. . . .

3. Source Term Considerations j a

3.1 I ntr od uction . . . . .. . . . . . . . . . . . . . . : . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 1-6 . . .. . . . . . . . . . . . .

3.2 U se of Drywell Sprays .. ... . . . . ..... . . ........... . . .. .... . .... .... . . . ... ........ ..... .... ......... .. ... ..... . . . ..... l7 I 3.3 A DS Ope r ati on . . . . . . . . .. .. .. . . . . . . . .. . . . . . . . .. . . . . . . .. . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . .l.7 ......... i 3.4 R eco mme nd ation . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .l.8 ..........

4. Cost / Benefit Analyses l 1

4.1 I nt rod u c t ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . .l.8 ....................

4 .2 I ncor rect U nit s . . . . . .. .. . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . .18 .................

4.3 Replacement Power .. ..... ............. .... ..... . .. . .. .. .. .... . ... . . ... .. . ...... ... . .... .. ..... . .... . ...... ....... . . l9 I 4.4 incorrect Cost / Benefit Equation ............................................................................... -l9 ,

4.5 Other Observations . ... . .. . ... ..... . . ... . . . .... ... .. ..... . .... .. ..... . ... .. ...... .. ..... . . . ... ... .. ..... .. . .. .. . .. .. . l.10 4.6 Present Worth Analysis . . . . . .... . .. ... ... .. ........ . . . .. ..... ........... . ..... . ....... . ... ............ ... . . . . ... . .. l.10 4.7 Value . I mpact Ratio . .. .. . . . . ... . .. . . ........ ..... .... .. . ... .. . . .... . ... . ...... . ... ...... .. . .... . . ..... . ... . . ... . . . l.10 4 .8 Rec om me nd ation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . 111 ...................

5. NYPA TW Severe Accident Management Analyses I 5.1 I n t r od ucti on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .l.11 ...m.............

5.2 Stretching Out The Time .. ... .. . . . . ... . .. . .... . .. ... .. . ... . ... .. . . . .. . . .... . . ... . .. ... . . ... .. ..... . .. .. .. . . . . . . . .. l.12 5.3 Altemate Heat Removal Pathways .......................................................................... 113 5.4 R ec om me nd ation . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . 113 . . . . . . .. . . . . . . . . . . . ,

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6. NYPAIPE Analysis 1 6.1 - I nt r od u ction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .114 ..................

6.2 Recovery of Loss of Offsite Power .......................................................................... l.14 6.3 Recovery of Power Conversion System .................................................................. 114 i 6.4 Other Differences in the Analysis of l Recovery Actlons . ..... ... ... . . ... .. . . . .. ... . . . . . . . . . . . . . .... . . . ..... .. . . . . . . . . ........ ..... . . .. . . . .. . .. 1 15 l 6.5 Addit!onal Recovery Cre6t . ....... . . . . . . ... . .. . . . . .. . . . . . .. . . . . ... . .... . ... . .. . .. .. ... . . ..... .. . ... . .. . . .. .... . . 115 6.6 Preliminary Results . .. ... ... .. . . .. . ..... . . . .. . . .. .. . .. .. . . . .. . .. . . . . . ..... ... . .... . .. . . .. . . . ...... ... . ... . . . . . . . ...... 116 6.7 Recom mendation . . . .. . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . .. .. . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . .116 . . . .. . . . . . . . . . . . . . .

7. N otes a n d R eferon ces . . . . ... .. . . .. .. . . . .. .. . . . . ... .. . . . . .. ... . . . . . . . . . . .. . . . .. . . .. . . . .. . . . . .... . . . . .. . . . . l.27 . .. .. . . . . . . . . .

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.. ' ..- TABLES c9d FIGURES 1

Table No.  ; Title page i

1 . Stretching Out the Time to Reach Containment )

Fallure..................................................................................................... 1 17 2 RWCU With Offsite Power Available ~ ................................................................ 118 3 Recovery of Loss of Offsite Power ................ ........................ l.19

.4 Recovery of Power Conversion System .......... .................................. 119 5 Recovery of DC Safety Bus System ..................................... ......................... i.20 6 Important TW Sequences Which May Lead to  ;

Containment Overpressure . ...................................................................... . - l 20 NRC Bolling Water Reactors with Mark l Containments Table 2 Backfit Analysis for Proposed Hardened Vent

'Capabilitys.......................................................................................................... 1 21 +

Figure No. Title page 1 TW Sequence - B ase Case . . . ..... ..... . .. . ... . . .. . ..... . . ... . .. . . ... . . . . . .. ... . . . ... . . .... ......... . . 1 22 2 TW Sequence - .

Etfect of Additional Pool Mass ............................................................. ............ F23

(! 3 TW Sequences -

~ Etfect of Low Pressure Venting ..................................... ...... ............................. l.24 ,

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, 4 Reactor Water Cleanup System -

l Blow Down Mode to Condenser ...................................................................... l.25 5 Vent Une . Transition from Hard to Soft .......................................................... 1 26 6 Cost / Benefit Ratio vs. Installation Cost ........................ ................................. Ill 5 17 Cost / Bene fit Ratio vs. C DF .. . . . ..... .. ... .. . ... .... . .. . . .. . .. . . . . . . . . . .. ... . ... . . . .. ... .... . . ..... Ill.5.... . .

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Attachment I to JPN+90-055 L

L AUTHORITY COMMENTS ON THE NRC'S HARDENED VENT BACKFIT ANALYSIS ,

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1.- General Comments 1.1 - Introduction -

The Authority has conducted a review of the NRC's June 15,1990 letter which attempts to justify a requirement to install a hardened vent at the James A. FitzPatrick (JAF) plant. Installation of the hardened vent is intended to reduce the environmental consequences of a severe accident involving loss of long-term _ decay heat removal capability (TW accident sequences). The staff's

' analysis contains serious errors and omissions. This analysis forms an inadequate basis for the staff to decide to backfit a hardened vent at JAF.

1.2 - Policy lasues -

The issues at hand involve much more than technical differences or the expenditure of funds.  !

Decisions about plant modifications should be made in the context of the total risk profile of a -

plant. For example, an IPE may reveal certain weaknesses in some plant system that affect'a number of postulated accidents. It may be decided to modify that system to overcome these weaknesses. Such a modification could'also reduce the frequency of TW sequences-being initiated and, consequently, the benefits of installing a hardened vent. Such a reduction may even  ;

make Installation of a hardened vent not cost beneficial.'

As discussed later, it appears that the Peach Bottom experience in modifying its electrical  ;

distribution system is an example of just this kind of process occurring, After its electric power distribution system was modified, Peach Bottom's hardened vent was no longer cost beneficial.

The licensee should have the choice of which plant modification is best for its facility, should the

. IPE risk profile indicate the need for plant modification. Yet the NRC's present " rush to judgement" would deny us the full use of the insights to be gained from a completed IPE. Even if it were shown that TW sequences were major risk contributors, it is not a foregone conclusion that reducing this risk must be done through use of a hardened vent.

l In Generic - Letter 88 20, Supplement No. 2 " Accident Management Strategies for 1 Consideration on the Individual Plant Examination Process," the NRC encouraged utilities to  !

develop accident management strategies because "the NRC believes that a significant _ risk l l

reduction benefit can be achieved, with reasonable resource expenditure, by implementation of ...

selected accident management strategies." The Authority fully endorses this strategy which makes use of the Integrated plant capability to prevent or mitigate accidents. Generic Letter.88-20, Sup. 2 acknowledges that existing regulatory requirements or other NRC actions may hinder utilities' implementation of effective accident management measures, and further states that

" licensees are encouraged to inform the NRC of such situations so that NRC can consider the need for further clarifications or modification of these regulations or guidance." As discussed in greater detall below, the Authority considers the current imposition of hardened vents to be an inappropriate regulatory action. The NRC should rethink its desire to irnpose a hardened vent backfit at FitzPatrick in light of the accident management and IPE progre.ms currently underway l within the industry. j L

1 11 I.__ - . -. - -.

.w.

New York Power Authority Attachment I toJPN 90055 i

2. Legallasues 2.1 Plant Specific Requirement The NRC's backfit analysis contained with its June 15, 1990 letter is not a plant specific analysis. This analysis, which purports to justify the installation of a hardened vent for FitzPatrick, is a generic analysis and not the plant specific analysis roquired by the 10 CFR 50.109 backfit rule and chapter 0514 of the NRC manual, "NRC Program for~ Management of Plant Specific Backfitting of Nuclear Power Plants." The staff's backfit analysis is merely an extrapolation of ,

information in documents the NRC previously developed to support the generic Mark I '

Containment Performance Program and references no specific technical analysis. of the FitzPatdek plant. Essentially, tne NRC has recycled a generic analysis in an attempt to justify a plant specific backfit.

10 CFR 50.109 section (a)(3) states that a backfit will be imposed on a nuclear power plant only when the NRC determines that 'there is a substantial increase in the overall protection of the public health and safety or the common defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection." [ emphasis added] This requirement for a plant specific analysis is reiterated in the NRC manual chapter 043 titled " Regulatory Analysis." It also was reiterated by <

the staff in the workshops conducted by the NRC on the backfit rule shortly after its revision in ,

1985. For example, at the May 8,1986 workshop conducted in King of Prussia, PA, Mr. Olmstead, then the NRC's Executive Legal Director, explained that a generic backfit analysis could be used to justify a plant specific backfit only if it was customized for the plant in question.

"If the generic backfit Is imposed by orders for your individual facility, 30.109 requires backfit analysis for that facility. The staff may rely on the IndMdual - on the generic-by-type [ sic) backfit analysis,' but you are free to raise your facility as an exception to that analysis, and we would require them to perfect that analysis before the order would go into place.'

The NRC staff subsequently affirmed at the same workshop that a I;censee is entitled to a plant specific backfit analys!s when a backfit is imposed on its plant with an order or license amendment. Because of various JAF plant specific factors such as the present unique hard pipe vent design and layout, and a modified fire pump system configuration, the FitzPatrick plant must be considered an exception to any generic backfit analysis.

Section 2.2 of the staff's analysis discusses the reduction in core damage frequency associated with hardened vents in Mark I containments. That discussion is based on a memorandum from Brian W. Sheron to Ashok C. Thadani, " Reduction in Risk from the Addition of Hardened Vents in BWR Mark l Reactors" (October 19,1989). That memorandum, however, merely grouped the plants with Mark I containments into four general categories - incorporating only a few plant specific characteristics such as licensed power level and the surrounding population. No plant specific technical analyses were performed. Page 1 of the enclosure to the Sheron memorandum confirms this, stating: 'It is entirely impractical, within the time frame and resources of the present study, to perform a plant specific analysis for each reactor. Rather, the approach taken was to group the BWR Mark I reactors into 4 categories ...." The NRC's backfit analysis further acknowledges that "[f]or FitzPatrick, the (core damage frequency) was estimated using the [probabilistic risk assessment (PRA)) results of a plant with similar features." There is no I2 m

New York Power Authority Attachment I to JPN 90-055 I i

discussion of how JAF differs from the plant the staff used, nor is there an assessment of the .

safety significance of these differences.

I Indeed, there are important differences at JAF that the staff's analysis did not account for. For l

example, the present vent path for FitzPatrick is believed to be unique in that it consists of hard piping from the containment to the inlet transition piece of the Standby Gas Treatment (SBGT) system filter train, located outside of the reactor building pressure boundary. Hard piping continues from the outlet of the SBGT filter train to the top of the main stack. This piping has a i pressure rating of 150 psig which is significantly above the postulated primary containment failure pressure. Vent failure at the transition piece will not damage plant equipment other that the SBGT liself, nor will it limit operator access to equipment contained within the reactor building. (See Figure 5.) Previous NRC concems that plant operators might not follow emergency operating i procedures for venting because of possible impact on plant access and equipment operability are, therefore, not applicable at JAF. The primary benefits gale 9d from installing a fully hardened vent are limited to protecting the SBGT filter trains and redirecthg a potential ground level release to an elevated release. Other JAF plant specific factors are discussed elsewhere in this Attachment.

There was no consideration of this vent's physical layout or construction, however, in the staff's backfit analysis. To provide the true plant specific analysis required by 10 CFR 50.109, the f

staff must consider the plant. specific configuration of the FitzPatrick vent and factor it into a backfit determination "for that facility." 10 CFR 50.109(c)(8) requires consideration of "the potential impact of differences in facility type, design or age on the relevancy and practicality of the proposed backfit." A generic mathematical analysis of core melt frequency for Mark I containments generally falls to meet this requirement, in the absence of a FitzPatrick plant specific analysis, it is not possible for the NRC to conclude, based upon its generic analysis, that a hardened vent for FitzPatrick would result in a reduction in core damage frequency of 4.5E 5 per reactor year. As developed In Attachment IV, preliminary NYPA JAF IPE results show a TW containment failure frequency of 3.17E 7 per reactor year, and a significantly lower core damage frequency. The NRC's figure is inappropriate for consideration of any FitzPatrick backfit and compromises the accuracy and reliability of the staff's subsequent cost benefit analysis.

2.2 Excess Costs '

The staff's backfit analysis has inappropriately considered certain costs. In particular the Authority points out that it is not appropriate to include averted costs in the backfit analysis.1, The consideration of averted costs requires the NRC to proceed beyond a reasonable delineation of its authority under the Atomic Energy Act. The backfit rule directs the NRC to consider in a backfit analysis "any other information relevant and material to the proposed backfit ...." However, the NRC is not a public utility commission. Its mandate is to regulate the radiological and environmental health and safety of nuclear power plants. It is not authorized to review the prudence of business judgements which are at the discretion of plant management and are not related to the public health and safety (e.g. economic costs of replacement power). Whether a proposed backfit is or is not justified, must be determined based on the safety benefit and the

" direct and indirect costs of implementation" which in 10 CFR 50.109(c)(5) includes continuing costs and plant downtime for installation.

Consideration of averted onsite property damage costs is especially inappropriate. The Authority is already required to insure the FitzPatrick plant against onsite property damage in the event of an accident.10 CFR 50.54(w) requires the Authority to hold an insurance policy with a

" minimum coverage limit for the reactor statien site of either $1.06 billion or whatever amount of I3

- _ _ - .. - . - - _ ~ - .- . . - --.

.' . New York Power Authority Attachment I to JPN 90-055 insurance is generally available from private sources.... 2 Therefore, since the Authority must purchase Insurance coverage for onsite property damage in the event of an accident, it would amount to double-counting for the NRC to include the averted costs for onsite property damage in its backfit analysis for a hardened vent at FitzPatrick. Additionally, vent overpressure at JAF is not expected to lead to onsite damage, except for the limited effects on the SBGT train and its enclosure.

2.3 Altematives, The NRC has failed to consider many valuable alternatives to installing a hardened vent, in its October,1989 letter to the Commission, the Authority identified a number of such attematives.

Section 50.109(a)(7) of the backfit rule contemplates the consideration of attematives to a backfit before it is imposed on a licensee. However, the backfit analysis prepared by the NRC to support a hardened vent for the FitzPatrick containment fails to consider other possible alternatives such as: 1) the currently proceduralized use of the Reactor Water Cleanup system, or 2) some other procedure or hardware modification at FitzPatrick, ,

Moreover, the staff has not provided an adequate Justification for why it believes that the issue must be resolved before completion of the FitzPatrick IPE. Since there is no near term safety concem, an acceptable alternative would be to wait for the comp'.etion of the IPE so that a decision on the need to install any modification can be made in light of the whole risk profile of the plant. Preliminary JAF IPE results do not warrant any hardware modifications due to TW event sequences.

2.4 Additional Considerations Reference 8 of the NRC's June 25,1990 letter provides information on how the family of Mark I plants were divided into four groups. This approach was based on placing plants with similar RHR configurations into the same group. For group 1 (Reactors without isolation condensers, two RHR heat exchangers per loop and two RHR pumps per loop capable of feeding the heat exchangers),

the q,eneric reduction in the TW core damage frequency, Acdf, is 2.3E 5/RY. However, plant specific analyses of Peach Bottom I and 11, both members of this group, yields a TW Acdf of 2.4E-6/RY when credit is given for the use of the control rod drive pumps. Here we see almost a factor of ten difference between generic values and plant specific values, all within the same group (Group 1),

( This wide spread between generic and plant specific values is critical. For example, the five other members of Group 1 (Brown's Ferry 1,2, and 3; Pilgrim; and Vermont Yankee) were judged to exceed the $1000/ person-rem averted criterion based upon the staff's cost benefit analysis and presumably would be subject to an order to install a hardened vent, if they did not volunteer.

However, Peach Bottom I and ll's cost benefit values of $543/ person rem averted were well below the $1000/ person-rem averted criterion.3 With generic values, Peach Bottom I and II's cost benefit value rises to $7245/ person-rem averted. So the difference between generic and plant specific values, all within the same group, could spell the difference between a backfit order or not. If Peach Bottom I and 11 did not have their own plant specific values and had to rely upon the staff's generic values, the licensees would be in a much weaker position to question the legitimacy of the NRC analyses. JAF has been assigned to group 2 with a Acdf of 4.5E-5/RY as a generic value. Yet JAF's preliminary plant specific TW containment failure frequency value is 3.17E 7/RY, and like Peach Bottom's, falls well below its. group generic number. Again, this " rush to l4

l New York Power Authority Attachment I19 JPN 90-055 Judgement' without a completed IPE would have denied the Power Authority full opportunity to establish the need or lack of need to modify JAF, let alone specifically to install a hardened vent.

As stated earlier, these four groups were established on the basis of different RHR system configurations. In spite of major RHR configuration differences, the staff's evaluation led to a spread among the group generic values that was rather small. All groups fell within a factor of 2.2 of the four group average value. The spread between the Peach Bottom plant specific number and its group generic number is much larger than the groupto aroup spread. JAF's TW .

frequency is more than two orders of magnitude smaller than its grey value. Furthermore, Peach ,

Bottom's plant specific TW Acdf value differs from its group velue nastly because of its AC/DC l power configuration, it is clear that other plant specific characteristics, like the AC/DC power configuration, can be more important than the RHR configuration differences which form the basis for establishing the different groups, it is possible that if plants were grouped according to their l AC/DC power configuration, completely different conclusions might be reached. Therefore, the l very process of forming these br groups is questionable.

The Sheron memoranda i raises some other very important points. It acknowledges "The analysis also assumes that core melt will occur on loss of containment cooling and containment failure. This is not necessarily the case .... Such considerations are very plant specific and would take extensive resources to investigate " The Sheron memorandum credited the operation of the control rod drive pump (located in Peach Bottom's turbine building) In its analysis of Peach Bottom and judged that the physical location of this system warranted lowering Peach Bottom's TW Acdf. Similar core cooling credit was not given to the other plants for their plant specific features, thereby potentially overestimating the importance of the TW sequences.

This "very plant specific" coro cooling issue is quite important to JAF. It is very unlikely that -

failures in the present JAF vent pathway will result in any loss of core cooling. As stated earlier, the JAF plant is already "hard piped' through a major portion of its vent pathway, it does not make a transition to " soft' sheet metal until it reaches the Standby Gas Treatment (SBGT) inlet expansion fitting, which is located outside of the reactor building pressure boundary. The SBGT's concrete enclosure is adjacent to the reactor building and is provided for physical protection and radiatien shielding. The consequences of a " burst" vent path at JAF might be the destruction of its SBGT train and blowout of the doors from the SBGT enclosure to the outside environment. Note also that this release is virtually nonradioactive. Figure 5 shows the SBGT equipment configuration and its relationship to the reactor building pressure boundary. A fully hardened vent is not needed to protect other equipment or personnel at FitzPatrick. The $680,000 cost estimate submitted by the Authority in its October,1989 letter was to hard pipe a bypass line around the SBGT, totally within its enclosure, to ensure an elevated release point. Without any modification, FitzPatrick already has a " ground level

• hardened vent.

Additionally, the JAF fire protection system has been modified such that the fire pumps can be aligned to cool the core, provide additional suppression pool Inventory or provide a containment >

spray function. The fire pump system is located in the plant's Screenwell area, north of the turbine building. This is on the opposite side of the plant from the SBGT enclosure. Because of the physical location of the fire pump system, it is expected to be operable even after vent failure or even containment failure. If anything, even greater core cooling credit than that given to Peach Bottom is due to JAF for this system configuration, but none was given.

Plant specific analyses require detailed reviews of actual physical layouts and the actual design characteristics of plant systems. Plant walkdowns, reviews of design documents, reviews of actual plant operating experience, and site characteristics are essential in a plant specific analysis. No such reviews were conducted by the staff for JAF.

t 15

. . NewYork Power Authority Attachment i13 JPN 90055 1

We conclude from the above that the staff has not performed the required plant specific  !

analysis of JAF and the group approach has not been demonstrated to be an acceptable I substitute. In fact, using group analyses can itad to incorrect decisions about the need to backfit l a plant, and even the type of backfit that is most appropriate.

Lastly, even if it were shown that a hardened vent at JAF had a cost benefit ratio greater than

$1000 per person tem averlod, there is the additional requirement that a substantial improvement in safety must result from such a modification. In the NUREG 1150 analysis of Peach Bottom, the change In the overall core damage frequency due to adding a hardened vent was limited to about a factor of 2.6. This rather small change in the core damage frequency is due to the fact that j venting does not diminish the contribution of other accident sequences, such as ATWS and i station blackout, to the overall core melt frequency. The importance of reducing the TW cdf can only be evaluated when the cdf contribution of all other accident sequences are also known, in addition to being ineffective in reducing the overall core damage frequency by a substantial amount, venting does not appreciably change the containment failure frequency for non/TW, internal event, sequences. For example, page 9-5 of NUREG 1150, Volume One states:

" Intentional venting of the containment was considered as an accident management strategy to prevent containment overpressurization failure of the containment for both Peach Bottom and Grand Gulf. The frequency weighted mean probability of ^

venting that precludes subsequent containment failure is approximately 10% at Peach Bottom and 4% for Grand Gulf. The values are small, mostly because of the high probability of early fa!Iure mechanisms for which venting is ineffective. "

A ten percent change in the internal event induced containment failure frequency is insignificant.

The Commission has not defined, numerically, what constitutes a substantial improvement in overall safety. However, such a judgement can only be made in the context of the whole risk profile. This requires a far more complete analysis than the one presented in the NRC's June 15, 1990 letter. A completed plant specific IPE would provide comprehensive data which can be used to evaluate the need for a hardened vent for that plant.

2.5 Recommendation The staff has neither provided the requisite plant specific analysis nor demonstrated a substantial ,

improvement in safety in its analysis. The Authority recommends that the staff make these determinations on the basis of the completed JAF IPE. )

3. Source Term Considerations 3.1 Introduction Assuming that TW sequences even lead to core damage, there appears to be a number of defects in the source term analyses used in the staff's June 15,1990 letter. These defects result in overestimating the person-rems per TW sequence release and, therefore, the benefits in the cost / benefit analysis.

l 1-6

New York Power Authority Attachment I to JPN 90-055 3.2 Use of Drywell Sprays j

Drywell sprays have a dual role in severe accident management. First, in sequences where  !

core damage and reactor vessel failure precede containment failure, drywell sprays appear to be j useful in reducing the probability of containment failure through drywell shell attack. Should a containment integrity be lost due to shell attack, the hardened vent then has no effect. '

Second, drywell sprays have the wider application of reducing the source terms in all core damage / reactor vessel failure sequences. One source term reduction process is to remove airborne radionuclides from the containment atmosphere. Additionally, a layer of water over the  ;

resulting "corium" would serve to reduce the likelihood of core - concrete interactions and would act as a filter for radioactive material that might evolve out of the corium, in SECY 89-017, page 10, it was judged that such drywell sprays would reduce the source term in the range of 2 to 10.

l Interestingly, the very use of driwell sprays to prevent a loss of containment integrity through shell attack and to enable venting to be effective, also reduces the benefits of venting through source '

term reduction.

Drywell sprays also reduce the source term in those unlikely situations where containment failure leads to a lose of ECCS capability and subsequent core damage and reactor vessel failure.  ;

As noted earlier, the FitzPatrick fire pumps can provide a containment spray function. Even if initial high drywell pressure limits the fire pump flow rate, after containment failure the contalnment i pressure should be low enough so that fire pump driven drywell sprays could fully inject water into the containment, thereby reducing the source terms and resulting offsite consequences. As stated before, the location of the fire pumps and their control systems at JAF make it highly unlikely that they would be affected by containment failure, The NRC's June 15,1990 letter overestimatos the FitzPatrick person-rems per TW release because it did not account for the use of dryweli sprays and, in particular, the use of fire pump driven drywell sprays.

3.3 ADS Operation Depressurization of the reactor coolant system prior to reactor vessel failure can also reduce the source term. This is because depressurization cools the intemal surfaces of the reactor system which results in better trapping of radionuclides. The Authority has calculated the impact of depressurization on the source term for a Peach Bottom station blackout sequence. These ,

results were presented at the NRC's Mark I workshop in Baltimore, Maryland and again through an NRCInvited paper at the 16th Water Reactor Safety Information Meeting.4 A source term reduction factor of about 10 was calculated. Witn this source term reduction due to depressurization, the person rems averted would be about 30% o' that used to estimate the ,

population exposure given in the NRC's June 15,1990 letter.

The Authority re:,ults parallel a statement in SECY 89-017, pg. 7 which states, " Enhanced RPV depressurization system reliability would also delay containment failure and reduce the quantity and type of fission products ultimately released to the environment". The Authority has already taken steps to enhance JAF's ADS capability by sigaificantly increasing the reliability of the ADS nitrogen supply. In TW sequences there is generally no concern with electrical power availability and preliminary reviews show that high containment temperatures that might affect the ADS cables would not occur before reactor vessel failure, i.e., not until well after the ADS function had been completed. If long term containment pressurization caused the ADS valves to close, this containment backpressure would be receved once containment integrity was lost. ADS functionality should then return upon reduction of the containment pressure. Because of the reduced decay heat production at the time when core cooling is assumed to be lost (i.e., after I7

New York Power Authority l Attachment I to JPN 90-055 containment failure), there is ample time for the ADS to operate prior to fuel damage. The ADS I will at:o be operable when the drywell sprays reduce drywell pressure below ti,e ADS closure i pressur:.

The NRC's June 15,1990 letter overestimates the source term by not accounting for the use of the ADS system to depressurize the reactor vessel prior to vessel failure, thereby reducing the source term.

3.4 Recommendation

• The staff's TW sequence person-roms should be recalculated to account for both of these reductions in the source term.

4. Cost / Benefit Analysee 4.1 Introduction The Commission's hardened vent cost / benefit analyses is incorrect in three ways. One cause of incorrect results, which is discussed above, is the use of improper input numbers; specifically, the use of invalid TW core damage frequency numbers and improper person rem values pet TW telease. Beyond these PRA related input errors there are economic factor input errors. The Authority previously pointed out in its October 27,1989 letter that the cost / benefit equation itself is faulty. Several examples of defects in the cost / benefit analysis are given below.

4.2 hcorrect Units '

Attachrt,Snt lit of the NRC's June 15,1990 letter on the staff's backfit analysis, SECY 89 017, and NUREGlCR 3568, 'A Handbook for Value-Impact Assessment *, use the fol!owing equations in their cost /banefit analyses:

Vph = NT (Op x R) and Vop = NdfU i

in the first equation:

Vph = Value of public health risk avolded for not benefit method (dollars)

N = Number of affected facilities T = Average remaining lifetime of affected facility (years)

Dp = Avoided public dose per reactor year (person rem / reactor year)

R = Monetary equivalent of unit dose (dollars / person rem)

Checking the units of the top equation yleids the following results:

(hian Rern) (doum) doHars = (years) (Reactor Year) (Man-Rem) or

"" (Rec to ' car)

These units do not balance because the above equation is missing a term, such as the availability factor, which has the dimensions of (reactor years)/ year. Assuming availability factor is considered, the top equation should be rewritten as:

l-8

___m

New York Power Authortty Attachment I to JPN 90055 Vph = NT(Dp x R)AF where:

AF = Availability Factor (reactor. year / year)

Implementing this correction would reduce V h by about one quarter. A similar error occurs in the equation for Vop and it too should be reduced by about one quarter. Both of these errors lead to overestimating the benefits in the cost / benefit equation.

4.3 Replacement Power it appears that the replacement power figure in the June 15,1990 letter was based on the FitzPatrick plant running at full power 365 days per year for the remainder of its operating life. The replacement power figure should be multiplied by its availability factor. This is necessary because replacement power is already needed when the plant is normally shutdown. Loss of the plat would only entall additional power purchases during the time when the plant would normally be expected to be at power. This error also leads to overestimates of the benefits in the cost / benefit equation.

4.4 Incorrect Cost / Benefit Equation Appendix A of the NRC's June 15,1990 letter, SECY 89-017, and NUREG/CR-3568 use this equation:

Cost Benefit = Averted Exposure Installation Cost . Averted Onsite Cost The Authority's letter to the NRC, " Installation of a Hardened Wetwell Vent", (JPN 89-070, October 27, 1989) pointed out that this equation is invalid, it can lead to both positive and negative values and becomes infinite should the installation cost equal the averted onsite cost.

The Authority recommended that the Commission refer to NSAC 143 " Questionable Techniques i Usod in Cost. Benefit Analyses of Nuclear Safety Enhancements", September,1989. This EPRI sponsored document also found this cost / benefit equation to be inappropriate. A copy of this '

EPRI document is enclosed as Attachment 11.

To further demonstrate the fact that this equation is in erro; we performed two sensitivity studies, in these studies, the availability factor was not modified a. suggested in sections 4.2 and 4.3 above. These sensitivity analyses can be found in Attachment 111 to this letter, in the first of those analyses we show that if this equation is used, then under some circumstances, increasing the assumed core damage frequency can lead to a smaller cost / benefit ratio. For example, at the assumed FitzPatrick TW core damage frequency of 4.5E 5/RY, the staff calculates an absolute value of the cost / benefit ratio of 15,386.5 If the TW core damage frequency had a slightly higher value,5.5x105/RY, and all other factors remaining unchanged, the value of the cost / benefit ratio decreases to 7,140. Common sense tells you that the cost / benefit ratio should increase when the core melt frequency increases, in the second case we show that in some circumstances increas'ag the Installation costs can result in a larger absolute cost / benefit ratio. For example, increasing the installation costs for FitzPatrick from $680,000 to $750,000, with all other factors unchanged, increases the absolute cost / benefit ratio from 15,366 to.44,910. Higher installation costs should lower th0 cost / benefit ratio, not raise it.

Lastly, the same absolute cost / benefit ratio can be obtained with multiple values of the core damage frequency. A particular cost / benefit ratio should be uniquely determined by a single value of the reduction in core damage frequency when all other parameters are held constant.

4 I9 j

New York Power Authority Attachrnent Ito JPN 90055 The same phenomenon occurs for the installation oosts. All of the above show that the staff's cost / benefit equation is illogical. The equation is fundamentally flawed 4.5 Other Observations There are several other observations that further support the fact that the staff's cost / benefit equation is incorrect and even ' unstable.* Referriry 'o Table 2 of the NRC's January 8,1990 report, ' Mark l Plant Specific Enhanced Venting Ce, Jility Regulatory Analysis" (attached on page 121) we examined the NRC's application of this formula to the total family of Mark I plants.

This table has both positive and negative coet/ benefit ratios which swing from 32,783 to

+18,730. Just comparing absolute values, the range goes from l32,783l to l522l, a ratio of 32,783/522 = 63. How could the family of Mark I's vary so much in their cost / benefit ratios?

Their power levels all fall into a narrow band. Their remaining lifetimes are within a factor of 1.6 of i

the average valuel the assigned Group TW frequencies are, with the exception of Peach Bottom which has a plant specific number, within a factor of 2.2 of the average value: the assumed installation costs fall withln a tight band with the highest value only 1.9 times larger than the average value. All of these plants are based on the same source term. Onsite losses includ; g replacement power, again except for Peach Bottom, are never more than 2.8 times the average value. The highest population (50 miles) site is only three times the average value.

The overall picture is one of general similarity among the cost / benefit input numbers for the family of Mark I plants. Yet when these generally similar Input numbers are substituted into the cost / benefit equation, the results vary enormously. JAF, which has a lower than average power level and a surrounding population that is but 40% of the average Mark I, ends up with one of the larger cost / benefit ratios (using absolute values). Generally similar input numbers should ,

produce a narrow range of cost /beaefit ratios.

Earlier we pointed out the need to multiply a number of input factors by the availability factor.

We have Irevestigated the impact on the cost / benefit ratio for FitzPatrick by just correcting Vop, Vph, and the replacement power cost figu o by multiplying them by an assumed availability factor of 0.75. When this was done, the JAF cost / benefit ratio changed enormously, from 15,366 to

+ 13,636. If the assumed availability factor was 0.86, the JAF cost / benefit ratio would be infinite.

