ML20153D882

Assessment of Consequences of Seismic Event at Big Rock Point Plant

ML20153D882
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	08/29/1988
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
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ML20153D691	List: ML20153D687 ML20153D882 ML20153D891
References
NUDOCS 8809060071
Download: ML20153D882 (321)

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V Ad ASSESSMENT OF THE CONSEQUENCES OF A SEISMIC EVENT AT THE BIG ROCK POINT PLANT CONSUMERS POWER COMPANY 8809060071 00002Y PDR ADOCK 05000155 P PNU NUO333-2628A-BQ01-NLO4

)

i ATTACHMENT l i

Consumers Power Company

, Big Rock Point Plant  !

Docket 50-155  ;

AN ASSESSMENT OF THE CONSEQUENCES

, OF A SEISMIC EVENT AT THE BIG ROCK POINT PI. ART August 29, 1988 i

l 1

333 Pages OC0588-0011-NLO4

TABLE OF CONTENTS Page No.

I. Introduction 1 II. Selection of Initiating Events 10 III. Identification of Important Systems and Equipment Fo11osing a Seismic Event 16 IV. Big Rock Point Plant Response To a Seismic Event (Event Tree Description) 28 V. Seismic Capacity, Failure Modes, Effects, Repair and Recovery 50 VI. Methodology for Identification of the Seismic "Weak-Links" at Big Rock Point 83 VII. Fault Tree Logic Summary 94 VIII. Results and Conclusions 95 APPENDIX A -

Fault Trees /

APPENDIX B -

Seismic Fragility of Big Rock Point Core Assembly /

and Reactor Vessel Supports SMA Report #13703.01, /

October 1985 /

ADDENDUM -

Evaluation of Alternate Shutdown Building /

Hodification Omission /

NUO383-2628A-BQ01-NLO4

1 I. INTRODUCTION Review of the seismic design of the Big Rock Point Plant began in 1979 as a part of Systematic Evaluation Program (SEP) Topic III-6. Since that time, relatively significant analyses and modification of the Big Rock Point Plant have been completed. Evaluations performed to / ate have included 15 major structures and systems; anchorage of 56 equipment items which could have an impact on safety-related electrical equipment during a seismic event; analysis o' various major mechanical equipment important to plant response following an earthquake; and, seismic qualification of cable tray and conduit raceway systems. Capital  ;

expenditures to date have totaled uore than $2 million excluding work performed in house or the coat of modifications. Reevaluation of Big Rock Point seismic resistance therefore, has teen the single most resource intensive of the more than 100 original SEP topics. It is also clear from the Nuclear Regulatory Staff's (Staff) draft Safety Evaluation Report (SER) dated October 19, 1982, that seismic reevaluation of Big Rock Point is also the single SEP topic with which the Staff had the most difficulty in coming to a clear conclusion.

In December, 1982 the Staff requested Consumers Power Company to t investigate alternative methods of resolving the differences of opinion as to the seismic design adequacy of the plant. These alternative methods could consider the comparison of seismic risks at the Big Rock Point site with those considered acceptable at other more typical nuclear power facilities and an evaluation of the consequences of failures as a result of an earthquake or combinations ther60f. In /

June 1983 the original weak-link analysis was submitted to the staff / .

without any recommendations or conclusions. In November of 1985 /

Consumers Power Company respoaded to the staff's request with what /

is effectively an assessment of the consequences of a seismic event /

at the Big Rock Point Plant. /

In that report the weakest links at Big Rock Point were identified. / I

{ The identified pieces of equipment were assumed to fail at 0.0G ground /

acceleration since their true seismic capacity was unknown. Consumers /

Power Company committed to upgtede or analyze these pieces of equipment /

during the 1988 refueling outage to improve their seismic capacity. /

t Since this time, these items have been delayed per letter to the NRC /

dated May 6. 1988 due to required completion of higher plant priority /

projects and manpower availability. /

T In this report the next set of weak-links whose seismic capacities /

are greater than 0.0G have been identified and proposed fixes /

developed. The weak-links identified are listed in Table yIII-5. /

The conclusions from the Seismic Weak-link analysis have been / I presented to the Big Rock Point Technical Review Group (TRG). /

These proposed fixes will be ranked by TRG when cost estimates for /

implementation are available. /

t NUO383-262SA-BQO1-NLO4

2 i

In evaluating the effect of an earthquake at Big Rock Point, conse-quences can be viewed in two different ways. As indicated by the Staff's request there are the consequences of the failure of various plant components and their effect on the transient response of the plant to these seismically-induced failurest there are also the con-sequences of the earthquake with respect to the health and safety of i the public should sufficient systems and equipment in the plant fail with a resultant significant radiological release from the site.

Seismic hazards levels are specified for each nuclear site by the NRC to minimite the potential for such a radiological release. The  ;

seismic-hazards curve for the Big Rock Point site is presented in NUREG-CR-1582, "Seismic Hazards Analysis" and is graphically displayed along with the hazards curve from the Big Rock Point PRA in Figure I-1 of this report. The Staff has defined the design basis ground acceler- i ation for a given nuclear facility to be that which occurs less l 1 frequently than once every thousand years. If Big Rock Point were a  ;

new facility being constructed 'oday, the design basis grcund accelete-tion would be less than 0.11g (referring to the Staff's hazards curve).

Gfven that Big Rock Point is significantly different than a typical or eserage facility being built or in operation today in that it is much sma31er, consideration should be given to applying design basis ground ,

acceleration criteria such that the risk or level of protection

afforded the public, fs commensurate with that which is considered acceptable at a newer typical or average facility.

l The public dose following a significant event can be viewed in one of t two ways that dose to the public as a whole; or, that dose an individual may receive as a result of the relecse. The dose to the public as a v'. tole per curie released has been published for a variety of plants in NUREG-CR-1497 "Radioactive Materials Release from Nuclear

, Power Plants" and NUREG-CR-1498 "Pepulation Dose Cotmaitment Due to  !

Radioactive Releases from Nuclear Power Plant Sites in 1977." A l susanary of the information presented in these documents is shown in l Table I-1. If the average dose per curie released for these 33 plants represents that dose which could be expected from an average or typical  :

plant *, then it can 1,e seen that a curie releaJed at the Big Rock Point site results in a dose to the public of a factor of 14 lovat than  ;

a dose from an average or typical site. This is due to the lower j

population density and its distribution in the Big Rock Point area.  !

Factoring in that the Big Rock Point Plant design included only 10% of i i

i the radionuclide inventory of the typical plant, one can see that the

! risk to the public as a whole is a factor of 140 less severe for a given  !

i earthquake at the E,ig Rock Point site than at a typical site. Risk to  !

t individuals in the Big Rock Point vicinity is a different master,

however. 'the potential dose to an individual is independent of the {

population density or its distribution. Therefore, the individual risk [

4 for a given earthquake is reduced only by the fractional fission product  :

inventory found at Big Rock Point as compared to another site, 10".  !

t i The risk to an individual in the Big Rock Point Plant area it, therefore 1 a factor of 10 less than that for an individual near a typical facility, i r

t

• Assumes meteorology, ratio of dose to inhalation, aad terrain similar for i these plants.  ;

NUO383-2628A-BQU1-NLO4

3 Using only the individual risk perspective, one can conclude that requiring the design basis ground motion for the Big Rock Peint Plant to be equivalent to that which would be applied to a never typical facility results in the mandatory implementation of s level of protec-tion to the health and safety of the public at least a factor of ten more restrictive than at the newer facility. Such a requirement is not necessary, given the public riak posed by Big Rock. CPCo and the staff concluded that less quantitative, rigorous approaches to determining the seismic risks were appropriate. (CPCo to NRC, June 1, 1983).

Qualitatively, an assessment of a power plant's resistance to an earthquake can be made by using Table I-2 of this section, which gives a brief description of the effects on objects at various Modified Mercalli Intensity (MMI) categories. Using the original plant design seismic loading f actor, low end Category VI ef fects may be expee'.ed.

If the Staffs hazard curve value for Big Rock Point (if Big Rock were considered as a new facility). high end Category VI - low end Category VII effects may be expected. Since Big Rock Point is constructed of structural steel and reinforced concrete (with masonary catagorized as masonary 'C') the expected damage would include some cracks to the masonary 'C' walls and broken glassware, but no mejor structural damage is expected to occur. This, therefore, supports the conclusion that qualitative assessments of the need for seismic upgrading of the Big Rock Point Plant are appropriate.

An alternate approach in performing a seismic consequence analysis is ,

to forego the use of hazards nirves altogether and attempt to determine I the maximum size earthquake the facility can withstand based on knowledge of plant transient response and relative structural capacity of plant systems and components. This is the approach taken in generating this report. A brief dsscription of this approach follows and is presented in greater detail in later sections of the report.

Identifying the design features of the plant most susceptible to the earthquake, those trandients which were felt most important or most likely to occur following a seismic event were selected. Three transients were chosen for study: loss of offsite power; a medium steam line break inside containment; and an Anticipated Transient Without Scram (ATWS). Loss of offsite power was chosen because of its ef fect on essential',y every mitigating system at Big Rocx Point. Other transients affect only portions of the systems availabic to shut down and cool the reactor. The medium steam line break inside containment was chosen for the Loss of Coolent Accident (LOCA), because adequate core cooling is dependent on most systems. Steam line breaks inside  :

containment also require the use of enclosure sprays to maintain containment atmosphere within the environmenta'. qualification envelope, i A medium break was selected because it requires the satisfactory '

functioning of the Reactor Depressurization System (RDS) to prevent core damage.

I l

l l NUO383-2628A-BQ01-NLO4 l

i t

4 The logic behind each of these transients had been developed previously in the PRA. The loss of offsite power and medium steam line break event trees were extracted from the PRA and modified to reflect the use of only those systems which could potentially be shown to survive the earthquake. The logic (by which failure of the systems were noted) in the event trees was also extracted from the PRA. Fault trees for each of these systems were reviewed and substantially simplified. Av an example, the RDS tree was revised to include only a single train of power supplies, sensors, actuation cabinets, and depressurization valves because all four trains are essentially identical to each other in terms of their function, location and structural features. In other words, if one train fails as a result of the seismic event, this study assumed the likelihood of a similar failure in the other traine was quite high. Components which may have dissimilar seismic resistance (such as the two diesel generators) or have functional dissimilarities in the way they operate (such as the fire pumps and their power supplies) were not combined in this manner. Passive components and l structures normally unimportant during these transients but whose failures may be made important as a result of ground motion were added to the trees (such as masonry walls).

By combining the fault trees for each core damage sequence, dependencies between the important systems were identified and a complete list of cut sets

• for each sequence was produced. All the /

combinations of all the failures which must occur following an /

earthquake at Big Rock Point were thus tabulated for each core damage /

sequence identified in the event trees. /

A conservative estimated ground accelera:1ea which would result in the failure of each component in the fault treep was determined. Applying these accelerations to the cut set members, the acceleration at which all members of a cut set will fail was determined. This acceleration was the acceleration at which the strongest component in the cut set failed and represented the seismic resistance of that cut set. The seismic resistances for all the cut sets were ranked with respect to the size of the earthquake necessary to satisfy the cut set. Those cut sets which were satisfied at the smallest ground accelerations represent the seismic "weak-links" at the Big Rock Point Plant. These weak-links are the components and structures at which proposed modifi-cations for upgrading the plant should be aimed. Modifications addressing members of the more seismically resistant cut sets are of little benefit unless the weaker links are also cddressed.

It was recogn' red that the best estimate ground accelerations for many of the components identified in the attachments were not available.

• A minimal cut set is a smallest combination of component failures which, if they occur, will cause the top event to occur, VII-15 NUREG-0492 - Fault Tree handbook - January 1981.

NUO383-2628A-BQ01-NLO4

5 However, a great deal of information was extracted from the Big Rock Point design and review analyses performed to date as a part of the SEP.

Where this information was not available, a conservative app'roximation of the failure acceleration was used, in some cases even a value of zero. The more components with which this approach was taken the more artificially inportant seismic events would become. The weakest of the weak-links combined with the acceleration at which the transient occurs represented an estimate of that size earthquake which would result in a significant release. This estimate obviously does not represent a rigorous aeterministic analysis. Nor does it represent a quality PRA evaluation. Risk is presented in units of ground acceleration t.ad no mention of probability is made. The acceleration at which the weakest link fails cannot be translated to a core damage probability by use of a hazards curve because the effect of random failures on top of seismically induced failures has,not been incorporated in this analysis and no effort has been made to address the effects of the uncertainties associated with the assumed component capacities.

Although this approach can be used to rank the relative importance of various combinations of failures against one another, it is, also very useful in focusing any future efforts in assessing or upgrading the seismic capacity of the Big Rock Point Plant.

The rer:rt is broken up into several sections besides the introduction.

These include the following:

Section Description II Selection of Initiating Esents Identifies the appropriateness and completeness of selection of the loss of offsite power, medium steam line bc 'k and failure to SCRAM (ATI.*S) in evaluating seismic risks at Big Rock Point.

III Identification of Important Systems and Equipment Introduction to Big Rock Point systems, structures and components important in the mitigation of the loss of offsite power and medium steam line break events.

I NUO383-2628A-BQ01-NLO4

6 IV Big Rock Point Plant Response to an Earthquake Detailed description of plant transient response to steam line

, break and loss of offsite power events. Introduction of event tree logic for these events.

V Seismic Capacity of BRP Structures and Components Identification of assumed capacity for each structure and component important during a seismic event.

Failure mode and effects analysis for each component. Discussion of assumptions made with respect to repair and recovery following an earthquake.

VI Methodology Descr (cion of methods developed for 44entification of Big Rock Peint seismic "weak-links".

VII Fanit Tree Logic Summary /

System logic models important to loss of offsite power and main stear,line break events.

VIII Results and Conclusions Prioritized listing of seismic "weak-links" at Big Rock Point ranked from weakest to strongest.

Commitments to upgrade riant.

Appendix A Fault Trees /

/

Appendix B Seismic Fragility of Big Rock /

Core Assembly and Reactor Vessel Supports. SMA Report 13703.01 /

Addendum Effects of Alternate Shutdor. /

Building Modificatfor Omission /

NUO383-2628A-BQ01-NLO4

i 7

1 TABLE I-1 Ferson-%es Curies 10~' Feroon-ree FoFulation 2-80 km (Whole body)_ (Noble gas) Curie Flant Tm ,

Fower (Hwe) 436 1.6+5 0.09 1.39+4 0.65 f Arkansas FWR 1.34+4 2.09 Big Rock skit 63 1.3+5 0.28 1067 6.7+5 2.7 1.66+5 1.62 Brevns Ferry SWR 799 1.9+5 6.3 2.46+5 2.56 Erunswick BWR 2.23*4 3.14 Calvert Cliffs FWR 850 2.4*6 0.7 1.1+6 0.063 3. G t't 1.66 Cook FWR 1054 778 1.8*5 0.017 1.27+3 1.34 Cooper BWR

]

825 2.2+5 0.014 - 3.35+3 ' .42 Crystal River FWR 0.63 i

FWR 906 1.8+6 0.006 1.27+3

) Davin Besse 180.0 3.3+5 461 Dresden SWR 1003 6.4+6 j 3.87+3 7.7

Duane Arn';'.d BWR 545 5.7+5 0.3 BWR 821 8.3+5 0.54 2.33*4 2.3 Fitag,atrick 1.15 FWR 478 7.5+5 0.044 , 3.31+3 Fort Calhova 1.75

]

FWR 490 1.2+6 0.056 3.2+3 Cinna 70.5 FWR 582 3.4*6 2.2 3.12+3 Conn Yankee 1.9+3 5.26 i

I! arch SWR 786 2.8+5 0.1 FWR 873 1.6+7 12.0 1.6+4 75.0 Indian Folnt .