We view this as yet another indication of the instability of the staff's cost / benefit equation.

4.6 Present Worth Analyses The stat!'s response number 6 in Appendix B of its June 15,1990 letter misrepresents what the Authority stated in its October 27,1989 letter (JPN 89-070). We specifically stated that we were not discounting health effects. The EPRI document, NSAC 143, offers a review of this issue and explains how the present worth process should be applied to future expenses, 4.7 Value Impact Ratio The NRC statt uses two economic indicators to judge whether or not a potential plant modification should be backfit, the cost benefit ratio and the value impact ratio. Defects in the cost benefit approach are discussed in section 4.4 and in Attachment 111.

The value-impact ratio is superior to the cost benefit ratio in that it avoids the pitfalls of including onsite losses, negative values, infinite values, numerical Instability, etc. Further, the value impact ratio is consistent with the approach recommended by the EPRI document, NSAC-l10

1

. . New York Power Authority Attachment I to JP6 30-055 i

143. In the Authority's view, the value impact ratio should be the sole economic indicator used in backfit decisions.

The value impact ratio, VI, is the monetized averted person-rom divided by the installation cost. Although the Authority supports the use of the value-impact ratio, one correction needs to )

be made to assure consistent units. As discussed earlier in Section 4.2, calendar years have to be  !

converted to reactor years by multiplying by availability factor. For example, with an availability  !

factor, AF, of 0.75 reactor years / year,25 more calendar years equal 25 x .75 = 18.75 reactor-  ;

years. The staff's estimated total person rems at JAF, using the assumed TW frequency of 4.5E- l 5/RY, changes from 1638 to 1228, for an AF of 0.75 RY/Y.

)

As described in Section 6 and in Attachment IV, the plant specific TW frequency for JAF is 3.17E107/RY. This is the frequency of containment overpressure and, based on arguments presented elsewhere, the JAF TW core damage frequency would be much smaller. i Conservatively equating the TW containment failure frequency to core damage frequency and taking no credit for source term reduction from either the drywell sprays or ADS operation (see Section 3), the JAF hardened vent total averted person rems for TW sequences is:

1228 = 8.65 person-rems At $1000 per person-rem averted, the worth of a hardened vent at JAF is limited to $8,654. l The value impact ratio for a hardened vent at JAF for TW secuences is: 8.65/.68 = 12.7 person rems per million dollars, )

1 1

4.8 Recommendation '

The NRC's cost / benefit equation can generate positive, negative, and even infinite results.

4 Since both costs and benefits can only be positive, non zero integers, the ratio of these two quantitles must be positive, as well as determinate. All of the above demonstrates that no backfit order can be reasonably issued that relies on the present staff cost / benefit methodology. We recommend that the NRC's valueimpact methodology be used to replace the cost benefit analysis now in use by the staff. In addition, this methodology should be corrected for the  ;

availability factor, and only plant specific TW cdf's be used. Since the JAF value impact ratio is belc.v 1000 person-rems /million dollars, a hardened vent at JAF is not justified.

5. NYPA TW Severe Accidsut Management Analysee 5.1 Introduction TW sequences are distinctive. Neither core cooling nor containment integrity could be lost until a long period of time has elapsed. Whereas in ATWS and station blackout sequences the important operator actions have to take place within a minutes to hours time frame, TW sequences are characterized by time frames stretching from many hours to-days. During this extensive time period there is no particular loss of plant status information, nor is there any loss of normal access to most equipment in the plant. Further, onsite electric power is assumed to I always be available and offsite power may be available also. Most difficuttles such as electric breakers that fall to close, most operator errors, miscalibrations, improper or undesirable plant configurations, etc., would be temporary and would be overcome well within the time available. l Restoration of any one of the several normal heat removal pathways to the ultimate heat sink terminates these sequences. Additionally, comparatively simple steps such as adding water to 1 11 l

1 j

I New York Power Autno;t/ l Attachment I to JPN 90-055 the suppression pool can be taken to appreciably stretch out the time to reach the containment overpressure condition. This additional time would enhance the likelihood of recovering one or more normal heat removal pathways.

, Recovery actions are not limited to restoring normal heat removal pathways. Alternate heat

, removal pathways can also be established. In all cases, such alternates would slow down the l

containment pressure rise rate, They would in many circumstances, also be capable of bringing the plant to thermal equilibrium and prevent containment overpressure. Steps, such as increasing

the amount of water in the suppression pool, are also valuable when used with alternate heat ,

removal pathways, e.g. more time would be available to establish such altemate pathways and '

the required heat removal rate would be smaller, i During the past year the Authority has been investigating the TW sequences. Part of this investigation has been an element of our JAF IPE development. The other part has been dedicated to thermal-hydraulic studies of hypothetical TW severe accident management strategies, it is recognized that these studies involve some actions which may go beyond, or are in conflict with current EOPs (and EPGs). The Authority views these analytical studios as useful

, methods to evaluate optional courses in its severe accident management program. Just as we believe it is premature to make judgements about the hardened vent, it is similarly premature to pursue implementation of these strategies with the NRC and industry groups before the entire risk ,

profile is understood on a plant specific basis. This is especially true since the risk of the TW sequence at FitzPatrick is very low without taking credit for the studied actions. Therefore, i

interpretation of some material in this section should be that it is or may be possible to even further reduce or prevent the risks of the TW sequence without the use of the vent. 1 2

Our TW severe accident management studies fall into two related efforts, those actions which can stretch out the time to reach containment failure pressure and those strategies used to establish attemate heat removal pathways.

I 5.2 Stretching Out The Time ,

Figure 1 on page 122 entitled *TW Sequences Base Case," is a MARCH code analysis of the JAF plant where no particular action is taken to recover from the TW sequence. Core cooling is provided throughout this time period and normal plant procedures, such as reactor depressurization upon high suppression pool temperature, are followed. This base case shows that it takes a long time,39.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, to reach the assumed containment failure pressure of 132

( psia. This assumed failure pressure is based on earlier NRC Mark I analyses. Figure 2 shows the I

impact of adding more water to the suppression pool, such as via the fire pump system. For an addition of 300,000 gallons, a containment pressure of 132 psia is not reached for 68.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (almost 3 days). This mass, and more, could be added without exceeding structural limits, flooding the wetwell vent line or the drywell/wetwell vacuum breakers.

Another strategy is low pressure venting of the containment through SBGT early in the accident sequence. In this analysis, containment venting was assumed to be initiated when the drywell pressure reached +2.0 psig and then terminated 900 minutes later. At the time of termination the pressure in the vent would be comparatively low and no failure of the vent pressure boundary, including the SBGT housing, would be expected. Figure 3 shows the impact of low pressure venting. This strategy would delay containment overpressure failure by about 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />.

Suppression pool water addition and low pressure venting can be combined. We did not analyze this combination, but estimated its effects.

l 1

112 l l

. . New York Power Authority Attachment I to JPN 90-055 i Results are summarized in Table 1 on page 117. We conclude that simple means exist to stretch out the time to reach containment overpressure conditions. TW events, which are already slowly evolving, can easily have their time to containment failure doubled. This provides additional time in which to recover systems or components which have prevented decay heat removal, i

5.3 Alternate Heat Removal Pathways l A number of possible aitomate heat removal pathways can be used to overcome many TW sequences. Examples include the use of: i

1. Reopening a pair of MSIV's in one steam line

2. RWCU -

3. MSiv Drain Unos

4. RHR in the Steam Condensing Mode

5. Suppression Pool in a " Food and Bleed" Mode

6. Combinations of the above These alternate heat removal pathways can be used in conjunction with the strategies to stretch the time out.

At this time we have mainly concentrated on the use of the RWCU, the reactor water cleanup system. Figure 4 is a simplified drawing of the RWCU in the configuration that was used in this analysis. This accident management configuration has the advantage of being a configuration that the plant operators use routinely. Blowdown to the condenser is a routine startup evolution, used to balance CRD flow into the reactor vessel. Even if an operator error was made in the valve lineup of this system, ample time exists to correct it. In order to analyze this system we extended our MARCH code to model this system.

TatAs 2 on page 118 is a synopsis of some of our MARCH code RWCU calculations. The impacts of pool water addition and the pressure to which the reactor was depressurized were examined. Our major conclusion is that the use of the RWCU system prevents containment overpressurization in all cases when offcite power is available, except when there is a primary system LOCA such as one or more stuck open relief valves. Even when no offsite power is available, significant benefits can be obtained from the RWCU system. At this time, we are examining possible success paths with the RWCU system that will also prevent containment overpressure even when offsite power is permanently unavailable.

5.4 Recommendation TW sequences represent a basic ' study" in plant recovery. Unless there is permanent damage, almost all unfavorable conditions should be recoverable within the time available.

Accident management with the equipment at hand, rather than design changes, resolves many TW accident sequences. Preliminary JAF IPE results show that the existing plant configuration results In a TW accident sequence probability at a level which already assures the health and safety of the general public.

l13

. . New York Power Authorfty Attachment I to JPN 90-055

6. NYPA IPE Analysis 6.1 Introduction A plant specific TW analysis is being conducted for the James A. FitzPatrick Nuclear Plant.

Major portions of this analysis are derived from the ongoing JAF IPE. Actual plant operating data, spanning the time period from January,1976 to December,1989, have been utilized. Plant systems, and their support systems have been modeled using a fault tree methodology. Plant specific component failure data were derived from FitzPatric'K operating experience, and generic common cause failure beta factors have been taken from EPRI NP3967

• Classification and Analysis of Reactor Experience involving Dependent Events.' Using the SETS code, accident sequences are being quantified by combining the boolean equations derived from the system failure modes associated with these sequences, and the resultant equations reduced to form minimal cutsets. The dominant minimal cutsets were then reviewed to provide the basis for both identifying and quantifying recr*/ cry actions. l The following informebon has been extracted from this effort, it is preliminary and subject to change. There are a r'Jmber of areas where JAF's IPE analysis of TW sequences are similar to the June 15, 1990 NRC values. However, there are also differences. A partial list of such differences is given in the following sections.

6.2 Recovery of Loss of Offslie Power (LOSP)

A TW sequence initiated because of a LOSP can be recovered by the restoration of that power. With offsite power restored, the operator will have the Power Converdon System (PCS) available for decay heat removal. Therefore, for those cutsets where offsite power restoration is appilcable, a non-recovery term la incorporated for failure to restore offsite power. The values of probability of non recovery of offsite power are given in Table 3 on page 119. The values were derived on the basis of the methodology presented in NUREG/CR 5032, *Modeling Time to Recovery and initiating Event Frequency for Loss of Offsite Power locidents at Nuclear Power Plants,' and NUREG/CR 4550, Vol. 4 ' Analysis of Core Damage Frequency, Peach Bottom, Unit 2.'

The staff selected a 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> non-recovery factor. However, et 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> the drywell pressures and temperatures are below design basis accident conditions. As shown in Section 5 above, adding water to the suppression pool delays the containment pressure / temperature buildup. With a 300,000 gallon addition it would take over 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> to reach design basis accident pressure / temperature conditions. Even without adding more pool water, it takes longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to overpressurize the containment. Therefore, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> recovery period is fully justified.

NYPA's use of a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> LOSP non recovery factor of 4E 3 is appropriate and conservative.

The staff used the overly conservative 13 hou non recovery factor of 1.3E 2 in its analysis. The more appropriate 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> value is 31% of the .3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> value and, therefore, LOSP TW sequences are correspondingly less important that those calculated by the staff.

6.3 Recovery of Power Conversion System (PCS)

(

i A TW sequence initiated because of a loss of the PCS can be recovered by restoration of PCS.

Therefore, for those cut sets where PCS restoration is applicable, a non-recovery term is incorporated for failure to restore PCS. The values for probability of non recovery of PCS are given in Table 4 on page I 19. The JAF 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> PCS non recovery factor is 7E 4. The data are l 14

i

. . New York Power Authority

Attachment I to JPN 90455 from NUREG/CR-4550, Vol.1, ' Analysis of Core Damage Frequency

Intemal Events

! Methodology.'

Note that the staff used a 13 hour1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> PCS non-rooovery factor of 1E 2. Using the more appropriate 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> non recovery factor as discussed above results in a PCS non-recovery facter of only 7% of that used by the staff. Use of the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> figure significantly reduces the importance of TW sequences caused by loss the PCS.

6.4 Other Differences in the Analysis of Recovery Actions There are several other areas in which the JAF IPE analysis of TW sequences differs from the group analycis performed by the NRC staff. A partlaf list of such differences are discussed below.

The staff used a valve non recovery factor of 0.3. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> valve nonrecovery factor used in the JAF IPE is 0.1.

In allloss of feedwater sequences, the NRC analysis assumed MSIV closure. In the FitzPatrick l plant design, MSIV closure would not occur unless there was a simultaneous loss of both the HPCI and RCIC systems concurrent with the loss of foodwater.

In TW sequences that initiate from a failure of a safety AC bus (TAC 10500 and TAC 10600 sequences), the JAF IPE takes credit for the use of the power conversion system since the MSIVs would remain open.

The staff uses a DC power non-recovery factor of 0.25. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> non-recovery factor used in the JAF analysis is 1E 3. A TW sequence initiated because of the loss of a DC safety bus can be recovered by the restoration of DC hardware. Therefore. for thm cutsets where DC bus restoration is applicable, a non-recovery term is incorporated for a failure to restore a DC bus.

The values for probability of non recovery of DC hardware as a function of time are given in Table i 5 on page 120. The data are from NUREG/CR-4550, Vol.1, ' Analysis of Core Damage  !

Frequency. Intemal Events Methodology.'

6.5 Additional Recovery Credit Although credit has been taken for some recovery actions, additional credit may be warranted.

Several areas already appear promising.

The generic beta factor approach has been used to estimate common cause failure probability. However, this approach is probably too conservative to be used in TW sequences.

Many of the contributors to the beta factor are completely recoverable in the very slow moving TW sequences. Application of recovery analyses to the beta factors would reduce the importance of common cause failures.

Present staff and Authority analyses do not account for the JAF fire pump system to overcome certain common cause failures, such as overcoming common cause RHRSW pump failures.

Present analyses may not fully account for defeating the logic system that controls MSIV closure, as appropriate. This could lead to reopening of a main steam line through one pair of MSIVs and transferring steam to the condenser. This could be a very valuable recovery technique for sequences involving a stuck open safety relief valve. Additionally, stuck open safety relief valves may be closed by operator action in the control room (manipulating the control switch) or by pulling that SRV's control power fuse. Raising the containment pressure or reducing the reactor pressure may also force SRV closure as the pressure differential between the reactor and the containment falls below 50 psid. No credit was taken for these recovery actions.

115

- - New York Power Authortty Attachment I to JPN 90055 '

The staff and the Authority have not quantified actions taken to stretch out the time to reach containment overpressure conditions. This additional time increases the likelihood of accident recovery. Neither has the staff or the Authority quantified the establishment of alternate heat removal pathways which could terminate the TW accident sequence. 1 The above comments are not meant to be exhaustive. However, they clearly indicate that present evaluations of the TW sequence have undervalued recovery actions and therefore  ;

overestimated the TW frequency. l 6.6 Preliminary Hesults ,

Table 6 on page 120 identifies a number of important TW sequences and their calculated frequencies. Note that these are the frequencies of causing containment overpressure and are not core damage frequencies. As stated earlier, it is highly probable that TW Induced containment failures will not lead to a loss of core cooling at JAF. The total containment failure frequency from TW sequences at JAF is 3.17x10-7/RY. The TW core damage frequency is some fraction of this number.

Even assuming the JAF core damage frequency is equal to the containment failure frequency, it is clear that the NRC's group 2 generic core damage frequency of 4.5x10-5/RY,when compared to the JAF coro damage frequency, significantly overpredicts the importance of JAF TW i sequences. Nots Bat, like the staff, the Authority is using the expressions ' core damage i frequency" and ' core melt frequency" interchangeably. This is conservative since not all core damage events lead to core melt, let alone reactor vessel failure or containment failure. i it should be recalled that credit for other recovery activities, as discussed in section 6.5, have not been quantified. Application of these additional recovery factors lead to even lower JAF TW containment failure frequencies. It should also be recalled that even if a TW sequence did lead to containment failure and core damage, the source terms would be smaller that those used by the ,

staff (See section 3).

Detailed preliminary information on the Authority's JAF TW analyses can be found in Attachment IV.

6.7 Recommendation Before any hardened vont order is issued, the staff and the Authority should meet to review the Authority's JAF TW analyses.

t 116 1

. . New York Power Authorhy Attachmert I to JPN-90-055 Table 1 l

STRETCHING OUT THE TIME TO REACH CONTAINMENT FAILURE (March Code Analyses)

A. Increasethe Suppression Pool Mass Containment Pool Mass, Failure Time, 8

Condition (10 Pounds) (Hours)

Base Case 6.62 39.2 Base Cate 7.86 56.7

+ 150,000 gallons  ;

Base Case 9.12 68.3

+300,000 gallons B. Use Low Pressure Drywell Venting Venting Ouration, Containment Failure Condition ' (Minutes) Time, (Hours)

Base Case 0 39.2  !

Low Pressure 900 50.0 ,

DrywellVenting C. Combine A and B, above -

Venting Duration, Containment Failure Condition (Minutes) Time, (Hours)  ;

Base Case 0 39.2 Base Case + 900 67 (Estimate) 150,000 Gallons +

Low Pressure DrywellVenting Base Case + 900 79 (Estimate)  ;

300,000 Gallons +

Low Pressure DrywellVenting -

117 l

_ . _ _ . , -_ .___. _ _ . , _ _ . , . _ _ . . . . ~ . . - ._. . _ _ _ . . - . . _ ,

New York Power Authority Attachment I to JPN.90-055 i

i Table 2 ,

RWCU WITH OFFSITE POWER AVA!t.ABLE (March Code Analyses) l A. Depressurize Reactor to 65 PSIA (Applying HCTL Curve Operating Procedure)  !

RWCU Flow PoolWater Containment Time of 1 Rate, GPM Added, (Gal.) Peak Press., (PSIA) Peak, Hrs, j 180 0 84 75 l 180 150,000 80 100 I

180 300,000 77 117

+ Low Press. Vent B. Effect of Umited Reactor Depressurization RWCU Flow PoolWater Reactor Containment Time of i Rate Added Pressure, Paak Press., Peak, (GPM) (Gallons) (PSIA) (PSIA) (Hrs) 180 0 65 84 75 180 150,000 150 78 117 180 0 400 66 100 I

c l18

, -. y . . ~ . + -4,, m.. - , . - - . , - - , . , , , .+-.,e- , .- -

New York Power Authority

]

Attachment I to JPN 90055 1 1

l Table 3 l l

RECOVERY OF LOSS OF OFFSITE POWER .

l Time to Recover Probability of Non Recovery 1 30 min. 0.306 48 min. 0.224 60 min. 0.192 68 min. 0.176 '

77 mir- 0.161 80 min. 0.157 l 150 min. 0.096 180 min. 0.081 190 min. 0.077 5 hrs. 0.048

7 hrs. 0.032 .

8 hrs. 0.027 9 hrs. 0.023 10 hrs. 0.020 13 hrs. 0.013 15 hrs. 0.010 i 16 hrs. 0.009 18 hrs. 0.007 23 hrs. 0.004 27 hrs. 0.004

.. .... .. ..... ... .. ............ .. ....... ...t. . . ..... .... .... ......

Table 4 RECOVERY OF POWER CONVERSION SYSTEM

• l Time to Recover Probability of Non Recovery l

5 10 min. 1.0 10-20 min. 1.0 20 30 min. 1.0 30-40 min. 0.9 40-60 min. 0.6 60 70 min. 0.4 24 hrs. 0.06 24 hrs. 0.0007 l

l19 l l

New York Power Authorny Attachment I to JPN 90-055 Table 5 RECOVERY OF DC SAFETY BUS SYSTEM Time to Recover Probability of Non Recovery 5

• 0 min. 0.9 1%" min. 0.9 20-3L+ 'n. 0.8 30-40 min. 0.7 40-80 min. 0.6 60-70 min. 0.6 70120 min. 0.4 24 hrs. 0.1 4-6 hrs. 0.05 6-8 hrs. 0.01 812 hrs. 0.002 .

24 hrs. 0.001 Table 6  !

IMPORTANT TW SEQUENCES WHICH MAY LEAD TO CONTAINMENT OVERPRESSURE Initiating Events Sequence Frequency * (RYJ)

T1 5.76E 8 T2 1.27E.7 T3A 4.08E-9 T3B 3.01E 9 T3C < 1.0E 9 TAC 10500 1.24E 9 TAC 10600 < 1.0E-9 '

TDC-A 2.72E 8 TDC-B 3.05E 8 A 1.76E 8 S1 3.96E 8 S2 8.82E 9 i

Total 3.17E 7

• Each initiator frequency is calculated by summing the Individual dominant (>1.0E 9) sequence frequencies for that initiator.

These containment failure frequency probabilities are preliminary and include recovery

- factors.

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• New York Power Authorny Attachment Ito JPN 90066

7. Notes and References

1) The NRC's draft safety goal policy Indicated that the staff would not use averted costs in ]

backfit determinations. See Transcript of Ragion i Backfitting Workshop at 177 78, Transcript of Region 11 Backfitting Workshop ct 46-47. 1

2) In its promulgation of this regulation, the NRC recognized that the $1.06 billion figure was clearly adequate to ensure decontamination and decGiwsssioning to the extent necessary to protect public health and safety. See 52 Fed. Reg. 28,963 and 28,964 (1987). l l

3) See Table 2, Column 0,

• Mark l Plant Specific Enhanced Venting Capability Regulatory 1 Analysis," Jan. 8,1990, attached. Note that this table uses a Peach Bottom I and ll TW Acdf of 3.6E-6/RY Our reference to this table is not an endorsement of the staff's cost / benefit analyses.

4) See *ls Mark l Shell Failure Really important?' H. Spector (NYPA) and P. Bientarz (RMA),

NUREG/CP 0097, Vol. 5, pages 503 524, March 1,1989.

5) Note that negative cost benefit ratios are, in effect, treated as absolute values in the NRC  !

cost benefit analyses.' (See NUREG/r,R 3568) We continue to believe that negative cost / benefit ratios and infinite cost /t anefit ratios are not correct and are the result of an j improper cost / benefit equation.

t l

I l 27 l

i i

ATTACHMENT 11 to JPN-90055 l NSAC-143 i OUESTIONABLE TECHNIQUES USED IN COST-BENEFIT l ANALYSES OF NUCLEAR SAFETY ENHANCEMENTS I I

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3 i

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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333  :

DPR 59 .

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ATTACHMENTlli to JPN 90055 HARDENED VENT COST / BENEFIT SENSITWITY STUDIES i

NewYork Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50333 DPR 59

New York Power Autherfty Attachment 11113JPN 90-055 l

JAMES A. FITZPATRICK NUCLEAR POWER PLANT HARDENED VENT COST / BENEFIT SENSITIVITY STUDIES 1

This attachment provides sensitMty studies of the Cost / Benefit ratio analysis contained in Appendix A to the June 15,1990 NRC letter.

The method for performing these Cost / Benefit sensitivity studies is to vary one parameter in l the cost / benefit equation within a range of interest while holding all other parameters constant.

The two parameters selected for these sensltivity studies are the hardened vent installation cost and the reduction in core damage frequency attributed to the hardened vent.

For the purpose of these sensitivity studies, the data are those provided by the NRC in Appendix A, Table 2, Column O of their June 15,1990 letter. These studies do not apply the )

availability f actor correction identified in Section 4.2 of Attachment 1.

I NRC Staff's Cost / Benefit Equation '

The cost / benefit of installing a modification is expressed as the ratio of the averted public exposure to the difference between the installation cost of the modification and the averted on site i cost (including replacement power costs). This formula is expressed as follows:  !

Cost / Benefit =  %

instl cost - (Vo p + Rept pwr)

Where:

Vpg = value of public health risk avoided for not benefit method ($)

insti cost = cost to Install modification ($) ,

Vop = discounted cost of onsite property damage ($) t Rept pwr = replacement power costs ($)

These terms are calculated as follows:

Vpg = NT (Dp x R)

Where: i l

N = Numberof affectedfacilities(1) l T= Remaining lifetime of affected facility (25 years) l Dp = Avolded public dose per reactor year (a core damage frequency x averted i offsite exposure (1.46 x 108 person-rem))

Mbnetary equivalent of unit dose ($1000 per person rem)

Vop = NdFU Where:

N= Number of affected facilities (1) dF = reduction in core melt frequency (per reactor year)

U= present value of onsite property damage occurring with frequency F ($/RY)

And: U = (C/m)[(e'"(0)/r2 )(3,, r[t(na(n)j(j,,e) l l

1 page 1111

0 ' Q

' New York Power Authority Attachment lil to JPN 90055 Where:

C = cleanup and repair cost ($1.0 billion) m = period of time in which damage costs are paid (10 years) r= discount rate (10%)

t(i) = years before reactor begins operation (0) t(f) = years remaining in plant lifetime (25 years) .

For the variables of interest in this sensitivity study, the Cost /Bonefit equation reduces to:

(1) Installation cost CB= $1,642,500 (instlcost $786,575)

(2) Reduction in Core Damage Frequency attributed to the hardened vent (CDF)

CB= $3.65

$6.8 $1.7479 Db7 Where CDF is expressed in terms of 10-5 Base Case Base case assumptions:

Installation cost = $680,000 Reduction in CDF = 4.5 x 10~5 per reactor year, in equation (1)

CB= $ 1,642,500

_($680,000 $786,575)

= $ 1,642,500 S 106,575

= 15.41 in equation (2)

CB= $3.65

$6.8 $1.7479 4.5 j

= $3.65

$1.5111 $1.7479

= $ 3.65

$-0.2358

= 15.41 page ill 2

e .

New York Power Authority Attachment til to JPN=90-055 Thir, shows agreement betwem the two formulas applied to the sensitivity study. The NRC C-B r8io is given in Table 2 as it,X6. (The factor of 1000 difference is due to normalizing to  ;

different values. The NRC's C-B is OMdered beneficial at a value of >1000 and the Authority's C-B is considered beneficial at >1.) ihe minor difference in the nutneric value appears to be related to truncation errors.

installation Cost Sensitivity Analysis Figure lil 1 and the following table show the change in Cost / Benefit ratio when the installation cost of the hardened vent is varied from $500,000 to $1,000,000. All other parameters are held constant.

Installation Cost Cost / Benefit Ratio

$500,000 5.731

$000,000 -6.803

$680,000 (Base case) 15.412

$700,000 18.972

$750,000 44.908

$800,000 122.346

$900,000 14.481

$1,000,000 7.696 1

When the installation cost of the hardened vent equals $786,575 the Cost / Benefit ratio is ]

Indeterminate. Note that when the installation cost increases from the base case of $680,000 to

$750,000, the absolute cost / benefit ratio actually increases in value. A higher installation cost should lower the cost / benefit ratio.

Core Damage Frequency Sensitivity Analysis Figure 1112 and the following table shows the chan in core damage frequency is varied fromto2.0 6.0 xx 10-5 10 geper in Costyear.

reactor / Benefit All otherratic when parameters are held constant.

Core Damage Frequency / RY Cost / Benefit Ratio 2.0 x 10-5 2.21 2.5 x 10-5 3.76 3.0 x 10-5 7.04 3.5 x 10 5 18.72 page lil 3

New York Power Authorfty Attachment 111 to JPN-90055 Core Damage Frequency / RY Cost / Benefit Ratio 4.0 x 10-5 76.20 4.5 x 10'S (Base case) 15.41 5.0 x 10-5 .g,41 5.5 x 10'S 7.14 6.0 x 10 5 5.94 When the reduction in core damage frequency equals 3.89 x 10-5/RY the Cost /Bonefit ratio is indeterminate. Note that when the core damage frequency increases from the base case value of 4.5 x 10 s/RY, the absolute cost / benefit ratio decreases. A larger oore damage frequency should .

Increase the cost / benefit ratio, l Conclusion l These sensitivity studies consisted of determining the change in the Cost / Benefit (C B) ratio i for installation of a hardened vent when two parameters (hardened vent installation cost and reduction in core melt frequec.7) are varied one at a time, within reasonable limits from the nominal values contained in the June 15,1990 NRC letter. Contrary to the intuitive understanding that the absolute C/B ratio of a modification should increase with decreased installation cost and increase as the core damage frequency increases, the opposite can happen. These studies show that the change in C B is relatively insensitive to these parameters and is primarily driven by the difference between the installation cost and the averted on site cost. When these two values approach each other, the C B ratio tends towards infinity (either positive or negative) regardless of either the cost of the modification or the reduction in core damage frequency. Both of these 86nsitivity studies show that if only the absolute value of the cost benefit ratio is considered, then multiple values of the core damage frequency (and insta!!ation cost) can yield the same cost benefit ratio. The C B equation is fundamentally flawed and should not be used as the basis for determining the desirability for installing a plant modification.

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i page lll-4

i Figure 6 COST /CENEFIT RATIO VS. INSTALLATION COST u-e e- l  ;

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ATTACHMENTIV to JPN 94055 PREUMINARYJAF IPE RESULTS l

l i

NewYork Power Authority JAMES A. FITZPATRICK NUCLEAR POWER ?LANT Docket No.50 333 DP i

e

TABLE of CONTENTS Attachment IV' Preliminary JAF lPE Results-pa22 )

1.

Introduction...................................................................................................................... :1 i

2. Accident in itlators ................................. ............ ..... .... ........ ..... . ....... . ... . .. .. . ...... .... ........ ... 1. .

3. , Accident Sequence Development and Quantification ............................................... 3 I

T 1 Ind uced TW Sequences ....... .... . ....... . ... . . .............. . ......... ...... .. ........ .... ... ... ...... ... ... .... . . .. 5 T2 I nd uced TW Sequences ... .. ..... . ......... .. ............. ..... ...... .. ........ . ... ... ..... . ... .... . ........ ........ .. )

8 )

T3A Induced TW Sequences .....................= ........................................................ 11 T3B induced TW Sequences ...... ..... .... ... ............. ....... .... .... ....... ......... ...... ... .. ... .... .... .. . .. ... 11  !

T3C I nduced '"WJeq uences . ... .... .... ....... . ...... .. . ... ....... ... .... . ... ..... ...... ................................, 11 j TAC.10500 I nN4 TW Sequences .... . .. ...... .. .......... . ..... . .. .. ..... . . .. ... .. ..... . . . .. ..... ......... .... ... 12 i TAC.10600 Ino -.;eu TW Seq uences ........... ....... .. ....... ....... . .... . . .... .. . .... ..... . . .. . . .. ............ .. . . 15  !

TDC A I nduced TW Seq uences ............................ .... .... ..... . . ...... .... .. . . ... . .... . ... . . . ... .. ....... ... . . 20  ;

T DCB I nduced TW Sequences . ... ....... ............. ... .... .... . . .. .. ... ... . . . .. .. ...... . . . .. .. . . .... ... .... ......... 24 Large LOCA (A) Induced TW Sequences ......................................................................... 26  ;

intermed'. ate LOCA (S1) Induced TW Sequences ............................................................ 27 i Small LOCA (S2) Induced TW Sequences ....................................................................... 29 j i

4. R ecov ery An al ysi s . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. ... . .. . . . . . .. . . .. . . . . . . . . . . . . . . . . 33 . . . . . . . . . . . . . . . l. . . . . . .. .

Recovery of Loss of Offsite Power (LOSP) . ....................................................................... ~33 Recovery of Power Conversion System (PCS) ................................................................ 33  !

Recovery of DC Salety Bus System ................................................................................ 34- _I Probability of Fallure to Repair Components .................................................................... 34 Probability of Failure to Align Failed Components - ........................................................... 35 Determination of HEP for Recovery of Failure to Prevent-Automatic Switchover of HPCI from CST to Torus ................................................... 37 ",

5. '

Concluslons..................................................................................................................... . - 39 i

..... ........ . .... .. . ................... ...... .......... .. ..... . . .... ... . . ..... ... ........ t FIGURES and TABLES Figure No. Title  !

PEle 1 Observed Repair Times and Theoretical Distribution -

Pumps.......................................................................................................... 40 2 Observed Repair Times and Theoretical Distribution -

Valves........................................................................................................... 41 l

s. i

REURES and TABLES (cont'd)

Table No.- Title page 1 J AF Initiating Event Frequencies ............................................................................... 2 2 Minimal Cutsets for Dominant Sequences Before Recovery T 1 Induced TW Cutsets ...... .... ............. ......................... .................. .... ....... ......... 42 T2 Induced TW Cutset s . .......... ....................... ........... ..... .. ................ . ......... ....... 48 Intermediate LOCA Induced "W" Cutsets .......................................................... 55 Small LOCA Induced "W" Cutsets ..................................................................... 57 TDCB I nduced TW Cutsets ... .... ... ...... ............................... .. .. ............. ........ ....... . 66 3' TW Sequences Before and After Recovery Loss of Offsite Power (T 1 ) ....................... ... ......... . ...... . ........ .......... ... . . ..... ..... . . . ... 69 Loss of Power Converslon System (T2) ............................................................ . .72 i Tranalents with PCS Available (T3A) .................................................................. 78.

Loss of Feedwater Transient (T38) ..................................................................... 84  :

Inadvertent Open Relief Valve (T3C) ..... ............................................................ 90 i Loss of Safeguard AC Bus 10500 (TAC.10500) .................................... ............ 92 Loss of Safeguard AC Bus 10000 (TAC.10600) ................................................. {

95 1 Loss of Safeguard DC Bus DC-A (TDC.A) ......................................................... 99 Loss of Safeguard DC Bus DC B (TDC-B) ......................................................... 103  :

Lar ge LOCA (A) . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..105 i

I ntermediate LOCA (S 1) ......... .... ...... ......... .. .. .... ...... .. .... .... ....... ..... ...... ............. 106  !

Small LOCA (S2) ... .. .... . .... .... .... . .. .. . ..... ... ... . ..... =....................................... 107 I 4 Recovery of Locs of Offsite Power ................................................................, ... 110 5 Recovery of Power Conversion System ......................................................... . 111 6 Recovery of DC Safety Bus System .................................................................. 112 7 Probability of Falling to Repair Battery Panels

"*nTime(t)............................................................................................. 113 i 8 e _; Falling to Repair Heat Exchangers M ni n Ti me (t) . . . . . . . .. . . . ... . . .. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . .. . . .. . 114 .....................,

9- robai.,.lity of Falling to Repair AC Buses l i'

Within TP (t)............................................................................................. 115 10 't

. of Ci -t Heat Removal Sequence wa M............................................................................................... 116 Appendix A WE e Nees '

t a, v. %. .e Powe r (T 1 ) . . . . . . . . .. . . . . . . . . . . . . . .. . . .. . . . . . . . . . . . . . . . . . . . .... .. .. ..... . . ..... . . . . . . ..A.1 ......

Loss of Power Conversion System (T2) ............................................................ A.2 Transientswith PCS Available(T3A) . .................................................. A3 Loss of Feedwater Transients (T38) . ................................................................ .. A.4 Incdvertent Opon Relief Valve (T3C) .................................................................. A-5 s Lar ge LOC A (A) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . .. . .. .A6 . . . . . . . . . . . . . . . . .1. . . . .. . . .

Intermediate LOCA (S 1) .. ... .. ........ .. ....... . .... . ... ... .. ....... . .. ... .. ...... .... ... .. . .... A.7 . . .. ........

Small LOC A (S2) . . . . . . .. . . . .. . . . . . . . . . .. . .... . . . .. ..... . .. . . . .. . .. . .. . . . . . .A.8 . . .. . . . . . . . . . . . . . . . . . .

Loss of Safeguard AC B us 10500 ................................. ..................................... A9 Loss of Saleguard AC Bus 10600 ...................................................................... A 12 Loss of Battery Control Board BCB.2A .............................................................. A.15 Loss of Battery Control Board BCB.28 ............................................................. A.17 11 am

I

1. INTRODUCTION A plant specific TW analysis was. conducted for the James A.

FitzPatrick Nuclear Plant. -Major portions of-this analysis were derived from the ongoing JAF IPE. l Actual plant operating data, '

spanning the time = period from January, 1976 to December, 1989, were utilized. Fault tree models-were developed for plant systems, and their support systems. Plant specific JAF-component experience.

failure data were derived from plant operating Generic common cause beta factor data were taken from EPRI-NP3967 " Classification and Analysis of Reactor.

~ Experience: Involving. Dependent Events" were used. :Using the SETS code, accicket- sequences were then quantified by combining the boolean obdatigna derived from the system failure modes-associated tdto thbse sequences, and the resultant equations reduced to-ToPA winimal cutsets. The dominant minimal cutsets  !

were then revi3wed to provida the basis for both identifying and quanti!ying recovery. actions.  !

The Analysis of TW sequences can bo divided into three.

phases that match the sequence chronology:

A. Indentification of Accident Initiators B.- Accident Sequence Development and Quantification C. Recovery Analysis L

2. ACCIDENT INITIATORS 1 Accident initiators are typically divided into two 1

categories; transients and LOCAS. The transient cate further divided into general and special initiators. gory is Based on actual plant reviews of JAF operating trips, general-initiating event frequencies have been' estimated for the following-plant transients: Tl T2,.T3A,.T3B, and T3C.

category of' support systems inktiators, a loss of a vitalFor ACthe or DC bus (TAC or TDC) were selected as "special initiators". -Loss of vital AC and or DC' bus was selected for evaluation because failures involving either of these systems will lead to a. plant trip and a concurrent degradation of the safety system (s) required to bring the plant to a safe shutdown.

The LOCA category is divided into three subcategories; A-(large LOCA), S1-(intermediate LOCA), and S2-(small LOCA).

Table 1 frequencies. presents the initiators along with their 1

TABLE 1 JAF INITIATING EVENT FRECUENCIES INITIATOR '- MEAN FREQtENCY Source

(/ Reactor Yeer) .

T1: Loss of offsite Power (LOSP) 0.057 (1) .;

T2: Loss of PCS Transients (MSIV, or Turbine Sypass Failure) 0A8' (2)

T3A: Transients With Condenser Initially Available 4.7 . (2)

T38: Loss of Feeduster With Main condenper Aveliable 0.39 (2)

T3C: 10RV (Inadvertently open Relief watve) 0.0964 (2)

TAC: Transient Caused by Loss of Safety AC bus ' ~3 5.0x10 - (3)

'TDC: Transient caused by Loss of Safety DC bus 5.0x10 -3 (3) u

~

A: Large LOCA I 1.0x10 .(3)

S1: Intermediate LOCA - 3.0x10 (3)

S2: Sumil LOCA 3.0x10~ -

-(3) ,

Note (1) JAF plant specific date with Bayesian egdete.

Note (2) JAF plant specific data frass actunt operating history.

Note (3) NUREG-1150 Peach Bottant Study (Table 4.3-3, NUREG/CR-4550, voltsee 4, Rev.1 port 1).

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3. ACCIDENT SEQUENCE DEVELOPMENT AND QUANTIFICATION Event trees-were constructed to identify the' appropriate TW seguences in terms.of the systems, components, tor functions'

which have succeeded or failed.-

The.following define the event tree Wsading used in the development of! transients and LOCAs evene trees. 1 l

C -

Failure of,the. Reactor Protection' System (RPS). )

B1 -

Offsite AC power (115 kv) fails from random failures B2 -

Emergency onsite'AC power fails from random failures.

P -

SRVs fail to reclose from; random failures (P1: 1 l SORV, P2: 2 SORVs, P3: 3 SORVs).  !

Q -

Power Conversion System is unavailable because of .;

random failures. -

Q1, -

. Main Condenser unavailable.

U1: -

HPCI fails from random failures..

UlX -

HPCI suction fails to remain on the Condensate Storage Tank'(CST)

U2 -

RCIC fails from random failures.

X1 -

Failure to depressurize the primary system via SRVs or the Automatic Depressurization System (ADS). ,

o X2 -

Failure to depressurize-the primary system to .

i allow Residual Heat Removal' Shutdown Cooling Mode Q

.(RHR-SDC) to operate.-

"V1 -

Condensate System fails fromerandom failures.

J V2 -

Core Spray System fails-from; random failures.

L V3 -

Low Pressure Coolant Injection '(LPCI)--mode of RHR '

fails from random failures.'-

V4 -

RHR-service water cross-tie fails to inject coolant because of random failures.

3 1

-~ - .

4 i

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W1 --

RHR suppression pool cooling mode fails-from random i

fallures.- -]

W2 -~ RHR drywell~ spray mode fails from random-failures. -i 1

W3 -

RHR Shutdown Cooling Mode fails from random i failures. .

-1 1

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Sequences were developed for each of the initiators presented in Table 1. The following "TW" sepences result in states where containment integrity is threatened. In defining a sequence, a slash (/) in front of-an event designator indicates the event is l a success and not a failure.-

T1 Seouences i Loss of Offsite Power (T1) induced "TW" sequences: -l SEQUENCE T1-4 T1*/C*/B2*/P*/Ul*/UlX*W1*W2*/X2*W3 A-loss of offsite power transient occurs (T1), which closes the i MSIVs. The reactor scrams and at least one division of J emergency onsite AC power is(/C) supplied to either safeguard buses 10500 or 10600 (/B2). The SRVs open and -.reclose (/P) to relieve the pressure that results'from the power imbalance caused when the turbine tripped. With the MSIVs closed, feedwater is-lost r and the reactor water level starts to decrease. At Level 2 126" above top of active fuel (TAF), HPCI and RCIC are automatica,lly r initiated. HPCI successfully starts to inject (/U1) to provide #

core water level control.. As the initiator-(T1) prevents the '

-reopening of the MSIVs (condenser vacuum cannot be restored until-offsite AC power is recovered), the main condenser will be unavailable for decay heat removal. Decay heat will therefore be  !

transferred (as steam) to the suppression pool (torus) via the SRgs, increasing the pool temperature. At a pool temperature of 95 F the operatcr will initiate suppression pool cooling, but this action is not successful (W1). At-this point he will initiate the torus spray mode of RHR, but as this is not sufficient to prevent containment over-pressurization. Therefore the drywell spray mode of RHR will then be initiated. However, this also fails to operate (W2). At this' point the operator depressurizes the reactor successfully (/X2) and aligns the RHR system into its shutdown cooling mode. This last mode of decay heat removal'is also unsuccessful (W3). Thereby threatening containment integrity.

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( SEQUENCE T1-5 T1*/C*/B2*/P*/Ul*/UlX*W1*W2*X2 i l- This sequence is the same as sequence T1-4, except that'the operator fails to depressurize the reactor (X2)-and cannot attempt the-shutdown cooling mode of RHR.

SEQUENCE T1-9 T1*/C*/B2*/P*/Ul*UlX*W1*W2*/X2*W3 This is the same as sequence T1-4 except thgt HPCI. fails (UlX) due to a torus temperature greater than 200 F.

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SEQUENCE T1-10 T10/C0/B20/P2/U1CUlXcWicW20X2 Same.as-sequence T1-9 except that the operatorLfails to l depressurize the-reactor (X2) and cannot attempt the shutdown l

' cooling mode of.RHR.

-SEQUENCE T1-14 T1*/C*/B2*/P*Ul*/U2*Wi*W2*/X2*W3 Same as sequence T1-4 except-that HPCI'has failed (Ul)~due to random mechanical faults; RCIC is then used .(/U2) - for reactor vessel level' control.

SEQUENCE T1-15 T1*/C*/B2*/P*Ul*/U2*Wi*W2*X2 Same as' sequence T1-14, except that:the operator fails to deprassurize,the reactor (X2) and cannot-attempt the shutdown cooling mode of RHR. .j

' SEQUENCE T1-19 T1*/C*/B2*/P*Ul*U2*/X1*/V2*W1*W2*/X2*W3 'i A loss'of'offsite power transient: occurs-(Tl), which closes the MSIVs. The reactor scrams emergency onsite AC power =is(/C) and at least one. division of. supplied to either safeguard b 10500 or 10600 (/B2). The SRVs open and _reclose (/P) to relieve a the pressure.that results from the power imbalance caused when l the turbine tripped. With the MSIVs closed, feedwater is lost u and' reactor. water level starts to decrease.. At Level.2, 126" above TAF), HPCI and RCIC automatically initiate but' both(fail to operate (Ul*U2). The primary system is successfully allowing the Core spray system to operate depressurized

(/V2) .: As the (/X1)iator init (T1) prevents the-reopening of the MSIVs  !

(condenser, vacuum cannot be restored-until offsite AC power is recovered), the main condenser is not available for decay. heat Decay heat is then transferred (as steam) to the removal. _'

suppression pool (torus) via the SRVs, ingreasing the pool

. temperature. At a-pool temperature of 95 F the operator will ,

. initiate suppression pool cooling, but'thistaction is not 1 successful (W1). At this point the operator initiates the-torur, spray mode of.RHR,ebut'this mode of operation-is not sufficient  !!

to prevent containment over-pressurization. Therefore, the 1 drpell spray mode of-RHR is then initiated.;However, this also i fails to operate (W2). Jd: this point the operator depressurizes  !'

.the1 reactor successfully (/X2) and aligns the RHR system into-its shutdown cooling mode.1.-This last mode of' decay-heat removal is also-unsuccessful (W3)is threatened.The sequence results in a= state where containment'integr:.ty SEQUENCE T1-20 T1*/C*/B2*/P*Ul*U2*/X1*/V2*Wi*W2*X2 Samelas sequence T1-19, except that the operator fails to  !

depressurize the reactor (X2) and cannot attempt the shutdown cooling mode of RHR.-

1 SEQUENCE T1-24 T1*/C */B2 */P*Ul*U2 */X1*V2 */V3 *W1*W2 */X2 *W3 Same as-sequence T1-19 except that Core Spray fails (V2) due to random-mechanical faults; LPCI is then used (/V3) ' for reactor vessel-level control.

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SEQUENCE T1-25 T1*/C*/B2*/P*Ul*U2*/X1*V2*/V3*Wi*W2*X2 Same as sequence T1-24, except that the operator fails to depressurize the reactor (X2)-and Jannot attempt'the shutdown i cooling mode of RHR.

SEQUENCE-T1-29 T1*/C*/B2*/P*Ul*U2*/X1*V2*V3*/V4*W1*W2*/X2*W3 Same as sequence T1-24 except that LPCI has failed (V3) due to random mechanical faults; the operator uses.the RHRSW cross-tie

(/V4) for reactor vessel level control.

SEQUENCE T1-30 T1*/C*/B2*/P*Ul*U2*/X1*V2*V3*/V4*Wi*W2*X2 i

.Same as sequence 1T1-29, except-that the. operator fails to' depressurize the reactor (X2) and cannot attempt-the shutdown , ]

1 cooling mode of RHR. I SEQUENCE T1-33 T1*/C*/B2*P1 ALloss of offsite power transient occurs (T1), which closes the MSIVs. The. reactor scrams ( and at least1one division of.

emergency onsite AC power is/C) supplied to either safeguard buses 10500 or 10600 (/B2). The SRVs open to relieve the pressure that results from the power imbalance caused when the turbine tripped, ,

but one SRV fails to close - (P1) , creating a steam loss of coolant accident. The sequences transfers .to the - small. (S2) LOCA tree for further development.

SEQUENCE T1-34 T1*/C*/B2*P2 Same as sequence T1-33, except that two SRVs fail to reclose-(P2) and the sequence transfers.to-the intermediate (S1) LOCA event tree-for further development.

SEQUENCE T1-35 T1*/C*/B2*P3 Same as sequence T1-34, except that three SRVs falloto reclose

-(P3) and the sequence: transfers to the"large (A) LOCA event tree for furtherfdevelopment. t i

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, i T2 Sequences "TW" sequences induced by plant transients (T2), Power Conversion System (MSIVs) is not available:-

SEQUENCE T2-4' T2 */C*/Bl*/P*/U1/UlX*Wi*W2 */X2 *W3 1 A. transient occurs that-causes a complete.lossJof-the power conversion system.-(T2). The reactor scrams (/C), and the AC power from the 115kv bus provides adequate power (/B1). The.SRVs open and reclose- (/P) to -relieve pressure that results from the power: imbalance caused when the turbine trips.- With the MSIVs closed,-feedwater.is lost and reactor water level starts to i decrease. At Level:2,.(126" above TAF), HPCI and RCIC ~

. automatically initiate. HPCI successfully starts to inject (/U1) to provide core water level control. With the MSIVs closed, the condenser is also not available for decay heat removal'. Decay heat will is-transferred-(as steam) to the suppressionipool

. (torus) via the SgVs, increasing the-pool temperature.. At a pool

' temperature of 95 F thefoperator initiates suppression pool

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cooling, but this action is not successful (W1). At this point the operator initiates the torus spray mode =of RHR,-but this mode ofm operationiis not sufficient-to prevent containment  ;

over-pressurization. Therefore, the drywell'. spray mode.of RHR-is then initiated. However, this also fails.to operate (W2). At

'this point-the operator depressurizes the reactor successfully ,

(/X2 This)last. node-of decay. heat removal is also unsuccessful'(W3).and aligns the RHR syste '

This sequence:results'in a-state where containment integrity is ,

threatened.

SEQUENCE-T2-5 T2 */C*/Bl*/P*/Ul*/UlX*Wi*W2 *X2 Same as sequence T2-4, except that.the operator fails;to depressurize the reactor-(X2)-and cannot' attempt.the' shutdown cooling mode of RHR. 3 SEQUENCE T2-4a T2*/C*/Bl*/P*/Ul*UlX*Wi*W2*/X2*W3 l-

.Same as sequence T2-4.except that HPCI fails (UlX) due.to a torus temperature greater than 200 0 F.. This sequencefresults in a state where containment ~ integrity is threatened. ,

SEQUENCE T2-Sa T2 */C*/Bl*/P*Ul*UlX* *Wi*W2 *X2 Same as sequence T2-4a, except that-the operator fails to l depressurize.the reactor (X2) and cannot-attempt the shutdown cooling mode of RHR.~

SEQUENCE T2-9 T2*/C*/Bl*/P*Ul*/U2*Wi*W2*/X2*W3 E 'Same-as sequence.T2-4 except that HPCI fails (U1) due to random-mechanical- faults;: RCIC is then- used' (/U2) for reactor vessel level control.=

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.1 SEQUENCE T2 T2*/C*/Bl*/P*Ul*/U2*W1*W2*X2-Same'as sequence T2-9, except that the operator ~ fails to j'

'depressurize the' reactor (X2)'and-cannot attempt the shutdown cooling mode of.RHR. )

-3 SEQUENCE T2-14 T2*/C*/Bl*/P*Ul*U2*/X1*/Vl*Wi*W2*/X2*W3-A:transjent' occurs thaticauses a complete loss of.the power -i conversion system.(T2).. The reactor scrams (/C)_,1and .the AC '

power from the 115kv bus provides adequate power (/B1) . The SRVs j open and reclose (/P) ~ to relieve ' pressure .that resultu' from the power imbalance caurad when the turbine trips.- With the MSIVs ]

-closed,-feedwater is lost and reactor' water level starts to- -

L decrease. At Level'2, (126" above TAF), HPCI and RCIC j L

automatically initiate but both fail to operate (Ul*U2). The j primary system is 'successfully depressurized :(/X1) to allow the Condensate System to operate (/V1). Since the MSIVs are closed,'

the condenser is not available for decay heat removal, decay heae i will be transferred (as steam).to the suppression pool (torus)  !

• 1"9 th* pool temperature. At a pool H via the SRVs, temperature inc5**the operator initiates. suppression pool

~of'95.F. .

1 cooling, but this action:is not successful (W1). At this point the operator? initiates the torus spray mode of RHR,.but this mode of operation is not sufficient to prevent containment J

4 over-pressurization. Therefore the drWell~ spray mode of RHR is j then initiated. However, this a,lso fails to' operate (W2). At ,

this point the operator depressurizes the reactor:successfully 1

(/X2 This)last mode of decay heat removal is also unsuccessfuland aligns (W3the RHR, system into its' This sequence now results:in a state where: containment.integri)ty is threatened.

SF"JENCE T2-15 T2*/C*/Bl*/P*Ul*U2*/X1*/V1*Wi*W2*X2 l Same as sequence-T2-14, except that the operator: fails to ,

depressurize~the reactor (X2) and cannot attempt the shutdown I cooling mode of RHR.

SEQUENCE T2-19 T2 */C*/ Bl*/P*Ul*U2 */X1*Vl*/V2 *W1*W2 */X2 *W3 1

Same as sequence T2-14 except that the Condensate SystemLhas i failed (V1) due to random mechanical faults; Core Spray is then I used (/V2) for reactor vessel level control. l SEQUENCE T2-20 T2 */C*/Bl */P*Ul *U2 */X1 *Vl */V2 *Wi*W2 *X2 Same as sequence T2-19, except that the operator fails.to

! depressurize the reactor (X2) and'cannot attempt the shutdown cooling mode of RHR.

SEQUENCE T2-24 T2 */C*/Bl*/ P*Ul*U2 */X1*Vl*V2 */V3 *Wi*W2 */X2 *W3 l Same as sequence T2-19 except that Core spray fails-(V2) due to l random mechanical faults; LPCI is then used.(/V3) for reactor i vessel level. control.

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-SBQUENCE T2-25--T2*/C*/Bl*/P*Ul*U2*/X1*Vl*V3*/V3*Wi*W2*X2 l

.Same as-sequence T2-24,Lexcept that4the operator' fails to depressurize- the reactor: (X2) and:cannot attempt the shutdown cooling mode-of RHR.

-SBQUENCE T2-29 T2*/C*/Bl*/P*Ul*U2*/X1*V1*V2*V3*/V4*W1*W2*/X2*W3 Same as sequence T2-24 except that LPCI (V3) has failed due to

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random mechanical faults; the operator uses the RHRSW cross-tie

(/V4) for reactor vessel level control.

.l SEQUENCE-T2-30, T2 */C*/ Bl */P*Ul*U2 */X1 *V1*V2 *V3 */V4 *W1 *W2 *X2 Same as sequence T2-29, except that the operator fails to

, depressurize the reactor (X2) and cannot attempt the shutdown ,

cooling-mode of RHR.

SEQUENCE T2-33 T2*/C*/Bl*P1 ,

A transient occurs that causes a completeLloss of the power conversion system (T2). The reactor scrams'(/C), and the AC .

power from the 115kv bus provides adequate power (/B1). The SRVs open to relieve the pressure that results from the power

, imbalance caused when the' turbine tripped, but one SRV fails to close (P1) , creating a steam loss of coolant accident. The

sequences transfers to the small-(S2)' LOCA tree for further development.

SEQUENCE T2 T2*/C*/Bl*P2 l Same as sequence T2-33, except that'two SRVs fail'to reclose (P2) and the sequence transfers to the intermediate (S1) LOCA event tree for further development.

SEQUENCE T2-35 T2*/C*/Bl*P3 Same as sequence T2-34, except that three SRVs fail to.reclose (P3) and the sequence transfers to the large (A) LOCA event tree ,

l for further development, l SEQUENCE T2-36 T2*/C*B1 L .

A t.ransient occurs that causes a complete loss of the power 4

conversion system (T2). The reactor scrams (/C) ,- and- the AC power from the<115kV fails to provide adequate power (B1). These 1N se@ences are transferred to the loss of off-site (T1) event tree for further development.  :

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T3A'Secuences

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"TW" sequences induced by' plant transients (T3A), where the Power Conversion System is. initially available:

- SEQUENCE T3A-2 T3A*/C*/Bl*Q A transient-occurs that causes a turbine-trip (T3A). The reactor {

scrams (/C), and the AC power from the 115kv. bus is:available

- (/B1) . The MSIVs subsequently fail to remain openL(Q) due to 1 random mechanical faults. This sequence transfers to the T2 event tree-(loss >of PCS) for further development.  ;

SEQUENCE T3A-3 .T3A*/C*Bl*/B2 A-transient scrams occurs that causes a turbine trip-(T3A). -The-reactor

(/C), but AC power from 115kv is inadequate (B1). Onsite i emergency power. is established (/B2) ._ This sequence transfers to

-the T1 event tree (loss of off-site power) for further development.

i T3B Secuences "TW" sequences induced by Loss of Feedwater event'(T3B) sequences: i SEQUENCE T3B-9 T3B*/C*/Bl*Q1 i A. loss of~feedwater event occurs (T3B) that causes a reactor-scram. The reactor: scrams' (/C) , and AC power from the 115kv is

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available (/B1) . The MSIVs subsequently: fail to remain open (Q1) .- These sequences therefore transferred to the T2-event tree (loss of PCS) for.further_ development.

SEQUENCE T3B-10 T3B*/C*Bl*/B2 A loss of feedwater event occurs (T3B) that causes a reactor scram. .The reactor scrams (/C), and AC power from-the 115kv bus is inadequate (B1), but onsite emergency power is- established

(/B2) . These sequences therefore transferred-to-the>T1 event tree (loss of off-site) for further; development.

T3C Secuences '

-valve:

"TW" sequences induced by an-inadvertent stuck _open relief SEQUENCE T3C-2 T3C*/C*/Bl*Q An inadvertent stuck open relief valve event occurs (T3C). The reactor scrams'(/C) either manually or by the RPS, and the AC power from the 115kv bus is adequate (/B1) . The MSIVs subsequently-fail to remain open (Q). These. sequences therefore

' transferred to the S2 (small break LOCA) event treet for further' development.

11 M-simis-mm i e i iiin ^ ^^'-

DEQUENCE T3C-3' T3C0/C*210/22 An inadvertent stuck open relief valve event occurs (T3C). The reactor'scrans (/C either manually or b from the 115kv bus)is inadequate- power --(y-:B1the RPS, and AC power power is establishe6 (/B2). These sequence)s,therefore but onsite emergency  ;

transferred to the S2 (small break LCC.$.) event tree at-point "Z2" l for further development.  !

TAC-10500 Secuences "TW sequences induced by.a loss.of the 10500:AC (TAC-10500) vital safeguard bus:

SEQUENCE TAC-10500-4 TAC-10500 */C */Bl *Q */P*/Ul */UlX *Wi* W2 i

A loss of the conditions 10500 that may AC leadsafeguard bus'scram.

to a= manual occurs (TAC-10500) creating l The control rods are successfully inserted-into the core-(/C), and AC power from the -1 115ky bus is available (/B1) . Subsequently, condenser vacuum is 1 lost leading to MSIV closure. With the MSIVs closed,-the reactor pressure is controlled by the SRVs opening and closing successfully (/P) ... - With the MSIVs closed,- feedwater is also lost' ,

and' reactor water level starts to. decrease. At Level 2, (126" l above TAF), HPCI and RCIC are automatically initiated. HPCI successfully injects-(/U1) to provide reactor water level control. Since the condenser is not available for decay heat removal, decay heat-will be transferred.(as steam) to the.

suppression pool (torus via the.SRVs temperature. At a pool) temperature of inc5***in9 th* P 1 95 F the operator initiates suppression pool cooling, but this action is not successful (W1). The operator will then initiate the torus spray i mode of RHR,ization, the drywell-spray mode of RHR is thenbut over-pressur as this is not sufficient to initiated. However, this also fails to operate (W2).- The shutdown cooling mode-(SDC) of RHR is unavailable due to the initiating event. However, the LPCI mode through the.RHR heat i

exchangers is available. Nevertheless this mode of decay heat is also unsuccessful With no decay heat removal: pathway established this se(W3).quence now results in a state where containment integrity is throatened.

SEQUENCE TAC-10500-5 TAC-10500*/C*/Bl*Q*/P*/Ul*UlX*Wi*W2 Same due toas sequence TAC-10500-4, a torus-temperature except greater than that200HgCI F. Thishas-failed (UlX) sequence results in a state where containment integrity is threatened.

SEQUENCE TAC-10500-20 TAC-10500 */C*/Bl *Q */P*Ul*/U2 *W1*W2 Same as sequence TAC-10500-4, except that HPCI has failed (U1) due to random mechanical faults; RCIC is then used (/U2) for reactor vessel level control.

SEQUENCE TAC-10500-23 TAC-10500*/C*/Bl*Q*/P*Ul*U2*/X1*/Vl*Wi*W2 Same as sequence TAC-10500-20, except that HPCI and RCIC have both failed (Ul*U2); the primary system is successfully depressurized (/X1) to allow the Condensate System to operate

(/V1) for reactor vessel level control.

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SEQUENCE TAC-10500-26 . TAC-10500*/C*/Bl*Q*/P8 Ul*U2*/X1*V1*/V2*W1*W2 Same as sequence TAC-10500-23, except that Condensate System has

, failed.(V1).due:to random mechanical faults; Core spray is then used (/V2) for reactor vessel level control.

i SEQUENCE TAC-10500-29 TAC-10500 */C*/ Bl*Q*/P*Ul*U2 */X1*Vl *V2 */V3 *W1 *W2 Same as se@ence TAC-10500-26, except that Core spray fails (V2) ,

due to random mechanical faults; the operator uses LPCI (/V3) 3 for reactor vessel level control'. '

SEQUENCE TAC-10500-32 TAC-10500 */C*/ Bl *Q*/P*Ul*U2 */X1 *V1*V2 *V3 */V4 *Wi*W2 Same as sequence TAC-10500-29,-except;that LPCI has failed (V3) p due to random mechanical faults; the operator,uses RHRSW cross-tie:(/V4) for reactor vessel level control.

SEQUENCE TAC-10500-34a TAC-10500*/C*/Bl*Q*P1-t A. loss-of the 10500 AC safeguard-bus occurs'(TAC-10500) creating conditions that may lead'to a manual scram. . The control rods are' successfully inserted into the core (/C)', and. AC power from the 115kv - bus is' available - (/B1) .. Subsequently, condenser vacuum is' lost leading to MSIV closure. As the MSIVs closed,-the SRVs open

,to relieve reactor pressure, but one SRV fails to close (P1),

creating,a: loss of coolant accident. This sequence-transfers to ,

the small (S2) LOCA-tree for further development. .

SEQUENCE. TAC-10500-34b TAC-10500*/C*/Bl*Q*P2 L Same as sequence TAC-10500-34a, except that'two SRVs fail.to reclose (P2) and the sequence is transferred to the intermediate

_ (S1). LOCA event tree for further1 development.

SEQUENCE TAC-10500-34c TAC-10500*/C*/Bl*Q*P3' l

Same as sequence TAC-10500-34a, except that three SRVs. fail to .

I reclose (P3).- At this point, the sequence probability is less l that 1.0E-8; therefore, this sequence is not developed further.  ;

o . 1 SEQUENCE TAC-10500-37 TAC-10500 */C* Bl*/B2 */ P*/Ul*/UlX*W1*W2 i

Same as sequence TAC-10500-4, except that offsite power is lost (B1) but onsite emergency powerLis established (/B2). HPCI' provides core cooling (/U1), but containment'over-pressure

, protection fails (Wi*W2). This sequence results in a-state where-j containment integrity is threatened.

SEQUENCE TAC-10500-40 TAC-10500*/C*Bl*/B2*/P*/Ul*UlX*Wi*W2 ,

l Same as sequence TAC-10500-37, except that HPCf has failed (UlX) due to.high torus temperature greater than 200 F.

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SEQUENCE TAC-10500-51 TAC-10500*/C*Bl*/B2*/P*Ul*/U2*W1*W2 Same as sequence TAC-10500-37, except that HPCI fails (U1) due to random mechanical faults; RCIC is then used (/U2) for reactor vessel level control.

SEQUENCE TAC-10500-54 TAC-10500*/C*Bl*/B2*/P*Ul*U2*/X1*/V2*W1*W2 Same as sequence TAC-10500-51, except that HPCI and RCIC have both failed (Ul*U2); the primary system is successfully depressurized (/X1) to allow the Core spray system to operate

(/V2) for reactor vessel level control.

SEQUENCE TAC-10500-57 TAC-10500*/C*Bl*/B2*/P*Ul*U2*/X1*V2*/V3*Wi*W2 Same as sequence TAC-10500-54, except that the Core spray fails (V2) due to random mechanical faults; LPCI is used (/V3) for reactor vessel level control.

SEQUENCE TAC-10500-60 TAC-10500*/C*Bl*/B2*/P*Ul*U2*/X1*V2*V3*/V4*W1*W2 Same as sequence TAC-10500-57, except that LPCI failed (V3) due to random mechanical faults; the operator uses the RHRSW cross-tie (/V4) for reactor veasel level control.

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I TAC-10600 Secruences TW sequences = induced by-a loss of the-10600 AC (TAC-10600) vital

- j safeguard bus: '

SBQUENCE TAC-10600-16 TAC-10600 */C*/ Bl *Q*/P*/Ul*W1 *W2 */X2 *W3 A loss of the 10600'AC~ safeguard bus occurs (TAC-10600) creating conditions that may lead to a manual scram. The control rods are i successfully inserted into the core (/C), and AC power from the 115kv '

bus is available=(/B1). Subsequently, condenser. vacuum is lost

. leading to MSIVs closure. As the MSIVs close, reactor pressure is l controlled by the SRVs opening and closing successfully (/P). Since  ;

the MSIVs are closed, feedwater is lost and reactor water level- i starts to decrease. At Level'2, (126" above TAF , HPCI and RCIC are automaticallyJinitiated. HPCI'successfully injec)tsSince the-condenser reactor water level control.- not available for decay heat' removal, decay heat will be transferred (as steam) to the suppression pool (torus) via the SRVs 0 increasing the pool j

1 temperature. At a pool temperature of 95 F the operator' initiates suppression pool cooling,"but this is not' successful (Wl). The operator will'then initiate the torus spray mode of RHR, but as this is not sufficient to prevent containment over-pressurization the d m ell spray mode of RHR is then be initiated.. However, this also .l fails to operate (W2). The operator-then depressurizes-the reactor successfully,(/X2) and aligns the RHR system into its shutdown-cooling mode. This last mode of decay heat removal is also.

unsuccessful'(W3). With no decay heat removal pathway established l this sequence now results in-a state where containment integrity is threatened.  !

l SEQUENCE TAC-10600-16a TAC-10600*/C*/Bl*Q*/P*/Ul*W1*W2*X2 l l

Same as sequence TAC-10600-16, except thatothe operator fails to j depressurl:e the reactor-(X2) and cannot attempt the shutdown cooling mode of RHR.