0.87 FWR 535 6.0*5 .021 2.4+3 6:cuaunce 3.76 50 3.3+5 1.6 4.25+4 IaCrosse SWR 2.86+2 3.5 r.tice Yankee FWR 825 5.7+5 .01 FWR 1530 2.5+6 220.0 6.2+5 35.5 Millstone 1&2 6.87+3 2.9 Honticello FWR 536 2.1+6 0.2 SWR 610 8.3+5 .098 3.5+3 2.8 Nine Mile Ft. 3.56+4 1.94 FWR 860 7.4+5 0.69 Oconee 740 1.0+6 1.5-3 59.9 2.5 Palisades FVK 7.11+4 7.0 Peach Botton BUR 1065 4.1+6 5.0 670 4.4*6 52.0 4.13+5 12.6 F11 grim FUR 47.5 FWR 520 2.1+6 0.32 673.0 Frairie Island 5.1 sk1t 789 6.7+5 1.3 2.56+4 Quad Cities 0.024 4.76+2 5.0 Robinson FWR 665 6.4+5 777 2.9+5 0.58 2.54+4 2.28 St Lucie FWR 1090 4.9*6 0.038 19.6 194.0 Salen* FWR 29.5 FWR 1040 7.C+6 9.5 3.22+4 Zion

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10 II. SELECTION OF INITIATING EVENTS In this section, a discussion of the possible initiating transients which could occur at the Big Rock Point Plant as a result of a seismic event will ne presented. An initiating event is a failure which requires active response of systems and/or the operator in order to prevent an accident sequence. A number of initiating events were identified as a part of the Big Rock Point PRA (Appendix XII). The development of this list of events was an iterative process utilizing a variety of industry sources (see Table II-1), Big Rock Point Plant specific incidents and the event trees and fault trees of Appendixes I and II of the PRA. This list is considered sufficiently complete to use as a basis for studying the response to a seismic event of the Big Rock Point Plant.

Unless the fragility of the equipment required to fail in order to initiate a transient is assessed, the assumption must be made that any of the= initiating transients identified in Appendix XII of the PRA is possible. It is desirable to limit the number of systems required for study following a seismic event due to the complexity and cost of evaluating the seismic resistance of stru tures and components associated with those systems. For this reason an attempt will be made to categorize the events in PRA Table XII under limited number of transient and accident headings and choose a most limiting event for each category that requires suf ficient systems and equipment to completely characterize the Big Rock Point Plsnt operating response to an earthquake regardless of the transient that occurs.

Four major category headings will be chosen under which the t ransients and accidents will be compiled: Anticipated Transients; LOCA; reactivity transients; and, external events. The last category, external events, is the category in which earthquakes reside. Other external events include wind, tornado, fire, toxic chemical accidents, airplane crashes, flooding, etc, all* believed to be extremely remote simultaneous with an earthquake. This category will not be considered further.

Tables II-2, 3 end 4 contain a, tabulation of transients under each of the three remaining categories. The lists begin with what is considered the most limiting transient representative of all the other transients in that catego 7 The three most limiting transients are: Loss of Offsite Power; Medium Steam Line Break Inside Containment; and, ATWS for the anticipated transient, LOCA and reactivity transient categories, respectively. A statement as to why each event listed is believed to be conservatively bounded by the most limiting transient is presented following the transient description.

Only one transient, simuitaneous closure of the recirc loop valves was not categorized. Based on both the availability of ac power in order to initiate this transient and the relatively significant amount of time the operator 'aas to terminate the transient by reopening the valves or providing a makeup water source, this transient is probably less limiting than a loss of offsite power. However, as the power to the control circuitry for these valves is normally disabled through the hand switch, NUO383-2628A-BQ01-NT,04

11 a closure of these valves resulting from the earthquake was not considered possible and the transient was not categorized.

NUO383-2628A-BQ01-NLO4

l t

12 ,

e t

TABLE II-1 -

INDUSTRY AND BRP TRANSIENT INITIATOR REFERENCES GE Standard SAR i

GE Generic ATVS Report NEDO-10349 i I

WASH - 1400  !

EPRI Report on ATWS EPRI NP-801 Standard Review Plan Chapter 15 f

t Systematic Evaluation Program Topics - Item XV  !

GE Service Information Letters ,

. i t

EPRI LER Data Base i

BRP Control Room Logbooks and SCRAM Reports i

BRP LER3 I

i l

i

(

F i

i I

t I

NUO383-2628A-BQO1-NLO4

13 TABLE II-2 ANTICIPA1ED TRANSIENTS POTENTIALLY RESULTING FROM AN EARTHQUAKE Description Discussion Loss of Offsite Power (LOSP) Most Limiting Transient Turbine Trip Considered as a Part of LOSP

Load Rejection Considered as a Part of LOSP Loss et Main Condenser Considered as a Part of LOSP MSIV Closure Considered as a Part of LOSP IPR Fails Closed Considered as a Part of LOSP Miscellaneous Plant Occurrences No Out-of-Tolerance Conditions Necessarily Exist Manual SCRAM No Out-of-Tolerance Conditions Necessarily Exist Spurious Nuclear No Out-of-Tolerance Conditions Instrumentation Necessarily Exist Loss of Feedwater Considered as a Part of LOSP I Loss of One Feed Pump Considered as a Part of LOSP Feedwater Controller Failed Considered as a Part of LOSP Closed Recirc Pump Trip considered as a Part of LOSP l Recirc-Pump Shaft Seizure Similar to Recirc Pump Trip 1 Only With One Pump Loss of Auxiliary Power Considered as a Part of LOSP Loss of DC Power DC Power Dependencies Considered as a part of LOSP Loss of Misc Power Panels Considered as a Part of LOSP Service Water Failure Considered as a Part of LOSP Instrument Air Failure Considered as a Part of LOSP l

i.

NUO383-2628A-BQ01-NLO4 i

14 TABLE II-3 LOSS OF COOLANT ACCIDENTS POTENTIALLY RESULTING FROM AN EARTHQUAKE Description Discussion Medium Steam'Line Break Inside Most Limiting Transient - Requires Cont (HSLB) Enclosure Spray, Core Spray, RDS and Post-Incident Cooling Feedwater Maximum Demand Assumed to Hydro Primary System and Results in Safety Relief Valve Sticking Open (Same as MSLB)

IPR Fails Open Blowdown through Turbine Requires Same Systems as MSLB Except Enclosure Spray Inadvertent Opening of Safety Requires Same Systems as MSLB Valve Spurious Opening of Bypass Same as IPR Failure Valve Spurious RDS Requires Same Systems as MSLB HELB in Recire Pump Room Requires Same Systems as MSLB HELB in Pipe Tunnel Requires Same Systems as MSLB Except Enclosure Spray Interfacing LOCA (Inside Cont) Requires Same Systems as MSLB Except Enclosure Spray Large LOCA (Inside Cont) Requires Same Systems as MSLB Except Enclosure Spray Medium LOCA (Inside Cont) Requires Same Systems as MSLB Except Enclosure Spray Large SLB (Inside Cont) Requires Same Systems as MSLB Except Enclosure Spray Small SLB (Inside Cont) Requires Same Systems as MSLB Large SLB (Outside Cont) Requires Same Systems as MSLB Except Enclosure Spray Medium SLB (Outside Cont) Requires Some Systems as MSLB Except Enclosure Spray Small SLB (Outside Cont) Requires Same Systems as MSLB Except Enclosure Spray NUO383-2628A-BQ01-NLO4

15 TABLE II-4 POTENTIAL REACTIVITY TRANSIENTS RESULTING FROM AN EARTHQUAKE Description Discussion ATVS Most Limiting Transient Loss of Feedwater Heater Momentary Power Transient Core Spray Injection at Homentary Power Transient Start-Up Rod Withdrawal at Start-Up Momentary Power Transient ,

Rod Withdrawal at Power Momentary Power Transient Rod Drop Momentary Power Transient Idle Recir Loop Start-Up Momentary Power Transient 1

9 I

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t NUO383-2628A-BQ01-NLO4

16 III. IDENTIFICATION OF IMPORTANT SYSTEMS AND EQUIPMENT FOLLOWING A SEISMIC EVENT Three transients were identified in Section II as being those incidents which would require sufficient systems and equipment to characterize completely Big Rock Point Plant operational response following a site seismic event. These transients were: a medium steam line break inside containment; a loss of offsite power; and an ATWS, In this section, a description will be developed of each system important for providing adequate core cooling following any of these three transients.

This description will include identification of major components making up each system, location of the equipment with respect to major plant structures and development of success or failure criteria associated with each system function. A list of major active components described in this section and supporting instrumentation and equipment is provided in Table III-1.

Fire Protection System To provide core cooling following a major LOCA or the actuation of the RDS, the Big Rock Point Plant is equipped with a low pressure core-spray system referred to as the Fire Protection System (FPS). The FPS is located in four distinct areas of the plant as shown on the system line diagram presented in Figure III-1. These four areas include the screenhouse, turbine building, reactor containment and core-spray heat-exchanger room located within the fuel cask loading dock structure.

The FPS consists of two fire pumps (one diesel-powered and one ac-powered).

Actuation of the FPS is automatic on low-steam drum level ($17" below steam drum center line) or low-fire-header pressure (<70 psi electric pump or <60 psi diesel pump). Manual actuation of the FPS can be accomplished from either the RDS Panel in the control room or from local control panels in the screenhouse. The FPS draws water from beneath the screenhouse and discharges to the yard piping through carbon steel and cast-iron piping components within the screenhouse.

The yard piping is buried cast iron. It encircles the turbine and reactor buildings and provides three paths into the plant through which core spray water can flow.

Two of the paths enter the plant through the turbine building. Turbine building fire piping is carbon steel and includes threaded joints and mechanical couplings. The core spray water passes through basket strainer filters prior to entering the containment through sections of welded piping in the turbine building pipe tunnel.

FPS inside the containment is primarily welded carbon steel piping.

Water flow to the reactor vessel following reactor depressurization is through two core spray piping paths. Each path contains two normally closed motor-operated valves which automatically open on combined low-reactor-water level (52'9" above the core) and low-reactor pressure (5200 psig). One pair of valves is de-powered (M07051 and M07061) while the other pair is ac powertd (M07070 and M07071). Operation of NUO383-2628A-BQ01-NLO4

17 the de powered valves permits water flow to a ring sparger through a penetration in the side of the reactor vessel. The ac powered valves feed a nozzle located in the reactor vessel head. Spray flow distribution from either of these sources is sufficient to provide adequate cooling of all the fuel assemblies.

A third path for FPS is actuated through the fuel-cask loading-dock area to the containment by opening a de powered motor-operated valve (M07072). Manual operation of the valve also can be accomplished locally.

Successful operation of the FPS during a LOCA requires the automatic J starting of at least one fire pump, the opening of either the ac or de motor-operated valves in the core spray lines and maintaining the integrity of the FPS from the fire pumps to the containmeSt building through the screenhouse, yard loop and turbine building. If time permits, the fire pumps can be started within minutes from the control room by operator action. If a significant amount of time is available, the pumps may be started locally in the screenhouse and any FPS failures which might have occurred in the turbine building can be isolated from the yard and bypassed by opening the de-powered valve (M07072).

Success of this alternate path is considered effective only if the fire piping leading to and in the containment are isolated from ruptured piping so as to prevent significant diversion of core spray water from the containment.

If coincident with a loss of offsite power, success of electrical components (electric fire pump, ac core-spray valves, and fire pump manual-start circuitry in the control room) requires the operation of the emergency diesel generator, the emergency electrical bus and associated electrical switchgear. Again, if significant time is available, the standby diesel generator can be started manually and pick up important loads after connection to the emergency bus.

Reactor Depressurization System In order to permit the functioning of the low pressure core spray system during the course of a transient, a method for lowering the pressure of the primary coolant system is necessary. At Big Rock Point, this function is provided by the RDS which allows a means of rapid depressurization of the primary coolant system should a loss of coolant inventory occur. The major components of the RDS are located in the containment, service building, turbine building and screenhouse.

The RDS is shown on a system line diagram of the primary system in Figure III-2. This system consists of four 6-inch diameter pipes each containing two-in-series normally closed valves. The first valve (CV4180 through 4183) is an air-operated fail-open isolation valve.

Air pressure is pruvided to the valve operator through a normally de-energized solenoid valve (SV 4980 through 4983). The second valve, referred to as the depressurization valve, is a pilot-operated solenoid valve which is closed when de-energized. Depressurization of the NUO383-2628A-BQO1-NLO4

18 primary coolant system through these lines requires energization of each solenoid valve resulting in the opening of the isolation valve and depressurization valve in each line. Three of the four paths must be opened in this manner to result in a sufficiently rapid depressurization of the reactor to permit core spray.

Actuation signals are provided through solid-state circuitry located in actuation and sensor cabinets in the service building. Sensors for satisfying RDS actuation logic include steam-drum level instrumentation, reactor level instrumentation (each located in the containment),

fire-pump discharge pressure instrumentation (located in the screenhouse) and the two-minute timers (located in the control room). Power sources for energizing the solenoid valves are derived from a set of UPS (uninterruptible power supplies) batteries located in the turbine building. System operation is accomplished as follows. A loss of primary coolant system inventory will result in a lowering of the level in the steam drum. At 17" below the center line of the drum, an actuation signal is sent to each of the two fire pumps in the screenhouse and a two-minute timer starts. The low-steam-drum level signal, fire pump discharge pressure (2100 psig) and the timing out of the two-minute timers are three of the four signals required to provide an actuation signal to the RDS valves. When the primary coolant level reaches 2'9" above the top of the core, the fourth signal is provided by the reactor level instruments completing the actuation logic and opening RDS valves depressurizing the reactor.

Success of this system requires proper functioning of two of four sets of steam-drum-level instruments, reactor level instruments, fire pump pressure switches and two-minute timers. The UPS power sources for each of the preceding sets of RDS instrumentation must be available and at least one fire pump must start. Proper functioning of each pair of valves in three of the four blowdown paths must also occur.

Time permitting, the operator can manually start a fire pump and actuate the RDS from the control room. Use of this manually initiated method of blowdown in conjunction with a loss of offsite power requires the functioning of the emergency or standby diesel generators, the emergency bus and associated switchgear, inclosure Spray System In the event that the earthquake results in the rupture of a steam line inside containment, the sticking open of a safety relief valve or the spurious actuation of an RDS train, it is possible that fluid escaping the primary coolant system could super heat resulting in an escalation of the containment atmospheric temperature above the 235'T environmental qualification temperature for important RDS and core spray system components. The purpose of the enclosure (ie, containment) spray system is to spray the containment atmosphere with water from the FPS, quenching the superheated steam and thereby lowering the containment atmosphere below the qualification temperature. The enclosure spray system is included as a part of the line diagram in Figure III-1.

NUO383-2628A-BQO1-NLO4

19 The enclosure spray system consists of two spray headers located at the top of the reactor building internal structure (ie, containment) with ,

each header controlled by a normally closed motor-operated valve. One '

valve (M07064) is automatically actuated by containment pressure switches located on the outside cable penetration area (~ 2.2 psig). '

This motor-operated valve requires a de power supply. An ac-powered motor-operated valve (M07068) controls the backup containment spray ,

header and is manually actuated from the control room should the de t valve fail to operate.

Success of the enclosure spray system requires that either of the two motor-operated valves open. The enclosure spray system additionally relies on the same FPS inside and outside containment that the core spray system requires for successful operation and includes the operation of either of the two fire pumps. Operation of the ac motor-operated valve requires that the emergency diesel generator, the emergency bus and the associated switchgear are operational when normal ac power is unavailable.

This system is not required when the primary coolant loss results from rupture of piping normally containing saturated liquid or a full blowdown of the RDS. ,

Post-Incident System Following RDS and core spray actuation, the containment will begin to fill with water coming from the primary coolant system, the core sprays and the enclosure sprays. On reaching an elevation in the containment between 587 ft and 590 ft (approximately 260,000 gallons), the operator is required to terminate water addition to the containment and initiate i decay heat removal by means of the post-incident system. Switchover to l post-incident recycle occurs on the order of 4 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> depending on the nature and location of the source of primary coolant inventory loss.