SEQUENCE TAC-10600-19 TAC-10600 */C */ Bl* Q */ P*Ul*/U2 *W1*W2 */X2 *W3

'Same as sequence TAC-10600-16 except that HPCI has failed (U1) due to L random mechanical faults; RCIC is then used (/U2) for reactor vessel level control.

l l l SEQUENCE TAC-10600-19a TAC-10600*/C*/Bl*Q*/P*Ul*/U2*Wi*W2*X2 i

Same as sequence TAC-10600-19, except that the operator fails to depressurize the reactor (X2) and cannot attempt the shutdown cooling l mode of RHR.

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-SEQUENCE TAC-10600-22 TAC-10600*/C*/Bl*Q*/P*Ul*U2*/X1*/Vl*Wi*W2*/X2*W3 A-loss of the 10600 AC safeguard bus' occurs (TAC-10600) creating .,

conditionsLthat may lead to a manual scram. The control rods.are 1 successfully inserted into the core (/C), and AC power from the 115kv bus is available. (/B1) . . Subsequently, condenser' vacuum is  ;

lost leading to MSIV closure. With the MSIVs' closed, reactor l pressure is controlled by the SRVs opening and closing successfully. (/P) . Since the MSIVs are closed, feedwater is-lost '

and-reactor water-level starts to decrease.- At Level 2, (126" above TAF) , HPCI and- RCIC are automatically ' initiated' but both fail to. operate (Ul*U2). The primary system.is successfully depressurized (/X1) to allow the Condensate System to operate

(/V1).. Since the condenser is not:available for decay. heat' removal, decay heat will be transferred- as steam) to the ,

suppression pool.(torus) via the SRVs inc(geasing the pool -

tem erature. At a pool temperature of.95 F the operator ini intes suppression. pool cooling,:but this action is not-successful (W1). The operator will'then: initiate-the torus spray 1 mode over-pressur of RHR,ization, the d m ell spray mode of RHR will-then.bebut as this is not sufficient t initiated. However,-this also fails to operate (W2). At this point the operator depressurizes the reactor successfully. (/X2) and aligns the RHR system into its shutdown cooling-mode. This last mMe of decay heat removal. is. also unsuccessful With no decay heat removal pathway. established this sequenc(W3). e now '

results.in a state where containment integrity is threatened.

SEQUENCE TAC-10600-22a TAC-10600 */C*/ Bl*Q*/ P*Ul*U2 */X1*/Vl*W1 *W2 *X2 L Same as sequence TAC-10600-22,- except,that the: operator fails to j-I depressurize the reactor (X2).the shutdown cooling mode of RHR.

SEQUENCE TAC-10600-25 TAC-10600*/C*/Bl*Q*/P*Ul*U2*/X1*Vl*/V2*W1*W2 */X2*W3 Same as sequence TAC-10600-22 except that-Condensate System has- j failed (V1)-due to random mechanical faults; Core Spray is then

- used (/V2). for reactor vessel level: control. J l

SEQUENCE TAC-10600-25a TAC-10600*/C*/Bl*Q*/P*Ul*U2*/X1*Vl*/V2*W1*W2*X2 Same as sequence TAC-10600-25, except-that the operator fails to depressurize the reactor (X2) and cannot. attempt the shutdown cooling mode.of RHR.

y

~

16 i-l i

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SEQUENCE TAC-10600-28 TAC-10600*/C*/Bl*Q*/P*Ul*U2*/X1*Vl*

V2*/V3*Wi*W2*/X2*W3 Same as sequence TAC-10600-25 except that the Core spray system fails (V2) due to random mechanical faults; LPCI is then used

(/V3) for reactor vessel level control.

SEQUENCE TAC-10600-28a TAC-10600*/C*/Bl*Q*/P*Ul*U2*/X1*Vl*

V2*/V3*W1*W2*X2 Same as sequence TAC-10600-28, except that the operator fails to depressurize the reactor (X2) and cannot attempt the shutdown cooling mode of RHR.

SEQUENCE TAC-10600-31 TAC-10600*/C*/Bl*Q*/P*Ul*U2*/X1*Vl*

V2*V3*/V4*Wi*W2*/X2*W3 i Same ac sequence TAC-10600-28 except that LPCI has failed (V3) due to random mechanical faults; the operator uses RHRSW cross-tie

(/V4) for reactor vessel level control.

SEQUENCE TAC-10600-31a TAC-10600*/C*/Bl*Q*/P*Ul*U2*/X1*Vl*

V2*V3*/V4*W1*W2*X2 Same as sequence TAC-10600-31, except that the operator fails to depressurize the reactor (X2) and cannot attempt the shutdown cooling mode of RHR.

SEQUENCE-TAC-10600-33a TAC-10600*/C*/Bl*Q*P1 A loss of the 10600 AC safeguard bus occurs (TAC-10600) creating conditions that may lead to a manual scram. The control rods are successfully inserted into the core (/C), and AC power from the 115kv bus is available (/B1) . Subsequently, condenser vacuum is lost leading to MSIV closure. With the MSIVs closed, the SRVs '

open to relieve reactor pressure, but one SRV fails to close (Pi),

creating a loss of coolant accident. This sequence transfers to the small (S2) LOCA tree for further development.

1 SEQUENCE TAC-10600-33b TAC-10600*/C*/Bl*Q*P2 l Same as sequence TAC-10600-33a, except-that two SRVs fail to reclose (P2) and the sequence transfers to the intermediate (S1) l LOCA event tree for further development. 1 SEQUENCE TAC-10600-33c TAC-10600*/C*/Bl*Q*P3 Same as sequence TAC-10600-33a, except that three SRVs fail to reclose (P3). At this point, the sequence probability is less that 1.0E-8; therefore, this sequence is not developed further.

17

SBQUENCE TAC-10600-46 TAC-10600*/C*Bl*/B2*/P*/Ul*Wi*W2*/X2*W3- ,

Same as sequence TAC-10600-16, except that offsite power is lost- ,

(B1) but onsite emergency . HPCI provides core cooling (/U1) power is established - (/B2) .,-but containment over-pressure .  !

protection fails (W1*W2*W3). This sequence'rcsults in a state

.where--containment integrity is threatened.

SEQUENCE TAC-10600-46a TAC-10600 */C* Bl */B2 *Q*/P*/Ul *W1*W2 *X2 Same as sequence TAC-10600-16, except that:the operator fails'to depressurize tho: reactor (X2) and cannot attempt-the shutdown cooling mode'of RHR.

SEQUENCE TAC-10600-49 TAC-10600*/C*Bl*/82*Q*/P*Ul*/U2*Wi*W2*/X2*W3 Same as sequence TAC-10600-46 except that HPCI fails - (U1) due 'to i random mechanical faults; RCIC is then used (/U2) - for- reactor vessel level control. '

SEQUENCE TAC-10600-49a TAC-10600*/C*Bl*/82*Q*/P*Ul*/U2*Wi*W2*X2 l Same as sequence TAC-10600-49, except that the operator fails to ,

i-depressurize the reactor (X2) and cannot attempt the shutdown: cooling '

mode of RHR.

SEQUENCE TAC-10600 TAC-10600*/C*Bl*/B2*Q*/P*Ul*U2*/X1*/V2*W1*W2*/X2*W3.

Same as sequence TAC-10600-49, exce and RCIC (U2) have failed; The primary pt that both:HPCI-(U1) system is;successfully-depressurized (/X1) allowing '.he Core-spray system to operate

(/V2): for reactor vessel level control.

l

< SEQUENCE TAC-10600-52a TAC-10600 */C* Bl*/ B2 *Q */P*Ul*U2 */X1*/V2 *W1*W2 *X2 j p -Same as sequence TAC-10600-52, except that the operator fails to i depressurize the reactor (X2) and cannot attempt'the shutdown cooling mode of RHR.

SEQUENCE TAC-10600-55 TAC-10600*/C*Bl*/B2*Q*/P*Ul*U2*/X1*V2*

l

/V3*W1*W2*/X2*W3 b Same as sequence TAC-10600-52 except that Core. spray. system fails L (V2) due to random mechanical faults; LPCI is then used (/V3) = for reactor vessel level control.

P I 18 4

I I

E

+--e y-,,,---e- 9 ,.p--- s ,eee-ee,

SEQUENCE TAC-10600-55a TAC-10600*/C*Bl*/B2*Q*/P*Ul*U2*/X1*V2*

'/V3*Wi*W2*X2 I

Same as sequence TAC-10600-55, except-that the operator fails to depressurize the reactor-(X2) and'cannot attempt the shutdown cooling mode of RHR.

SEQUENCE TAC-10600-58 TAC-10600*/C*Bl*/B2*Q*/P*Ul*U2*/X1*V2* 1 V3*/V4*Wi*W2*/X2*W3 .

Same as sequence TAC-10600-55 due to random mechanical faults; the operator uses RHRSW cross-texcept that LPCI fails (V3)ie (/l '

for reactor vessel level control.  ;

SEQUENCE TAC-10600-58a TAC-10600*/C*Bl*/B2*Q*/P*Ul*U2*/X1*V2* ,

V3*/V4*Wi*W2*X2 Same as sequence TAC-10600-58, except that the operator fails to. l depressurize the reactor (X2) end cannot attempt the shutdown-cooling mode of RHR.

l 19 1

TDCA Senuances The "TW' sequences induced by a loss of Battery Control Board BCB-2A (TDCA):

SEQUENCE TDCA TDCA*/C*/Bl*/P*/Ul */UlX *W1 *W2 */X2 *W3 A loss of battery control board BCB-2A occurs (TDCA) which creates conditions leading to a manual reactor scram. 'The control rods are successfully inserted into the core (/C), and-AC power from the 115kv bus is available (/B1).- Subsequently operator closes the MSIVs according to abnormal operating , an procedure F-AOP-45 (Loss of "A" DC Power System).-'With the MSIVs closed closing,.successfully reactor. pressure

(/P) .is controlled by the SRVs Since the MSIVs opening and are closed,-feedwater is-lost and reactor water level starts to decrease.- At Level 2, (126" above LTAF), HPCI automatically initiates,- (RCIC is powered off BCB-2A, and-therefore is not'available). HPCI successfully

. starts to inject (

'Since the condense r is/U1) notto provide reactor available for decaywater heatlevel control.-

removal, decay heat will-be transferred (as steam)-to the suppression. pool (torus) via.the SgVs increasing the pool temperature. At a-pool temperatureLof 95 F the operatoriinitiates suppression-pool' cooling,'but this action is not successful (W1). The operator will then initiates the' torus spray mode of RHR,obut us this is not sufficient to prevent containment over-pressurization the drywell spray. mode of RHR is then initiated. However, th1s fails, to-operate (W2). At this point the operator depressurizes the and align.the RHR system into its reactor successfully shutdown cooling mode. (/X2)is Th last mode -of; decay heat removal is also unsuccessful-(W3). .With no decay heat-removal pathway established this sequence now results in a state where containment-integrity is threatened.

9 SEQUENCE TDCA-4a TDCA*/C*/Bl*/P*/Ul*/UlX*W1*W2*X2 Same Os sequence TDCA-4, exce depressurize the reactor (X2)pt andthat theattempt cannot operator fails.to the-shutdown cooling mode of PHR.

SEQUENCE TDCA-7 TDCA*/C*/Bl*/P*/Ul*UlX*W1*W2*/X2*W3 Same as sequence TDCA-4, except that HPCI has failed (UlX) due to a torus temperature greater than 200 F. 0 SEQUENCE TDCA-7a TDCA*/C*/Bl*/P*/Ul*UlX*W1*W2*X2 Same as sequence TDCA-7, except that the operator fails to depressurize the reactor (X2) and cannot attempt the shutdown cooling mode of RHR.

20

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-SEQUENCE TDCA-10 TDCA*/C*/Bl */P*Ul*/X1 */Vl*Wi*W2 */X2 *W3 -

i

~

Same as sequence TDCA-4, except that HPCI -fails due to random

- mechaincal faults (U1) and:RCIC is unavailable due_to the loss of ,

BCB-2At the operator.successfully depressurizes.the reactor (/X1) to allow the condensate System'to operate (/V1) for reactor

• vessel level control.

SEQUENCE TDCA-10a TDCA*/C */Bl*/P*Ul*/X1 */Vi*Wi*W2 *X2 '

Same-as sequence TDCA-10,.except that~the operator fails to depressurize the reactor (X2) and cannot. attempt the shutdown cooling mode of RHR.

SEQUENCE TDCA TDCA*/C*/Bl*/ P*Ul*/X1 *Vl*/V2 *Wi*W2 */X2 *W3 Same-as sequence TDCA-10 except that the Condensate System has failed-(V1) due to random mechanical' faults; Core spray system is then used-(/V2) for reactor vessel level control.

SEQUENCE TDCA-13a TDCA*/C*/Bl*/P*Ul*/X1*Vl*/V2*W1*W2*X2 Same as sequence TDCA-13, except that the operator fails to depressurize the reactor (X2) and cannot attempt.the shutdown

! cooling mode of RHR.

SEQUENCE TDCA-16 TDCA*/C*/Bl*/P*Ul*/X1*Vl*V2*/V3*Wi*W2*/X2*W3 Same as sequence TDCA-13 except.that Core spray system fails (V2) due to random mechanical faults; LPCI is then used (/V3) for l reactor vessel. level control.

SEQUENCE TDCA-16a TDCA*/C*/Bl*/P*Ul*/X1*Vl*V2*/V3*Wi*W2*X2 l Same as sequence TDCA-16,'except'that the operator fails to l depressurize the reactor (X2)c and cannot attempt the shutdown l cooling mode of RHR.

l SEQUENCE TDCA-19 TDCA*/C*/Bl*/P*Ul*/X1*Vl*V2*V3*/V4*Wi*W2*/X2*W3 Same as sequence TDCA-16 except that LPCI fails (V3) due to random mechanical faults; the operator usessthe RHRSW cross-tie

(/V4)_ for reactor vessel level control.

t 21

-. . -,- - .-<- +

• 1 I

SEQUENCE TDCA-19a - TDCA*/C*/Bl*/P*Ul*/X1*V1*V2 */V3 *Wi*W2 *X2 Same as sequence TDCA-19, except that the~o A rator fails'to depressurize the reactor (X2) and cannot attempt the-shutdown cooling mode-of RHR.

SEQUENCE TDCA-22 TDCA*/C*/Bl*P1 A loss of. battery control board BCB-2A. occurs (TDCA) which creates conditions leading to a manual reactor scram. The control rods are successfully inserted into;the core-(/C), and AC power from the 115kv bus is available (/B1). Subsequently, an operator, closes.the MSIVs according to abnormal' operating procedure F-AOP-45 (Loss'of "A" DC Power System). With the MSIVs l closed,.the SRVs.open to relieve reactor-pressure, but one SRV I fails-to close (P1), creating a loss of coolant accident. This I sequences transfers to the small . (S2) LOCA- tree for further ]

development.

SEQUENCE TDCA-23 .TDCA*/C*/Bl*P2 Sameras-sequence'TDCA-22, exceptLthat'two SRVs fail to reclose j (P2) andLthis sequence transfers to.the intermediate (S1) LOCA ,

event tree for further-development.

SEQUENCE TDCA-24 TDCA*/C*/Bl*P3 Same as sequence TDCA-22, except that three SRVs fail to reclose I (P3) and the sequence is transferred to the large (A) LOCA' event tree for further development. I SEQUENCE TDCA-27 TDCA*/C*Bl*/B2*/P*/Ul*/UlX*W1*W2*/X2*W3 Same as sequence TDCA-4, except~that offsite power is lost (B1),

but onsite emergency power is established-(/B2). HPCI provides core cooling (/Ul), but containment over-pressure protection i fails This sequence results in a state where l contain(Wi*W2*W3).

ment integrity is threatened. j SEQUENCE TDCA-27a TDCA*/C*Bl*/B2*/P*/Ul*/UlX*Wi*W2*X2 Same as sequence 27, except that the operator fails to depressurize the reactor (X2) and cannot attemptLthe shutdown  !

cooling-mode of RHR.

SEQUENCE TDCA-30 TDCA*/C* Bl*/ B2 */ P*/Ul*UlX*W1*W2 */X2 *W3 Same^as sequence 27, except that HgCI fails (UlX) due to high torus temperature greater than 200 F.

4 I

l 22 s

__ ~ __ _ . - _ . . . _ _ . _ __ _ . . _ . . _ _ ._ _

r L

SEQUENCE TDCA-30a TDCA*/C* Bl*/B2 */P*/Ul*UlX *Wi *W2 *X2 L

Same as sequence TDCA-30, except that the operator fails to depressurize the= reactor-(X2) and-cannot attempt the shutdown cooling r%de'of RHR.

SEQUENCE TDCA TDCA*/C*Bl*/B2*/P*Ul*/X1*/V2*W1*W2*/X2*W3 Same as sequence TDCA-27,-except that-HPCI fails-due to random mechanical faults-(U1)-and RCIC is unavailable due to the loss of BCB-2A. 'The primary system is successfully depressurized (/X1)

. allowing the core spray-system to-operate (/V2)' for reactor vessel-level control. I 1

SEQUENCE TDCA-33a-jTDCA*/C*Bl*/B2*/P*Ul*/X1*/V2*W1*W2*X2 I Same as-sequence TDCA-33, except that the operator fails to-depressurizeithe reactor-(X2) and cannot attempt the shutdown

= cooling. mode-of RHR.

SEQUENCE TDCA-36 _TDCA*/C* Bl*/B2 */ P*Ul*/X1*V2 */V3 *W1*W2 */X2 *W3 Same as sequence'TDCA-33 except that Core spray attempt fails (V2) due to random mechanical faults; LPCI is then used (/V3) for  ;

= reactor. vessel level control. l SEQUENCE TDCA-36a TDCA*/C*Bl*/B2*/P*Ul*/X1*V2*/V3*Wi*W2*X2 Same as sequence TDCA-36, except that the; operator fails to depressurize the reactor _(X2)_and'cannot attempt the shutdown cooling mode of RHR.-

SEQUENCE ~TDCA-39 TDCA*/C*Bl*/B2*/P*Ul*/X1*V2*V3*/V4*W1*W2*/X2*W3 Same as sequence TDCA-36 except that LPCI fails (V3)Fdue'to i random mechanical faults; the operator uses the RHRSW cross-tie l

(/V4) for reactor, vessel level control.

SEQUENCE'TDCA-39a TDCA*/C*Bl*/B2*/P*Ul*/X1*V2*V3*/V4*Wi*W2*X2 Same as sequence TDCA-39, except'that the operator fails-to depressurize the reactor (X2) and cannot attempt the shutdown cooling mode of RHR.

23 -

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TDCB Sequences  !

"TW" sequences induced by a loss of Battery Control Board BCB-2 B - (TDCB) : :

SEQUENCE TDCB-4 TDCB*/C*/Bl*/P*/U2 *W1*W2

~

A loss of battery control-board BCB-2B occurso(TDCB) which H creates-conditions. leading to a manual reactor scram. The J

. control rods . are - successfully inserted into the core - (/C) , and AC - l power from the -115kv bus is- available (/B1) . Subsequently i operator closes the MSIVs according to-abnormal-operating , an iprocedure F-AOP-46 (Loss of "B" DC Power System).. With the MSIVs closed, reactortpressure is controlled-by the-SRVs opening and closing successfully (/P) . Since the MSIVs-are closed, feedwater

.is lost and reactor water level starts to decrease. At Level 2, (126" above- TAF) ,: RCIC, is automatically initiated (HPCI is ,

powered off BCB-2B,'and1therefore is not available). . RCIC' d

.successfully' starts to, inject 1(/U2) to provide reactor water- A leveltcontrol. 'Since the-condenser is not available for decay (

heat removal," decay heat will be transferred ~[as steam) to the:

• • '"9 th' pool suppression pool temperature. At a 1(torus) via the SRVs pool temperature of in955*F the' operator-

initiates suppression pool cooling, but this' action is not  :

successful (W1). The' operator will then initiate the torus spray

' over-pressur mode of,RHR,ization, the drywell-spray mode of-RHR is then.but this,is'not sufficient to prev initiated.' However, this also-fails to operate (W2). The t

' operator is prevented from implementing the-shutdown cooling- '

(SDC)-mode of RHR, because the initiator prevents the opening of-the outboard motor operated containment. isolation valve. 'However the LPCI mcde of decay heat = removal'is available, but this along with the: other modes: of RHR fails. With no decay heat removal pathway established-this sequence.results in a state where- a containment integrity is threatened.

SEQUENCE TDCB-7 TDCB*/C*/Bl*/P*U2*/X1*/Vl*Wi*W2 I Same as sequence TDCB-4, except that RCIC has failed due to random mechanical faults and HPCI is unavailable due to the lost of BCB-2B; the primar(U2)y system is successfully depressurized

(/X1) to allow the Condensate = System to operate (/V1) for reactor vessel level control.  ;

SEQUENCE TDCB-10 ~TDCB*/C*/Bl*/P*U2*/X1*V1*/V2*Wi*W2 Same as sequence TDCB-7, except that Condensate System (V1) fails due to random mechanical faults; core spray system ~is then used'

(/V2) for reactor vessel level control. .,

SEQUENCE TDCB-13 TDCB*/C*/Bl*/ P*U2 */X1*Vl*V2 */V3 *Wi*W2 except that Core spray system fails Same astosequence (V2) due random mechan TDCB-10,ical faults; LPCI is then used (/V3) .

for -

reactor vessel level control. ,

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SEQUENCE-TDCB-19 TDCB*/C*/Bl*P1 A loss of battery control board BCB-2B occurs (TDCB).which creates < conditions leading to:a manual reactor scram. -The

, control. rods-are successfull power from the 115kv bus tisavailable' y inserted into the core (/C), and

- Subsequently an1 AC '

operator-will close the MSIVs according(/B1).to abnormal operating procedure F-AOP-46 (Loss-of "B" DC Power System). With the MSIVs l closed,-the SRVs open to relieve reactor pressure, but one SRV fails.to close-(P1), creating a steam loss of coolant accident. 4 This sequences transfers to the small (S2) ICCA tree: for further development..

SEQUENCE TDCB-20 TDCB*/C*/Bl*P2 Same as sequence TDCB-19, except that.two SRVs fail to reclose. ,

(P2) and this sequence transfers to the= intermediate-(S1) LOCA.

event tree for further development.

SEQUENCE TDCB-21 TDCB*/C*/Bl*P3 Same as sequence TDCB-19, except thatLthree SRVs fail ~to.reclose '

(P3) and this sequence transfers to the large (A) LOCA event tree Jfor'further development.

SEQUENCE TDCB-24 TDCB*/C*Bl*/B2*/P*/U2*Wi*W2 Same as sequence TDCB-4, except that offsite power is lost (B1),

~

but onsite emergency power is-established-(/B2).' RCIC provides core cooling (/U2), but containment over-pressure _ protection

' fails (Wi*W2). This sequence results.in a state where containment integrity is threatened.

SEQUENCE-TDCB-27 TDCB*/C*Bl*/B2*/P*U2*/X1*/V2*W1*W2 Same as sequence TDCB-4, except that RCIC fails due to random

. mechanical faults (U2)-and HPCI is unavailable due to the> lost of

.BCB-2B; the primary system is successfully depressurized (/X1) . to allow the Core spray system to operate (/V2) .for reactor vessel level control.

SEQUENCE TDCB-30 TDCB*/C*Bl*/B2*/P*U2*/X1*V2*/V3*Wi*W2 Same as sequence TDCB-27, except that the Core spray system fails (V2) due to random mechanical faults; LPCI is then used-(/V3) for reactor vessel level control.  ?

SEQUENCE TDCB-33 TDCB*/C*Bl*/B2*/P*U2*/X1*V2*V3*/V4*W1*W2 Same as sequence TDCB-27, except.that LPCI fails (V3) due to random mechanical faults; the operator uses the RHRSW cross-tie

(/V4) for reactor vessel level control.

l l

l 25 l

u ,

,n.,,. . - . . , . --. m ,,..g...

t A Seauences "TW"' sequences induced-by-large break LOCA-(A):

SEQUENCE A A*/C*/Bl*/V2 *Wi*W2 '

A'largex break LOCA occurs (A) some where in the primar

.which creates-conditions leading to a reactor-scram._ yTheystem reactor protection system successfully inserts the control rods into the core (/C), and' the AC power from- the 115kv bus is adequate (/B1) . - HPCI and RCIC are lost'on low reactor steam supply. Low pressure makeup is provided by the core: spray system pressure

(/V2) when either Level 1 459" TAF) or high drywell

-increasing :as(2.7apsig) is reached. W th--thecontainment temperature result of the break operator initiates suppression-pool cooling, but this.a,ction is=not'succ;ssful (W1).- The operator will then~ initiate the torus spray mode of RHR, but'as this is not sufficient to prevent containment over-pressurization,nthe d m ell spray mode-of.RHR will then=be- , i initiated. . However, this also fails-to operate At.this point theLoperator-will consider implementing the(W2). shutdown i

cooling: (SDC) mode of RHR. However, the initiator prevents the low . reactor water' level interlock (177" _ TAF): and high drywell pressure C2.7 psig i

.of operatj.on unsucc)essful.from clearing,. thereby preventing SDC ' mode- i With no decay heat removal pathway I established'this' sequence integrity is threatened.

results in a' state where. containment. 1 SEQUENCE A-6 : A*/C*/Bl*V2 */V3 *W1*W2 j Same as sequence A-3, except-that Core spray s due to random mechaincal faults; LPCI-is then used ystem fails. for

(/V3) (V2)-

reactor vessel level control. j SEQUENCE A-9 A*/C*/Bl*V2 *V3 */V4 *W1*W2 Same as sequence A-6, except that LPCI fails (V3) due'to random mechaincal faultsF the operator uses the RHRSW cross-tie;(/V4) for reactor vessel level control. >

SEQUENCE A-13 y

A*/C*Bl*/B2 */V2 *W1*W2 Same as sequence A-3, except that offsite power is lost (B1),

but onsite emergency power is established \/B2). The. core spray system-provides core cooling (/V2), but containment over-pressure protection fails (Wi*W2).

SEQUENCE A-16 A*/C*Bl*/B2*V2*/V3*W1*W2 Same aus sequence A-13 except that Core spray fails has failed

-(V2) due to random mec,haincal faults; LPCI is then used (/V3) for reactor vessel level control.

=26

-'y -%

SBQUENCE A-19 A*/C*Bla/82 *V2 *V3 */V4 *W1*W2 Same as sequence A-16, except that LPCI fails (V3) due to random mechaincal faults; the operator uses the RHRSW cross-tie (/V4) for reactor vessel level control.

S1 Seauences "TW" sequences induced by an intermediate break LOCA (S1):

SEQUENCE S1-3 A*/C*/Ble/Ule/V2 *W1*W2 An intermediate break LOCA occurs (S1) some where in the primary system which creates conditions leading to a reactor scram. The reactor protection system successfully inserts the control rods into the core adequata power(/B1).

(/C), andAtthe AC power Level 2, (126"from the 115kv above TAF bus isand HPCI RCIC are automatically initiated. However, RCIC is),not considered a viable system since it cannot provide sufficient flow to overcome the MCA. HPCI (/U1 makeup but will eventually fail as rea)ctor pressure decreasesinitially provides core below the HPCI-low steam supply setpoint.

initiation of the Core Spray (/V2). sy",L0m This necessitates for reactor level the control. Subsequently with containwent temperature increasing as a result of the break, the operator initiates sup pool cooling, but this action is not successful (W1)pression . The operator will then initiate the torus sprsy mode of RHR, but as this is not sufficient to prevent containment over-pressurization; therefore, the drwell spray mode of RHR is then initiated; however, this also fails to operate (W2). At this point the operator will consider implementing the shutdown cooling (SDC) mode of RHR. However, the initiator prevents the low reactor wtater level interlock (177" TAF) and high drywell pressure C2.7 psig) from clearing, thereby preventing SDC mode of operat:,on. With no decay heat removal pathway established this sequence is threatened. now results in a state where containment integrity SEQUENCE SI-6 S1*/C*/Bl*/Ul*V2*/V3*W1*W2 Same as sequence Sl-3 Guo to random mechainc,al faults;'LPCI is then used (/V3)except for that Core spray system

! fa reactor vessel level control.

SEQUENCE S1-9 S1*/C*/Bl*/Ul*V2*V3*/V4 W1*W2 Same as sequence S1-6, except that LPCI fails (V3) due to randor4 mechaincal faults; the operator uses the RHRSW cross-tie (/V4) for reactor vessel level control.

27

l 8800ENCE 81-13 81*/C*/Bl*Ul*/V2*Wi*W2 Same as sequence S1-3, except that HPCI fails due to random nochnnical depressurised faults (U1);

(/X1) the primary allowing the Core system s is successfully

(/V2) for reactor vessel level control. pray system to operate 8BQUENCE 81-16 81*/C*/Bl*Ul*V26/V3*Wi*W2 Same as sepence S1-13 due to random mechainca,l faultst LPCI is then used (/V3)except forthat Core spray system fa reactor vessel level control.

8BQUENCE 81-19 81*/C*/Bl*Ul*V2*V3*/V4*W1*W2 Same as sequence S1-16, except that LPCI fails due to random mechaincal faultn; the operator uses the(V3) RHRSW cross-tie

(/V4) for reactor vessel-level control.

SEQUENCE 81-23 81*/C*Pl*/B2*/Ule/V2*W1*W2 Same as sequence S1-3 exce but onsite emergency p,ower $t that offsite(/B2).

s established power HPCI is lostinLtia CB1)lly provides reactor makeup (/U1), followed by successful operation of the Core spray system (/V2), but containment over-pressure protection fails (Wi*W2)ity where containment integr is threatened.This sequence results in a state 88QUENCE 81-26 81*/C*Bl*/B2 */Ul*V2 */V3 8W1*W2 Same as sequence S1-23

.due to random mechainca,l faults; LPCI is then used-(/V3)except for that Core spray system fa reactor vessel level control.

BBQUENCE 81-29 S18/C*Bl*/B2*/Ul*V2*V3*/V4*W1*W2 Same as sequence S1-26, except that LPCI fails (V3) due to random mechaincal faults; the operator uses the RHRSW cross-tie

(/V4) for reactor vessel level control.

SEQUENCE S1-33 S1*/C*Bl*/B2 *Ul*/V2 *W1*W2 Same.as sequence S1-23, except that HPCI fails due to random depressurizedfaults mechanical  ; the

(/XI) al(U1) primary lowing thesystem Core is successfully spray system to operate

(/V2) for reactor vessel level control.

SEQUENCE S1-36 S1*/C* Bl*/B2 *Ul* V2 */V3 *W1*W2 Same as sepence S1-23, except that Core spray system fails (V2) due to random reactor vessel mechaincal level control. faultst LPCI is then used (/V3) for 28

,....-_i

SEQUENCE S1-39 S1*/C*Bl*/B2*Ul*V2*V3*/V4*W1*W2 Same as sequence S1-36, except that LPCI has failed (V3 due to random mechaincal faults; the operator uses RHRSW cross-) tie

(/V4) for reactor vessel level control.

S2 Secuences I

"TW" sequences induced by a small break LOCA (S2):

SEQUENCE S2-5 S2*/C*/Bl*Q*/Ul*W1*W2*/X28W3 A small break IcCA occurs (S2) somewhere in the primary system which creates condition 6 leading to a reactor scram. The reactor protection system successfully inserts the control rods into the core (/C), and the AC adequate (/B1). Subsequently, condenserpower fromvacuum the 115kv bus due is random mechanical faults leading to MSIVs closure. Since theto is lost MSIVs are closed, feedwater is lost and reactor water level starts to decrease. At Level 2, (126" above TAF , HPCI and RCIC are automatically initiated. HPCI successfully i)nject (/U1) to provide reactor water le, vel control. Subsequently, with containment temperature increasing as a result of the break, the operator but this is not successfulinitiates (W1). suppression The operator pool will cooling then in 1tiate the torus spruy mode over-pressurization containment of RHR, but this is not sufficient to prevent the drywell spray mode of RHR is then initiated. However, thks also fails to operate (W2).

At this point the operator depressurizes the reactor successfully (/X2) and aligns the RHR system into its shutdown cooling mode. This last mode of decay heat removal is also unsuccessful (W3 .

established this) sequence now results in a state wherewith no decay heat removal pathway containment integrity is threatened.

SEQUENCE S2-6 S2*/C*/Bl*Q*/Ul*W1*W2*X2 Same as sequence S2-5, except that the operator fails to depressurize cooling mode ofthe reactor (X2) and cannot attempt the shutdown RHR.

SEQUENCE S2-10 S2*/C*/Bl*Q*Ul*/U2*Wi*W2*/X2*W3 Same as sequence S2-6 except that HPCI fails (U1) due to random mechanical faults; RCIC is then used (/U2) for reactor vessel level control.

SEQUENCE S2-11 S2 */C*/Bl *Q*Ul */U2 *W1 *W2 *X2 Same as sequence S2-10, except that the operator fails to depressurize cooling mode ofthe reactor (X2) and cannot attempt the shutdown RHR.

29

88QUENCE 82-15 82*/C*/Bl*Q*Ul*U2*/X18/V1*Wi*W28/X2*W3 l l

Same as sequence S2-11, except that RCIC fails due to randon l

. mechanical faults (U2); the primary system is successfully l depressurized (/X1) allowing the Condensate System to operate

(/V1) for reactor vessel. level control.

SBQUENCE 82-16 82*/C*/Bl*Q*Ul*U2*/X1*/V1*Wi*W2*X2 f Same as sequence S2-15, except that the operator fails to depressurize the reactor (X2) and cannot attempt the shutdown cooling mode of RHR. j SEQUENCE 82-20 82*/C*/Bl*Q*Ul*U2*/X1*Vit/V2*W1*W2*/X2*W3 1

Same as sequence S2-15, except that Condensate System (V1) fails ,

due to random mechanical faults; Core spray system is then used l

(/V2) for reactor vessel level control.

SEQUENCE S2-21 82*/C*/Bl*Q*Ul*U2*/X1*Vl*/V2*W1*W2*X2 )

Same as sequence S2-20, except that the operator fails to depressurize the reactor (X2) and cannot attempt the shutdown I cooling mode of RHR.

SEQUENCE S2-25 S2*/C*/Bl*Q*Ul*U2*/X1*V1*V2*/V3*Wi*W2*/X2*W3 SameasseguenceS2-20, exce t that Core s ray system fails duetoranommechanicalfauts;LPCIistgenused(/V3) for(V2) reactor vessel level control.  ;

SEQUENCE S2-26 S2*/C*/Bl*Q*Ul*U2*/X1*Vl*V2*/V3*Wi*W2*X2 Same as sequence S2-25, except that the operator fails to depressurize the reactor (X2) and cannot attempt the shutdown cooling mode of RHR. ,

SEQUENCE S2-30 S2*/C*/Bl*Q*Ul*U2*/X1*Vl*V2*V3*/V4*W1*W2*/X2*W3 Same ss sequence S2-25, except that LPCI (V3) fails due to random mechanical faults; the operator uses the RHRSW cross-tie

(/V4) for reactor vessel level control.

l l

SEQUENCE S2-31 S2*/C*/Bl*Q*Ul*U2*/X1*Vl*V2*V3*/V4*W1*W2*X2 Same as sequence S2-30, except that the operator fails to depressurize the reactor (X2) and cannot attempt the shutdown cooling mode of RHR. I 30 1

I

l I

SBQUENCE 82-37 82*/CeBl*/B2*Q*/Ul*W1*W2*/X2*W3 Same as sequence S2-5, except that offsite power is lost (B1),

but onsite emergency power is established (/B2). HPCI successfully provides coolant makeup (/U1), but containment over-pressure protection fails (W1*W2*/X2*W3).

SBQUENCE 82-38 82*/C*Bl*/B2*Q*/Ul*Wi*W2*X2 )

Same as r,equence S2-37, except that the operator fails to depressurize the reactor (X2) and cannot attempt the shutdown

! cooling mode of RHR.

] SEQUENCE 82-42 82*/C*Bl*/82*Q*Ul*/U2*W1*W2*/X2*W3 Same as sequence S2-37 except that HPCI fails (U1) due to random  !

mechanical faultst RCIC is then used (/U2) for reactor vessel  !

level control. )

SEQUENCE 82-43 82*/C*Bl*/B2*Q*Ul*/02*W1*W2*X2 i

Same as sequence S2-42, except that the operator fails to I depressurize the reactor (X2) and cannot attempt the shutdown i cooling mode of RHR.

SEQUENCE 82-47 S2*/C*Bl*/B2*Q*Ul*U2*/X1w/V2*W1*W2*/X2*W3 l 1

Same as sequence S2-42, except that RCIC has failed due to t 4 random mechanical faults (U2); the primary system is  !

l successfully depressurized (/X1) allowing the Core spray system  !

(/V2) to operate for reactor vessel level control. j i l SEQUENCE S2-48 S2*/C*Bl*/B2*Q*Ul*U2*/X1*/V2*Wi*W2*X2 Same as sequence S2-47, except that the operator fails to

, depressurite the reactor (X2) and cannot attempt the shutdown

• cooling mode of RHR. l d

SEQUENCE S2-52 S2*/C*Bl*/B2*Q*Ul*U2*/X1*V2*/V3*W1*W2*/X2*W3

) Same as sequence S2-47, except that Core spray system fails due to random mechanical faults (V2); LPCI is then used (/V3) for reactor vessel level control.

)

SEQUENCE S2-53 S2*/C*Bl*/82*Q*Ul*U2*/X1*V2*/V3*W1*W2*X2 Same as sequence S2-52, except that the operator fails to I depressurize the reactor (X2) and cannot attempt the shutdown cooling mode of RHR. ,

l 31 l

1

8BQUENCE 82-57 82 */C* Bl*/82 *Q*Ul*U2 */X1 *V2 *V3 */V4 *W1 *W2 */X2 *W3 Same as sequence S2-52, except that LPCI fails due to random mechanical faults (V3); the operator uses the RHRSW (/V4) for reactor vessel level control.

SEQUENCE 82-58 82*/C*Bl*/B26Q*Ul*U2*/X1*V2*V3*/V4*W1*W2*X2 Same as sequence S2-57, except that the operator fails to depressurize the reactor (X2) and cannot attempt the shutdown cooling mode of RHR.

i i

32 t

The acc2 dent cequcnc 3 d volcped in tho cv0nt trcO3 w0ro analyzed and quantified to identify the sequences that contribute moot to the TW loss of containment heat removal frequency. Thi dominant sequences cutsets before recovery are presented in Toble 2. Table 3 summarizes all of the sequences that lead to loss of containment heat removal. The sequences with a frequenc/ less than 1.0E-9 were eliminated from further analysis. For sequences with a frequency greater-than 1.0E-9, specific operat,or recovery actions were applied, the non-recovery TJobability being based on either 24 or 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> sequence timo, as appropriate. After recovery factors were 1 incorporate 6, the dominant sequences.were re-quantified. The dominant scquence frequency after recovery actions are also summarized in Table 3. The recovery analysis is described in more detail in Section 4. j l

4. Recovery Analysis 1

The approach for developing the applicable non-recovery l probabilities applied to the dominant sequences is described in the following sections.  !

1 4.1 Recovery of Loss of Offsite Power (LOSP)

A TW sequence initiated because of a LOSP can be recovered ,

by the restoration of offsite power. With offsite power restored, the operator can recover the Power Conversion System (PCS) for decay heat removal. Therefore, for those cutsets where offsite power restoration is applicable, a non-recovery term is incorporated for failure to restore offsite power. The probability of non-recovery of offsite power is given in Table 4 as a function of time. The values were derived from the 4

methodology presented in NUREG/CR-5032, "Modeling Time to Recovery and Initiating Event Frequency for Loss of Offsite j Power Incidents at Nuclear Power Plants" using data for the i Peach Bottom Nuclear Power Plant. (NUREG/CR-4550, Volume 4,  ;

" Analysis of the Damage Frequency, Peach Bottom, Unit 2").

4.2 Recovery of Power Conversion System (PCS)

A TW sequence initiated because of a loss of the PCS can be recovered by the restoration of PCS. Therefore, for those cutsets where PCS restoration is applicable, a non-recovery term is incorporated for failure to restore PCS. Table 5 presents the probability of failing to recover loss of the PCS as a function of time. The data are from NUREG/CR-4550, Vol. 1,

" Analysis of Core Damage Frequency: Internal Events Methodology."

33

4.3 Recovery of DC Safety Bus System A TW sequence initiated because of a loss of a DC safety bus can be recovered by the restoration of DC hardware. i Therefore, for those cutsets where DC bus restoration is applicable, a non-recovery term is incorporated for a failure to restore DC bus. Table 6 presents the probability of failing to recover loss of the DC bus as a function of time. The data are from NUREG/CR-4550, Vol. 1, " Analysis of Core Damage Frequency.

Internal Events Methodology."

4.4 Probability of Failure to Repair or Allan l C22ponents There are additional actions the operating crew can take to recover a TW sequence. These activities involve either

" repairing" a failed component or manually " aligning" a failed component. These two different activities are discussed below. ,

4.4.1 Probability of Failure to ReDair Components The TW sequences are dominated by a failure of four types of components: pumps, valves, battery panels and heat  !

exchangers. The probability of repairing a pu,mp or valve as a function of time is given in Figures 1 and 2, respectively.

These values are from WASH-1400. For a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> sequence time, the probability of failing to repair a valve is 0.1, the failure to repair a pump is 0.3. '

1 The probability of failing to repair a heat exchanger, ,

battery panel and AC power bus are presented as a function of time in Tables 7, 8, and 9, respectively. The average time to -

repair was obtained from the 1984 IEEE Guide to the Collection and Presentation of Electrical, Electronic, Sensing Component, and Mechanical Eculpment Reliability Data for Nuclear Power Generating Station (IEEE Std 500 Reliability Data). The mean value for time to repair battery panel was judged to be approximately 2.5 hrs based on repair times for circuit breakers and relays. The mean value for time to repair heat exchangers was 21.42 hrs. The mean time to repair an AC bus was determined to be 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. The probability of failing to repair within a given amount of time (t) was computed on the basis of an -

i Appendix II exponential modelThe 39, 386 && 404. inferred from exp-formula, WASHt/ 1400,ime avg t to repair,)pages, gives the approximate probability of falhure to repair as a function of time.

34 ,

l 1

4.4.2 Erobability of Failure to Alich Failed ConDonents In a transient in which loss of containment heat removal has occurred, the EOPs direct operators to perform a number of actions. In carrying out the EOPs, valves or other components

, could fail to operate for a number of reasons. .For such failures to be recovered, plant personnel may be required to manually align, close, or open a given component. The generic Human Error Probability (HEP) for failing to recover such a l failure is determined below.

When a valve or other component fails to behave as expected upon manipulation from the control panel, operators should be able to detect the failure immediately because indicator lights on the control panel will not change as expected. Furthermore, since there would likely be 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> or more for the failure to be detected before containment failure would occur, there are numerous other indicators that operators would monitor that should lead them to suspect the possibility of a stuck valve (e.g., no flow, no downstream pressure, or no mitigating effect) or other component failure. Taus, there are numerous indications for such an event and in 40 hrs as many as four different crews would have the opportunity to detect, diagnose, and correct the problem.

Once the problem is detected, walking to the failed component and manually manipulating it could conservatively be accomplished within 30 minutes. Thus, 30 minutes is more than sufficient for diagnosing the problem. Given the numerous indications likely to reveal a problem and the number of operators that would have the opportunit the specific problem, a very conservativ/ to detectfor e estimate anddiagnosis identify failure (according to the ASEP HRAP nominal diagnosis model) would be 1E-5.

The HEP for actually going out and manually aligning the i component once the failure had been detected is based on several facts. First, since the failure occurs after operators are following certain EOPs, the post-diagnosis actions must be considered dynamic. That is, according to ASEP HRAP, a safety system failure while the crew-is using an EOP, creates an environeant calling for a more complicated degree of interaction between the onerators and the equipment. However, because of the familiarity the the time operators available have to with respond the loss to the failure,inment of the conta heat removal sequence, and the fact that only one primary safety system has failed, it can be assumed that the operator is under only moderately high stress.

35 i

l

l Two basic actions are involved in accomplishing the task.

The operators must direct someone(s) to go to the problem loca2 ion and have him or her adjust the component as needed.

That person (s) must ther. walk to the location and perform the task. The HEP for the operator failing to make the assignment (according to ASEP HRAP, Table 8.5) is 0.05. However, because of the numerous indications of any failure to perform the post-diagnosis actions and the amount of time available for the detection, recovery credit can be given for the supervisor and three independent the error. Thus, checks (for the HEP different for crews)is tasks isfailing to detect the post-diagnos computed as follows:

HEP post-diagnosis =(0.05 + 0.05)x0.5x0.5x0.5x0.5

=6.3E-3 l

The HEP for the diagnosis portion of the task was previously determined to be 1E-5. The total HEP for aligning a component is obtained by summing the HEP diagnosis and the HEP post-diagnosis values.

HEP TOT =1E-5+6.3E-3

=6.3E-3 The error factor for this value is 5, with the 6.3E-3 being a median value. The mean value for this term is equal to 1E-2.

(1.67x6.3E-3 based on equation found in NUREG/CR-4550 Vol.1,

Rev. 1, " Analysis of the Core Damage Frequency Internal Events Hethodology", page 8-18).

a s

36 l

~

A --.

___.- . . - - - . -- . . - -- . - . - . - _~ ..

4.5 Determination of HEP for Recovery of Failure To Prevent Automat:,c Switchover of HPCI Prem CST to T.StDAE In a transient in which loss of containment heat removal has occurred, operators are directed by the EOPs (specifically FitzPatrick EOP-2, RPV control) to prevent the automatic l switchover of the HPCI pump suction from the CST to the torus.

If control room personnel fails'to perform the steps necessary to override the automatic switchover, high water level in the torus will induce the switchover in approximately one hour. If ,

the automatic switchover occurs, control room personnel will i have approximately six more hours in which to reverse the switchover before damage occurs. That is, they have six hours l in which to diagnose that the switchover needs to be reversed ,

and to open the HPCI CST suction valve and close the HPCI torus i suction valve. The fact that the switchover has occurred is l

, immediately indicated by indicator lights for the valves I involved. To determine the HEP for this task, the HEPs for diagnosis and task actions must be determined and added together as described in NUREG/CR-4722, " Accident Sequence Evaluation-Program Human Reliability Analysis Procedure" (ASEP HRAP) .

In order to reverse the automatic switchover, an operator l must walk to panel 9-39 in the relay room and lift the transfer l logic leads BB-19 and BB-21. He then must reopen the CST suction valve and close the torus suction valve. A very conservative estimate of the time to reach panel 9-39 and make the necessary changes is 30 min. Thus, at least 5 hrs are available for diagnosing what must be done. Given the fact that the need for override of the swithover is clearly specified in the EOPs, that indications that automatic switchover has occurred appear on the control panel and that the operators are not time stressed, the HEP for diagnosis would be negligible.

However, during this time significant attention is being paid to ,

other containment parameters (which are in jeopardy during these sequences), there is a remote possibility that operators may inadvertently miss the diagnosis. Thus, the HEP for diagnosis .

will be made on the basis of the Nominal Diagnosis Model from '

NUREG/CR-1278, per (ASEP HRA). Use of the-nominal diagnosis  ;

model provides a balance between the long time available and the '

possibility of being burdened with other concerns.- On the basis of this model the diagnosis HEP would be no greater then 1.0E-3 (medium joint HEP, ASEP HRAP, Figure 8.1).

I l

37 l

- , , - , - - . . . - . . - - . - - . -- - 0. . ~.

The HEP for the post-diagnosis actions is determined based on the following. First, the operators are trained in the use of the EOPs. In addition, more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> are available to perform the tasks. The operators are tamiliar with such accident sequences, and only a single safety system has failed (i.e., RHR). Thus the situation is classified (in terms of ASEP HRAP) as requkring step-by-step activities with only moderately high stress.

Two basic activities are involved in completing the task.

The operator must go (or assign someone to go) to panel 9-39 and then perform the appropriate adjustments. Therefore, two errors are possible. According to Table 8-5 in.ASEP HRAP the HEP for an operator failing to go or assign someone to go 1s equal to 0.02.

Similarly, an HEP of 0.02 is also assigned to the task of actually performing the changes at the panel 9-39. However, because the shift supervisor and other control room personnel '

would be aware of the need for the switch override through the EOPs and experience, recovery credit can be given for the shift supervisor and a third party independent check failing to detect the failure to accomplish the override (Table B-5, ASEP HRAP .

Thus, to obtain the total HEP for the post-diagnosis actions),

the HEPs for the two original operator actions must be summed end then multiplied by 0.2 for probability of non-recovery by the supervisor and by 0.2 for non-recovery by a third party.

The resulting HEP is therefore computed as follows:

-HEP post-diagnosis actions = (0.02+0.02)x0.2x0.2 4

= 0.04x0.2x0.2 l

= 1.6E-3 The post-diagnosis HEP of 1.6E-3-is then added to the previously determined HEP for the diagnosis part of the task (IE-3) to obtain the total HEP for overriding the HPCI switch.

l Therefore, HEP TOT = HEP diagnosis + HEP post-diagnosis Actions

= lE-3+1.6E-3

[ = 2.6E-3 i

The error factor for this value according to the ASEP HRAP Table B-5 is equal to 5 with 2.6E-3 being a median value. This i mean value is, therefore, equal to 4.22E-3.

38-i i

l,

)

t

B

5. conclusigna MARCH code analyses calculate that at least 24 to 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />  ;

of recovery time will be available to recover from a loss of decay heat removal capability as a result of a TW sequence. +

During this. period the probability that the heat removal capab111ty increases with time. plant will recover Taking into account recovery actions that call for the  !

remedying of failures or errors, the frequency wh bealossofcontainmentheatremovalis3.17x10~jchtherewill

/ reactor  :

year. This calculation is based upon a recovery analyses for

, the dominant. sequence. A list of Fitzpatrick initiators and associated TW fre@encies are presented in Table 10. Our l 4

frequency prediction is conservative because no credit was taken- '

for the use of existing plant system to nitigate a TW sequence.

For example, the use of the plant's fire water system to provide i cooling water to the'RHR heat-exchanger and the operation of the reactor water cleanup (RWCU) system to remove decay heat from the reactor were not quantified.  ;

h I

i 3

b 39 l

l l  !

' _ _ _ _ ______ _ __ _ _ . . _ _ - _ _ _ _ . - .- --- .. . ~ . _ _ . . . . '

~ +- 4

,,=

. . 1 i

I. L 1- ~

J f . ..

~

I

-, e -

t I -

3

I

' I .t

\ -l I 1- t a g i i

• e -

g I . t P

' = 'i e

I w 3 ,g-i

. I y

. ee -

-t ,

1- g

, ==s .I: ,;

g , 1- -t g . I

!. . a e. _- g = l i l o l I -I i 1 , l I g f ' I

.. _ l 8-

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+ g= g- I g .

4 I 1 '

, or -

" l' .I ,

j I I 1 i -3

, ' I 't- g

e. =

t e i

i .i : i tii. t. i i i - i i nij- i -i 4 i i nsig

= ,,,, .. ,  ;

?

FIGURE 1 Observed Repair Times and Theoretical Distribution - Pumps  ;

i' g.

3 5 .

A

- e 4 Me-..gi.rewM*, eeW 6 s e ww9-54.T P'M ait'er=+1gaw,v--'ey ehMe- beFaN**W"'"'-D'Tt :i'f3'Wehe'*L w' W "We.*ve@W-(v i N'@W9'-T"W41'M--PWM*DWM**s'U4-4M"P- "w&9-D- gt8L6'W'e- Cw asw e'W W* We'P'a-W% 4*w'T 8arn 64' 1M uh 4ee rvu-w*--a . -DepuuluAde w+ b w--s .miew._--Aa. m'v er r w "

g I I 8

s

' t i I I i I f g i t I I ,

8 I -

I ' = *

,e -

I I s t 3 I ,a t **L'f5 i .., ,

! . l l I I

' u i 1 I e t es -

1 I t a a i g H 8

, I g a 8* , I g I  : ,

.. i i ,

-m g  : ,

l 8 i es -

g i t I I I e "

g i I e i t i i i i i iii i i i i i iii i- i + i i i ii v.

FIGURE 2 Observed Repair Times and Theoretical Distribution - Valves

. . . - . ._ _ _ . _ _ _ _ - _ _ - _ - _,-_ __ .- _. _ _ , . - _ -- - , . - - - . . . - - - . , . - ~ . . - . _ . . . - . - - . . - - - _ . -.

l SHEET 1 TABLE 2. MINIMAL CUTSETS FOR DOMINANT SEQUENCES BEFORE RECOVERY I

l EVENT NAME DESCRIPTION PROB.

(Note 1)

1) T1-4

1) T1 LOSS OF OFFSITE POWER INITIATOR 7.01E-05 RSW-CCF-VF-2MOVS CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/P SRV'S RESEAT

/C REACTOR PROTECTION SYSTEM t

2) T1 IDSS OF OFFSITE POWER INITIATOR 5.24E-05 LCI-CCF-PF-4MDPM RHR PUMP COMMON CAUSE FAILURE UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/P SRV'S RESEAT

/C REACTOR PROTECTION SYSTEM

3) ESW-MAI-MA-LOOPB ESW LOOP "B" UNAVAIIABLE DUE TO MAINT. 3.29E-05

" T1 LOSS OF OFFSITE POWER INITIATOR LCI-XHE-RE-FM3AP FAIL TO RESTO PM-3A PATH CMPTS AFT MAIN UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/P SRV'S RESEAT

/C REACTOR PROTECTION SYSTEM

4) T1 IDSS OF OFFSITE POWER INITIATOR 2.52E-05 ESW-MDP-FR-P2B PUMP 46P-2B FAILS TO CONTINUE RUNNING LCI-XHE-RE-PM3AP FAIL TO RESTO PM-3A PATH CMPTS AFT MAIN UlX-SUCC- SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/P SRV'S' RESEAT

/C REACTOR-PROTECTION SYSTEM

5) T1 LOSS OF OFFSITE POWER INITIATOR 2.52E-05 ESW-MDP-FR-P2A PUMP 46P-2A FAILS TO CONTINUE. RUNNING LCI-XHE-RE-PM3DP FAIL TO RESTO PM-3D PATH CMPTS AFT MAIN UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/P SRV'S RESEAT

/C REACTOR PROTECTION SYSTEM

6) - ESW-MAI-MA-LOOPA ESW IDOP "A" UNAVAILABLE DUE TO MAINT. 2.30E-05 T1 LOSS OF OFFSITE POWER INITIATOR LCI-XHE-RE-PM3DP FAIL TO RESTO PM-3D PATH CMPTS AFT MAIN UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/P SRV'S RESEAT

/C REACTOR PROTECTION SYSTEM

SHEET 2 EVENT NAME DESCRIPTION PROB. (Note 1)

ESW-MAI-MA-IDOPB ESW LOOP "B" UNAVAIIABLE DUE TO MAINT. 1.58E-05

7) T1 - IDSS OF OFFSITE POWE R NITI R -

LCI-HTX-VF-HE-2A IDOP A HEAT EX6GE1 -2A UJ.TRE UlX-SUCC SUCpESSFU".c ALIGNMENT O HPC SUCT TO CS

~/P SRV S RES. MT

/C REACTOR P:ROTECTION SYSTEM  ;

RSW-CCF-PF-4MDPS C I O W PUMPS UlX-SUCC SUCpESSFUL ALIGNMENT OF HPCI SUCT TO CS

/C REA CTION SYSTEM

9) T1 ESW-MDP-FR-P2B IDSS OF OFFSITE POWER PUMP 46P-2B FAILS TO R

ING 1.21E-05 LCI-HTX-VF-HE-2A IDOP A HEAT EXCHANGER E-2A I UlX-SUCC SUCpESSFUL SRV S RESEAT ALIGNMENT OF HPCI S CT TO CS

/P

/C REACTOR PROTECTION SYSTEM

10) T1 ESW-MDP-FR-P2A LOSS OF OFFSITE POWER INITIATOR PUMP 46P-2A FAILS TO CONTINUE RUNNING 1.21E-05 -

e. LCI-HTX-VF-HE-2B IDOP B HEAT EXCHANGER E-2B FA w UlX-SUCC SUCpESSFUL ALItennt.,nr OF HPCI TO CS- -

/P SRV S RESEAT t

'/C REACTOR PROTECTION SYSTEM

11) ESW-MAI-MA-LOOPA ESW LOOP "A" UNAVAILABLE DUE TO MAINT. 1.11E-05 T1 IDSS OF OFFSITE POWER INITIATOR  ;

LCI-HTX-VF-HE-2B IOOP B HEAT 3XCHANGER E-2B FAILURE '

UlX-SUCC SUCpESSFUL ALIGNMENT OF HPCI SUCT TO CS

/C RNCTO CTION SYSTEM ESW-XHE-RE-ESW3B I RES RE ES BA TEST LCI-XHE-RE-FM3AP FAIL TO RESTO PM-3A PATH CMPTS MAIN UlX-SUCC SUCpESSFUL ALIGNMENT OF HPCI SU TO CS P SRV S RESEAT C REACTOR PROTECTION SYSTEM C _m. _ _ _ _ _ _ _ _ _ _ -:____.________ -+ _ - ___w.=-

4en-w. ""' <=ee+ -w-== +w.+ 3-ee- e ,e w-- -w ' ' * - - iw-m-'v~ =- e =w.- -*-*=w ,w.--v+= e-m -

w ---w, m -.-w-e, ee-, e+., .e.-o-,-. - --..

SHEET 3 4

EVENT NAME ,

DESCRIPTION PROB. (Note 1)

13) T1 LOSS OF OFFSITE POWER INITIATOR 1.00E-05 IE $ 2 E E E E E = b A3 kb^ RES b Per M A P PH y. TO CSliAIN UlX-SUCC SUCpESSFUL ALIG O f HkI SU

/P SRV S RESEAT

/C REACTOR PROTECTION SYSTEM

14) T1 ESW-XHE-RE-ESW3A LOSS FAILURE OF TO OFFSITE_RE RESIU POWER INITIATOR 46ESW-3A AFTER TEST 1.00E-05 LCI-XHE-RE-FM3DP FAIL TO RESTO PM-3D PATH CNPTS AFT MAIN UlX-SUCC SUCpESSFUL ALIGNMENT OF HPCI SUCT TO CS

/P SRV S RESEAT

/C REACTOR PROTECTION SYSTEN

15) T1 AC4-XHE-MC-UVRIA LOSS OF OFFSITE POWER INITIATOR 1.00E-05 MISCALIBRATgN OF BUS gS g g g UXS2 C Sh F ALI F HPCI SUCT TO

/P SRV RESEAT

/C REACTOR PROTECTION SYSTEM i

I i

i

_ _ - . - , _ _ . -- - . - . _ - - - . - . - - - _ - - _ . _ _ _ _ _ - _ _ + _ . _ . . _ _ a , ,~ .n. --,-.v+. - -w a w~. . ----...n. . - --- -- . - - - . . . . . _ _ ~ - . - . , , . . - - - . . - . _ . - - - _ _ _ . - . _ - - _ - - . -

SHEET C EVENT NAME DESCRIPTION PROB. (Note 1)_

1) T1-14

1) T1 RSW-CCF-VF-2MOVS LOSS OF OFFSITE POWER INITIATOR CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN 1.33E-06 CI-MAI-MA-HPCIU HPCJ P SRV SUNAVAILABLE RESEAT DUE TO MAINTENANCE C REACTOR PROTECTION SYSTEM

2) T1 LCI-CCF-PF-4MDPM IDSS OF OFFSITE POWER INITIATOR RHR PUMP COMMON CAUSE FAILURE 9.96E-07 CI-MAI-MA-HPCIU HPCJ P SRV S UNAVAIIABLERESEAT DUE TO MAINTENANCE C REACTOR PROTECTION SYSTEM

3) ESW-MAI-MA-IDOPB ESW IDOP "B" UNAVAIIABLE DUE TO MAINT. 6.26E-07 T1 IDSS OF OFFSITE POWER INITIATOR LCI-XHE-RE-PM3AP FAIL TO RESTO PM-3A PATH CMPTS AFT MAIN CI-MAI-MA-HPCIU HPCJ UNAVAILABLE DUE TO MAINTENANCE

/C R CTION SYSTEM

4) T1 IDSS OF OFFSITE POWER INITIAM R 4.79E-07 RSW-CCF-VF-2MOVS CCF OF RHRHX SW DISCH VLVS FAIL M OPEN HCI-TDP-FS-HCIPM TUR DRIV PUMP FAILS TO START UPON DEMAN

/P SRV'S RESEAT

'C

/ REACTOR PROTECTION SYSTEM

5) T1 M W-MDP-FR-P2A IDSS OF OFFSITE POWER INITIATOR PUMP 46P-2A FAILS TO CONTINUE ING 4.78E-07 7, TI-XHE-RE-FM3DP FAIL TO RESTO PM-3D PATH MAIN-HCI-MAI-MA-HPCIU HPCJ UNAVAILABLE DUE TO MAI ANCE

/P SRV S RESEAT

/C REACTOR PROTECTION SYSTEM

6) T1 LOSS OF'OFFSITE POWER INITIATOR 4.78E-07 ESW-MDP-FR-P2B PUMP 46P-2B FAILS TO CONTINUE RUNNING LCI-XHE-RE-FM3AP FAIL TO RESTO PM-3A PATH CMPTS AFT MAIN CI-MAI-MA-HPCIU HPCJ P SRV SUNAVAILABLE RESEAT DUE TO MAINTENANCE C REACTOR PROTECTION SYSTEM

, _ _ _ _ ___ . _ . . _ _ _ _ _ _ _ . . _ . _ _ _ _ _ _ . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _~ _ _ _ _ . . . . . . _ . . . . . _ . . _ . . _ _ - _ _ _ _ . _ . . _ . _ _

SHEET 5 EVENT NAME DESCRIPTION ~ PROB. (Note 1)_

7) ESW-MAI-MA-IDOPA ESW IOOP "A" UNAVAI BLE DUE TO MAINT. 4.38E-07 T1 IDSS OF OFFSITE pgITIA* X)R

-HP P A@A B TO MAI INANCE P SRV S RESEAT

/C REAC'IT)R PROTECTION SYSTEM

8) T1 RSW-CCF-VF-2MOVS IDSS OF OFFSITF POWER INITIATOR CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN 4.21E-07 HCI-TDP-FR-HCIPM TUR DRIVE PUMP FAILS TO CONTINUE RUNNIN  ;

/P SRV'S RESEAT '

/C REACTOR PROTECTION SYSTEM

9) T1 LCI-CCF-PF-4MDPM IDSS OF OFFSITE POWER INI RHR PUMP COMMON CAUSE FAI TOR 3.58E-07 CI-TDP-FS-HCIPM TUR DRIV PUMP FAILS TO ST UPON DEMAN P SRV'S RESEAT C REACTOR PROTECTION SYSTEM

10) T1 LOSS OF OFFSITE POWER INITIATT)R 3.15E-07 LCI-CCF-PF-4MDPM RHR PUMP COMMON CAUSE FAILURE HCI-TDP-FR-HCIPM TUR DRIVE PUMP FAILS 'IT) CONTINUE RUNNIN

/P SRV'S RESEAT

/C REACTOR PROTECTION SYSTEM

$ 11) TI IDSS OF OFFSITE POWER INITIATOR 3.04E-07 RSW-CCF-VF-2MOVS CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN CI-PHN-FU-SPEXH FAI KSOF STEAM SUPPLY / EXHAUST PATHS P SRV RESEAT C REACTOR PmirEvnON SYSTEM ESW-MAI-MA-LOOPB ESW IDOP "B" UNAVAILABLE DUE 'IT) MAINT. 3.00E-07

12) T1 LOSS OF OFFSITE POWER INITIATOR LCI-HTX-VF-HE-2A IDOP A HEAT EXCHANGER E-2A FAILURE CI-MAI-MA-HPCIU HPCJ P SRV SUNAVAILABLE RESEAT DUE TO MAINTENANCE C REACTOR PROTECTION SYSTEM

.l i

^

__ ___ _ ___ _ - _ - = _ _ _ _ _ - _ _ _ _ . _ _ _ _ _ _ - - - _ .. .. . .- -. - _ . -. - _ - - - _ ____ _ -.-

SHEET G EVENT NAME DESCRIPTION PROB. (Note 1)