The major components associated with post-incident recycle are located in the fuel-cask loading-dock area and shown in the line diagram presented in Figure III-1.

Post-incident system initiation consists of starting one of the two core spray pumps, drawing water from the bottom of containment and pumping it through the core spray heat exchanger and core spray system piping back to the core spray and containment spray headers. Either i core spray pump can be started remotely by a hand switch from the control room. To remove heat from the containment water, FPS water is I supplied to the shell of the core-spray heat-exchanger through motor-operated Valve M07066 (an ac-powered valve remotely actuated from the control room). Fire water addition to the containment is terminated by closing hand-operated Valves VFP-29 and VFP-30 located in the turbine  :

building, r

NUO383-2628A-BQ01-NLO4

l l

l 20  !

l l

l Successful operation of this system requires the availability of a single fire pump, screenhouse fire piping (the yard loop), core spray piping in the fuel-cask loading-dock area, core spray piping inside containment, the functioning of M07066 and the operation of a single core-spray pump. Operation of the ac-powered equipment (core spray pump, M07066, and the electric fire pump) will depend on the emergency or standby diesel generators, emergency bus and miscellaneous 480 V ac distribution equipment. if coincident with offsite power failure.

Failure of remote operation of M07066 can be overcome by remote operation of M07080, which is parallel to M07066. Local manual operation of either valve is an option. Rupture of underground yard piping also can be bypassed by isolating the yard piping from the screenhouse and connecting fire hose from the hose manifold on the side of the screenhouse to Valve VPI-10 in the shell side of the heat exchanger.

Operation of the system is assumed to be required for a minimum of one month following its initiation at which time natural heat losses from the primary coolant system and containment are sufficient to ensure removal of decay heat.

Primary Coolant System Isolation l Should the transient which occurs following the seismic event not tesult in a significant loss of primary coolant inventory within the containment, it is beneficial to ensure that no such loss is occurring outside the containment by isolating the main steam lite. This function is accomplished by closing the MSIV (M07050). The MSIV is a de-powered motor-operated gate valve located just inside the contain-ment shell (see Figure III-2). Automatic operation of this valve requires that electrical equipment associated with the reactor protection system (RPS) be functional (in the event of a loss of offsite power voltage to the RPS will be lost, automatically sending a /

signal closing the HSIV) and that de power is available. (The de power /

source is located in the alternate shutdown building). If time permits, /

! the valve can be actuated manually from the control room, or frem the /

alternate shutdown building. /

Emergency Condenser System I Assuming a LOCA inside containment is not occurring and successful pri-mary system isolation is accomplished, the emergency condenser system is used for removal of decay heat and ulcinately the cooldown of the primary coolunt system. The emergency condenser is shown in both line diagrams found in Figures III-1 and III-2. The emergency conden-ser system consists of a large tank (ie, shell) of water in which two tube bundles are immetsed and through which steam from the primary system flows by natural circulation, condenses and returns to the steam d rum.

Each emergency condenser tube bundle has a normally open ac-powered inlet valve (M07052 and M07062) and a normally closed de-powered outlet i valve (M07053 and M07063). The de-powered motor-operated valves receive a signal to open when a high reactor pressure is experienced NUO383-2628A-BQ01-NLO4

21 (21450 psia) or a loss of station power occurs. Either emergency condenser tube bundle is capable of removing sufficient heat from the primary coolant system to accommodate decay-heat generation and prevent the actuation of steam drum safety relief valves (lowest valve set point is 1550 psi). The emergency condenser shell contains a stored water supply capable of removing the equivalent of four hours of decay heat. To prevent a reactor pressure rise to the safety relief valve set point due to depletion of the emergency condenser shell inventory, a source of makeup water are available through a manually actuated de powered valve (SV4947) which can be operated from the control room or the alternate shutdown building. The emergency condenser makeup line draws its water supply from the enclosure spray headers upstream of the de power-operated Valve M07064.

Success of the emergency condenser requires automatic or manual operation of either of the two de-operated outlet valves, manual actuation of the i makeup valve f rom the enclosure spray headers, the de power source is available from the alternate shutdown. building, and maintaining the integrity of the nakeup pipe, core spray piping inside and outside containment, yard piping, screenhouse fire piping and at least one fire

pump. As was assumed during operation of the core apray system, if turbine building piping failures occur and fire water is supplied to the containment through M07072, makeup to the emergency condenser shell is successful only if ruptured turbine building piping is isolated from the FPS to prevent diversion of fire water.

If coincident with a loss of offsite power, emergency condenser operation has a dependency on emergency ac power sources (either diesel generator, emergency bus and switchgear) only for cperation of the electric fire pump.

Control Rod Drive Makeup The purpose of the control rod drive system following an earthquake is to supply high-pressure low-volume makeup to the primary coolant system in the long term to overcome shrinkage due to cooldown and normal primary coolant system' leakage. Shrinkage of the primary coolant by itself will not result in uncovery of the core but when combined with suf ficient primary coolant leakage (<1 gpm unidentified and <10 gpm identified at 1350 psi reactor pressure) could result in extremely low reactor water levels. These levels are not expe:ted to occur for a day or more after shutdown of the reactor at these leakage rates. The core spray system can be manually initiated following cooldown of the reactor to less than 150 psi, but the control rod drive system is the preferred source of makeup as it can be initiated at any time during the cooldown, even with the reactor at elevated pressures.*

l Major components of control rod drive makeup are located in the reactor and turbine buildings. The system includes a 25,000 gallon

• It is emphasized that control rod drive makeup is a backup to core spray.

l NUO383-2628A-BQ01-NLO4

i 32 capacity condensate storage tank (which is normally more than half full), welded carbon steel piping located underground, the condensate pump room and the pipe tunnel in the turbine building, and the recirc pump room in the reactor building. Water flows through the underground piping to an air-operated valve (CV4090) which opens on a loss of instrument air or a loss of the condensate pumps to the suction of the control rod drive pumps (25 gpm positive displacement pumps) inside containment. From there the water is pumped to the reactor through the control rod drive mechanisms and the reactor cleanup system.

Success of this system requires periodic operation of at least one control rod drive pump and the opening of the normally closed Valve CV4090 11 the condensate pumps are not running or instrument air is not available. Control rod drive pump suction and discharge piping must remain intact as must the condensate storage tank.

Reactor Trip The electrical equipment and hardware required to ensure shutdown of the reactor by automatic rapid control rod insertion following an earthquake includes the RPS, control rod drive SCRAM piping, air-operated SCRAM valves and the CRD mechanism. Both channels of RPS circuitry are located in the control room and include the logic and relays for terminating power to the SCRAM valve solenoids. Reactor protection sensors most likely to trip the reactor following an earthquake include any or all of the following: reactor pressure (21400 psia); low steam drum level (58" below drum center line); low reactor level (52'9" above the core); high condenser pressure (28" Hga); high enclosure pressure

($1 psig); loss of voltage to reactor protection system (552 V); high neutron flux (2120); or manual SCRAM depending on the nature of the transient which is initiated by the earthquake.

Control rod drive hydraulic equipment which must function following the earthquake includes the piping between the control rod drive mechanism and the SCRAM dump tank and the air-operated SCRAM discharge valves (CVNC10) including the solenoid valves from the instrument air header (SVNC27). The hydraulic piping need not remain intact following the earthquake but must not fail such that flow from the control rod drive mechanism to the SCRAM dump tank is prohibited (such as by crimping of the piping). With the reactor at pressure (>450 psig), the reactor pressure alone is sufficient to permit control rod insertion without the aid of the SCRAM piping to the control rod drive mechanisms or the nitrogen-filled accumulators.

Success of this system requires the tripping or loss of power to the RPS terminating power to the SCRAM solenoid valves causing them to close venting the air from the SCRAM valves. The SCRAM valves fail open on loss of air venting the top of the control rod drive piston to the SCRAM dump tank forcing the control rod into the core due to the resulting large differential pressure across the control rod drive piston. Normal core geometry must be maintained by reactor internals to permit insertion of the control rods.

Success of this system has no dependency on the availability of any ac or de power supplies.

NUO383-2628A-BQ01-NLO4

23 TABLE 7.11-1 BIG ROCK POINT PLANT IMPORTANT ACTIVE COMPONENTS FOLLOWING A SEISMIC EVENT Additional Instrumentation Major Active Instrumentation and Eqtipment Useful in Manual Components Required to Activate Component Activation of Components Electric Fire PS615 (Fire Pump Discharge LIRE 19A & B Pump Pressure) (Drus Level Indication)

LT3184 A-3187 (RDS Drum Level) LT-3188 RDS Actuation Cabinet 6.1 (ASDB Drum Level /

Fire Pump Control Panel C17 Indication) /

UPS Batteries Diesel Fire PS612 (Fire Pump Discharge LIRE 19A & B Pump Pressure) (Drum Level Indication)

LT3184-3187 (RDS Drum Level)

RDS Actuation Cabinet 5.1 Fire Pump control Panel UPS Battertes Core Spray Pumps 1A and 2A Buses LT3171 & 3175 2B Bus (Reactor Building Level) l l Control Rod 1A and 2A Buses LT3180-3187 l Drive Pumps 2B Bus or ASDB & LIRE 19A & B l Disc-1441 (Drum Level /

Instrumentation) /

l Emergency Diesel Control Panel C18 Cenerator EDC Batteries UPS A Batteries Undervolt.ge Relay Standby Diesel Standby EDC Batteries Generator Standby EDC Transformers I

i l

l NUO383-2628A-BQ01-NLO4 L .

4 24 T6BLE III-1 (Continued)

Additional Instrumentation Major Activi' Instrumentation and Equipment Useful in Itanual

, g Componente Required to Activate Component Activation of Cogognte, s l

l' SV4984-4987 RDS Actuation & Sensor LSRE09 A-D

'E (Depressurization Cabinets UPS Batteries (Reactor Level) 1 Valves) LT3184-3183 (Drum Level)

}A LT3180-3187 (Reactor Level) g PS789-796 (Fire Pump Pressure) 2 Minute Timers CV4180-6183 SY4980-4913

-(Isolation Actuation & Sensor Cabinets Valves UPS Batteries L73184-318% (Drum Lovel)

LT3160-3187 (Reactor Level)

PS789-796 (Fire Pump Pressure) 2 Minute Timers CVNC10 SYNC 27A & 8 (SCRAM Val,es) Reactor Protection Channels A & B Reactor Protection Sensors:

PS664-667 (Enclosure Pressure)

LSRE09A-D (Reactor Level) 1,$RE06A-D (Drum level)

PSRE07A-D (Reactor Pressure)

PS654-657 (Condenser Pressure)

Y1ux Monitets:

- RH01A-B (UV Contacts)

M07070 and 7071 LSRE098. D F&M (Reactor Level)

(Core Spray) PSIC113. D. T&H (Reactor Pressure)

Contact Relay for H07070 and 7071 Bus 25 H07051 and 7061 LSRE09A, C. EEG (Reactor Level)

(Core Sprey) PSIC11A, C E&G (Reactor Level) de Bus D01 Station Batteries hT0333-2628A-BQ01-NLO4

25

, TABLE III-l (Continued)

Additional Instrumentation Major Active Instrumentation and Equipment Useful in Manual Components Required to Actuate Cemponents _ Actuation of Components

'M07064

.(Enclosure Spray PS636A PS7064A & B& B )(Enclosure Pressure) de Buses DIO, D02 and D01 Station Batteries M07068 Conte Relay for M07068 PI367 (Enclosure Spray) *. (Enclosure Pressure)

M07066 Cs .act Reley for M07066 LT3171 and 3175 (Core Spray Htx) 2A Bus (Reactor Building Level) 2B Bus M07080 Contset Relay for M07060 LT3171 and'3175 (M07066 Backup) 2A Bus (Reactor Building Level) 2B Bus M07072 de Buses DIO, D02, and D01 L (Backup Core Spray Statien Batteries Supply)

M07053 and 7063 PSRE07A-D (Reactor Presefre) PSID28E (Emergency Condenser) Bus D12 PIA 49 Alternate Shutdown System (Reacter Pressure) /

Batteries /

SV4947 de Buses D12 Alternate LS3550 /

(Emergener s:ndenser Shutdown System Batteries (Emergency Condenser /

Makeup) Shell Lerel)

PIA 49 (Reactor Pressure)

PT 1m3 (ASD Reactor Pressure) l l

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28 IV. BIG ROCK POINT PLANT RESPONSE TO A SEISMIC FVENT (EVENT TREE DESCRIPTION)

In this section, plant response to two of the three initiating eveats identified in Section II is described in detail. The loss of offsite power and medium steam line break inside containment transient descriptions take the form of event trees. Logic for these trees was presented initially in Appendix I of the PRA. This logic was modified for the purpose of this study to reflect only those systems which were felt to be easiest to show to be resistant to the forces of the earthquake, those systeme are deocribed in Section III. More detail was added to the "loss of offsite power tree" than exists in the PRA for ease of identifying depentencies between systems within the tree and simplifying the procmas of identifying the seismic weak links where they exist in the plant design. For loss of offsite power sequences leading to the use of long-term cooling methods involving the emergency condenser, a detailed long-term cooling event tree was developed. This I

tree reflects the reed for long-term cooling makeup;from the control rod drive system ar.d reflects the use of the RDS, the core spray system and the post incident cooling system as redandant systems to this makeup source. The following transient descriptions begin with an eveat tree diagram, definitions of event tree headings and are followed by a discussion of each branch point of tha event tree.

The third transient, ATWS, is not develaped in this section. The potential for this transient to occur will be presented in Section V by identifying those equipment failures which must occur to result in SCRAM failure. Discussion of additional evaluations which will ensure that a failure to SCRAM will not occur will then be presented in Section VIII.

NUO383-2628A-BQ01-NLO4

29 Seismically Induced Medium Steam Line Break Insf.de Containment MSLS i E55 *. RDS

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\.E MSt.B- MED\UM STEAM LINE EREAK 7 SS -ENCLOSURE SPRAY SYSTEM Rbs -REACTOR DEPRt.SSUR)2 ATION SYSTEM CS CORE SPR AN SY ST EM PT. 5 POST iMCl~ BENT SYSTEM l

l 1

i FIGURE IV - 1 NL'0383-2628A-BQ01

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30  !

BP 1 Enclosure Spray At Branch Foint 1 (Figure IV-1) a break in a steam line has occurred as a result of an earthquake. Offsite power is assumed to have been lost also as a result of the earthquake. Because the steam superheats as it leases the primary system the conted.nment temperature begins to rise toward the environmental qualification temperature of 235'F. On containment pressure attaining 1.7 psig a signal is sent via PS636A & B and PS7064A & B to open M07064 actuating enclosure sprays. Success at this branch point implies the availability of M07064, the containment pressure switches, the valve's de power supply, the integrity of fire piping in the containment, turbine building, yard and screenhouse, and the operation of either of the two fire pumps. The ac fire pump uill require the operation of the emerge.cy diesel and 2B bus. Automatic starting of the fire pumps on low drum level or low fire header pressure due to enclosure spray actuation will be required. Should M07064 fail to open the operator may manually actuate the Backup Enclosure Spr:y Valve M07068 which is an ac powered valve with dependencies on emergency power.

Failure to deliver water to the enclosure spray header is assumed to result in the exceeding of the environmental qualification temperature resulting in the failure of important RDS or core spray equipment by Sequence LE. Successful enclosure spray leads to RDS operation at Branch Point 2.