13) T1 IDSS OF OFFSITE POWER INITIATOR 2.69E-07 Db -

BhB P L A TS B D BCB-2

/P SRV S RESEAT

/C REACTOR PROTECTION SYSTEM

14) T1 IDSS OF OFFSITE POWER INITIATOR 2.66E-07 "5Y ME[ S HP8 U P

H b SRV S RESEAT LABLE N MA NTEN

/C REACTOR PROTECTION SYSTEM

15) T1 ESW-MDP-FR-P2A IDSS OF OFFSITE POWER INITIAMR PUMP 46P-2A FAILS TO CONTINUE RUNNING 2.29E-07 LCI-HTX-VF-HE-2B IDOP B HEAT EXCHANGER E-2B FAIILTE CI-MAI-MA-HPCIU HPCJS UNAVAIIABLE DUE TO MAINTENANCE P SRV RESEAT C REACTOR PROTECTION SYSTEM

SHEET 7 EVENT NAME DESCRIPTION PROB. (Note 1)

1) T2-4

1) RSW-CCF-VF-2MOVS CCF OF RHRHX SW DISCH VLVS FA::L "O OPEN 7.01E-05 T2 IDSS OF POWER CONVERSION SYS ::NI'IATOR UlX-SUCC SUCCESS ALIGNMENT OF HPCI SUC" TO CS

/C

/P REAg"S SRV IVRRESEA CTION SYSTEM LCI-CCF-PF-4MDPM RHR PUMP COMMON CAUSE FAILURE 5.24E-05

2) T2 IDSS OF POWER CONVERSION SYS INITIATOR '

UlX-SUCC SUCCESS W L ALIGNMENT OF HPCI SUCT TO CS

/C REACTOR PROTECTION SYSTEM

/P SRV'S RESEAT

3) RSW-CCF-PF-4MDPS COMMON CAUSE FAILURE OF RHRSW PUMPS 1.40E-05 T2 LOSS OF POWER CONVERSION SYS INITIA'IUR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/C REApTOR PROTECTION SYSTEM

/P SRV S RESEAT

4) LCI-HTX-VF-HE-2A LCI-HTX-VF-HE-2B LOOP A HEAT EXCHANGER E-2A FAILURE IDOP B HEAT EXCHANGER E-2B FAILURE 2.85E-06 T2 IDSS OF POWER CONVERSION SYS INITIATOR '

UlX-SUCC SUCCESSFUL ALIGNNENT OF HPCI SUCT TO CS e /P S b '

5) LCI-HTX-VF-HE-2B LOOP B HEAT EX - E-2B FA::LUltE 1.24E-06 T2 IDSS OF POWEt ION SYS ::NI'IATOR UlX-SUCC SUCCESSW L Alt,ana. OF HPCI SUC' TO CS RSW-RCK-NO-MV89A VALVE Cupradj :,CTIONCIRCU:SYSTEM-

T NO ourrur

/C REA TOR SRVgS PRO:

RESEAT

/P

6) LCI-HTX-VF-HE-2A LOOP A HEAT EXCHANGER E-2A FAILURE 1.24E-06 T2 IDSS OF POWER CONVERSION SYS INITIATOR NO-MV89B VE un T PU

/C REAgTORPROTkCTIONSYSTEM SRV S RESEAT

/P

SHEET C EVENT NAME DESCRIPTION PROB. (Note 1)

FAIL TO RES70 PM-3A PATH CMPTS AFT MAIN 1.10E-06 ID REL 86-1HOEB03 PATH FAILURE

7) AC4-RLY-NO-HOEB3LCI-XHE-RE-PM3AP T2 IDSS OF POWER CONVERSION SYS INITIATOR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/C REA SRVpTOR PROTECTION SYSTEM

/P S RESEAT

8) LCI-XHE-RE-FM3AP FAIL TO RESTO PM-3A PATH CMPTS AFT MAIN- 1.10E-06 AC4-RLY-NO-HOEB1 REL 86A-1HOEB01 PATH FAILURE T2 LOSS OF POWER CONVERSION SYS INITIATOR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/C REA R PROTECTION SYSTEM

/P SRV S RESEAT

9) T2 LOSSOFPOWERCONVERSIONSYSINyATOR 5.42E-07 ChNO-MV89A A $RC O PUh RSN-RCK-NO-MV89B VALVE CONTROL CIRCU:. NO OUTPUT REAg CTION SYSTEM

10) LCI-HTX-VF-HE-2A IDOP A HEAT EXCHANGER E-2A FAILURE 5.30E-07 AC4-RLY-NO-HOEB3 ID REL 86-1HOEB03 PATH LOSS OF PONER Cuny FAI ION SYS NITIATOR T2 UlX-SUCC IG OF HPCI UCT TO CS

"" 8'8'""

# SUCCESSFUL si?9 SRESER ANc

11) LCI-HTX-VF-HE-2A LOOP A HEAT EXCHANGER E-2A FAILURE 5.30E-07 AC4-RLY-NO-HOEB1 REL 86A-1HOEB01 PATH FAILURE T2 IDSS OF POWER CONVERSION SYS INITIATOR -

UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCt TO CS C REA SRVpTOR PROTECTION SYSTEM P- S RESEAT

t SHEET D i EVENT NAME DESCRIPTION PROB. (Note 1) --

l

' 12) LCI-XHE-RE-FM3DP FAIL TO RESTO PM-3D PATH CMPTS A]T MAIN 2.69E-07 T2 IDSS OF POWER CONVERSION SYS INL' 7IATOR UlX-SUCC SU SFUL ALIGNMENT OF HPCI SUC'? TO CS DC1-BDC-ST-10500 PAN FAULTS BY ANY IDAD 10500 I

' /C REA R PROTECTION SYSTEM

/P SRV S RESEAT

13) LCI-XHE-RE-PM3DP FAIL TO RESTO PM-3D PATH CMPTS A1T MAIN 2.69E-07 T2 IDSS OF POWER CONVERSION SYS INI*7IAMR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUC'? TO CS DCl-BDC-ST-DC-A4 PANEL FAULTS BY ANY IDAD 71DC-A4

/C REA SRVgTOR PROTECTION SYSTEM 1

/P S RESEAT FAIL TO RESTO PM-3D PATH CMPTS AFT MAIN 2.69E-07

14) T2LCI-XHE-RE-FM3DP LOSS OF POWER CONVERSION SYS INITIATOR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS -

DCl-BDC-ST-BCB2A PANEL FAULTS BY ANY IDAD BCB-2A

/C REACTOR PROTECTION SYSTEM

/P SRV'S RESEAT LCI-XHE-RE-PM3AP FAIL TO RES*IV PM-3A PATH CMPTS AFT MAIN 2.69E-07

15) T2 IDSSOFPOWERCONVERSIONSYSIgITIATUR D ST-10600 PA FA 60 e, /C REA R SRV yS RESEAT C ON SYS o /P 4

a s

1 i

e -

---

._____-_n..__----. -_.-__- . _ - - _ . -- _ . . . . - . _ - - . . _ _ _ - - - - - - ~ <

SHEET 10-EVENT NAME DESCRIPTION -

PROB. (Note 1)

1) T1-33-S2-37 CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN 7.45E-06

1) RSW-CCF-VF-2MOVS T1 LOSS OF OFFSITE POWER INITIATOR P1 ONE STUCK OPEN RELIEF VALVE

./C REACTOR PROTECTION SYSTEM

2) LCI-CCF-PF-4MDPM RHR PUMP COMMON CAUSE FAI 5.57E-06 T1 IDSS OF OFFSITE POWER INI ATOR 1 P1 ONE STUCK OPEN RELIEF VA

/C REACTOR PROTECTION SYSTEM RSW-CCF-PF-4MDPS COMMON CAUSE FAILURE OF RHRSW PUMPS 1.49E-06

3) T1 LOSS OF OFFS TE INITIATOR

/b R C CT O SS

• ) IfI:llH:?F:lif:2i 1285 a lifiFHeili!!8"" E:li FilliiH . 2E-07 T1 IDSS OF OFFSITE POWH INITIATOR P1 ONE STUCK OPEN RELIEF VALVE

/C - REACTOR PROTECTION SYSTEM

5) LCI-HTX-VF-HE-2A IDOP A HEAT EXCHANGER E-2A FAILURE 1.32E-07 T1 IDSS OF OFFSITE POWER INITIATOR P1 ONE STUCK OPEN RELIEF VALVE

" RSW-RCK-NO-MV89B VALVE CONTROL CIRCUIT NO OUTPUT

/C REACTOR PROTECTION SYSTEM

6) LCI-HTX-VF-HE-2B LOOP B HEAT EXCHANGER E-2B FAILURE 1.32E-07 T1 IDSS OF OFFSITE POWER INITIATOR P1 ONE STUCK OPEN RELIEF VALVE RSW-RCK-NO-MV89A VALVE CONTROL CIRCUIT NO OJTPUT

/C REACTOR PROTECTION SYSTEM

7) AC4-RLY-NO-HOEB1 REL 86A-1HOEB01 PATH FAILURE 1.17E-07 LCI-XHE-RE-FM3AP FAIL TO RESTO PM-3A PATH CMPTS AFT MAIN T1 IDSS OF OFFSITE POWER INITIATOR

. P1 ONE STUCK OPEN RELIEF VALVE 4~

/C REACTOR PROTECTION SYSTEM T1

~

~

kL IDSS OF OFFSITE POWER P

ITIAIUR AFT MAIN P1 ONE STUCK OPEN RELIEF VALVE

/C REACTOR PROTECTION SYSTEM i

l

SHEET 11 EVENT NAME __

DESCRIPTION PROB. (Note g

9) AC4-RLY-NO-HOEB3 ID REL 86-1HOEi303 PATH FAILURE 1.17E-07 LCI-XHE-RE-FM3AP FAIL TO RESTO PM-3A PATH CMPT3 AFT MAIN T1 LOSSS OF

' OFFSITE POWER INgA'IVR JC REC u"R FR8H"riWHS?4M C

10) T1 P1 IDSS OF OFFSITE POWER INITIATOR ONE STUCK OPEN RELIEF VALVE 5.75E-08 RSW-RCK-NO-MV89A VALVE CONTROL CIRCUIT NO OUTPUT RSW-RCK-NO-MV89B VALVE CONTROL CIRCUIT NO OUTPUT

/C REACTOR PROTECTION SYSTEM

11) AC4-RLY-NO-HOEB3 ID REL 86-1HOEB03 PATH FAILURE 5.62E-08 LCI-HTX-VF-HE-2A LOOP A HEAT EXCHANGER E-2A FAILURE T1 LOSS OF OFFSITE POWER INITIATOR P1 ONE STUCK OPEN RELIEF VALVE

/C REACTOR PROTECTION SYSTEM

12) AC4-RLY-NO-HOEB1 REL 86A-1HOEB01 PATH FAILURE 5.62E-08 LCI-HTX-VF-HE-2A IDOP A HEAT EXCHANGER E-2A FAILURE T1 IDSS OF OFFSITE POWER INITIATOR P1 ONE STUCK OPEN RELIEF VALVE

/C REACTOR PROTECTICid SYSTEM km 13) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 5.62E-08 M LCI-HTX-VF-HE-2B LOOP B HEAT EXCHANGER E-2B FAILURE T1 IDSS OF OFFSITE POWER INITIATOR P1 ONE STUCK OPEN RELIEF VALVE

/C REACTOR PROTECTION SYSTEM

14) DCl-BDC-ST-BCB2B PA.NEL FAULTS BY ANY IDAD BCB-2B 2.86E-08 LCI-XHE-RE-FM3AP FAIL TO RESTO PM-3A PATH CMPTS AFT MAIN T1 IDSS OF OFFSITE POWER INITIATOR P1 ONE STUCK OPEN RELIEF VALVE

/C REACTOR PROTECTION SYSTEM

15) LCI-XHE-RE-PM3DP FAIL TO RESTO PM-3D PATH CMPTS AFT MAIN 2.86E-08 T1 IDSS OF OFFSITE POWER INITIATOR P1 ONE STUCK OPEN RELIEF VALVE DCl-BDC-ST-10500 PANEL FAULTS BY ANY ICAD 10500

/C REACTOR PROTECTION SYSTEM

SHEET 12 EVENT NAME DESCRIPTION PROB. (Note 1)

1) T2-34-S1-3

1) RSW-CCF-VF-2MOVS CCF OF RHRMX SW DISCH VLVS FAIL TO OPEN 1.01E-07 T2 IDSS OF POWER CONVERSION SYS INITIATOR P2 TWO STUCK OPEN RELIEF VALVES

/C REACTOR PROTECTION SYSTEM

2) LCI-CCF-PF-4MDPM RHR PUMP COMMON CAUSE FAILURE 7.54E-08 T2 LOSS OF POWER CONVERSION SYS INITIATOR P2 TWO STUCK OPEN RELIEF VALVES

/C REACTOR PROTECTION SYSTEM

3) RSW-CCF-PF-4MDPS COMMON CAUSE FAILURE OF RHRSW DUMPS 2.01E-08 T2 IDSS OF POWER CONVERSION SYS INITIATOR P2 TWO STUCK OPEN RELIEF VALVES

/C REACTOR PROTECTION SYSTEM

4) T2P2 IDSS OF POWER CONVERSION SYS INITIATOR TWO STUCK OPEN RELIEF VALVES 7.36E-09 LCI-MOV-CC-MV16A N.C. MOV16A DOES NOT OPEN LCI-MOV-CC-MV16B N.C.-MOV16B DOES NOT OPEN

/C REACTOR PROTECTION SYSTEM l 5) LCI-HTX-VF-HE-2A IDOP A HEAT EXCHANGER E-2A FAILURE 5.49E-09

' T2 IDSS OF POWER CONVERSION SYS INITIATOR P2 TWO STUCK OPEN RELIEF VALVES

$ LCI-MOV-CC-MV16B N.C. MOV16B DOES NOT OPEN  ;

/C REACTOR PROTECTION SYSTEM -

6) LCI-HTX-VF-HE-2B T2 IDOP B HEAT EX LOSS OF POWER CO ER E-2B FAILURE ION SYS INITIATOR 5.49E-09 P2 TWO STUCK OPEN REL EF VALVES LCI-MOV-CC-MV16A N.C. MOV16A DOES NOT OPEN

/C REACTOR PROTECTION SYSTEM

7) LCI-HTX-VF-HE-2A LCI-HTX-VF-HE-2B "DOP A ifFAT EXCHANGER E-2A FAILURE

_OOP B HEAT EXCHANGER E-2B FAILURE 4.10E-09 T2 EDSS OF POWER CONVERSION SYS INITIATOR P2 TWO STUCK OPEN RELIEF VALVES

/C REACTOR PROTECTION SYSTEM

8) T2P2 IDSS OF POWER CONVERSION SYS INITIATOR TWO STUCK OPEN RELIEF VALVES 3.40E-09 LCI-RLY-FU-K84A INTERIDCK RLY 10A-K84A PATH FAULTS I-MOV-CC-MV16B N.C. MOV C REACTORl'ROTECTION

.16B DOES NOT SYSTEMOPEN i

_ _ ._- .- _ _ . _. . -.. . . _ . ... , . - . ~ - -.

SHEET 13 EVENT NAME DESCRIPTION PROB. (Note 1)

9) T2P2 IDSS OF POWER CONVERSION SYS INITIATOR TWO STUCK OPEN RELIEF VALVES 3.40E-09 LCI-MOV-CC-MV16A N.C. MOV16A DOES NOT OPEN LCI-RLY-FU-K84B INTERLOCK RLY 10A-K84 ATH FAULTS

, /C REACTOR PROTECTION SY

10) -LCI-HTX-VF-HE-2A LOOP A HEAT EXCHANGER E-2A FAILURE 2.54E-09 T2 IDSS OF POWER CONVERSION SYS INITIATOR P2 TWO STUCK OPEN RELIEF VA LCI-RLY-FU-K84B INTERLOCK RLY 10A-K84B TH FAULTS

/C REACTOR PROTECTION SYS IDOP B HEAT EXCHANGER E-2B FAILURE 2.54E-09

11) T2LCI-HTX-VF-HE-2B LOSS OF POWER CONVERSION S INITIATOR P2 TWO STUCK OPEN RELIEF VA S LCI-RLY-ETJ-K84A INTERLOCK RLY 10A-K84A TH FAULTS

/C REACTOR PROTECTION SYSTEM

12) T2 IDSS OF POWER CONVERSION SYS INITIATOR 2.39E-09 P2 TWO STUCK OPEN RELIEF VALVES RSW-RCK-NO-MV89A VALVE CONTROL CIRCUIT NO OUTPUT LCI-MOV-CC-MV16B N.C. MOV16B DOES NOT OPEN

/C REACTOR PROTECTION SYSTEM E

13) T2 P2 IDSS OF POWER CON 7ERSION SYS INITIATOR TWO STUCK OPEN RELIEF VALVES 2.39E-09 RSW-RCK-NO-MV89B VALVE CONTROL CIRCUIT NO OUTPUT LCI-MOV-CC-MV16A N.C. MOV16A DOES NOT OPEN

/C REACTOR PROTECTION SYSTEM

14) LCI-HTX-VF-HE-2A T2 IDOP A HEAT EXCHANGER E-2A FAILURE IDSS OF POWER CONVERSION SYS INITIATOR-1.79E-09 P2 TWO STUCK OPEN RELIEF VALVES RSW-RCK-NO-MV89B VALVE CONTROL CIRCUIT NO OUTPUT

/C REACTOR PROTECTION SYSTEM

15) LCI-HTX-VF-HE-2B T2 IDOP B HEAT EXCHANGER E-2B FAILURE IDSS OF POWER CONVERSION SYS INITIATOR 1.75rE-09 P2 TWO STUCK OPEN RELIEF VALVES RSW-RCK-NO-MV89A VALVE CONTROL CIRCUIT NO OUTPUT

/C REACTOR PROTECTION SYSTEM

q, SHEET 14

. EVENT NAME DESCRIPTION PROB. (Note 1)

1) S1 1) RSW-CCF-VF-2MOVS CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN 7.76E-05 S1 INTERMEDIATE IDCA INITATOR

/C REACTOR-PROTECTION SYSTEM-

2) LCI-CCF-PF-4MDPM RHR PUMP COMMON CAUSE FAILURE 5.80Z-05 S1 INTERMEDIATE LOCA INITATOR

/C REACTOR PROTECTION SYSTEM

3) RSW-CCF-PF-4MDPS COMMON CAUSE FAILURE OF RE!SW PUMPS 1.55E-05 S1 INTERMEDIATE LOCA INITATOR

/C REACTOR PROTECTION SYSTEM

4) S1 INfERMEDIATE IOCA INITATOR 5.66E-06 LCI-MOV-CC-MV16A N.C. MOV16A DOES NOT OPEN LCI-MOV-CC-MV16B N.C. MOV16B DOES NOT OPEN

/C REACTOR PROTECTION SYSTEM

5) LCI-HTX-VF-HE-2B- IDOP B HEAT EXCHANGER E-2B FAILURE 4.22E-06 S1 INTERMEDIATE IDCA INITATOR LCI-MOV-CC-MV16A N.C. MOV16A DOES NOT OPEN

/C REACTOR PROTECTION SYSTEM

.6) LCI-HTX-VF-HE-2A LOOP A HEAT EXCHANGER E-2A FAILURE 4.22E-06 S1- -

INTERMEDIATE LOCA INITATOR m

LCI-MOV-CC-MV16B N.C. MOV16B DOES NOT OPEN

./C REACTOR PROTECTION SYSTEM

7) LCI-HTX-VF-HE-2A IDOP A HEY sf' RANGER E-2A FAILURE 3.15E-06 LCI-HTX-VF-HE-2B IDOP B HEAT %> CHANGER E-2B FAILURE S1 INTERMEDIATE LOCA INITATOR

/C REACTOR. PROTECTION SYSTEM

8) S1 INTERMEDIATE ~LOCA INITATOR 2.62E-06 LCI-MOV-CC-MV16A N.C. MOV16A DOES NOT OPEN I-RLY-FU-K84B INTERLOCK RLY 10A-K84B PATH FAULTS C - REACTOR PROTECTION SYSTEM ey. --ue ,,ar. p..-% .- - - w ,c- gaa, i :.~'-...

-ar- -a-w n.- -e--u -_s -_m--* --, . . - . - ____w--. --,------ar--- _ _m..

. .. , . , . . .. __q

=mu

~ '1

. SHEET 15 EVENT _NAME DESCRIPTION PROB. (Note 1) . _

9) S1 INTERMEDIATE LOCA INITIATOR 2.62E-06 LCI-RLY-FU-K84A INTERIhCK RLY 10A-K84A PATH FAULTS LCl-MOV-CC-MV16B N.C. MOV16B DOES NOT OPEN

/C REACTOR ~ PROTECTION SYSTEM

10) LCI-HTX-VF-HE-2A IDOP A HEAT EXCHANGER E-2A FAILURE 1.95E-06 S1 INTERMEDIATE IDCA INITATOR LCI-RLY-FU-K84B INTERIDCK RLY 10A-K84B PATH FAULTS

, /C REACTOR PROTECTION SYSTEM

11) LCI-HTX-VF-HE-2B IDOP B HEAT EXCHANGER E-2B FAILURE 1.95E-06 S1 INTERMEDIATE IDCA INITATOR INTERIDCK RLY 10A-K84A PATH FAULTS LCI-RLY-FU-K84A

/C REACTOR PROTECTION SYSTEM-

12) LSW-RCK-NO-MV89B -INTERMEDIATE IDCA INITATOR 1.84E-06 VALVE CONTROL CIRCUIT NO OUTPUT LCI-MOV-CC-MV16A N.C. MOV16A DOES NOT OPEN

/C REACTOR PROTECTION SYSTEM

13) S1 INTERMEDIATE IDCA INITATOR -1.84E-06 RSW-RCK-NO-MV89A- - VALVE CONTROL CIRCUIT NO OUTPUT LCI-MOV-CC-MV16B N.C. MOV16B DOES NOT OPEN

/C- REACTOR PROTECTION SYSTEM

14) LCI-HTX-VF-HE-2A IDOP A HEAT EXCHANGER E-2A FAILURE 1.37E-06 ui S1 INTERMEDIATE IDCA INITATOR m RSW-RCK-NO-MV89B VALVE CONTROL CIRCUIT NO OUTPUT

/C REACTOR. PROTECTION SYSTEM e

a ..,.m.__., __y

.- - .e m. .O_- .=_u._.a -mem-w---

. -.o-hm-_ -,%-v--.a -,..~ e--_eamm %,.A.,_Aa.___ m. si . . - - . . _ - - . .g.__.

SHEET 165 .

EVENT NAME DESCRIPTION PROB. (Note 1),

1) S2-5

1) DCl-BDC-ST-BCB2B PANEL FAULTS .BY ANY LOAD BCB-2B 2.97E-07 LCI-XHE-RE-FM3AP FAIL TO-RESTO PM-3A PATH CMPTS AFT MAIN S2 SMALL IDCA INITIATOR

/C REACTOR PROTECTION SYSTEM

2) TBC-RCK-NO-P2B PUMP 37P-2B CONTROL CIRCUIT FAILURE 1.94E-07 RSW-CCF-VF-2MOVS CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN SMALL LOCA INITIATOR S2

/C -

REACTOR PROTECTION SYST N .

RS b= - MhVS C O NXkWDSCH SMALL'IDCA INITIATOR L FA L OPEN S2-

/C REACTOR PROTECTION SYSTEM

4) ACO-MAI-MA-XFRT3 TRANSFORMER T3 IN MAINTENANCE 1.92E-07 RSW-CCF-VF-2MOVS CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN S2 SMALL LOCA INITIATOR-

/C REACTOR-PROTECTION SYSTEM

5) DCl-BDC-ST-BCB2B PANEL FAULTS BY ANY IDAD BCB-2B 1.91E-07 S2 ' SMALL IDCA INITIATOR LCI-MOV-CC-MV16A N.C.-MOV16A DOES-NOT OPEN

/C REACTOR PROTECTION SYSTEM.

• TRANSFORMER' T2 IN MAINTENAITCE 1.90E-07

" 6) ACO-MAI-MA-XFRT2 RSW-CCF-VF-2MOVS- CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN

-S2 SMALL IDCA -INITIATOR

./C REACTOR PROTECTION SYSTEM

7) TBC-MDP-MA-P2B PUMP 37P-2B UNAVAILABLE DUE TO MAINT. 1.67E-07 RSW-CCF-VF-2MOVS CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN S2 SMALL LOCA' INITIATOR

/C REACTOR PROTECTION SYSTEM-

~

8)Tgg;ggg;gp2$$$PM L E PUMb CSM00 PCAUSE hA M ^

88 iMhtORNOffMIersYSrEM E .. _ __ _. _ _ _ __ _ ..__ __ __ _ _ _ _ _ _ _ __ _

SHEET 17 EVENT NAME DESCRIPTION

. PROB. (Note 1)

9) TBC-RCK-NO-P2A PUMP 37P-2A CONTROL CIRCUIT FAILURE 1.45E-07 LCI-CCF-PF-4MDPM RHR PUMP COMMON CAUSE FAILURE -

S2- SMALL LOCA IN1TIATOR-

/C REACTOR PROTECTION SYSTEM

10) ACO-MAI-MA-XFRT3 TRANSFORMER T3 IN MAINTENANCE 1.43E-07 LCI-CCF-PF-4MDPM RHR PUMP COMMON CAUSE FAILURE S2 SMALL LOCA INITIATOR

/C REACTOR PROTECTION SYSTEM ,

11) DCl-BDC-ST-BCB2B- -PANEL FAULTS- BY ANY LOAD BCB-2B 1.43E-07 '

LCI-HTX-VF-HE-2A IDOP A HEAT EXCHANGER E-2A FAILURE S2 SMALL LOCA INITIATOR

/C REACTOR PROTECTION SYSTEM '

12) ACO-MAI-MA-XFRT2 TRANSFORMER T2 IN MAINTENANCE 1.42E-07 LCI-CCF-PF-4MDPM RHR PUMP COMMON CAUSE FAILURE S2 SMALL LOCA INITIATOR

/C REACTOR PROTECTION SYSTEM.

13) TBC-MDP-MA-P2B PUMP 37F -2B UNAVAILABLE DUE TO MAINT. 1.25E-07 LCI-CCF-PF-4MDPM RHR PUMF MMMON CAUSE FAILURE S2 SMALL LOCA INITIATOR

/C REACTOR-PROTECTION SYSTEM

$ 14) DCl-BDC-ST-BCB2B PANEL FAULTS BY ANY LOAD BCB-2B 8.84E-08 S2 SMALL IDCA INITIATOR LCI-RLY-FU-K84A INTERLOCK RLY 10A-K84A PATH FAULTS

/C REACTOR' PROTECTION SYSTEM

15) TBC-XHE-RE-P2B FAILURE TO RESTOPE 37P-2B AFTER MAINT. 6.98E-08 RSW-CCF-VF-2MOVS CCF OF-RHRHX SW DISCH VLVS FAIL TO OPEN S2 SMALL LOCA INITIATOR

/C REACTOR' PROTECTION SYSTEM-4 1

i

. _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- _ _ _ _ _ _ _ _- . - -_- --a. . -_- -

__ - - . . .- __ . . ~ _ _ _ - _ _ - _

SHEET 18 EVENT NAME _

DESCRIPTION PROB. (Note 1)

1) S2-37

1) AC4-RLY-NO-HOEA1 ~ REL 86A-1HOEB01 PATH FAILURE - 1.22E-06

. LCI-XHE-RE-FM3DP FAIL TO RESTO PM-3D PATH CMPTS AFT MAIN fC bbR Ok'bf bYSTEM

2) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 5.86E-07 LCI-HTX-VF-HE-2B- LOOP B HEAT EXCHANGER E-2B FAILURE S2 SMALL LOCA INITIATOR

/C REACTOR PROTECTION SYSTEM

3) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 2.55E-07 '

S2 SMALL IDCA INITIATOR . . i '

RSW-RCK-NO-MV89B VALVE CONTROL CIRCUIT NO OUTPUT

/C REACTOR PROTECTION-SYSTEM

4) ACO-MAI-MA-XFRT3 TRANSFORMER T3 IN MAINTENANCE 1.92E-07 RSW-CCF-VF-2MOVS CCF OF RHRHX SW DISCH VLVS FAIL TO OPEN S2 SMALL IDCA INITIATOR -

/C REACTOR PROTECTION SYSTEM-

5) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 1.71E-07 S2 SMALL IDCA INITIATOR LCI-RCK-NO-RP-3D SWGR CONTROL CIRCUIT NO OUTPUT

, /C REACTOR PROTECTION SYSTEM

6) ACO-MAI-MA-XFRT3 TRANSFORMER-T3 IN MAINTENANCE 1.43E-07 LCI-CCF-PF-4MDPM RHR PUMP COMMON CAUSE FAILURE S2 SMALL IDCA INITIATOR

/C REACTOR PROTECTION SYSTEM

7) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 1.09E-07 S2 SMALL LOCA INITIATOR AC4-SBR-DN-10640 CKT BRKR-10640 DOES NOT OPERATE PROPERL

/C REACTOR PROTECTION SYSTEM

8) AC4-RLY-NO-HOEB3 ID REL'86-1HOEB03 PATH FAILURE 1.09E-07 AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE S2 SMALL -IDCA INITIATOR

/C REACTOR PROTECTION SYSTEM o

a 4

. _ _ _ _ _ _ _ . _ - . _ _ _ . _ _ . - _ . _ _ _ . _ - _ - - - - - _ - . _ _ _ _ _ _ _ _ _ - . _ - 'e - -____r

- _ _^^- _ - - _ - _ '___--- - A___.___ _ ._ ___ _:____.____a __-=. _ = _ _ _ _ _ - _ . = - - - - = , - -_on-.-- . . . . _____a x-

g, SHEET 19 EVENT NAME DESCRIPTION PROB. (Note 1)'

AC4-RLY-NO-HOEB1 REL 86A-1HOEB01 PATH FAILURE 1.09E-07

9) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE S2 SMALL IDCA INITIATOR

/C REACTOR PROTECTION SYSTEM-

10) AC4-RLY-NO-HOEA1 REL 86A-1HOEB91 PATE FAILURE 8.79E-08

-S2 -

SMALL IDCA ::NITIATOR LCI-MDP-FR-RP-3D - PUMP P-3D S"OP RUNNING GIVEN PUMP START

/C REACTOR PRO"ECTION SYSTEM

11) AC4-RLY-NO-HOEA1 REL'86A-1HOEB01 PATH FAILURE 8.22E-08 S2 SMALL IDCA INITIATOR LCI-CKV-CC-CV42D CHECK VALVE VCW-30AN-42D DOES NOT OPEN

/C REACTOR PROTECTION SYSTEM-

12) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 5.41E-08 72 SMALL IDCA INITIATOR W-MOV-CC-MV89B V gV MO g g g g N g GEN

13) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 4.99E-08 S2 SMALL -IDCA INITIATOR LCI-RCS-OC-22B12 RLY 10-A-K22B CNTCT 1-2 DOES NOT RMN OP

/C REACTOR PROTECTION SYSTEM

• 14) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH-FAILURE 3.99E-08 S2 SMALL IDCA INITIATOR RSW-MSW-FU-SW48B SW 10A-S48B PATH FAULTS

/C REACTOR PROTECTION SYSTEM

15) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 3.99E-08 S2 -

SMALL LOCA INITIATOR I-MSW-FU-10S3D SW 10-S3D PATH FAULTS C REACTOR PROTECTION S'fSTEM

.. . . . . . .- ._ - - ..r -u - .. ., in.- .- - . t ui i-.- -i ...i.n -. . . . . . . . . '. ..mu....m:.. . ~- - .

g - . ..m

-SHEET 20 E -

b Sb8gRIqlgg PROB. (Np e 1)

-1) S2-42

1) AC4-RLY-NO-HORA 1 ' REL 86A-1HOEB01 PATH FAILURE 2.65E-08

-DCl-BDC-ST-BCB2B -- PANEL FAULTS BY ANY LOAD BCB-2B S2 SMALL LOCA INITIATOR

_/C REACTOR PROTECTION SYSTEM

2) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 2.32E-08 LCI-XHE-RE-PM3DP . FAIL TO-RESTO PM-3D PATH CMPTS AFT MAIN

.S2 SMALL LOCA INITIATOR HCI-MAI-MA-HPCIU HPCI UNAVAILABLE DUE TO MAINTENANCE

/C REACTOR PROTECTION SYSTEM

3) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 1.11E-08 LCI-HTX-VF-HE-2B IDOP B HEAT - EXCHANGER E-2B FAILURE S2 SMALL IDCA INITIATOR-HCI-MAI-MA-HPCIU HPCI UNAVAILABLE DUE TO MAINTENANCE

/C REACTOR: PROTECTION SYSTEM

4) AC4-RLY-NO-HOEA1 .REL 86A-1HOEB01 PATH FAILURE 8.34E-09 LCI-XHE-RE-PM3DP FAIL To RESTO PM-3D PATH CMPTS AFT MAIN S2 SMALL IDCA INITIATOR -

HCI-TDP-FS-HCIPM TUR DRIV PUMP FAILS TO' START UPON DEMAN

/C REACTOR-PROTECTION SYSTEM

en 5) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 7.33E-09 e .LCI-XHE-RE-FM3DP FAIL TO RESTO PM-3D PATH CMPTS AFT MAIN S2 SMALL IDCA INITIATOR.

HCI-TDP-FR-HCIPM TUR DRIVE PUMP FAILS TO CONTINUE RUNNIN

/C REACTOR. PROTECTION. SYSTEM

6) AC4-RLY-NO-HOEAl- REL 86A--1HOEB01 PATH FAILURE 4.85E-09 S2 .

SMALL LOCA INITIATOR RSW-RCK-NO-MV89B VALVE CONTROL CIRCUIT NO' OUTPUT HCI-MAI-MA-HPCIU HPCI UNAVAILABLE DUE TO MAINTENANCE

./C REACTOR. PROTECTION SYSTEM-

7) AC4-RLY-NO-HOEA1 REL'86A-1HOEB01 PATH FAILURE 4.85E-09 '

LCI-XHE-RE-PM3DP ' FAIL TO~RESTO PM-3D PATH CMPTS AFT MAIN-

S2 .

' SMALL IDCA' INITIATOR HCI-PHN-FU-SPEXH FAILURE OF STEAM ~ SUPPLY_/JXHAUST PATHS

/C REACTOR' PROTECTION SYSTEM i

i i

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .-- _ _ _ _ ,_ ~ . . . - . - . _ - .- ____ - - - _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ _ _ _ . _ _ _ . _ _ _ _ __

SHEET 21'-

EVENT NAME DESCRIPTION . PROB. (Note 1)'

8) AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH FAILURE 4.00E-09' LCI-HTX-VF-HE-2B LOOP B HEAT EXCHANGER E-2B FAILURE S2 SMALL IDCA INITIATOR HCI-TDP-FS-HCIPM TUR DRIV PUMP FAILS TO START UPON DEMAN

/C REACTOR PROTECTION SYSTEM r

9) ACO-MAI-MA-XFRT3 TRANSFORMER T3 IN MAINTENANCE 3.64E-09 RSW-CCF-VF-2MOVS CCF OF RHRHX SW DISCH VLVS FAIL'TO OPEN I S2 SMALL LOCA INITIATOR HCI-MAI-MA-HPCIU HPCI UNAVAILABLE DUE TO MAINTENANCE

/C REACTOR PROTECTION SYSTEM AC4-RLY-NO-HOEA1 REL 86A-1HOEB01 PATH' FAILURE' -3.51E 10) 'LCI-HTX-VF-HE-2B LOOP B HEAT EXCHANGER E-2B FAILURE S2 SMALL LOCA INITIATOR CI-TDP-FR-HCIPM TUR DRIVE PUMP FAILS TO CONTINUE RUNNIN C REACTOR PROTECTION SYSTEM 1

4 4

M i

1

-SHEET 22' EVENT NAME DESCRIPTION PROB'. (Note 1)

1) TDCA-4

1) - LCI-XHE-RE-PM3DP -- FAIL TO RESTO PM-3D PATH CMPTS AFT MAIN 3.34E-03 TDCA IDSS OF BATTERY CONTROL BD 2A INITIATOR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/C REA SRVgTOR FROTECTION SYSTEM S RESEAT.

/P

2) LCI-HTX-VF-HE-2B TDCA -

LOOP.B HEAT EXCHANGER E-2B FAILURE IDSS OF BATTERY CONTROL BD 2A INITIATOR

'1.60E-03 UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS -i

/C -REACTOR PROTECTION SYSTEM

'P

/ SRV'S RESEAT

3) RSW-RCK-NO-MV89B VALVE CONTROL CIRCUIT NO OUTPUT 7.00E-04

( TDCA LOSS OF BATTERY CONTROL BD 2A INITIATOR C -

R OE Y

/P SRV S RESEAT. .

LCI-CKV-CO-CV42B CHECK VALVE'VCW-30AN-42B FAILS TO CIOSE 5.76E-04 -

4) TDCA LOSS OF BATTERY CONTROL BD 2A INITIATOR '

UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS -

/C REACTOR PROTECTION SYSTEM -

/P SRV'S RESEAT l 5) - TDCA LCI-RCK-NO-RP-3 D - SWGR CONTROL CIRCUIT NO OUTPUT 4.68E-04 o

I LOSS OF BATTERY CONTROL BD 2A INITIATOR '

UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/C REA OR PROTECTION SYSTEM SRV S RESEAT

/P

6) AC4-SBR-DN-10640 CKT BRKR 10640' DOES NOT OPERATE PROPERL 2.98E-04 TDCA IDSS OF BATTERY CONTROL BD 2A INITIATOR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/C REACTOR PROTECTION SYSTEM i

'P

/ SRV'S RESEAT u-___________________________-______________-__ . _ - . _ _ _ _ - _ . _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - _ . . __

.._._-.______a .

l l-l " SHEET'23 1

EVENT NAME ,

DESCRIPTION PROB. (Note 1)

7) LCI-CKV-CC-CV42D CHECK VALVE VCW-30AN-42D DOES NOT OPEN . 2.25E-04 TDCA LOSS OF BATTERY CONTROL.BD 2A INITIATOR

.UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

-REAgTOR PROTECTION SYSTEM.

/' C l /P SRV S. RESEAT

! VALVE MOV89B DOES NOT OPEN . 1.48E-04 l 8) TDCARSW-MOV-CC-MV89B LOSS OF BATTERY CONTROL BD 2A INITIATOR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO .CS l /fi f HR09SREER2'ECmNSYSm T

9) LCI-RCS-OC-22B12 TDCA RLY 10-A-K22B'CNTCT 1-2 DOES NOT RMN OP LOSS OF BATTERY CONTROL BD 2A INITIATOR 1.37E-04 UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO 'CS'

/C REA SRVgTOR PROTECTION SYSTEM

/P S RESEAT I. .

10) RSW-MSW-FU-SW48B SW 10A-S48B PATH FAULTS 1.09E-04 TDCA IDSS OF BATTERY CONTROL BD 2A INITIATOR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS

/C REA SRVgTOR PROTECTION SYSTEM-S RESEAT

/P TDCE LS BAT R O D A-UlX-SUCC SUCCESSFUL ALIGNMENT OF HECI SUCT TO CS '!

/C REA SRVpTOR PROTECTION SYSTEM-

/P S RESEAT..

HX INLET VLV MOV65B FAIL CLS 4.07E-05

12) TDCALCI-MOV-PG-HV65B IDSS OF BATTERY CONTROL BD 2A(PLUGGED) INITIATOR i UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS C REAgTOR PROTECTION SYSTEM P SRV S RESEAT i

i

' SHEET 24 EVENT NAME DESCRIPTION PROB. (Note 1)

13) LCI-XVM-PG-XV45D VALVE VGW-30AN-45D FAILS CLSED (PLUGGED 3.25E-05 IDSS OF BATTERY CONTROL BD 2A INIT::ATOR

~

.TDCA UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT "O CS

/C REAgTOR PROTECTION SYSTEM

/P- SRV S RESEAT .!

14) RSW-XVM-PG-XV24B MANUAL VALVE 24B. FAILS-CIDSED iPLUGGED1 3.25E-05 TDCA IDSS OF BATTERY CONTROL BD -2A INITIATOR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS'

/C REAgTOR PROTECTION SYSTEM SRV S RESEAT

/P 15)-RSW-XVM-PG-XV11B LP ~ B MNUL VLV 11B FAILS CIDSED (PLUGGED. ~3.25E-05 -

TDCA IDSS OF BATTERY CONTROL BD 2A INITIATOR UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT TO CS t

/C REAgTOR PROTECTION SYSTEM

/P SRV S RESEAT.

DCl-FUS-NO-NRP3D DC : FUSE BIDWN (NEGATIVE 1.95E-05

16) TDCA IDSS OF BATTERY CONTROL) BD 2A INITIATOR

'UlX-SUCC SUCCESSFUL ALIGNMENT OF HPCI SUCT 'IV CS

/C REAgTOR PROTECTION SYSTEM

/P SRV S RESEAT

'I

'1

_ _--__ __ _._ ___. _ ___.___ _ ___ __ _____ __ _-_., _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . __ 1

. SHEET 25 EVENT NAME DESCRIPTION

---

_- -_ _ - _ - _ _ PROB.'.(Note 1)

1) TDCB-4 '!

1) LCI-XHE-RE-FM3AP FAIL TO RESI'O PM-3A~ PATH CMPTS AFT MAIN 3.34E-03 TDCB LOSS OF BAT'ERY CONTROL-BD B INITIATOR

/C REA

/P SRVgTORS RESEAT PRO"ECTION SYSTEM

~

TDCB F TTkR O b N ATOR '

/C REAgTOR PROTECTION SYSTEM '

/P SRV S RESEAT

3) RSW-RCK-NO-MV89A ' VALVE CONTROL CIRCUIT NO OUTPUT .

7.00E-04 TDCB LOSS OF BATTERY CONTROL BD B INITIATOR

/C REA

/P SRVgTORS RESEATPROTECTION SYSTEM

4) LCI-CKV-CO-CV42C TDCB CHECK VALVE'VCW-30AN-42C FAILS TO CLOSE' 5. 76E-04 :

IDSS OF BATTERY CONTROL BD-B INITIATOR

/C REA SRVgTOR PROTECTION SYSTEM

/P S RESEAT.

5) LCI-RCK-NO-RP-3A TDCB SWGR CONTROL CIRCUIT'NO OUTPUT 4.68E-04

.IDSS OF BATTERY CONTROL BD B INITIATOR

/C REA OR PROTECTION SYSTEM m '/P SRV S RESEAT o .

b ID S REA BA ERY NT k NIT R

/C SRVgTOR PRO'ECTION SYSTEM.

/P S RESEA'

~

7) LCI-CKV-CC-CV427.

TDCB CHECK VALVE VCW-30AN-42A DOES NOT OPEN 2.25E-04 .

IDSS OF BATTERY CONTROL BD B INITIATOR

! /C REA SRVgTORS RESEATPROTECTION SYSTEM ,

1 /' P ,

8) RSW-MOV-CC-MV89A VALVE MOV89A DOES NOT OPEN '1.48E-04 i TDCB IDSS OF BATTERY CONTROL BD B INITIATOR C REA P SRVgTORS RESEATPROTECTION SYSTEM i

.- , , , _ _ - , , _ . . , .a , , _ _ . . . . _ _ _ . . .-

' SHEET 26 EVENT NAME DESCRIPTION- PROB.-(Note 1)~

9) LCI-RCS-OC-19A12 TDCB RELAY 10A-K19A CONT 1-2-DOES NOT RM OPE LOSS OF BATTERY CONTROL BD B INITIATOR 1.37E-04

/C REAgTOR PROTECTION SYSTEM

'P

/ SRV S RESEAT

10) RSW-MSW-FU-SW48A SW 10A-S48A PATH FAULTS -

1.09E-04 '

TDCB LOSS OF BATTERY CONTROL BD B INITIATOR

/C REAgTOR' PROTECTION SYSTEM SRV S RESEAT

/P

11) SPC-MSW-FU-SW17A SW 10A-S17A PATH FAULTS -

1.09E-04 TDCB IDSS OF BATTERY CONTROL BD B INITIATOR

/C REAgTOR PROTECTION SYSTEM

'P

/ SRV S RESEAT _.

12) LCgMOV-PG-MV13A .

gSggMgggAg yggD 4.07E-05

/C 'REAgTOR' PROTECTION SYSTEM

'P

/ SRV S. RESEAT

13) LCI-MOV-PG-MV65A VALVE MOV65A-FAILS CIDSED PLUGGED) 4.07E-05 l TDCB IDSS OF BATTERY CONTROL BD(B-INITIATOR

/C REACTOR-PROTECTION SYSTEM- +

'P

/ SRV'S RESEAT. -

3 LCI-MOV-PG-MV12A VALVE MOV12A' FAILS CLOSED (PLUGGED) 4.07E-05

14) TDCB IDSS OF BATTERY CONTROL BD B INITIATOR

/C ~REAgTOR PROTECTION SYSTEM ~

/P SRV S RESEATc '

15) RSW-MDP-FR-MP-1A PUMP P-1A~ STOPS RUNNING-GIVEN PM START 3.26E-05 RSW-MDP-MA-MP-1C PUMP P-1C IN MAINTENANCE TDCB . IDSS OF BATTERY CONTROL BD B INITIATOR

/C REACTOR PROTECTION. SYSTEM SRV'S RESEAT

/P

16) LCI-XVM-PG-M151A SUP~PL SU VLV MOV151A FAIL CLS-(PLUGGED 3.25E-05 TDCB LOSS OF BATTERY CONTROL BD B INITIATOR

- C -REAgTOR PROTECTION SYSTEM P SRV S RESEAT-

- , _ . . . . . _ . . . . , . . . ~ . , _ _ _ , _ _ _ _ . - _

- - - . ~ . . - - - __.______________-_____________-.____:__.- -_-___:________-__-__-___

g;_.,_,_.. __

Not'e (1). The' probability of each minimal cutsets does not include the multiplication of the' individual initiating event' frequency.

4 h

CD I

A 9

.~._ - - - - . - -- .

i Sheet 1 I. . - TABLE 3 l

TV SEQtENCES BEFORE AND AFTER REQWERY LOSS OF OFFSITE poler (T1) ,

i SEQUENCE SEQUENCE - SEERJENCE DESIGNATOR SEQUENE SEQUENE APPLICABLE SEQUENE CGWENTS NLMBER TYPE FREQUENCY FREQtENCY RECDWERY ACTIONS ELININRTED BEFGtE AFTER RECWERY RECOWERY T1-4 T1W T1*/C*/82*/P*/U1*/U1X*W1*W2*/x2*W3 3.43E-5 3.96E-8 Recovery of offsite poner in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. NO Notes Repair pump, velve faiture in 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. (1),(2)&(3)

Reelign w s in 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

T1-5 T1W T1*/C*/82*/P*/U1*/Ulx*W1*W2*x2 4.04E-9 1.80E-11 Recovery of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (1) i TI-9 T1W T1*/C*/82*/P*/U1*U1x W W2*/x2*W5 2.54E-8 1.52E-9 Recovery of offsite power in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. NO %te (1)

T1-10 T1W T1*/C*/82*/P*/U1*U1x*W1*W2*x2 <1.0E-9 None YES T1-14 T1W T1*/C*/82*/P*U1*/U2*W1*W2*/x2N3 2.11E-6 8.6EE-9 Recovery of offsite power in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. No notes Repelr piep, volve feitures in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. (1),(2)&(3)

$ Reetign couponents in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

T1-15 T1W T1*/C*/82*/P*U1*/U2*W1*W2*x2 <1.0E-9 None YES T1-19 T1W T1*/C*/82*/P*U1*U2*/X1*/V2*W1*W2*/x2*W3 8.95E-8 3.99E-10 Recovery of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (1)

T1-20 T1W T1*/C*/82*/P*U1*U2*/X1*/V2*W1*W2*x2 <1.0E-9 None YES T1-24 T1W T1*/C*/82*/P*U1*U2*/X1*V2*/v3*W1*W2*/x2*W3. <1.0E-9 None YES -l T1-25 T1W T1*/C*/82*/P*U1*U2*/X1*V2*/V3*W1*W2*x2 <1.0E-9 None YES T1-29 T1W . T1*/C*/82*/P*Ut*U2*/X1*V2*V3*/V4*W1*W2*/x2N3 <1.0E-9 None YES

-T1-30 T1W T1*/C*/82*/P*U1*U2*/X1*V2*V3*/V4*W1*W2*x2 - <1.0E-9 None YES t_ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ . . . . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~__ -- - c - ,*, , _______._.m_2__ _ _ _ . --___

I-l l Sheet 2 TABLE 3 TW SEQLEIICES BEFORE Ale AFTER RECIWERY LOSS OF OFFSITE poler (T1)

~

l SEQUENCE SEQUENCE SEQUENCE DESIGIIATOR SEQUEIICE SEQUEIICE APPLICABLE SEQUENCE CDUENTS IRMBER TYPE FREQUEIICY FREQUEIICY RECOWERY ACTIONS ELIMIIIRTED BEFORE AFTER RECDWERY RECOWERY

-l T1-33-S2-37 T1-P1-W T1*/C*/B2*P1*/U1*W1*W2*/X2*W3 9.58E-7 4.27E-9 Recovery of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. No ' Note (1)

T1-33-52-38 T1-P1-W T1*/C*/82*P1*/U1*W1*W2*x2 <1.0E-9 Ilone YES T1-33-S2-42 T1-P1-W T1*/C*/B2*Pl*U1*/U2*W1*W2*/X2*W3 7.07E-8 3.15E-10 Recovery of offsite pouer in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

YES Note (1) 11-33-S2-43 T1-P1-W T1*/C*/82*P1*U1*/U2*W1*W2*x2 <1.0E-9 ' mone YES T1-33-52-47 T1-P1-W T1*/C*/82*P1*U1*U2*/x1*/V2*W1*W2*/X2*W5 . <1.0E-9 None YES T1-33-S2-48 T1-P1-W T1*/C*/82*P1*Ul*U2*/X1*/V2*W1*W2*x2 <1.0E-9 None YES y T1-33-S2-52 T1-P1-W T1*/C*/82*P1*U1*U2*/X1*V2*/V3*W1*W2*/x2*W3 <1.0E-9 ' Ilone YES T1-33-S2-53 T1-P1-W T1*/C*/82*P1*01*U2*/X1*V2*/V3*W1*W2*x2 <1.0E-9 Ilone YES T1-33-S2-57 T1-P1-W T1*/C*/s2*P1*U1*U2*/X1*V2*V3*/V4*W1*W2*/x2*W3 d'.0E-9 Ilone YES f1-33-S2-58 T1-P1-W T1*/C*/82*Pl*U1*U2*/X1*V2*V3*/V4*W1*W2*x2 <1.0E-9 Ilone . YES T1-34-S1-23 T1-P2-W T1*/C*/82*P2*/U1*/V2*W1*W2 <1.0E-9 None YES T1-34-S1-26 T1-P2-W T1*/C*/82*P2*/U1*V2*/V3*W1*W2 <1.0E-9 None YES T1-34-SI-29 T1-P2-W T1*/C*/B2*P2*/U1*V2*V3*/V4*W1*W2 . <1.0E-9 ~ None YES T1-34-S1-33 T1-P2-W T1*/C*/82*P2*U1*/xt*/V2*a1*W2 <1.0E-9 .None YES i T1-34-St-36 T1-P2-W T1*/C*/82*P2*U1*/x1*V2*/V3*W1*W2 <1.0E-9 Ilone YES T1-34-St-39 T1-P2-W T1*/C*/82*P2*U1*/X1*V2*V3*/v4*W1*W2 <1.0E-9 None YES

~

_ . _ _ . _ _ . _ _ _ _ _ _ _ _ _ - _ _ _ _ - - _ c -_m '2_ _ ,

• _ _ _

. . Sheet 3

-TABLE 3 TW SEQtEdICES BEFtRE AIO AFTER REtXNERY LOSS OF OFFSITE POER (TI)

SEQUEIICE SEGIEIICE SEGJEIIE DESIGIIATOR SEGJEIICE MGJEIICE APPLICAOLE WREIIE CUWENTS Iltm0ER ' TYPE FItEGEIICY FetfetEIICY RECOWERY ACTIONS ELINIIIRTED BEFtRE AFTER RECOWERY RECDWERY g T1-35-A-13 T1-P3-W T1*/c*/s2*P3*/V2 NiaW2 <1.0E-9 Ilone YES T1-35-A-16 T1-P3-W T1*/c*/82*P3*V2*/V3*W1*W2 <1.0E-9 lione YES T1-35-A-19 T1-P3-W T1*/C*/82*P3*V2*V3*/V4*W1*W2 <1.0E-9 Ikne YES d

6*

, , - . , ,. ,--.#_.% . - -c.. ~. p. -,=,9 4%, . . - . - - y - * -

f 5 .--r .- _,- wa,

Sheet 4 l TMM 3 l _

TW SEQlKMCES BEFGtE Als AFTER RECDUERY LOSS OF POWER CONVERSION TRAIISIENT (T2) l

[

SEQUENCE SEQUENCE SEQUElICE DESIGIIATOR SEQUEIIE SEIKIEIICE -APPLICAELE EMENE CGUENTS Ill818ER TYPE FREQEIICY FREQ E NCY RE(XWERY ACTIONS ELIIIIIIRTED SEFORE AFTER REC 0WERY REIDWERY T2-4. 72W T2*/C*/81*/P*/U1*/U1X*W1*W2*/X2N5 7.58E-5 1.10E-08 Itecovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. .

NO Notes Repair pimp, volve feiture in 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />. (2),(3)&(4) .

Reetign eagwmnts in 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.

T2-5 T2W T2*/C*/81*/P*/U1*/U1X*W1*W2*X2 . 3.39E-S 2.37E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES leote (4)

T2-4e T2W T2*/C*/81*/P*/U1*U1X*W1*W2*/X2*W5 1.42E-7 8.52E-9 Recovery of PCS in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. 11 0 Ilotes Repair pimp, vetve fatture in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. (2),(3)&(4)

Reetign component in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

72-Se T2W T2*/C*/81*/P*/U1*U1X*W1*W2*X2 <1.0E-9 None YES.

T2-9 T2W T2*/C*/81*/P*U1*/U2*W1*W2*/X2N3 5.38E-6 3.77E-9 Recovery of PCS in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. No Isote (4) y T2-10 T2W T2*/C*/81*/P*U1*/U2*W1*W2*X2 <1.0E-9 Ilone YES-T2-14 T2W T2*/C*/81*/P*U1*U2*/X1*/V1*W1*W2*/X2*W5 3.22E-7 2.25E-10 Recovery of'PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES. ~ Ilote (4)

T2-15 T2W T2*/C*/91*/P*U1*U2*/X1*/V1*W1*W2*X2 <1.0E-9 None YES T2-19 T2W T 2*/C*/81 */P*U1 *U2*/X 1*V1 */V2*W1*W2*/X2*W3 3.36E-8 2.35E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

T2-20 T2W T2*/C*/81*/P*Ul*U2*/X1*V1*/v2*W1*W2*X2 <1.0E-9' IIone YES T2-24 - T2W T2*/C*/81*/P*U1*U2*/X1*V1*V2*/v3*W1*W2*/X2*W5 <1.0E-9 None YES T2-25 T2W T2*/C*/81*/P*U1*U2*/X1*V1*V2*/V3*W1*W2*X2 < 1.0E-9 None YES T2-29 72W T2*/C*/81*/P*U1*U2*/X1*V1*V2*V3*/v4*W1*W2*/X2 <1.0E-9 Ilone YES'

• W5 T2-30 T2W T2*/C*/81*/P*Ut*U2*/X1*V1*V2*V3*/V4*W1*W2*X2 <1.0E-9 Ilone ' TES

_ _ _ _ _ _ __. _ .__ _. _ _ . _ _ _u. _. _

Sheet 5 TABLE 3 TW SEctKMCES BEFCRE AfD AFTER REC 0WERY LOSS OF PohER CONWERSION TRANSIENT (T2)

SEQUENCE SEQUENCE SEQUENCE DESIGNATOR SEQUENCE SEQUENE APPLICA9LE SEGUENE CO W NTS NLMBER TYPE FREQUENCY FREetENCY REODWERY ACTIONS ELisitNRTED BEFCRE AFTER REC 0WERY RECDWERY T2-33-S2-5 T2P1W T2*/C*/s191*/U141*W2*/X2N3 5.06E-6 3.54E-9 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. NO Note (4).

T2-33-S2-6 T2P1W- T2*/C*/51*P1*/U1*W1*W2*x2 3.65E-8 2.55E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES . Note (4)

T2-33-S2-10 'T2P1W T2*/C*/81*Pl*U1*/U2*W1*W2*/X2*W3 3.31E-6 2.32E-9 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. No Note (4)-

T2-33-S2-11 T2P1W T2*/C*/B1*P1*U1*/U2*W1*W2*x2 3.65E-8 2.55E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4) l ,

T2-33-S2-15 T2P1W T2*/C*/81*P1*U1*U2*/xt*/v1*Wi*W2*/x2 N3 2.70E-7 1.89E-10 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4) i 72-33-S2-16 T2P1W T2*/C*/81*P1*U1*U2*/x1*/v1*W1*W2*M2 3.65E-8 2.55E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4) w w

T2-33-S2-20 T2P1W T2*/C*/81*P1*U1*U2*/X1*V1*/v2*W1*W2*/X2*W3 2.70E-7 1 89E-10 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

T2-33-S2-21 72P1W T2*/C*/81 41*U1*U2*/X1*V1*/v2 N1*W2*x2 3.65E-8 2.55E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. VES Note (4)

T2-33-S2-25 T2P1W T2*/C*/B1*P1*Ut*U2*/X1*V1*V2*/v3*W1N2*/X2*W3 3.65E-8 2.55E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

T2-33-S2-26 .T291W T2*/C*/81*P1*Ut*U2*/X1*V1*V2*/v3*W1*W2*x2 3.65E-8 2.55E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)  ;

T2-33-S2-30 T2P1W -T2*/C*/81*P1*U1*U2*/x1*vt*v2*v3*/v4*W1*W2*/X2 7.29E-8 5.10E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

• W3 T2-33-S2-31 T2P1W T2*/C*/s191*U1*U2*/X1*v1*/2*V3*/V4*W1*W2*x2 7.29E-8 5.10E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)  !

4 e

t P

I i

Sheet 6.

TABLE 3 TW SEctENCES BEFCRE ARO AFTER RECUWERY LOSS OF POWER CONVERSION TRANSIENT (T2)

SEQUENCE' SEQUENCE SEQUENCE DESIGNATOR SEQUENE SEQUENE APPLICABLE SERIENCE CtB8eITS -

utmBER TYPE FREQUENCY FREQtKNCY RECDWERY ACTIONS EListemTED BEFWtE AFTGP RECUWERY RECDWL fY T2-34-St-3 T2P2W T2*/C*/81*P2*/U1*/V2 N1*W2 1.22E-7 6 f3E-8 Repair pump, volve feiture in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. NO Isote (2)

T2-34-S1-6 T2P2W T2*/C*/81*P2*/U1*V2*/V3*W1*W2 1.17E-9 Ilone NO-T2-34-S1-9 T2P2W T2*/C*/81*P2*/U1*V2*V3*/V4*W1*W2 <1.0E-9' None TES T2-34-S1-13 T2P2W T2*/C*/81*P2*U1*/X1*/V2*W1*W2 1.22E None' No '

T2-34-St-16 T2P2W T2*/C*/81*P2*U1*/X1*V2*/V3*W1*W2 <1.0E-9 -- None' YES

-a T2-34-St-19 T2P2W .T2*/C*/81*P2*U1*/X1*V2*V3*/V4*W1*W2 <1.0E-9 None YES A

1 T2-35-A-3 T2P3W' T2*/C*/81*P3*/V2*W1*W2 1.68E-8 None No 72-35-A-6 T2P3W T2*/C*/81*P3*V2*/V3*W1*W2 <1.0E-9 Ncne YES T2-35-A-9 T2P3W T2*/C*/B1*P3*V2-V3*/V4*W1*W2 L<1.0E-9 None YES

~

Sheet 7 TABLE 3 TW SEQtKNCES BEFORE AND AFTER RECDWERY LOSS OF PGER CONVERSION SYSTEN (T2)

SEQUENCE SEQUENCE SEQUENCE DESIGNATOR SEtRIENE SEtRIENE APPLICASLE SEtklENCE CtpWENTS NLMBER TYPE FREGENCY FREGENCY RECOWERY ACTIONS- ELIMINATED BEFORE AFTER RECDWERY RECDWERY T2- M-T1-4 .T2W . T2*/C'81*/E?"/P*/U1*/U1M*W1*W2*/x2N3 2.03E-S 9.05E-11 Recovery of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES ' Note (1)

T2-36-T1-5 T2W T2*/C*s1*/s2*/P*/U1*/U1X*W1*W2*N2. <1.0E-9 None YES T2-36-T1-9 T2W T 2*/C*81*/82*/P*/U1 *U1 X*W1*W2*/x2*W3 <1.0E-9 None YES T2-36-71-10 T2W T2*/C*B1*/82*/P*/U1*UTX*W1*W2*x2 <1.0E-9 None YES -

72-36-71-14 T2W T2*/C*s1*/s2*/P*U1*/U2*W1*W2*/X2N5 1.24E-9 5.53E-12 Recovery of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (1)

T2-36-T1-15 T2W T2*/C'81*/82*/P*U1*/U2*W1*W2*X2 <1.0E-9 Mone. YES j T2-36-T1-19 T2W T2*/C*B1*/82*/P*U1*U2*/X1*/V2*W1*W2*/X2*W3 <1.0E-9 None .. YES T2-36-T1-20 T2W T2*/C*S1*/B2*/PN1*U2*xt*/V2*W1*W2*x2 <1.0E-9 None YES T2-36-T1-24 T2W T2*/C*B1*/82*/P*U1*U2*/X1*V2*/V3*W1*W2*/X2N3 <1.0E-9 None YES T2-36-T1-25 T2W T2*/C*B1*/B2*/P*U1*U2*/X1*V2*/V3*W1*W2*X2 <1.0E-9 None YES T2-36-T1-29 T2W T2*/C*81*/82*/P*U1*U2*/xt*V2*V3*/v4*W1N2*/x2 <1.0E-9 'None YES

• W3 72-36-T1-30 T2W T2*/C*s1*/82*/P*Ut*U2*/X1*V2*V3*/v4*W1*W2*x2 <1.0E-9 None YES

. _ . - _ _ _ _ _ _ _______.__ =_______ - ___ ___ - _ _ ___ __ __ - __ _ _ ._ _ _ _ _ _ - . -

Sheet 8

-TABLE.3 TW SEQLENCES BEFWtf AIO AFTER RECDWERY ,

LOSS OF PWER CONVERSION TRAllSIENT (T2)

SEQUENCE SEQUENCE SEQUENG DESIGIIATOR SEQUENCE SEEJEIICE APPLICA0LE SEQUENCE CIBU'ENTS Nt91BER TYPE FREGIKIICY FREQtEIICY RE DWERY ACTIONS ELIMIIIATED BEFORE- AFTER ~

RECOWERY RECDWERY l

T2-36-T1 S2-37 T251P1W T2*/C*81*/82*P1*/U1 N1*W2*/X2 N3 1.76E-7 7.85E-10 Recovery of offsite pouer in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Ilote (1)

S2-38 T2B1P1W T2*/C*81*/82*P1*/U1*W1*W2*X2 <1.0E-9 Ilone YES S2-42 T2B1P1W T2*/C*81*/82*P1*U1*/U2*W1*W2*/X2*W3 7.59E-9 ' 3.39E-11 Recovery of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Ilote (1)

S2-43 T291P1W T2*/C*81*/82*P1*U1*/U2*W1*W2*X2 <1.0E-9 None YES S2-47 T291P1W T2*/C*81*/s2*Pl*U1*U?*/X1*/V2 N1*W2*/X2*W3 <1.0E-9 leone YES 4

cs S2-48 T2B1P1W T2*/C*B1*/82*P1*U1*U2*/71*/V2*W1*W2*X2 <1.0E-9 lione TES S2-52 T2B1P1W T2*/C*B1*/82*P1*U1*U2*/X1*V2*/V3*Wi*W2*/X2*W3 <1.0E-9 None YES S2-53 T2B1P1W T2*/C*B1*/82*P1*U1*U2*/X1*V2*/V3*W1 h <1.0E-9 Ilone YES S2-57 T281P1W T2*/C*81*/82*P1*L11*U2*/X1*V2*V3*/V4*W1*W2*/X2 <1.0E-9 None YES

• W3 S2-58 T251P1W T2*/C*B1*/s2*P1*U1*U2*/X1*V2*V3*/V4N1*W2 <1.0E-9.. Ilone YES

• X2 9

^

e -l

~. _ , - - - . - . . , _ . - - -- . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .__ m__.___

Sheet 9

' TABLE 3 TW SEetKNCES BEFCRE AIS AFTER RECUWERY -

LOSS OF POWER CONWERSION TRAIISIENT (T2) '

SEQUEN SEIWENCE SEQUENCE DESIGIIATOR SEMIENCE SEIWEIIE APPLICASLE' SERGIE CtBRENTS IILMBER TYPE FREQUElICY FREetKIICY RECDWERY ACTIONS ELIMINRTED BEFORE AFTER RECOVERY RECOVERY T2-36-T1 S1-23 T2B1P2W T2*/C*B1*/B2*P2*/U1*/V2*W1*W2 <1.0E-9 Ilone . YES 51-26 T2B1P2W T2*/C*B1*/82*P2*/U1*V2*/V3*W1*W2 <1.0E-9 None YES St-29 T2B1P2W T2*/C*B1*/92*P2*/U1*V2*V3*/V6*W1*W2 <1.0E-9 Ilone YES S1-33 T2B1P2W T2*/C*B1*/82*P2*U1*/X1*/V2*W1*W2 - <1.0E Ilone YES.