BP 2 RDS As the blowdown progresses a low drum. level signal at 17" below drum center line will occur starting the fire pumps and a two minute timer in the RDS logic. Drum level will continue to fall to the low reactor water set point of 2'9" above the core. On reaching this level, coincident fire system pressure

>100 psi, low drum level and two-minute timer closure a signal will be sent to the solenoid valves on the RDS isolation and depressurization valves. Opening the valves rapidly depressurizing the resctor allowing low-pressure core spray. Successful system operation requires three of the four RDS valve trains to operate, two of four sets of drum

• level, reactor level, fire pressure and timer instrumentation to operate, and one of the two fire pumps to run.

Operation of the sensors and valves depends on the UPS, operation of the electric fire, pump depends on emergency ac power. Failure of this system is assumed to result in inadequate depressurization of the reactor and at least limited core damage before a pressure is low enough to allow core spray (sequence LR). Success leads to the need for core spray actuation.

BP 3 Core Spray A low reactor level coincident with rea: tor depressurization to <200 psig leads to actuation of ac and de-powered core spray valves permitting fire system flow through a spray nozzle in the vessel head or a ring sparger around the vessel perimeter. Flow >290 gpm through either of these lines is sufficient to cool an uncovered core. Success of thic ystem depends on opening either the ac or de-powered valves, the integrity of fire piping in the containment, turbine building, yard and screenhouse and the operation of at least one fire pump. The ac valves have a dependency on the operation of the diesel generator and the 2B bus. Failure of the core spray system leads to an inadequately NUO383-2628A-BQ01-NLO4 i

31 cooled core by sequence LC, success ultimately leads to the need for post-incident long-term cooling.

BP 4 Post-Incident System Water will enter the containment by way of primary coolant loss, core spray and enclosure spray. On the addition of ~ 260,000 gallons of water to the containment (387 ft elevation) the operator will place the post-incident system into service and isolate the fire water system water addition to the enclosure by closing hand-operated Valves VFP-29 and 30. Success of this system implies the operation of one core spray pump, the integrity of pump suction piping from the containment sump, discharge piping to the heat exchanger, removal of heat by way of the core spray heat exchanger and the integrity of post-incident system piping back to the core spray system inside containment.

Heat removal through the heat exchanger implies the addition of fire water to the heat exchanger shell through M07066 or parallel valve M07080 integrity of the yard loop or bypass with a fire hose, integrity of screenhouse fire piping and the operation of at least one: fire pump; 'M07066, the core spray pumps and the electric fire pump are assumed to require emergency ac power by way of the 2B bus and the emergency or standby diesel generators. Operation of the post-incident system is assumed to be required between 4 and 24 hout.' following the LOCA. Failure of this system leads to Sequence LLp.

NUO383-2628A-BQ#1-NLO4

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' 33

-EARTHQUAKE GENERATED LOSS OF OFFSITE POWER BP 1 DC Power At Branch Point 1 (Figure IV-2), an earthquake has occurred in which significant ground motion resulted at the Big Rock Point Plant Site. A loss of offsite power has been the consequence of this ground motion. Loss of offsite power may result at Big Rock Point following a seismic event due to a variety of failures which are considered to be "weak-links" with respect to the availability of offsite power during such an event. These "weak-links" include the 2400 V voltage regulator in the switchyard, ceramic insulators in the switchyard or at offsite substations, or various masonry walls to which the 2400 V cable to the Station Power Distribution System is mounted. At Branch Point 1 the success or failure of the 125 V de Distribution System is determined. Failure at this branch point implies that either the station batteries or the 125 V de MCC D01 has failed as a result of the earthquake, leaving the plant without a str. ion de power' source at Branch Point 45. The L functioning of. station de power leads to. Branch Point 2. .There was not a DC /

power fault tree developed, rather the DC power dependencies were modeled into /

the individual trees. /

BP 2 MSIV The first automatic action required will be isolation of the Primary System by /

closure of the MSIV. Successful MSIV actuation implies success of the Reactor /

Protection System in generating a containment isolation signal. Failure to /

close the MSIV is conservatively assumed to result in a blowdown outside the /

containment either due to a failure of the backup valves to the MSIV to close /

or due to the failure of steam piping in the pipe tunnel. The MSIV has it's /

de power source housed in the alternate shutdown building (ASDB). Successful /

MSIV closure requires the availablity of the de batteries in the ASDB. These /

failures lead to Branch Point 35. Successful MSIV closure leads to Branch /

Point 3. /

BP 3 Emergency Condenser Valves A loss of offsite power has occurred as a result of an earthquake and the Primary System has isolated by automatic clo.ute of the MSIV. Also generated as a part of station power loss was a signal to open the outlet valves to the emergency condenser (M07053 and M07063). This signal was generated by PSRE07 A-D which are ac-powered switches that fail closed on loss of voltage. Had station power been available, high reactor pressure (1450 psia) would have resulted in the closure of these switches. A manual demand to open the de emergency condenser valves can also occur if the operator is aware of the rising reactor pressure due to decay heat generation. Indication of reactor pressure is available to the operator in the Control Room via PIIA07 which is dependent on emergency ac power operation. The operator has six minutes from the time of isolation of the Primary System until the first safety relief valve lifts (1550 psia) in which to manually actuate the emergency condenser should automatic actuation fail. The emergency condenser valvcs alco have /

their power source housed in the ASDB. Successful val'e operation requires /

the de batteries available. Successful automatic or manual initiation of /

either of the emergency condenser outlet valves leads to Branch Point 4.

Failure of both of the valves to open leads to Branch Point 25.

NUO383-2628A-BQ01-NLO4

34 BP 4 UPS Following isolation of the Primary System, this event tree contains a heading for the UPS power supplies and actuation /senscr cabinets. The UPS will have an impact on several of the systems which follow in this tree, such as operation of the RDS, automatic loading of the emer:gency generator onto the 2B bus and the functioning of various instrumentation which the operator may require for manual initiation of these or other functions. Success of the UPS in surviving this earthquake leads to the sequences following Branch Pof.nt 5. Failure of the UPS leads to Branch Point 18.

BP 5 Emergency AC An earthquake has occurred, offsite power is unavailable, the Primary System is isolated and the emergency condenser is in service removing decay heat and cooling down the Primary System. Little, if any, dependency on ac power has been required up to this point in the tree. For the next four hours the plant will remain isolated with little loss.of-primary coolant inventory. By the end of this four-hour period, depletion of the water in the shell of the emergency condenser will have occurred, reactor pressure will have risen back to above normal operating conditions and a safety relief valve will begin to lift (1550 psia) limiting Primary System pressure unless action by the operator is taken to provide a source of makeup to the emergency condenser. Several of the systems following this branch point do have a significant dependence on ac power for success. At this branch point, therefore, the functioning of an emergency ac power source is required. Successful operation of the MSIV and emergency condenser valves, a significant amount of time to ensure that emergency ac power operation is available and successful actuation of either the emergency generator or the standby generator in loading the 2B bus will lead t~ ehe Branch Point 6. Failure of the 2B bus or both of the diesel genera: m s will lead to a relatively lengthy station blackout described at Branch roint 12.

BP 6 Fire Pumps The success or failure of the fire pumps is determined at this branch point.

The plant is cooling down on the emergency condenser and emergency ac power is l available. Success at this branch point assumes that the Fire System piping l in the screenhouse is intact and that either of the two fire pumps, diesel or electric-powered, is running. A loss of the piping integrity or failure of both fire pumps within the first fcur hours of the transient results in lif ting of a steam drum safety relief valve due to the inability to makeup to the emergency condenser shell from the Fire System. Gradual depletion of the Primary Coolant System inventory occurs for the next two hours at which time sufficient water has left the Primary System through relief valve actuation to begin to uncover the core. If the fire pumps are not recovered within this additional two-hour perfud it is assumed a core damage situation ensues via Sequence PC due to the inability to provide adequate core spray. Maintenance of the screenhouse piping integrity and successful operation of either fire pump-will lead to Branch Point 7.

NUO383-2628A-BQ01-NLO4

35 BP 7 Emergency Condenser Makeup Juccessful isolation of the reactor, operation of the emergency condenser and availability of a fire pump will allow use of the emergency condenser to cool the reactor to near shutdown conditions and maintain those conditions in the long term at Branch Point 8. Success of emergency condenser makeup implies the manual operation of de powered SV4947 allowing fire water flow from the Core Spray fystem to the emergency condenser shell. As this valve is operated manually, ',ndication that makeup to the shell is required should be available to the operator. This indication is assumed to be provided by an annunciat4~

of shell low-level via LS3550 or by an indication of rising primary system pressure on PIIA07 both being in the Control Room. The valve is powered from /

the de batteries housed in the ASDB. Successful valve operation implies that /

the power source is available. /

Emergency condenser makeup also relies on the integrity of core spray piping inside the containment,- the yard piping and piping between the yard and 4

- containment through the Turbine Building.or the Post-Incident System. Should Turbine Building piping be unavailable as a result of the earthquake, it must be isolated from the yard loop and containment piping to prevent diversion of fire water and a flow path to containment must be established through de-operated M07072. Credit for this operator action is taken in this tree given the amount of time available to perform these operations (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />).

Failure of the emergency condenser makeup by way of failure of the makeup valve, Control Room shell level instrumentation, containment yard or unisolatable Turbine Building piping failures, or failure to establish a flow path to containment through the Post-Incident System given a failure of Turbine Building piping is assumed to lead to Branch Point 9.

BP 8 Long-Term Cooling Cooldown of the isolated reactor on the emergency condenser has been successful' including makeup to the shell. 'The reactor is most likely at or near atmospheric conditions. Reactor cooling can continue indefinitely in this manner with the aid of makeup systems such as the control rod drive pumps or core sprays to accommodate any minor leakage which naturally occurs from the Primary Coolant System. Success or failure to cool the core at this branch point using these systems is developed in detail in the long-term cooling event tree which follows this discussion.

BP 9 RDS

, Primary System isolation and emergency condenser actuation have been successful and fire pumps are available, but emergency condenser makeup has failed resulting in the repressurization of the reactor on decay heat to the safety relief valve set point. Gradual depletion of the reactor inventory is occurring due to periodic relief valve operation. No credit for makeup to the reactor by control rod drive pump operation is taken dus

+ to uncertainty in its ability to operate in the steam environment created by relief valve actuation.

NUO383-2628A-BQ01-NLO4

36 is i Within six hours of the transient initiation, reactor inventory has reached the RDS actuation set point (2'9" above the core). Successful operation of the RDS at this stage of the transient leads to the.need for core spray at Branch Point 10. Failure of RDS implies failure of isolation (CV4180-CV4183) or depressurization (SV4980-SV4983) valves to open or failure of an automatic or manual actuation signal. Automatic actuation of the RDS is dependent on drum (LT3184-LT3187) and reactor level (LT3180-LT3183) instrumentation, fire pump pressure sensors (PS789-PS793) and RDS timers. Manual actuation is assumed to depend on the availability of reactor level indication to the operator. In addition to the RDS level, instrumentation control room annunciation of low reactor level is provided via Reactor Protection Switches LSRE09 A-D. Manual operation of the RDS from the Control Room also has dependency on ac power distribution equipment via Panel lY.

Failure of RDS at this branch point is assumed to lead to a gradual uncovering of the core at pressure by Sequence PEmR.

BP 10 Core Spray Depletion of primary coolant inventory to the point of RDS actuation requires core spray initiation. Successful core spray implies actuation of either of the sets of core spray valves (de-actuated M07051, M07061 or ac-actuated M07070 or M07071). Sensors required to actuate these valves include reactor level and Pressure Switches LSRE09 A-H and PSIG11 A-H. Credit is taken for manual initiation of important core spray functions by the starting of a fire pump but only if appropriate drum level instrumentation is available. Manual actuation of the core spray valves can also be accomplished from the control room although credit for this action is not included as a part of the logic of these trees.

As was the emergency condenser makeup, core spray is dependent on the integrity of core spray piping inside containment, yard piping and a path from the yard loop to the containment either through the Turbine Building or the Post-Incident i

System. The Turbine Building path is that path normally valved in for operation, but should rupture of this piping occur as a result of the earthquake, sufficient time is available to establish the path through the Post-Incident System by opening M07072 (more than six hours). Again, as with the emergency condenser makeup supply, core spray flow is considered adequate only if any ruptured fire piping which does exist in the Turbine Building is isolated from the yard loop and containment.

Success of this equipment leads to the need for a long-term cooling heat sink at Branch Point 11. Failure is assumed to lead to temporarily uncovered inadequately cooled core by way of Sequence PEmC.

BP 11 Post-Incident System j

RDS and core spray actuation having been successful, the containment will begin to fill with water coming from the Primary System, core and enclosure

, sprays. Reactor and steam drum level will recover quickly following core spray actuation and water will flow from the drum to containment through the NbO383-2628A-BQ01-NLO4 I

37 RDS valves. The operator may control flow to the Primary System from outside the containment by regulating flow with Valves VFP-29 and VFP-30 or M07072, thereby limiting the amount of water which enters containment. This tree assumes, however, that a containment water elevation between 587 feet and 590 feet is reached ultimately and the need for post-incident recycle and cooling is required.

Success of the Post-Incident System requires that at least one o' two core spray pumps be started pumping water from the bottom of containment, through the core spray heat exchanger to core spray piping inside the containment.

Failure of post-incident piping, core spray piping inside the containment, both of the core spray pumps or their station power buses or Reactor Building level instrumentation implies f ailure of the system. As assumed for the Core Spray System, diversion of water to rupture of piping in the Turbine Building will also fail the system unless this piping is isolated from the containment.

Heat removal by way of the core spray heat exchanger is required for success of the system. This heat removal is established by the opening of M07066 remotely from the Control Room or locally by hand. An ac motor operated valve, M07080, in parallel with M07066 can also be opened to supply Fire System water to the shell of the core spray heat exchanger. Success of this system implies the integrity of Post-Incident System piping to and from the heat exchanger, the heat exchanger shell itself and yard piping. Yard piping failures which occur may be bypassed with the inetallation of a fire hose from the hose manifold in the screenhouse to Hand-Operated Valve VPI-10 also in parallel to M07066.

Failure of this system to supply sufficient flow to the reactor to makeup for decay heat losses implies eventual core uncovery and inadequate core cooling by way of Sequence PEmLp.

BP 12 Fire Pumps Without AC Power A seismic-generated loss of offsite power has occurred, the Primary System has isclated, an emergency condenser cooling path has been established; station de power and the UPS are available but emergency ac power is unavailable due to loss of the 2B bus or both of the diesel generators. The Fire System is required to be functional to provide makeup to the emergency condenser and core spray if necessary. The success and failure of this system are identical to that described for the Fire System in Branch Point 6 except that only the diesel fire pump is available due to the failure of onsite ac power. In addition, the ability to start a fire pump manually from the RDS panel in the Control Room is no longer available due to this circuit's ac dependencies.

Failure of the diesel pump or screenhouse fire piping under these conditions is assumed to lead to core damage Sequence PQF. Time frame for reaching this state of inadequate core cooling is the same whether or not ac power is available (approximately six hours). Successful fire pump operation allows emeryency condenser makeup at Branch Point 13.

NUO383-2628A-BQ01-NLO4

38 BP 13 Emergency Condenser Makeup Without AC Power Primary power dependency of emergency condenser takeup is on the ASDB dc /

power. This dependency is a result of DC-Powered Makeup Valve SV4947. There /

is a minor ac power dependency at this branch point in that Control Room /

Primary System Pressure Indicator PIIA07 requires an ac power supply. /

The description of this branch point is therefore identical to the discussion presented in Branch Point 7 except that continuous reactor pressure instrumentation used as backup verification that emergency condenser makeup is occurring is not available. Failure of makeup leads to the need for RDS and core spray at Branch Point 15; successful operation leads to Long-Term Cooling Systems at Branch Point 14.

BP 14 Long-Term Cooling As indicated at Branch Point 8, detailed long-term cooling event tree has been developed and is presented-following discussion of this loss of offsite power tree. This tree includes a discussion of long-term ac power dependencies.

BP 15 RDS Without AC Power The success or failure of RDS given primary coolant level has reached 2'9" above the core is identical to that discussed in Branch Point 9 except that the ability to manually actuate a fire pump or manually depressurize the reactor from the Control Room is no longer available due to these circuits' dependencies on ac power. (RDS panel control switch and annunciation power is from Panel 1Y.) Failure of automatic RDS actuation leads to eventual inadequate cooling of the core at elevated pressure in a time frame similar to that which would occur with ac power available (; six hours) by way of sequence PQR.

Successful RDS operation leads to the need for core spray at Branch Point 16.

.BP 16 Core Spray Without AC Power As a result of the unavailability of any ac power source, the ac-dependent motor-operated valves (M07070 and M07071) of the Core Spray System will not be available to provide core cooling after RDS actuation. The de-powered valves (M07051 and M07061) will be available however as station de power is assumed to be unaffected. Therefore, with the exception of the ac valves and the te-powered Control Room instrumentation and actuation circuitry for manually starting a fire pump from the Control Room, the Core Spray System will function or fail as described in Branch Point 10. Failure of the Core Spray System will lead to an uncovered core at Sequence PQC, successful core spray will lead to Branch Point 17 and the need for long-term cooling.

BP 17 Post-Incident System The success and f ailure of the Post-Incident System in long-term cooling at this branch point are identical to those described for Branch Point 11 even though there are ac power dependencies in the operation of the core spray pumps and M07066. Recall from the discussion of this branch point that the path by which water fills the containment is through the RDS valves (well N1'0383-2628A-BQ01-NI.04

39 above the reactor core). The operator has the ability to monitor containment level (LT3171 and 3175) and reactor and drum level (LT3180-LT3183 and LT3184-LT3187) and regulate the flow to the Reactor Building from outside containment without ac power availability. It is assumed at this branch point that rather than terminating core spray makeup to the reactor entirely, allowint the core to become uncovered, the operator will regulate core spray flow allowing makeup to cover losses due to decay heat generation until an emergency ac power source for the core spray pumps can be recovered or offsite power is restored.

This method of core spray flow regulation can occur for days which is sufficient time to allow restoration or repair of ac power.

Failure of post-incident cooling equipment described at Branch Point 11 after emergency ac power failure then leads to an inadequately cooled core by Sequence PQLp.

BP 18 Emergency AC Without UPS An earthquake has occurred which has resulted in a loss of offsite power. DC power remained available, the MSIV closed and the emergency condenser outlet valves opened providing a heat sir.k for the Primary Coolant System.

It is assumed that the UPS have not survived, however, which has implications with respect to the manner in which ensrgency power is provided to the 2B bus, RDS actuation and Control Room indication of reactor and drum level.

At this branch point the success and failure of energency power $s discussed.

Automatic isolation of the 2B bus and loading of the diese) generator cannot be accomplished because of the switchgear dependency on UPS and de power.

However, given that the Primary System is isolated and the emergency condenser is in service, four hours minimum exists before safety relief valve actuation can o: cur and six hours minimum until the RDS actuation set points are satisfied.

Given these time frames, manual operation of the 2B bus electrical switchgear is extremely likely and the importance of the emergency power dependencies on the UPS is very small.

Success at this branch point implies the operation of either the standby or emergency diesel generators and the loading of this power to important equipment on 2B bus. Successful emergency ac power leads to Branch Point 19. Failure of emergency ac power leads to Branch Point 22.

BP 19 Fire Pumps Without UPS There is no dependency by the fire pumps on the UPS with the exception of the automatic starting of the pumps on low drum level as measured by RDS drum level instrumentation (LT3184-LT3187). Similar to the emergency gene rator switchgear this dependency is not important at this branch point givec the time frame in which makeup to the emergency condcaser (four hours folleving transient initiation) or core spray (six hours) is required. The success or failure of the fire pumps at this branch point is the same as that presented in the discussion for Branch Point 6. Successful fire pump actuation leads to Branch Point 20 eeergency condenser makeup. Failure of the fire pumps at this NUO383-2628A-BQ01-NLO4 l

f 40 branch point leads to eventual core damage Sequence PUF due to the failure of the makeup source to the emergency condenser and ultimately core spray failure.

BP 20 Emersency Condenser Makeup Without UPS.

The description of success or failure at this branch point is identical to that presented for Branch Point 7 as there is no dependency.on UPS by the fire water makeup to the emergency condenser. Success of this system leads to long-term cooling Braoch Point 21.

However, unlike Branch Point 7, failure of makeup to the emergency condenser by itself leads to inad.quate core cooling by way of Sequence PUEm. After having cooled the P.4 mary System adequately for four hours, the emergency condenser has ceased removal of heat from the reactor due to depletion of shell side wr.er. The reactor has repressurized and a safety relief valve is limiting p. essure at 1550 psia. Inventory depletion of the Primary Coolant System ir. occurring because no makeup water is available (like Branch Point 9,

n. < ~ ' no cont'colirod drive makeup.istassumed because of the steam environment inside containment). Approximately six hours following the earthquake and loss of power, the reactor water level reaches the RDS actuation set point (2'9" above the core). As no power supply is available for energizing the solenoid-operated RDS isolation and depressurization valves, the reactor remains at pressure and the core slowly becomes uncovered without the ability to provide low-pressure core spray injection.

BP 21 Long-Term Coolina ,

Similar to Branch Points 8 and 14, long-term cooling given the success of the emergency condenser is developed in a subsequent event tree.

BP 22 Fire Pumps With UPS and Emergency AC Failure For the same reasons stated in Branch Point 19," the success or failure of the fire pumps has limited dependence on the UPS. However, ac power availability does have a significant influence on this system. Therefore, the description of this branch point is essentially identical to that presented in Branch Point 12 with the failure of the diesel fire pump leading to core damage Sequence PUQF. Successful diesel pump operation allows emergency condenser makeup at Branch Point 23.

BP 23 Emergency Condenser Makeup Without UPS and Emergency AC There is no dependency of emergency condenser makeup cn UPS and only a minor dependency on emergency ac. This dependency was described in Branch Point 13 and involves the inability of the operator to use PIIA07 (primary coolant) as backup verification of the need for adding makeup to the emergency condenser shell. LS3550 (sh 11 level) is assumed to be available, however, as de power is still functional. Successful makeup addition to the emergency condenser shell at this branch point is still possible, therefore, and leads to the need

-for-long-term cooling systers at Branch Point 24. Failure of emergency condenser makeup piping, Makeup Valve SV4947 or shell level indication is NUO383-2628A-BQ01-NLO4

41 l

l l

aasumed to lead to core damage Sequence PUQEm for the same reasons presented in the description of Branch Point 20.

BP 24 Long-Term Cooling Development of a long-term cooling event tree follows the description of this event tree.

BP 25 Through 34 Emergency Condenser Valve Failure At this branch point a loss of offsite power resulting from a seismic event has occurred, de power is available and the MSIV has closed isolating the Primary System. The assumption is made at this branch point that neither of the emergency condenser outlet valves (M07053 and M07063) opened leaving the Primary Coolant System without a heat sink. Reactor pressure has risen to the set point of the first safety relief valve (1550 psia) and Primary System pressure is being limited in this manner. Assuming the reactor was operating at full power at the time of the transient', this reactor state will be reached within six minutes from initiation of the transient.

There are no dependencies on emergency condenser valve operation within the UPS, emergency ac power, fire pumps, RDS, core spray or Post-Incident System with the exception of de power (which is assumed to be available during these sequences). Emergency condenser makeup obviously will not be required following failure of the emergency condenser valves, hence no branches between Branch Points 27 and 28 or 31 and 32. There is a minor dependency of emergency condenser valve actuation on ac power in that PIIA07 can be an unambiguous indication that the emergency condenser valves need to be opened given that i they have failed automatically. This pressure indicator relies on automatic starting of the emergency diesel generator and operation of switchgear connecting it to the 28 bus. Given the limited time frame in which this instrument is being used no credit for manual actuation of the emergency diesel or standby diesel generators is taken in manual operation of these valves.

Given these limited dependencies, it can be seen that the reactor states and ,

sequence descriptions resulting from transient sequences leading to inadequate t core cooling as a result of emergency condenser makeup failure are identical to those which occur following emergency condenser valve failure, the only difference being the time required to reach the final reactor state.

NUO383-2628A-BQ01-NLO4 c

[

42 Emergency Condcaser Makeup Failure Sequences Corresponding Emernency Condenser Valve Failure Sequences Time of SRV Tine of Core Time of SRV Time of Core -

Sequence Actuation Uncovery BP Sequence Actuation Uncovery BP, PEmLp 4h Days 11 PEvLp 6m Days 30 PQEmLp 4 ~a Days 17 PEvQLp 6m Days 34 PEmC 4h 6h 10 PEvC 6a 2h 29 PQEmC 4h 6h 16 PEvQC 6a 2h 33 PEmR 4h 6h 9 PEvR 6a 2h 28 PQEmR 4h 6h 15 PEvQR 6m 2h 32 PF 4h 6h 6 PEvF 6m 2h 27 PQF 4h 6h 12 PEvQF 6m 2h 31 PUF 4h 6h 19 g

No Corresponding --- --- --

PUQF 4h 6h 22) Sequences --- --- --

PUEm 4h 6h 20 PEvU 6a 2h 25 4h 6h 23  !

PUQEm i

-i i

a 1

NUO383-2628A-BQ01-NLO4 4

43 BP 35 Primary Coolant System Isolation Failure On loss of offsita power a signal is generated by the Reactor Protection System to close all automatically-actuated containment isolation valves. At this point on the event tree the ASDB de power supply is assumed to have survived the seismic event but the MSIV has failed to close. No assumptions are being made with respect to the ability of backup valves to the HSIV to close, or if they do close, to prevent leakage from the unisolated main steam line to the main condenser, or of the integrity of the main steam line piping or its branch connections to remain intact following ground motion. It is conservatively assumed, therefore, that failure to close the MSIV following a loss of offsite power leads to a blowdown of the reactor to the Turbine Building or pipe tunnel. If large enough, such a blowdown could occur over the course of several minutes. The energy loss of such a blowdown is more than that added to the Primary System as a result of decay heat eliminating a demand on the emergency condenser valves to open or the need to makeup to the emergeacy condenser shell.

At this branch point then the success or failure of UPS is discussed. Failure of UPS implies the failure of the batteries, power supply or actuation / sensor cabinets. Failure of more than one UPS is sufficient to disable the RDS leading to an uncovered core with an insufficient blowdown to allow timely core spray initiation. If the blowdown continues outside containment or one or more RDS trains actuates properly, core spray injection can occur sventually although not before limited core damage has occurred. Failure of UPS leads to inadequate core cooling via Sequence PIU. Successful operation of at least three of the four UPS leads to Branch Point 36.

BP 36 Emergency AC Power Following MSIV Failure As the blowdown occurs due to failure of the MSIV to close, the emergency diesel generator will attempt to start and energize the 2B bus to provide puver for important equipment such as the-electric. fire pump and ac core spray valves. Failure of the diesel generator to start or load the 2B bus leads to Branch Point 41; successful energization of the emergency bus to Branch Point

37. Due to the potentially limited duration of this transient, no credit is taken for manual operation of the emergency or standby diesel generators.

BP 37 Fire Pumps Af ter MSIV Failure On attaining a low drum level (<17" below center line) a signal will be sent to start both fire pumps. Success at this branch point implies the starting of at least one of the two fire pumps and the maintaining of screenhouse fire piping integrity. This will allow fire pressure permissive (>100 psig) to the RDS actuation logic and provides a source of fire water for core spray.

Success leads to Branch Point 38 while failure leads to a lack of adequate core cooling after RDS and core spray actuation fails, Sequence PIF. No credit for operator action to start the fire pumps is implied, again due to the potentially limited time frame over which this transient occurs.

NUO383-2628A-BQO1-NLO4

44 BP 38 RDS With MSIV Failure As the blowdown outside containment continues, low drum level will initiate a two-minute timer and a low drum level signal to the RDS actuation logic.

l Further blowdown will result in reaching the low reactor level set point (2'9" above the core). When all four signals occur simultaneously in two of four RDS sensor trains, low drum level, low reactor level, high Fire System pressure and time out of the two-minute timer, the RDS solenoid valves will be energized opening the isolation valve and depressurization valve in each train. Failure of blowdown in more than one train is assumed to lead to an unsatisfactory depressurization of the reactor and limited core damage by Sequence PIR. Successful blowdown through three of the four RDS trains leads to the need for core spray at Branch Point 39. Again, no credit for operator action is taken due to the short duration of the transient, failure of automatically-operated components is assumed to lead to system failure.

BP 39 Core Spray After MSIV Failute The reactor has depressurized by RDS actuation after blowdown through the main steam line. The fire pumps and emergency power are successfully operating.

At low reactor water. level a signal was sent to the core spray valves, and on reaching 200 psig reactor pressure the ac- and de-operated valves receive a signal to open. Success of the Core Spray System at this branch point implies opening ot eithe- of the pairs of ac or de valves in the core spray lines maintaining the integrity of the core spray piping in the containment, in the Turbine' Building and the underground yard piping. No credit is given for isolation of the Turbine Building piping should it rupture as a result of the earthquake or establishing a backup fire water path to containment through the Post-Iacident System, again due te the limited time frame over which the transient occurs. Successful core spray leads to long-term cooling at Branch Point 40, failure leads to degraded core cooling Sequence PIC.

o .BP.40 Post-Incident System With MSIV Failure Although a result of water loss was effectively a loss of coolant accident outside containment, the Long-Term Cooling System description at this branch point is identical to that at Branch Point 10. Following a tranrient of much longer duration, one might argue that the main steam line is open outside containment and water overflow from the Primary System preferentially will pass through this line rather than through the RDS valves which are located at a higher elevation. This argument could lead to the potentially unconservative assumption that the Post-Incident System will not be needed for long-term cooling following a reactor blowdown outside containment. The assumption will be made, therefore, that repairs to main steam line components are made following the transient isolating the Primary System and ultimately requiring use of the Post-Incident System. F.ilure of this system is assumed to gradually lead to inadequata cooling by Sequence PILp.

BP 41 Through 43 A failure of the MSIV to close has resulted in a blowdown of the reactor to the Turbine Building. At this branch peint, de and UPS power are assumed to NUO383-2628A-BQ01-NLO4

45 cvailable but emergency ac power is not available duc either to a failure the emergency diesel generator or withia the 2B bus. The description of a:ch Points 41 through 43 is identical to Branch Points 37 through 39 with e'cxception that the electric fire pump and ac-powered core spray valves are t functional due to the loss of emergency power. Failure of the diesel fire mp er one of the de core spray valves by themselves will lead to Sequences QF cad PIQC. Sequence PIQR occurs just as described in Branch Point 38.

cc;ssful diesel fire pump, RDS and core spray actuation lead to long-term oling Branch Point 44.

>n 44 Post-Incident System e d2scription presented in Branch Point 11 is applicable to Post-Incident stem operation at this branch. Again the assumption is made that the st-Incident System is required, perhaps, as a result of a repair of the main saa line. The standby diesel generator is now considered available for tsntial use as backup to the emergency generator. Even if both diesels b ilcd 'as .a result of-the > earthquake, ' operator actiontlimiting the amount of tar cdded to containment will postpone the use of the Post-Incident System e es much as several days until repair or restoration of an ac power source r th2 core spray pumps is made available.

45 UPS With DC Power Failure o carthquake has rest'ted in a loss of offsite power. The assumption is da et this branch point that the station de power supply (station batteries ASDB batteries) also has failed.

2 powar failure automatically results in the inability to isolate the Primary sten by closure of the MSIV. As was assumed at Branch Point 35, a blowdown tha reactor to the Turbine Building results. A failure of the UPS has two portcnt consequences at this branch point. Failure of UPS A disables the ility.of automatic. loading of the diesel generator onto the 2B bus leaving powar source to the ac-powered core spray valves should they be needed. A pid blowdown outside containment will result in core uncovery without core ray es the power supply to the de core spray is also assumed to be disabled.