51-36 T2B1P2W T2*/C*B1*/92*P2*U1*/X1*V2*/V3*W1*W2 <1.0E Ilone YES w

-a

$1-39 T2B1P2W T2*/C*B1*/82*P2*U1*/X1*V2*V3*/V4*W1*W2' ' <1.0E-9 None .. YES T2-36-T1 A-13 - T?B1P3W T2*/C*B1*/82*P3*/V2*W1*W2 <1.0E-9 None YES -

A-16 T2B1P3W T2*/C*B1*/82*P3*V2*/V3*W1*W2 <1.0E-9 Ilone -YES A-19 T2B1P3W T2*/C*B1*/B2*P3*V2*V3*/V4*W1*W2 <1.0E-9 None .YES f

e+M-*..# ~=-=w.%.,U~.m.--un.~e.- -m.1v-ww .e_ %% _m , - tw,.m-e w=m _ m2.--_ . . - - - ~ -.~.--w___

_ _ _ _ _ _ _ _ _ =_____.___._____._.__._...u..z _ w._ _ .

Sheet 10 TABLE 3 TW SEQtENCES NEFtRE AS AFTER RECOWERY -

TRANSIENTS WITN POWE7 CONWERSION SYSTEM AVAILABLE (T3A)

SEWENCE SEQUENCE SEeJENCE DESIGNATOR SEeJENE SEeJENCE APPLICA0LE WGJENE CGNENTS NLMBER . TYPE FREQtENCY FREGLENCY RECUWERY ACTIONS ELIMINATED BEFORE AFTER REC 0WERY RECDWERY T3A-2-T2-4 T3AW T3A*/C*/81*O*/P*/U1*/uiX*W1*W2*/X2*W3 2.51E-6 1.76E-9 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. m Note (4)

T3A-2-T2-5 T3AW T3A*/C*/51*Q*/P*/U1*/U1X*W1*W2*X2 5.41E-9 3.79E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

T3A-2-T2-4a 73AW T3A*/C*/81*F/P*/U1*U1X41*W2*/X2*W3 2.26E-8 1.36E-11 Recovery of PCS in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. YES Note (4)

T3A-2-T2-Se T3AW T3A*/C*/B1*Q*/P*/U1*U1X*W1*W2*N2 <1.0E-9 None YES T3A-2-T2-9 T3AW T 3A*/C*/81*Q*/P*U1*/U2*W1 *W2*/X2N5 8.58E-7 6.01E-10 Recovery of PCS in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. YES Note (4)

T3A-2-T2-10 T3AW T3A*/C*/81*0*/P*U1*/U2*W1*W2*X2 <1.0E-9 None YES

$ T3A-2-72-14 T3m 73A*/C*/81*0*/P*U 1*U2*/X 1 */V1 *W1 *W2*/X2N5 '5.13E-8 3.59E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4) 73A-2-72-15 T3AW T3A*/C*/81*O*/P*U1*U2*/X1*/V1*W1*W2*X2~ <1.0E-9 None YES 'i T3A-2-72-19 T3AW T3A*/C*/81*0*/P*U1*U2*/X1*V1*/V2*W1*W2*/X2*W3 5.36E-9 3.75E-12 Recovery of oCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

L T3A-2-T2-20 T3Au T3A*/ C*/B l *Q*/P*U l *U2*/X 1 *V1 */V2*W1 *W2* X2 <1.0E-9 None YES T3A-2-T2-24 T3AW T3A*/C*/81*Q*/P*U1*U2*/X1*V1*V2*/V3*W1N2*/X2 <1.0E-9 None YES

• W3 T3A-2-72-25 'T3AW T3A*/C*/81*0*/P*U1*U2*/X1*V1*V2*/V3*W1*W2*X2 <1.0E-9 None YES T3A-2-T2-29 T3AW T3A*/C*/B1*O*/P*U1*U2*/X1*V1*V2*V3*/V4*W1*W2 <1.0E-9 None YES

• /X2N3 T3A-2-T2-30 T3AW T3A*/C*/81*Q*/P*U1*U2*/X1*V1*V2*V3*/V4*W1*W2 <1.0E-9 None YES'

• X2

Sheet 11  !

TABLE 3 TW SEQENCES BEF(RE AID AFTER RECOWERY ,

TRAIIS!ENTS WITN PMR CONVERSION SYSTEft AWAILASLE (T3A)

SEQUENCE SEQUENCS SEGJEIICE DESIGIIATOR SEERJENE SEGJEIICE APPLICA8LE l SEtRIEWE " CGUEIITS 28m'9 FED NLMSER TYPE FREGLEIICY FREQtEIICY RECOWERY ACTIONS SEF(RE - AFTER REUWERY RECOWERY T3A-2-T2 T3AQP1W T3A*/C*/8'*O*P1*/U1*W1*W2*/X2*W3 8.11E-8 5.67E-11 Recovery of PCS 3n 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Ilote (4)

S2-5 T3A-2-72 T3AQP1W T3A*/C*/81*t*P1*/U1*W1*W2*X2 5.84E-9 4.09E-12 Recovery of PCs in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

S2-6 T3A-2-T2 T3AGP1W T3A*/C*/91*Q*P1*U1*/U2*W1*W2*/X2*W5 5.30E-7 3.71E-10 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES mote (4)

S2-10 ,

T3A-2-T2 T3AQP1W T3A*/C*/91*O*P1*r:1*U2*W1*W2*X2 5.84E-9 4.09E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. . YES Note (4)

S2-11 T3A-2-T2 T3AQP1W T3A*/C*/81*Q*P18U1*U2*/X1*/v1*W1*W2*/X2*W3 4.34E-8 3.04E-11 Recovery of PCs in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES- Note (4)

S2-15 T3A-2-T2 T3AQP1W T3A*/C*/81*Q*Pl*U1*U2*/X1*/V1*W1*W2*X2 5.84E-9 4.09E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Ilote (4) q S2-16 e

T3A-2-T2 T3AQP1W T3A*/C*/81*O*P1*U1*U2*/X1*V1*/V2*W1*W2*/X298 4.34E-8 3.04E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES llote (4)

S2-20 T3A-2-T2 T3ACP1W T3A*/C*/81*0*Pl*U1*U2*/X1*V1*/V2*W1*W2*X2 5.84E-9 4.09E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

S2-21 T3A-2-T2 T3AQP1W T3A*/C*/81*Q*Pl*U1*U2*/X1*V1*V2*/v3*W1*W2*/X2 5.84E-9 4.09E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

S2-25 *W3 T3A-2-T2 T3AQP1W T3A*/C*/81*MP1*U1*U2*/X1*V1*v2*/v3*W1*W2*X2 5.84E-9 4.09E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES' Note (4)

S2-26 T3A-2-T2 T3AQP1W T3A*/C*/81*Q*P1*U1*U2*/X1*V1*V2*V3*/V4*W1*W2 1.17E-8' 8.19E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

S2-30 */X2*W5 T3A-2-72 T3AQP1W T3A*/C*/81*Q*Pl*U1*U2*/X1*V1*V2*V3*/V4*W1*W2 1.17E-8 8.19E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Ilote (4)

S2-31 *X2 T

4 5

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Sheet 12 -

TABLE 3 TW SEQENCES SEFWE Als AFTER RECOWERY -l TRANSIENTS WITH DCSdER CGrWERSION SYSTEN AVAILABLE (T3A)

,' SEQUENCE SEQUENCE SEQUENCE DESIGNATOR SEQUENE SEQUENCE APPLICABLE SEGUENE 055ENTS IRMBER TTPE~ FREGENCY FREQtENCY RECDUEltY ACTIONS - ELINtaRTED SEFORE AFTER RECOWERY RECDWERY T3A-2-T2 T3ACP2W T3A*/C*/B1*O*P2*/U1*/V2*W1*W2 1.7/E-8 1.24E-11 Recovery of PCS in 24 heirs. YES Note (4)

S1-3 T3A-2-T2 T3AQP2W T3A*/C*/B1*O*P2*/U1*V2*/V3*W1*W2 <1.0E-9 None YES St s T3A-2-T2 T3AaP2W T3A*/C*/Bl*0*P2*/U1*v2*V3*/VC*W1*W2 <1.0E-9 leone YES S1-9 T3A-2-T2 T3AQP2W T3A*/C*/81*Q*P2 N1*/X1*/V2*W1*W2 2.48E-9 1.74E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Not (4)

S1-13 1

73A-2-T2 T3AQP2W T3A*/C*/81*Q*P2*/U1*/xt*v2*/V3*W1*W2 <1.0E-9 leone YES S1-16 T3A-2-T2 T3AQP2W T3A*/C*/81*O*P2*/U1*/x1*v2*V3*/V4*W1*W2 <1.0E-9 None YES c) S1-19 0

T3A-2-T2 T3AQP3W T3A*/C*/81*O*P3*/v2*W1*W2. 1.48E-9 2.44E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

A-3 T3A-2-T2 T3AQP3W T3A*/C*/81*Q*P3*v2*/V3*W1*W2 <1.0E-9 Ilone YES A-6 T3A-2-T2 T3AQP3W T3A*/C*/81*O*P3*v2*V3*/V4*u1*u2 : <1.0E-9 'None YES A-9

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Sheet 13

- TABLE 3

- TW SEctENCES BEFORE AIO AFTER RECDWERY TRAh;!ENTS WITN POWER CONWERSION SYSTEM AURILABLE (73A)

SEQUENCE SECRENCE ' SEQUENCE DES 53114 TOR SEtREleCE SEQUENCE APPLICASLE' MeENE CtpWENT5.

IK548ER - ' TYPE FREEREtICY FREctflICY REO)WERY ACTIONS EL!stitIRTED ~

BEFtRE AFTER itEC0WERY REO)WERY T3A-3-71-4 T3AW T3A*/C*e1*/s2*/P*/U1*/U1X*W1*W2*/X2*W3 1.99E-7 '1.39E-10 Recovery of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. VES isote (1)

T3A-3-T1-5 73AW T W /C*81*/52*/P*/U1*/U1X*W1*W2*X2 <1.0E-9 Ilone YES ,

T3A-3-T1-9 T3AW T3A*/C*81*/82*/P*/U1*U1X*W1*W2*/X2*W3 <1.0E-9 IIone -TES 13A-3-T1-10 T3AW T3A*/C*e1*/s2*/P*/U1*U1X*W1*W2*X2 <1.0E-9 leone VES T3A-3-T1-14 T3AW T3A*/C*s1*/s2*/P*U1*/U2*W1*W2*/X2*W3 1.22E-8 8.54E-12 Recovery of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Isote (1)

T3A-3-71 T3AW T3A*/C*81*/82*/P*U1*/U2*W1*W2*X2 <1.0E-9 slone YES j cn ba T3A-3-T1-19 T3AW T3A*/C*s1*/s2*/P*U1*U2*/X1*/V2*W1*W2*/X2*W5 <1.0E-9 IIone YES T3A-3-T1-20 T3AW T3A*/C*81*/82*/P*U1*U2*/X1*/V2*W1*W2*X2 ~ <1.0E-9 None ~ YES

T3A-3-T1-24 T3AW T3A*/C"B1*/B2*/P*U1*U2*/X1*V2*/V3*W1*W2*/X2 -

<1.0E-9. Ilone YES i T3A-3-T1-25 'T3AW T3A*/C=5**/82*/P*U1*U2*/X1*V2*/V3*W1*W2*X2 . <1.0E-9 None YES T3A-3-T1-29. T3Au T3A*/C*81*/82*/P*U1*U2*/X1*V2*V3*/V4*W1*W2 <1.0E-9 teone YES

• /X2*W3 T3A-3-T1-30 T3AW T3A*/C*s1*/s2*/P*U1*U2*/X1*V2*V3*/V4*L1*W2*M2 <1.0E-9 Ilone YES o

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Sheet $3' TABLE 3 .

TW SEGLEENCES BEFWE ABC AFTER RECOWERY TRANSIENTS WITN paler ColNERSION SYSTEN AMILABLE (T3A)

SEQUENCE SEQUEN G SEQUENCE DESIGNATOR SEQUENCE SEQUENCE APPLICABLE SEGUENE CINOENTS utm8ER TYPE FREQUENCY FREGUENCY RECOWERY ACTIONS ELINtWRTED BEFGtE AFTER RECUWERY RECDWERY T3A-3-T1-33 T3A81P1W T3A*/C*81*/829 1*/U1*W1*V2*/x2*W5 5.21E-7 . 2.03E-11 Pecovery of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 'YES Note (1) 52-37 i 73A-3-71-33 T3A81P1W T3A*/C*81*/82*P1*/U1*W1*W2*x2 <1.0E-9 mone . YES

-S2-38 T3A-3-T1-33 T3A81P1W T3A*/C*81*/82*P1*U1*/U2*W1*W2*/X243 4.57E-8 2.03E-11 Reccery of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TES Note (1).

-S2-42 73A-3-T1-33 T3A81P1W T3A*/C*B1*/82 91*U1*/U2*W1*W2*x2 <1.0E-9 None YES

-S2-43 T3A-3-T1-33 T3A81P1W T34*/C*81*/82 91*U1*U2*/xt*/V2*W1*W2*/X2*W3 <1.0E-9 ' mone VES 47 T3A-3-T1-33 T3A81P1W T3A*/C*81*/8291*U1*U2*/X1*/V2*W1*W2*x2 <1.0E None VES co -S2-48 FJ T3A 3-T1-33 T3A81P1W T34*/C*81*/024 1*U1*U2*/x1*V2*/V3*W1*W2*/x2- <1.0E-9 None 'ES

-S2-52 *W5 T3A-3-T1-33 T3AB1P1W T3A*/C*81*/B2*P1*U1*U2*/x1*V2*/V3*W1*W2*x2 <1.0E-9 None YES

-S2-53 T3A-3-T1-33 T3As171W T3A*/C*B1*/B2*P1*U1*U2*/X1*V2*V3*/V4*W1*W2 <1.0E-9 None YES

-S2-57 */x2*W3 T3A-3-T1-33 T3AB1P1W T34*/C*81*/B2*P1*U1*U2*/xt*V2*V3*/V4*W1*W2*x2 <1.0E-9 None YES 58

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Sheet 19 in8LE 3 TW SEQUEuCES BEFORE AND AFTER ECUbERT

@ OF TEErwiEt IMIESIENTS (T35)

SEQUENCE SEQUEWCE SEGUENCE DES 19pT08t SEGUEhCE SEeUEWCE APPLICaeLE SEeUEWC.2 Ctpu erTS Nt# W R TYPE FWEetENCY FREelEWCT ECDUERY ACTIM E ELIntm TED SEFGtE AFTER IIEiDWERY IIECDUERT T30-10-71-4 T38W T38*/C*s1*/s2*c P/P*/U1*/U1R*W P W2*/x2*l6 6.75E-T 3.01E-9 secovery of offsite pouer in 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />. up note (1)

T3s-10-T1-5 T3eW T3e*/C*s1*/32*oP/P*/Ut*/U1x*W1*W2*X2 4.52E-10 more TES T38-10-T1-9 T30W T3e*/C*s1*/32*aP/P*/U1N1x*WPW2*/X2N3 <1.0E-9 Wore TES T3s-10-T1-10 T3ew T35*/C*s1*/s2'0 P/P*/U1*U1R*W1*W2*N2 <1.0E-9 #ene TES T3s-10-71-14 T3eW T38*/C*sP/s2*eP/P*UP/U2*W1 *W2*/X2N3 2.35E-7 1.65E-10 New/ of offsite power in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TES mote (1)

$ 738-10-T1-15 T38W T38*/Cas1*/s2'01*/P*U1*/U2*W1*W2*X2 <1.0E-9 mene TES T38-10-T1-19 T38W T38*/C*B1*/s2*01*/P*U1*U2*/X1*/V2*W1*W2*/X2 9.90E-9 6.95-12 Secowe y of offsite pouer in 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />. TES tote (1)

• W3 ,

T36-1041-20 738W T 38*/ C*B1 */82*01 */P*Ul *tJ2*/ x t */v2*WPu2*X2 <1.0E-9 mone TtS

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• /x2N3 T38-10-T1-25 T38W T38*/C*s1*/s2*c1*/P*UPU2*/X1*V2*/W3*W1*W2*X2 <1.0E-9 None TES T38-10-T1-29 T3eW T38*/C*sP/s2*01*/P*UPU2*/X1*V2*V3*/v641*W2 <1.0E-9 sone TES

• /x2*l6 T38-10-71-30 T38W T38*/C*S P/82*01*/P*U1*U2*/X Pv2*V3*/VP W1*W2 <1.0E-9 Wone NS

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Sheet 22 TABLE 3 TW SEQUENCES BEFORE ARD AFTER RECDWERY INADVERTENT (PEW RELIEF VALVE (T3C)

SEGUENCE SEWENCE SEtRJENE DESIGaATOR SEGUEMM SE4RJENE APPLICA0LE MSEHE CUWENTS utsgER TYPE FREGtKWCT FREGtKNCY RECOWER* ACTIONS EllaistTED SEFGtE AFTER RECDWERY RECOWERY 73C-2-$2-5 T3CW T!C*/C*/s1*tPN1*W1*W2*/M2*16 1.72E-7 1.20E-10 Recovery of PCS in 24 heers. 115 mote (4)

T3C-2-52-6 73 0s T3C*/C*/st*o*/Ut*W1N2*M2 1.24E-9 8.6EE-13 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TES note (4)

T3C-2-$2-10 T3Of T3C*/C*/st*o*U1*/U2*W1*W2*/X2N3 1.12E-7 7.84E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES mote (4)

T3C-2-52-11 T3Cy T3.**/C*/s1*tPU1*U2*W1*W2*X2 1.24E-9 8.60E-13 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES tote (4)

T3C-2-52-15 T3CW T3C*/C*/s t*o*U1*U2*/X1*/v1*E1*W2*/x2N3 9.19E-9 6.43E-12 Recovery of PCs in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Nete (4)

T3C-2-S2-16 T3Od T3C*/C* fs1*1PU1*U2*/X1*/V1*W1*W2*M2 1.24E-9 8.6tE-13 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TES mote (4)

$ 73C-2-S2-20 73CW T3C*/C*/51*o*U1*U2*/X1*v1*/v2N1*W2*/x2*W3 9.19E-9 6.43E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TES note (4)

T3C-2-52-21 T3CW T3C*/C*/st*o*U1*U2*/xt*vi*/v2*W1*W2*x2 1.24E-9 8.68E-13 Recewery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES note (4)

T3C-2-S2-25 T30# T3C*/C*/s t*tPU1 *U2*/r1*V1*v2*/V3*W1*W2*/K2N3 1.24E-9 8.68E-13 Recevery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TES mote (4)

T3C-2-S2-26 T3Cu 'T3C*/C*/81*o*U1*U2*/x1*V1*v2*/V3*w1*W2*X2 1.24E-9 8.68E-13 Recovery of PCS in 24 hnurs. YES Note (4)

T3C-2-52-30 73CW T3C*/C*/s1*o*u1*u2*/X1*V1*V2*V3*/V4*W1 N 2*/X2 2.48F-9 1.74E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES note (4)

• W5 T3C-2-S2-31 T30# T3C*/C*/st*tPU1*U2*/X1*V1*V2*v3*/v641*W2*x2 2.48E-9 1.74E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Note (4)

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Sheet 24 TABLE 3 TW SEQLENCES SEFmtE Ape AFTER RECOWERY LOSS OF SAFETY AC SUS 10500 (TAC-10500)

SEQUENE SEQUENCE SEQUENCE DESIGGIATOR SEGUEW E SEQUEWE APPLICASLE MWEWE 095W5T5 NLSIBER TYPE FREelfuCY FWEetENCY ECUWEWY ACTIONS ELintuRTED BEFGIE AFTER ,

RECOWERT REQWERY TAC-10500-4 TAW 1 AC5*/C*/81*o*/P*N1*/U1x*W1*W2 8.25E-T 5.7tE-10 accovery of PCs in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TES note (4) s TAC-10500-7 TACM TACS*/C*/81*o*/P*/U1*U1M*W1*W2 3.4E-9 2.09E-10 accovery of PCs in 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. TES Note (4)

TAC-10500-20 TACM TAC 5*/C*/B1*o*/P*U1*/U2*W1*W2 4.20E-7 2.96E-10 accovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. VES esote (4)

TAC-10500-Z5 TACW TAC 5*/C*/91*o*/P*U1*U2*/X1*/V1*W1*W2 5.15E-B 3.61E-11 secovery of PCs in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. VES note (4)

TAC-10500-26 TAOd TAC 5*/C*/s1*o*/P"Ut*U2*/X1*V1*/V2*W1*W2 1.72E-9 1.20E-12 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TES note (4)

TAC-10500-29 TAoi T AC5*/C*/s1*o*/'*U1 *U2*/X 1 *V1 *V2*/V3*41 *W2 <1.0E-9 None VES O TAC-10500-32 TACM TACS*/C*/81*o*/P*U1*U2*/X1*V1*V2*V3*/V4*Wi*W2 <1.0E-9 None VES u

TAC-10500- TAO # TAC 5%'11*ov1*/U1*W1*W2 1.14E-7 7.9 E-11 Recovery of PCS in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. VES uste (4) 34a-S2-4 TAC-10500- TACM TACS*/C*/E1*o*P1*'1*/U2*W1*W2 6.84E-9 4.79E-12 Recovery of PCs in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TES Note (4) 34a-S2-9 TAC-10500- TACM T ACS*/C* f81*o *P1*U1*U2*/x 1*/V1*W1 *W2 <1.0E-9 #ene TES 34a-52-14 TAC-10500- TACW TACS*/C*/81*o v i*U1*U2*/X1*V1*/V2*W1*WZ <1.0E None VES 34a-SZ-19 TAC-10500- TACW TACS*/C*/81*o*P1*U1*U2*/X1*V1*V2*/V3*W1*W2 <1.0E-9 None YES 34a-52-24

, TAC-10500- TACW TACS*/C*/s1*o41*Ut*U2*/X1*V1*V2*V3'"/V4*W1 <1.0E-9 mone YES 34a-52-29 *W2

Sheet 25 TABLE 3 TW SE0lKuC'S IIEFtutE AfD AFTER EEQNERY LOSS OF SAFETY AC gus 10600 (TAC-10000)

SEQUENCE SEGUEIICE SEGUFn DESIGIIATOR SEGUEleE lSEGUEW APPLICAELE WeEWE OpgesTS MUMOER TYPE FaEQUENCY FREstfuCY RECDUERY ACTIONS ELissteATED SEFCRE AFTER IEECOWERY IEECDUERY TAC-10500- TACW TAC 5*/C*/81*rP2*/U1*/V2N1*W2 2.38E-9 1.66E-12 Recovery of PCs in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. YES Ilote (4) 34trS1-3 TAC-10500- TAW TAC 5*/C*/31*O*P2*/U1*V2*/V3 N1*W2 <1.0E-9 NOME YES 346-51-6 j TAC-10500- TAQs TAC 5*/C*/91*o*P2*/U1*V2*V3*/V4*W1*W2 <1.0G-9 moue YES 346-51-9 TAC-10500- TAW TAC 5*/C*/s1*o*P2*U1*/X1*/V2*W1*W2 <1.0E-9 HOME TES 34b-S1-13 TAC-10500- TAW TACS*/C*/s1*o*P2*U1*/X1*V2*/V3*W1*W2 <1.0E-9 NOME VES c 34b-51-16 u

TAC-10500- TACW TAC 5*/C*/s1*O*P2*U1*/X1*V2*V3*/V4*W1*W2 <1.0E-9 WOIE TES 34b-51-19 TAC-10500- TAQs TAC 5*/C*/81*Q*P3*/V2*W1*W2 <1.0E-9 NOME YES 34c-A-3 TAC-10500- TAOi TAC 5*/C*/81*o*P3*V2*/V341*W2 <1.0E-9 NOME YES 34c-A-6 TAC-10500- TACW TAC 5*/C*/s1*Q*P3*V2*V3*/V4*W1*W2 <1.0E-9 NOME YES 34c-A _ ___ .__ __ . . . _ . . . _ _ _ _ . . . _ _ . _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - .- . . - _ _ - - - _ - . - . _ _ _ , - . - _ _ - _ _ . - _ _ _ _

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Sheet 27 TA9LE 3 TW SEQ EIICES K FWE AID AFTER RECDUERY LOSS OF SAFETY AC SUS 10600 (TAC-10000)

SEQUENE SEGUENCE SEGUENE DESIGIIATOIt SERIEWE WeelCE APPLICASLE WesIE CIpserTS IRDIRER TYPE FREm elCY FREGIENCY RECDUERY ACTIONS ELisellunTED BEFm E AFTER I RECOWERY RECDUERY TAC-10600-16 TAW TAC-10600*/C*/91*O*/P*/U1N1*W2*/x2N3 5.00E-8 3.50E-10 Recewry of PCS 5 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. TES Note (4)

TAC-10600-16e TA W TAC-10600*/C*/B1*0*/P*/U1*W1*W2*x2 <1.0E-9 NOME YES

! TAC-10600-19 TACW T AC- 10600*/C*/91*O*/P*U1*/U2*W1 *W2*/X2N3 <1.0E-9 IIONE YES TAC-10600-19e TA W TAC-10600*/C*/81*o*/P*Ut*/U2*W1*W2*x2 <1.0E-9 IIONE YES TAC-10600-22 TACM T AC- 10600*/C*/s t*o*/PN1*U2*/x-/v1*W1 *W2*/M2 <1.0E-9 NOIE YES

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• /X2*W5 TAC-10600-25e TAOf T AC- 106tW C*/81*0*/P*U1*U2*/X*V1 */V2*W1*W2 <1.0E-9 NOME MS

• x2 e i TAC-10600-28 TACW TAC-10600*/C*/81*S*/P*U1*U2*/X*V1*V2*/V3*W1 <1.0E-9 NOME YES

• W2*/x2*W3 TAC-10600-28e TACM TAC-10600*/C*/51*S*/P*U1*U2*/F ui % 2*/V3*W1 <1.0E-9 NOME YES

• W2*M2 TAC-?0600-31 TAot TAC-10600*/C*/51*o*/P*U1*c2*/# wi:v? v3*/ W <1.0E-9 WOME YES

• W1*W2*/X2*W5 TAC-10600-31e TACW T AC- 10600*/C*/91*O*/P*Ul*U2*/X*V1*V2*V3*/W <1.0E-9 NOC YES

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Sheet 29 TABLE 3 TW SEetENCES BEftBE AUS AFTER aECDUERY LOSS OF SAFETY AC BUS 10600 (TAC-10600)

SEGUENCE SEGUENC: SEGUEN E DES 19 eTOR SEOUEWE SERIEWE APPLICABLE SEG EN E CDUEWYS NISIBER TYPE FWEOLENCY FREstENCY RECUuEWY ACTIONS ELIWieATED BEFORE AFTER RECOWEWY RECDUERY TAC-10600-33b TAOf T AC - 10600*/C*/91*EPP2*/U1*/V2%f1*W2 <1.OE-9 uOME VES 3

TAC-10600-33b Taos TAC-10600*/C*/s1*o*P2*/U1*V2*/V3*W1*W2 <1.0E-9 NOME TES 6

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$ 16 TAC-10600-33b TACW TAC-10600*/C*/s1*trt2*U1*/X1*V2*V3*/W*WS.982 <1.0E-9 NONE VES

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Sheet 30 TMLE 5 TW SEMEuCES KFSE W AFTER EtXNERY LOSS OF SAFETY AC SUS 10642 (TAC-10000)

SEeUENCE SEGUENCE SEGUEWE DESIGNATOR SEefENE SEeUENE APPLICABLE SEWEum CesENTS IRSIBER TYPE FREstEuCY FEEELENCY REtXNERY ACTIONS ELI8eleATES BEFW E AFTER IIECOWERY NEtXNERY TAC-10600-46 TAOf TAC-10600*/C*s1*/82*/P*/U1*W1*W2*/M2N3 <1.0E-9 MOIE YES TAC-10600-46e TAOf TAC-10600*/C*S1*/82*/P*/U1*W1*W2*x2 <1.0E-9 NONE YES TAC-10600-49 TACW T AC - 10600*/C*S1 */s2*/P*U1 */U2Ni*W2*/1GMG <1.0E-9 NONE YES TAC-10600-49m TAOf T AC- 10600*/C*s1*/82*/P*U1*/U2N1*W2*K2 <1.0E-9 NOIE YES TAC-10600-52 TAOf TAC-10600*/C*s1*/82*/P*U1*U2*/X1*/V2*W1*W2 <1.0E-9 NONE TES

• /X2*W3 TAC-10600-52e TAOf TAC-10600*/C*s1*/s2*/P U1 N 2*/X1*/V2 N1*W2 <1.0E-9 IIDIE TES to *M2 (D

TAC-10600-55 TAOf TAC-10600*/C*s1*/s2*/PN1*U2*/X1*V2*/W3N1*W2 <1.0E-9 NOME TES

• /x2N5 TAC-10600-55a TApr TAC-10600*/C*s1*/82*/P*U1*U2*/X1*V2*/V3*Wi*W2 <1.0E-9 NOME YES

• x2 TAC-10600-58 TACW TAC-10600*/C*s1*/s2*/P*U1N2*/X1*V2*V3*/V4*W1 <1.0E-9 NONE YES

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(1) With offsite power resterted, operator can then restert PCS for CMt.

Therefore, for this sw to result in an unacceptatde outcome, e subseepsere, feiture of PCS af ter offsite pouer is recovered would have to occur.

(2) With the i. a fault repelred, eentairement heet renewet would be established and conteirment integrity and cootont in}ection would not be throetened.

(3) These feutts are recoverstne by an operator ameipulatire e component tocetty (e.g.) manuelty openirg a vetve or closing the breaker.

- Again, containment heet removet would be established and conteirament integrity and coolant injection would ret be threetened.

(4) With PCS restored, core heet removat is established. Therefore coolant injection and conteirament heet renwet would be provided.

(5) With DC pouer restored, core heet renewet is established. Therefore coolant injection and conteiramerit heet removal would be provided.

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l 4

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s

Table 6 Recovery of- DC Safety: Bus System a t

Time to Recover Probability of,Non-Recovery 5-10 min 0.9 10-20 min 0.9 20-30. min 0.8 30-40 min 0.7 40-60 min 0.6 60-70 min 0.6 70-120 min 0.4 2-4 hrs 0.1 4-6 hrs 0.05 6-8 hrs 0.01 8-12 hrs 0.002 24 hrs 0.001 112 i

Table.7 Probability of Failing to Repair Battery Panels Within Time (t) l Time Probability of Non-Recovery 1 hr 0.67- I 4 hr 0.20 9 hr. 0.03 12 hr 0.008 l 18 hr 0.0007 24 hr 0.0001-l i

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1 Table 8 1

Probability of Failing to Repair. Heat Exchangers Within Time (t) 4 T

Time Probability of Non-Recovery I

-1 hr .954 4 hrs- .850

-9 hrs .657 12 hrs .551

• 18 hrs .432 24 hrs .326 r

114

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Table 9  !

l Probability of Failing to Repair j AC Buses Within Time (t)

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-Time Probability of Non-Recovery i 1 hr o,93 4 hrs o,74 9 hrs 0.50 12 hrs o,4o j 18 hrs- 0.25 24-hrs 0.15 4

115

E-l-

I Table 10 Loss of Containment Heat Removal Initiator Frequencies i

Initiator Frequency

• T1 5.76E-8/RY T2 1.27E-7/RY T3A- 4.08E-9/RY T3B 3.01E-9/RY T3C

<1.0E-9/RY TAC-10500 1.24E-9/RY TAC-10600 <1.0E-9/RY l

TDC-A 2.72E-8/RY TDC-B 3,05E-6/RY- l A

1.76E-8/RY S1 3.96E-8/RY.

S2 8.82E-9/RY-Total 3.17E-7/RY

• Each frequency is calcult.ted by summing.the individual sequence frequencies that tre dominant (i.e., >1.0E-9) 116

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