11ura of more than one of the UPS will result in inadequate reactor prGssurization perhaps leading to limited core damage as discussed in Branch int 38. Success of three of the four UPS, at least one being UPS A, leads Brcnch Point 46. Failure of UPS A or any two UPS is assumed to lead to

! grcdsd core cooling by Sequence PDU.

46 Emergency AC With DC Power Failure ilura of the emergency diesel to start or failure of the 2B bus to energize scbics the power supply to the ac core spray valves. As stated previously, ould a rapid reactor depressurization occur, no credit for manual action in crting the standby diesel can be taken and core spray will be disabled by

. y of Sequence PDQ. Successful energization of the 2B bus leads to Branch int 47.

'0383-2628A-BQ01-NLO4 L_

((: 47 r:s LONG TERM COOLING EVENT TREE EM UPS EAC. CRb Rb5 C5 Lp

/

/

3 b I s - Pilp- LTC l j y pic - LTc2. /

pV R - LTc.3 /

1 /

/

9 g POL, LTCH /

j 9 Pcg. - LTC5 /

PQA LTC4 f

/

/

11 /

/

/

to Pl)V - 1.TC EUY /

/

/

I p0 4 - L T C.\ 0 ,

EM -EMERCENCY CONDENSER MAKEUP UPS -UNINTERUPTABLE POWER SUPPLY EMER AC -EMERCENCY AC POWER CRD -CONTROL ROD DRIVE MAKEUP RDS -REACTOR DEPRESSURIZATION SYSTEM CS -CORE SPRAY SYSTEM PIS -POST INCIDENT SYSTEM FIGURE IV - 3

48 Long-Term Cooling Given Emergency Condenser Operation Following a Loss of Offsite Power The previous tree described a loss of offsite power transient generated by a seismic event. Four branches of that tree lead to a need for long-term cooling assuming the Primary Coolant System was isolated, the emergency condenser was in service and emergency condenser makeup from the Fire System is successful. An assumption is made that offsite power may not be available for several days leading to the need to makeup to the Primary System allowing a recovery of normal level overcoming shrinkage due to the cooldown and normal Primary Coolant System leakage.

Shrinkage by itself is not sufficient to result in uncovering the core.

Combined with normal Primary System leakage from packing, seals, etc, over several days there exists the potential for attaining very low Primary Coolaat System levels, however. For this reason an additional system, Control Rod Drive Makeup, is required for use during long-term cooling in addition to the

! emergency condenser. Failure of the CRD makeup, like the emergency condenser, will be assumed to lead to water levels below the low reactos level set point and a need for RDS, core spray and post-incident recycle. Failure of this makeup supply is not expected to result in these low levels for over a day allowing a relatively long time for the operator to establish this makeup supply. In fact, if the Primary Coolant System pressure is maintaired below 150 psi by the emergency condenser, the Core Spray System can be used to provide the required makeup. As the CRD System is the preferred source of makeup and as it can be actuated even with the reactor at elevated pressures, the tree which follows was developed conservatively ignoring the potential Core Spray System option.* The core spray can be added if needed by inserting an additional heading following CRD makeup in the long-term cooling tree.

BP 1 UPS Long-term cooling has been established for the Primary System through use of the emergency condenser. This tree (Figure IV-3) containo a branch point for UPS, similar to the loss of offsite power tree to account for its effect on various important system functions such as providing Control Room indication for reactor drum level and actuation of the RDS if necessary. UPS failuro leads to Branch Point 10 whereas success results in Branch Point 2.

BP 2 Emesgency AC Success of emergency ac implies cperation of the energency diesel or standby diesel generator and energization of the 2B bus. This system is necessary to provide a power source for the CRD pumps and is also important in providing power for the electric fire pump, ac core spray valves and various Control Room indications. Emergency ac failure leads to Branch Point 7; successful energization of the 2B bus leads to the operation of the CRD pumps at Branch Point 3.

• It is emphasized that control rod drive makeup is a backup to core spray.

NUO383-2628A-BQ01-NLO4

l l

l 49 BP 3 CRD Pumps Successful operation of this system implies that one of the two CRD pumps be operating from the emergency power source, that the condensate storage tank be intact, CV4090 be open, CRD suction and discharge piping be intact and a path to the reactor (most likely through the scram header) be available. Failure of this system is assumed to lead to the need for RDS, core spray and post-incident cooling around a day following the loss of offsite power, Branch Point 4. If UPS is unavailable as in Branch point 10, loss of CRD assumas /

core damage as UPS is not available for RDS. Periodic operation of the system /

implies successful long-term cooling. /

BP 4 Through 6 RDS, core spray and post-incident coolang are required in these branches.

Description of these branch points is identical to that presented for Branch

' Pointe 9schrough 11 in the loss of offsite power tree except that rather than emergency condenser makeup, CRD makeup failure has occurred. Failure of these systems leado to inadequate core cooling via Sequences PYR, PYC and PYLp.

BP 7 Through 9 Loss.of emergency power source automatically implies the inability to makeup with the CRD pumps and the need for RDS, core spray and post-incident cooling.

The description of these branches is identical to Branch Points 15 through 17 of the loss of offatte power tree except rather than emergency condenser makeup, CRD makeup failure has occurred. Failure of these syatema is assumed to lead to inadequate core cooling by Sequences PQR, PQC and PQLp.

BP 10 Emergency AC With UPS Failure Given the length of time associated with this branch of the transient (on the order of a day), there is little dependency of ac power on the UPS. This branch point is therefore identical to Branch Point 2. Success of ac power leads to CRD makeup at Branch Point 11, failure is assumed to lead to core damage. Sequence PUQ fails because of the inability to makeup to the reactor or actuate the RDS.

BP 11 CRD Pumps This branch point is identical to Branch Point 3 except that reactor icvel I

instrumentation powered from the UPS is not available. Failure of CRD makeup is assumed to lead to inadequate cooling by way of Sequence PUY because of the inability to makeup to the reactor with the CRD System or actuate the RDS. The RDS dependancy upon UPS (BP-12, sequence PUYR) is redundant to sequence PYR, as the RDS log!c model contains all the dependencies on UPS.

NUO383-2620A-BQ01-NLO4

50 V. SEISMIC CAPACITY, FAILURE MODES, EFFECTS, REPAIR AND RECOVERY This section contains a table identifying each plant component important to the systems described in Sections III and IV. In addition to the components included as a part of the systems, descriptions, plant structures and equipment not normally considered important to the functioning of the system but which may be made important as a result of ground motion are identified, such as major plant structures and masonry walls. The effects of the failure of each component are assessed as well as the expected failure mode. As the purpose of this study was to rank the importance of these plant omponents in some manner, an assumed capacity for each of these c .;ponents was also determined. A compilation of this information is presented in Table V-1.

The components chosen were those identified as being important to the systems used to perform important plant shutdown functions in the event tree headings of Section IV. The effects of the failure of each of these components were based on the system-success-and-failure-criteria presented in the system descriptions and event tree discussions.

Important passive components whose structural failure could lead to loss of any of these system components were identified previously in plant walkdowns such as those conducted for IE Bulletins (electrical equipment, anchorage, masonry walls).

In Table'V-1, equipment dependencies on passive components were identified for major plant equipment and their power sources. To have a more complete system failure analysis; equipment dependencies on /

cable routings were included in the models. For the equipment identi- /

fied, all the cables listed in the schemes for the various pieces of electrical equipmont were identified. The raceway routing of these cables was then listed and a general walkdown of the raceways was conducted. Of primary concern during the walkdown evaluations were items that may fail the cables by falling onto the cable trays, conduit, or equipment. For instance, the walkway grating by the screenhouse entrance door (SCHDW) was assumed to fail the diesel and (lectrical fire pumps, due to cables which run underneath this walkway.

The seismic capacity of these items was assumed to be that of the structure to which it is attached. For the example above, the walkway was assigned a ground acceleration of 0.500g which is the value of the screenhouse structure.

The capacity of the raceways has been evaluated by the Seismic Qualification Utilities Group (SQUG). The group concluded:

"...that the existing raceway systems in SEP plants possess substantial inherent seismic resistance and that the seismic qualification of raceway systems is not a significant safety issue."

Based upon this conclusion, the seismic capacity of the cable trays was set equsi to the capacity of the structure it is mounted to. Pending final NRC approval of the SQUG conclusions, no further analysis of the cable trays will be undeetsken.

NUO383-2628A-BQ01-NLO4 1

51 i

'Therefore, the bases for the capacities for the components listed in Table V-1 are discussed in this section as well as assumptions made with respect to potential operator response in recovery and repair of the component failures should they occur. The procedure for using the information presented in Table V-1 is discussed in Section VI.

1 A. Assumed Capacity In order to evaluate the seismic capacity of Big Rock Point structures, equipment and components on a consistent basis, information from many sources was reviewed. Information sources included analyses, tests, specifications, historical equipment seismic reports and judgments made by experienced engineers. '.h e seismic capacities determined for Big Rock Point structures, equipment, and components are based upon the following i

1. Detailed structural analyses of as-builts,

2. Detailed structural design analyses, i

3. Comparison based upon historical performance with respect to i items of similar physical appearance and functional requirements, and

4. Visual inspection and evaluation based upon experienced judgments.

Assessments with respect to the adequacy of structures, equipment and components at some level of seismic input were reliably estimated by i i the above four means. However, the specification of a lower-bound free-field seismic excitation at which an item fails was more difficult.

Therefore, a significant amount of judgment was required to normalize all sources of information and to unify all data on a consistent basis.

The results of detailed analysis of as-builts and seismic designs provided margins with respect to code (ASME, ANSI, AISI, AISC, ACI, i

> etc) requirements. An understanding of the margins with respect to the

• codes themselves provided a basis for establishing mean capacities and i variances associated with structural integrity or reliability. A mean capacity minus one standard deviation provides what is referred to in i Table V-1 as an "assumed capacity." l

' An evaluation of items on the basis of historical performance may well be directly amenable to similar potential calculations as to those employed for items analyzed. For some items, such as welded pipe, very i few failures due to seismic excitation have occurred Thus, statistics l are not suitable. Therefore, the "assumed capacity" is the lowest free I

field seismic excitation at which a failure has been known to occur.

1 Such an "assumed capacity" will be much more conservative (lower) than that derived statistically from distribution of failures.

Evaluation of iteme by inspection was done by comparison with items j which had a historical performance data base or which had been analyzed ,

NUO383-2628A-BQ01-NLO4  ;

i

52 at Big Rock Point. An assignment of an "assumed capacity" was based on statistics or lower bounds of historical performance data.

Examples of items evaluated by the general four groupings above are contained in the follcaing table:

BIG ROCK POINT SELECTIVE ASSUMED CAPACITIES Item Basis for Evaluation Masonry Walls Detailed Analysis to As-Builts Building Structures Detailed Analysis to As-Builts Station Power Battery Racks Detailed Analysis Design Fire Piping - Screenhouse Detailed Analysis Design Electrical Equipment Historical Performance Welded Piping Historical Performance Diesel Generator Unit Inspection - Historical Performance

- Turbine '.Bttiding Fire Piping . - . Inspection - Comparison with Analyzed Pipe All assumed capacities reflected an undetstanding of item location, overall building and local response, and the site specific spectral

. shape. Scaling of capacities from Regulatory Guide 1.60 (Design Response Spectra for Seismic Design of Nuclear Power Plant) to those of the site specific spectra was based on an overall scale factor of 1.5 where applicable. For plant structures, the assumed capacity was associated with a potential lack of serviceability. The capacities were elevated from the results obtained by D'Appolonia (Report 78-435, August 1981 - Volumes I - X, Seismic Safety Margin Evaluation - Big Rock Point Nuclear Power Plant Facilities). Serviceability is generally defined as code limits except for secondary steel members in highly redundant structures where code limits were handled more liberally. A scale factor of 1.5.for spectral shape between Reg Guide 1.60 and the Big Rock Point site specific spectra was used to relate D'Appolonia results to the site specific spectra. The increase in allowables of 1.6 for the safe shutdown earthquake (NRC-Standard Review Plan 3.8.4) was used for some steel member cilowables in evaluatin'g capacitics.

For concrete, no ductility factor or Standard Review Plan allowable increase was employed. Thus, concrete allowables were those determined strictly from the ultimate strength design method per AC1-349-76.

The assumed capacities for masonry and for some other equipment items in Table V-1 were determined from fragility analysis similar to chat of Kennedy et al (Probabilistic Seismic Safety Study of an Existing Nuclear Power Plant, Nuclear Engineering and Design 59 (1980) 315-338).

The assumed capacities for equipment not limited by building capacity are based on a minus one standard deviation from the estimated mean capacity. The standard deviations are composite values based upon both random variabilities and variabilities associated with the uncertainty of the mean value.

NUO383-2628A-BQ01-NLO4

. - - = _ . - -

53 B. Operator Initiated Recovery and Repair It was not assumed that because a particular structure or component listed in Table V-1 did not survive an earthquake that the function provided by the failed component was unavailable for the duration of the transient. Numerous options for repair or recovery of systems are available to the Big Rock Point operator depending on the nature of the failures and the time frame in which the failed systeu functions are required. In this section, the ground zules for assumed operator response in overcoming component failures identified in Table V-1 are presented. These ground rules were used as a basis for identifying operator response following an earthquake for which credit was taken in the event tree branch point descriptions of Section IV.

Recovery of failed systems was assumed to be possible if sufficient major components of a system which provided a given function survive the earthquake but were snot operating because of secondary failure modes, An example of this situation vould be the failure of the diesel generator to start because undervoltage relays did not generate any actuation signal but, given sufficient time, manual operation of the generator or standby generator could be accomplished providing power to the emergency bus. The following assumptions were made in recovering a system which was not adequately performing its function due to saismically induced failures:

1. No operator action w s *-c.umed to occur while ground motion was occurring. This assumption was applied to any recovery actions required to occur over the first few minutes of the transient.

As examples, no credit for manual RDS or core spray actuation was taken if a rcpid blowdown of the reactor was occurring due to LOCA or MSIV isolation failures.

2. Limited manual re+;u'.ery of systems which have survived the earthquake is assumed during the first few minutes following '

the earthquake. Examples of these operator actions include manual operator actuation of the emergeacy condenser outlet valves or enclosure spray valves. These actioi:5 4.w atsumed to be possible only for those systems that do not have to function while ground motion is in progress, whose act uattor, can occur from the control room; and, are considered successful ordy if instrumentation is available to the operator indicating the need to actuate the system and has also survived the earthquake.

3. Manual recovery of systems from outside the control room was considered possible only if a relatively significant time frame was available to the operator to perform these actions (eg, hours). Such actions included operation of the standby diesel generator or isolation of ruptured fire system piping in order to establish a flow path to the reactor building through the post incident system. These types of operator action were considered possible only if the systems were currently designed NUO383-2628A-BQ01-NLO4

54 to be operated in this manner (ie, no credit was taken 'or "modifying-' a system for unusual operation within a time frame

, of several hours). Again, indication that these types, of actior',were necessary needs to be available to the operator.

4. hepair or modification of a system was not considered possible in a time frame on the orde of days. An example of this type of operator action was demenstrated in the dependency of the core spray pumps on ac power. If the emergency bus or 1A and 2A buses failed to survive t.he earthquake due to the collapse of a enrby block wall, the only way power could be supplied t >

these pumps, if required, would be to proiide temporary jumpers from one of the diesel generators directly to the pump motor, bypassing norrnal station power distribution equipment. Credit for this type of repair or modification was taken only when

'ignificant time frames were available for repairs or mods, i

f 4

1 l

P NUO383-262t. It t:LO4 n-- _ _ . - - -

55 TABLE V-1 C0tPONENT FAILURE PODES, EFFECTS AND CAPACITIES Assumed Capacity (ZPCA)

Component Consequences of Failure and Failure Mode Containment Bu 'Iding Loss of components anchored to .4g - interpretation of D'Appolonia Internal Structure reactor .suilding concrete structures Report 78-435, Volune II (ie, piping, instrumentation, electricst Shear Failure of Reinforced cabinets el=ctrical and pneumatic valves) Concrete Wall Fu21 Cask Loading Dock failure of Pos2 Incident System .4g - interpretation of D' App.?.onia components lo ed in core spray Project Report 78-435, Volume VI Column Interaction Turbine Building m Failure of components anchcred to .25g - interpretation of D'Appsionia

n. Pedestal pedestal and pipe tunr+.1 structures Project Report 78-435, Volume III (ie, core spray and ChD pump suction piping) Sliding of Turbina Pedestal

b. - Steel Superstructrre Failure of components anchored to steel .32g - interpretation of D'Appolonia members of turbine building (ie, piping, Project Report 78-435, Volume III turbine building crane, electrical cabinets) Failure of Overall Steel Superstructure

c. Fotndction Failure of components anchored to concrete .35g - interpretation of D'Appolonia structures in the turbine building other Project Report 78-435, Volume III than turbine pedestal or pipe tunnel Failure of Foundation Below (ie, electrical cabinets, motor control Steel Superstructure centers, electrical instrumentation batteries Ss vice Building Loss of equipment anchored to service .35g - interpretation of D'Appolonia building structures (ie, control room electrical Project Report 78-435, Volume liI panels, and electricel instrumentatfon, Failure of Steel Structures RDS actuatica/ sensor cabinets)

Screenhouse Loss of components anchored to screenhouse .5g - interpretation of D'Appolonia and diesel generator roca strnctures Report 78-435. Volume VIII (ie, diesel generator elect:Ical panels, electrical instrumentation batteries, crane)

NUO383-2628A-BQ01-NLO4

56 TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Consequences of Failure and Failure Mode Steck Los of turbine building, service building .2Sg - interpretatisa of D'Appolonia and ctrical penetration components Project Report 78-433, Volume IV (ie, control room, RDS and station power Tensile Stress in Vertical Rebar room equipment) i Block Walls (

M100.01 Loss of CRD pump suction piping in .33g - interpretation of SMA condensate pump room Report 13703.01-R003 Loss of Blocks or Col.1 apse M100.02 Loss of condensate piping between 1.4g hot well and condensate storage tank (CRD pump suction piping)

M100,03 Loss of UPS A (RDS power supply and power .63g.

supply for emergency diesel generator circuit breaker)

M100.04 Loss of UPS B .53g M100.05 Lose of threaded fire piping in turbine building .53g M100.06 Loss of UPS A and 3 .33g M100.07 Loss of UPS C and D .37g M100.06 Loss of UPS A and C .31g M100.09 Loss of UPS B and D .31g M100.10 Loss of UPS A .33g M100.Il Loss of UPS C .37g M100.12 Loss of UPS C .16g NUO383-2628A-BQ01-NLO4 1

Si TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Conaequencer of Failure and Failure Mode MIDO.13 Loss of UPS D and threaded fire piping .16g in turbine building M100.14 Loss of de power source (station batteries. .53g and 125 V de Bus D01, D02 and DIO)

M100.15 Loss of threaded fire piping in turbine building .53g M100.16 Loss of de power source (station batteries. .Dg 125 V de Bus D01, D02 and DIO, core spray pipi.ng and 2400 V cable (offsite power))

M100.17 } Loss of 2400 V cable snd contactors for motor >>l.0g operated valves MO 7072 and MO 7064 M100.18

} Loss of station ac distribution equipment .13g (2400 V Bus, IA, 2A and 2B Buses)

M100.19 Loss of 2400 V cable and core spray piping .53g in turbine building r;l de power sources adjacent to black wall M100.16 M100.20 loss of core spray piping in turbine building .12g M100.21 Loss of 2400 V Cable (offsite power) .llg and threaded fire pipint; in turbine building)

M100.22 Loss of PIS-core spray HX cooling water .300g - (Sas11 vall near ground /

level, assign medium value of 0.3g) /

M10^.23 Fails cables to control room from alternate .110g - (Similar to MIOO.21 given shutdown building. fe., Emergency condenser smallest value of block walls) make-up MSIV and Emergency condenser valves.

UUO383-2628A-AOOl-NLO4

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

4 1

3 j -

58 i

TABLE V-1 (Continued)

Assumed Capacity (ZPCA) i Component Consequences of Failure and F:ilure Mode

K200.Pf5 Loss of electrical porer to MO 7066 .4g - (same as fuel cask i loading dock due to j size of wall and encased j all four sides by loading j dock concrete structure) 4 2400 V Voltage Regulator (2) Loss of offsite power .25g Historical capacity l of unanchored switch yard equipment l

l Falls free supports Emergency Condenser LOCA, RDS valve and fire piping damage .12g - interpretation of SMA Shall Report 13702.01-ROC 3 Support failure Nonregenerative LOCA .13g - interpretation of SMA '

i Hact Exchanger Report 13702.01-R002 i Support failure Recctor Cooling Water Fails CRL pump iI .40 (Used containment building /

Haat Exchanger internal structure). /

t l

Cleanup (2) LOCA .Ilg - interpretation of SMA Denineralizer Report 13702.01-R003' Support failure Rescror A7WS 0.2g - interpretation cf SMA /

Internals Report 13702.01 /

Support Plate Alignment pins /

shearing. /

CRDM Discharge } A11~ > .3g - interpretation of SMA /

Piping Report 13702.01 /

Crimping /

LOCA > .3g - interpretation of SMA /

Report 13702.01 /

Rupture /

NUO383-?628A-BQ0!-NLO4

59 TABLE V-1 (Continued)

Assumed Capacity (ZPCA)

Component Consequences of Failure and Failure Mode Core;aic Inst.lators Loss of offsite power .2g Historical capacity Fire Piping in Screenhouse Inability to cupply water to yard piping .%5g - interpretation of Cata1We (Cest iron components) and failure of automatic RDS actuation Calculation Book 2076, October, 1981 Rupture / leakage Yerd Piping Inability to supply water to turbine building .2g - Interpretation of D'Appolonia or containment through post incident system Report 78-435, Volume IX Rupture / leakage Fire Piping in Tt_.bine Inability to supply water to containment through .15g - Inspection Buildirg turbine building ard potential diversion Failure due to internal Threaded and of water from fire piping in containment if load of basket Victolic Couplings coincident with failure of Valves VPI 301 or 302 strainers Walded Piping .25g (Same as turbine pedestal as penetrates pipe tunnel wall)

Fire Piping in Core Spray Inability to supply water to containment through .4g (Same as fuel cask H2ct Exchanger Room Post incident system or shell side of loading dock due to core sprsy heat exchanger historical performance o';

welded carbon steel piping)

Rupture / leakage Fire Piping Inside Inahility to provide makeup to emergency .4g (Same as reactor Containment condenser shell, core spray, or building internal structure containment spray due to historical performance of carbon steel piping)

Rupture / leakage r

NUO383-2628A-BQ01-NLO4

60 TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Consequences of Failure and Failure Mode Electric and Diesel Inability to supply water to fire piping .25 - interpretation of Catalytic Fire Pumps (Pumps 6 and 7) in screenhouse and failure of automatic Engineering Evaluation Book 2076 RDS actuation October 1981 Rupture / leakage of cast iron casing Cere Spray Pumps Inability to recycle water from the .4g (securely anchored to (Pumps 2A and B) containment sump to core sprays for fuel cask loading dock long-term heat removal af ter a foundation)

LOCA or RDS actuation Loss of function or rupture / leakage Control Rod Drive Inability to makeup to reactor in long .4g (securely anchored to Pumps arid Piping ters to accommodate normal primary coolant reactor building internal I=sid9 Containment system leakage structure)

(Pumps 4A and B) Loss of function or rupture / leakage Containment Piping outside .25g (anchored to pipe tunnel walls)

Rupture leakage Emergency Diesel Failure to supply automatic emergency .5g (securely anchored to Generator ac power to 2B Bus diesel generator room floor)

Toss of function Sttniby Emergency Failure to supply backup emergency ac power Not characterized Diesel Generator to 2B Bus (manually actuated) Loss of function Emergency Diesel Failure of emergency diesel to run >l.0g (Buried tank)

Fuel Supply Leakage / rupture Diesel Fire Pump Failure of diesel fire pump to run >1.0g (buried tank)

Fuel Supply Leakage / rupture NUO383-2628A-BQ01-NLO4

l 61 TABLE V-1 (Continued)

Assumed Capacity (ZPCA)

Component Consequeneca of Failure and Failure Mode Stcndby Diesel Failure of standby diesel generator to run Not characterized Fu21 Supply Leakage / rupture Core Spray Heat Failure to remove decay heat or diversion .4g (securely anchored to Exchanger of post incident system water during fuel cask loading dock long-term cooling after LOCA or RDS actuation foundation)

Leakage / rupture Stction Batteries Loss of power to major de powered .35g (securely anchored to electrical equipment (125 V de Bus DOI) turbine building foundation)

Loss of function UPS Batteries Loss of power to RDS sensors and solenoid .35g (Securely anchored to) valves (UPS A also supplies power to emergency turbine building foundation diesel generator switchgear) Loss of function Emergency Diesel Loss of power to emergocy generator .5g (securely anchored to Gznerator Batteries starting circuit and motor starter diesel generator room floor)

(EDG failure to start) Loss of function Standby Diesel Loss of power to standby diesel generator Not characterized G:nerator Batteries motor starter (standby EDG failure to start) Loss of function Diccel Fire Pump Loss of power to diesel fire pump starting .5g (securely anchored to Betteries circuit and motor starter (diesel fire screenhouse floor) pump failure to start) Loss of function 125 V de Bus Loss of power to major de powered electrical .35g (securely anchored to /

equipment (Bus D02, de core spray valves) turbine building foundation) /

Loss of function / overturning /

125 V de Bus (D02) Loss of power to major de powered equipment .35g (securely anchored to I (Bus DIO, station annunciators, mise station turbine building foundation)

. power switchgear) Loss of function NUO383-?628A-BQ01-NLO4

62 TABLE V-1 (Continued)

Assumed Capacity (ZPCA)

Component Consequences of Failure and Failure Mode 125 V de Bus (2D) Loss of power to major de powered equipment .60 (mounted rigidly to ASDB wall)

(emergency .ondenser outlet valves MSIT, emergency condenser make-up valve) 125 V de Bus (DIO) Loss of power to de powered enclosure spray valve .35g (securely anchored to turbine and backup fire water supply valve to core spray building foundation) through PIS Loss of function 480 V Bus IA and 2A Loss of power to ac powered core spray pumps .35g (securely anchored to turbine core spray heat exchanger valve, CRD pumps building foundation)

Loss of function / overturning 480 V Bus 2B Loss of power to ac powered core spray valves, .35g (securely ancho ed to turbine electric fire pump Bus IA and 2A, auto throvover building foundation) panel Loss of function / overturning Bus I and 2 and Causes secondary failure of auto throvover .35g (securely anchored to turbine tetion power panel and contactors for core spray valves and building foundation) ac powered containment spray valve Overturning Auto Throvover Penel Loss of power to panels lY and 3Y .35g (securely fastened to turbine (C05) building north wall, turbine building foundation)

Loss of function lY Panel (BUS 17) Loss of ac power to control room instrumentation .32g (securely anchored to turbine (RDS manual Octuation switches, reactor pressure building structural steel) indication, RPS drum level indication, emergency Loss of function condenser shell level indication)

RDS Sansor and Loss of power to RDS valves, sensors and control .35g (securely anchored to computer Actuation Cabinets room indication (reactor and drum level) room floor, service building) automatic and manual actuation failure of RDS and Loss of function / overturning fire pump start circuitry (auto only)

NUO383-2628A-BQOI-NLO4

63 TABLE V-1 (Continued)

Assumed Capacity (ZPCA)

Component Consequences of Failure and Failure Mode Penals C01, CO2 and C40 Loss of indication and control for main control .35g (securely anchored to control room and RDS panels room floor, service building)

Loss of function / overturning  !

Panal C30 Causea loss of instrumentation on south wall of .4g (securely anchored to reactor steam drum enclosure reactor pressure switches for building internal structure) emergency condenser valves and core spray valves, Overturning reactor level switches for core spray valves and RDS control room reactor pressure and drum level indication Electric and Diesel Fire Failure of diesel and electric fire pumps .5g (securely anchored to screen- l Pump Control Panels to start house structure)

(C17 and C09) Loss of function Emergency Diesel Failure of emergency diesel generator to start .5g (securely anchored to Generator Control diesel generator room wall)

Panc! (CIS). Loss of function Stendby Diesel Failure of standby diesel generator to .15g (historical performance Gens-ator Setup and supply emergency power to 2B Bus of unanchored transformers-Sten m Transformers SSMRP)

Loss of function l

l UPS Battery Chargers Failure of UPS power supplies to RDS .35g (securely anchored to turbine building foundation)

Overturning onto UPS batteries Station Battery Chargers Failure of DC power supplies .35g (securely anchored to turbine building foundation)

Overturning onto station batteries or DC distribution panels l

NUO383-2628A-RQ01-NLO4 m_.

1 64 TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Consequences of Failure and Failure Mode HSIV Failure to isolate primary system potential .40g (okay as pressure boundary)

(No 705J) blowdown outside containment Mslv is sufficiently restricted that it will not impact surrounding st ructures.

Emergency Condenser Loss of primary system decay heat sink Not characterized (okay as Outlet Valves pressure boundary). Fails to (No 7C53, MO 7063) open because of operator impact on surrounding structure due to seismic motion Emergency Condenser Loss of primary system decay heat sink .4g (securely anchored to Makeup Valve reactor building internal (SV 4947) structure)

Fails to open Ac Core Spray Valves Loss of redundant core spray nozzle Not characterized (okay as (MO 7C70, MO 7071) pressure boundary.) Fails to open because of operator impact on surrounding structure due to seismic motion DC Core Spray Valves Loss of primary core spray ring sparger Not chacacterized (okay as (MO 7051, MO 7061) pressure boundary). Fails to open because of operator impact on surrounding structure due to seismic motion Ac cod dc Enclosure Failure of RDC and core spray equipment following .4g (well anchored to reactor Sprey Valves IOCA due to exceeding environmental qualification bui1 Jing internal structure motion (MO 7068, MO 7064) cr.velope limited by supports with reasonable clearance to surrounding structures >6")

NUO383-2628A-BQ01-NLO4

65 TABLE V-1 (Continued)

Assumed Capacity (ZPCA)

Component Consequences of Failure and Failure Mode C:ro Spray IIcat Inability to supply remotely actuated cooling 4g (securely anchored to foundation Exchanger Valves water to side of core spray heat exchanger (loss of fuel cask loading dock, motion (MO 7066, MO 7080) of long term cooling heat sink following LOCA limited by supports and good or RDS actuation clearance from surrounding structures)

Fails to open B ckup Fire Supply Inability to supply remotely actuated cooling .4g (securely anchored to foundation to Containment water to fire system inside containment through of fuel cask loading dock, motion (MO 7072) post incident system (loss of emergency condenser limited by supports and reasonable or core spray makeup if coincident with turbine) clearance with surrounding structures)

Fails to open Cora Spray Check Core spray or emergency condenser water makeup Not characterized Valves diversion to turbine building if coincident with Internal failure causing

~

(VPI 301, 302) turoine building welded fire piping failure back flow RDS Isolation Vr.1ves Loss of ability to depressurize reactor allowing .4g (securely an. chored to (CV 4180 thru 83) low pressure core spray reactor building internal structure catalytic)

Failure to open RDS Isolation Solenoid Loss of ability to vent air from CV 4180-83 .4g (securely fastened to RDS (SV 4980 thru 83) isolatloc valves)

Failure to open RDS Depressurization Loss of ability to depressurize reactor .4g (securely fastened to reactor Velves allowing low pressure core spray building internal structure)

(SV 4984 thru 87)

CV4090 Control rod drive pump suction valve CV4090. 0.25g (given valve of turbine Loss of reactor make-up via loss of control building pedestal). Valve is rod drive suction line. mounted to pipe tunnel wall.

NUO383-2628A-BQ01-NLO4

66 TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Consequences of Failure and Failure Mode SV4894 Solenoid valve for CV4090. Loss of reactor 0.25 (given valve of turbine make-up via loss of control rod drive suction building pedestal).

line.

1&C Transformers Loss of Panel 1Y and contactors for ac core .35g (securely enchored to north spray and enclosure spray valves wall of station power room-turbine buildir.g foundation)

Fall from supports Sectie Power Room Lo=s of Panel 1Y .32g (secured by turbine building Coolang Unit structural steel)

Falls from supports Tool Crib less of UPS B and D Not characterized Overturn of tool cabinet Screen loss of UPS B and D .32 (securely attached to steel /

superstructure of turbine /

building) /

Lights Near Station Shorting of station batteries .32g (secured to turbine Batteries building structural steel)

Falls from anchors Lights Near UPS Shorting of UPS batteries .13g (capacity of most fragile Bet teries UPS blockwall - secured to steel on ceiling)

Falls from anchors Steel Esclosure for Loss of RDS drum level Trs-*nnitters C&D .4g (securely anchored to reactor Drum Level Mirror building internal structure) and Emergency Falls from anchors Light NUO383-2628A-BQ01-NLO4 1

a 67 TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Consequences of Failure and Failure Mode Vent Ducts in SFP Loss of RDS reactor level Transmitter A .4g (secured to reactor building Hect Exchanger Room internal structure)

Falls from support Hypochlorite Tank Loss of screenhouse fire piping .5g (securely anchored to screen-house structure)

Falls from support Circulating Loss of fire pumps due to flooding .2g - Inspection W: tar Piping in Rupture / leakage Scraenhouse H2cter. Lights. Battery Loss of diesel generator batteries .5g (securely anchored to diesel Charger in Diesel generator room wall)

Generator Room Falls from supports Diesel Generator Failure of diesel generator to run .5g (securely anchored to diesel cooling Water Head generator room structure)

Tcak and Muffler Turb'ine Building Crane Failure of turbine building fire piping .32g - interpretation of D'Appolonia modeled as a part of turbine building structure Fails with turbine building structural steel Racetor Building Crane Failure of fire system pip.ng in containment .23g - SMA R. port 13703.0lR003 or LOCA gantry legs buckling Rail overturns Clecnup Demin Hoist Failure of enclosure spray valves Not characterized Load /hofst falls from rail NUO383-2628A-BQ01-NLO4

68 TABLE V-1 (Continued)

Assumed Capacity (ZPCA)

Component Consequences of Failure and Failure Mode Screenhouse Trolley Failure of fire system piping in .5g captured by screen-screenhouse house structural steel Falls from rafla RDS Hofst Failure of isolation Valves CV 3182 and .4g (securely anchored to reactor 3183 building internal structure with large gussets and baseplate Overturns

! IT Emergency Condenser Fails emergency condenser level instrument piping .4g (securely anchored to reactor Baca (drains shell) buildfeg internal structure with large gussets and baseplate)

Overturns Reactor Building Crane LOCA pins potential failure of core spray .22g interpretation of Whiting Analysis piping inside containment LT3180-3183 RDS steam drum level instrumentation .4g (reactor building internal leads to inability - automatically actuate structure) anchored to south RDS, automatically * :rt fire pumps Ond wall of steam drve start RDS drum levea Indication in control room Loss of '~nction LT3184-3187 RDS reactor level instrumentation .4g (reactor building internal leads to inability to automatically actuate structure) anchored to wall in spent RDS. Also results in less of RDS reactor fuel pool heat exchanger room level indication in control room Loss of function LSRE09A-II Peactor level switches. Leads to inability .4g (reactor building internal to automatically actuate ac or de core spray structure) anchored to south valves wall of steam drum and spent fuel pool heat exchang.er room wall NUO383-2628A-BQUI-NLO4

69 TABLE V-1 (Continued)

Assumed Capacity (ZPCA)

Component Consequences of Failure and Failure Mode LS3550 EmerFency condenser shell level switch .12g (emergency condenser shell) leads to failure of emergency condenser attached to piping on side low level annunciation in control room of emergency condenser shell LIRE 19A and B Drum level indicatore leads to loss .35g (turbine building column of ac powered drum level indication uplift) in control room LTRE20A Steam drum level transmitter. Leads to loss 4.g (reactor building internal structure) located in Panel C30 of ac powered drum level indication in control room anchored to floor near personnel lock Loss of function LEREO8B Steam drum level element. Leads to loss 4.g (reactor building internal of ac powered drum level indication in control structure) attached to east room end of stean drum Loss of function PSICl l A-II Reactor pressure rwitches. Leads to loss .4g (reactor building internal of automatic opening of core spray structure) anchored to south valves on low reactor pressure wall of drum enclosure and fuel pool heat exchanger room walls Loss of function PIIA07 Reactor pressure. Leads to loss of reactor .35g (turbine building column pressure indication in control room uplift) located in Panel CO2 in control room Loss of function PSID28E Reactor high pressure switch. Leads to loss .4g (turbine building column of high reactor pressure annunciation uplift) located on control room in control ro'ra Panel CO2 Loss of function NUO383-2628A-BQ01-NLO4

70 TABLE V-1 (Continued)

Assumed Capacity (IPCA)

Component Consequences of Failure and Failure Mode PSRE07 Reactor pressure switches. Loss of automatic .4g (reactor building internal actuation of emergency condenser outlet valves structure) located on south wall of steam drun enclosure Loss of function PS636A and B Enclosure pressure. Loss of automatic Not characterized. Mounted actuation of enclosure spray valve to angle iron in cable penetration area Loss of function PS7064A and B Enclosure pressure. Loss of automatic Not characterized. Mounted actuation of enclosure spray valve to angle iron in cable penetration area Loss of function Pf512 and 615 Fire pump discharge pressure. Loss of .5g (screenhouse) located automatic starting of fire pumps on low in fire pump control discharge pressure panels anchored to screenhouse floor Loss of function PS789-796 Fire pump discharge pressure. Loss of .5g (screenhouse) anchored automatic actuation of RDS to screenhouse wall Loss of function PI367 Enclosure pressure. Loss of control .35g (turbine building column room enclosute pressure indication uplift) located in control room Panel CO2 Loss of function PT174 Enclosure pressure. Loss of control .35g (turbine building column room enclosure pressure indication uplift) anchored to wall in cable penetration area Less of function NUO383-2628A-3Q01-NLO4

71 TABLE V-1 (Continued)

Assumed Ct.pacity (ZPCA)

. Component Consequences of Failure and Failure Mode LT3171 and 3175 Enclosure level. Loss of control room .4g (reactor building internal indication of enclosure water level structure) anchored to wall in recirculation pump room entrance area Loss of function HSPM07 Fire piarp hand switch. Loss of ability .5g (screenhouse) located to manually start fire pumps locally in fire pump control panels anchored to screenhouse floor loss of function HSVEC1 Hand switch for SV4947. Loss of ability .35g (service building) to makeup to emergency condenser shell located in control room Panel COI Loss of function HSPBRDS RDS push buttons. Loss of ability .35g (service building) to start fire pumps or manually located in control actuate RDS from control room room Panel C40 Loss of function HS7053 Emergency condenser outlet valve hand .35g (service building) switch. Loss of ability to manually located in control actuate emergency condenser room Panel COI Loss of function HS7068 Ac enclosure spray valve hand switch. .35g (service building)

Loss of ability to manually actuate loccted in Panel CO2 HO7068 from contrcl room in control room.

Fail open NUO383-2628A-BQOI-NLO4

72 TABLE V-1 (Continued)

Assumed Capacity (ZPCA)

Component Consequences of Failure and Failure Mode HS7066 Core spray heat 2xchanger valve hand switch. .35g (service building)

Loss of ability to remotely actuate M07066. located in Panel CO2 in control room.

HS7080 Core spray heat exchanger back-up valve hand .35g (service building) switch. Loss of ability to remotely actuate located in Panel CO2 M07080. In control room.

HSPM04 Hand switch for control rod drive pumps. .35g (service building)

Loss of CRD pumps. located in con:+ ?1 room.

HSPM02 Hand switch for core spray pumps. Loss of .35g (service building) core spray pumps. located in Fanel COI.

HSSDL Reactor depressurization system hand switch .35g (service building)

(drum and reactor level) loss of RDS located in Panel C40.

HSRDS Reactor depressurization system hand switch. .3'g (service building)

Loss of RDS located in Panel C40.

CB3550 LS3550 circuit breaker. Loss of emergency .32g (turbine building condenser level indication in control room superstructure) located in panel in station power room Fail open CBlYFI PIIA49 circuit breaker. Loss of reactor .32g (turbine building superstruc-pressure indication in control room ture) located in Panel lY anchored to wall in station power room Fail open CBlYLI Circuit breaker in the IY panel for reactor 0.32 (turbine building super-structure) located in 1Y panel, mounted to vertical column in station power room.

hT0383-2628A-BQ01-NLO4

73 TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Consequences of Failure and Failure Mode CB7072 M07072 circuit breaker. Loss of power .35g (turbine building to MO7072 foundation) located in Ptnel DIO located in station pwer room Fail open

/

DISC-1441 Loss of CRD PPf1 (ASDB connection to pump) .40g (secured to vall in /

EDC room) /

CBlYPB Circuit breaker to RDS control panel. Loss .35g (turbine building of ability to manually actuate fire pumps foundation) located in or RDS from control room Panel lY anchored to wall in station power room.

Fall open PNL3Y PIIA07 circuit breaker (in panel 3Y). Loss .35g (turbine building or pressure indication in control room. foundation) located in Panel 3Y anchored to wait in station power rooc Fail open CB7053 Circuit breakers for emergency condenser .35g (turbine building outlet valves. Ioss of ability to foundation) located in Panel DOI actuate emergency condenser anchored to station povar room Fail open NUO383-2628A-BQ01-NLO4

74 TABLE V-1 (Continued)

Assumed Capacity (IPCA)

Component Consequences of Failure and Failure Mode CB7051 Circuit breakers for de powered core .35g (turbine building spray valves. Loss of core spray flow foundation) located in Panel D01 tnrough M07051 and 65 anchored to station power room floor Fail open CBEDC Emergency diesel generator circuit breaker. .35g (turbine building Loss of abiffty to energize emergency foundation) located in Bus 2B bus via EDC anchored to station power room floor Fail open CB7070 Circuit breakers for ac powered core .35g (turbine building spray valves. Loss of core spray flow foundation) located in Bus 2B through M07070 and 71 anchored to station power C;PMO6 Ac fire pump circuit breaker. Loss of ac .35g (turbine building fire pump foundation) located in Bus 2B anchored to floor in station power room Fall open power room Fail open CBSDC Circuit breakers for the stand-by diesel generator 0.35g (turbine building Loss of stand-by diesel generator. foundation) located in Bus 2B in station power room.

NUO383-2628A-BQ01-NLO4

75 TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Consequences of Failure and Failure Mode C37068 Circuit breaker for ac enclosure spray valve. .35g (turbine building Loss of ability to actuate M07068 foundation) located in Bus 2B anchored to floor in station power room. Fail open.

CBPM04 Circuit breakers for control rod drive pumps. 0.35g (turbine building Loss of CRD pumps. foundation) located in Bus IA and Bus 2A in station power room.

CBPM02 Circuit breakers for core spray pumps. Loss 0.35g (turbine building of core spray pumps. foundation) located in Bus IA and 2A in station power room.

CB7064 Circuit breaker for de enclosure spray valve .35g (turbine building Loss of ability to actuate M07064 foundation) located in Bus DOI anchored to floor of station power room.

Fail open.

NUO383-2628A-BQ01-NLO4

76 TABLE V-1 (Cor.tinued) 1.ssumed Capacity (ZPGA)

Component Consequences of Failure ~ and Failure Mode CBIA2B 1A-28, 2A-2B tie breakers. Loss of ability to .35g (turbine building

.11sconnect 2B bus from auxiliary buses foundation) located in Buses IA and 2A anchored to floor of station pcwer room Fail closed CB2A2B Tie breaker for Bus 2A to 2B. Loss of ability 0.35 (turbine building to connect the 2A bus to the EDC. foundation) located in Bus 2A in station power room.

CB7050 Circuit breaker for ftSIV. Loss of ability .35g (turbine building to isolate primary system foundation) located in Bus D01 anchored to floor of station power room 0B7066 Circuit breaker for core spray heat exchanger .35g (turbine building valve M37066. Loss of ability to remotely foundation) located in Bus 2A, actuate 2107066. anchored to floor of station power room.

Footnotes:

(2)See Figure V-1 for location of block walls. (Note: block wall M100.PIS is in the fuel cask loading dock area and, therefore, not shown in Figure V-1.)

12} Failure of these components leads to one of the three initiating events.

NUO383-262RA-BQ01-NLO4

- . _ , . _ _ , _ . - ,..- - _ _ . . _ _ . , _ _ , - . - - .1.

77 TABLE V-1 (Coctinued)

Assumed Capacity (ZPGA)

Conw>onent _ _ _ _ .

Consequences of Failure and Failure Mode CB7080 circuit breaker for core spray heat exchanger .35g (turbine building back-up valve rt07080. Loss 02 ability to foundation) located in Bus 25, remotely actuate N07080. anchored to floor of station power room.

Access Hatch Condenter circulating water pump access .5g (given value of of screenhouse hatch in screenhouse. Failure of electric structure).

fire pump, RDS switches PS789.

Air Duct Air duct in core spray test tank area. .4g (given value of containment Failure of containment level indication building internal structure).

(LT3171 and LT3175).  ;

i Computer Equipment Computer, Jesks and pris.ters in computer .35g (given value of service building) i room. Loss of RDS cabinets in computer room. .

Computer Room Ceiling Computer room ceiling tiles and supports. .35g (given value of service building)

Loss of RDS cabinets in computer room.

Computer Room Wall Plaster wall in computer room. Loss of .35g (given value of service building) l cables from 3DS cabinets in computer room to control room.

l

.35g (given value of service building)

Telephone Room kall Wall between computer room and telephone i room. Loss of cables from RDS cabinets I in computer room to control room.

I I

l Footnotes:

1 (1)See Figure V-1 for location of block walls. (Note: block wall M100.PIS is in the fuel cask loading dock area i

and, therefore, not shown in Figure V-1.)

l (2) Failure of these components leads to one of the three initiating events.

NbO383-2628A-BQ01-NLO*.

1

78 TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Consequences of Failure and Failure Mode Emergency Lights and Lights and emergency eye wash units near .5g (given value of screenhouse)

Eye Wash Stations the diesel fire pump and emergency diesel generator batteries. Loss of this equipment.

File and Storage File drawer and test cabinet, storage .35g (g.ven value of service building)

Ccbinets cabinets, behind the RDS cabinets in the compute r room. Loss of RDS.

Floor Grating a. Grating over rad drive access area. .4g (given value of containment Loss of containeent level switches building).

LS3562, 3564 and 3565.

b. Grating in electrical penetration room, .4g (given value of containment inside containment. Loss of LSRE09 building).

c. Grating near personnel lock area. Loss .4g (given value of containment of rod drive puaps and LT3180-3184. building).

d. Grating at steam drum enclosure access .4g (given value of containment area. Loss of LS3550 and M07053. building).

c. Grating from upper reactor cooling water .4g (given value of containment heat exchanger room to electrical penetration building).

room. Loss of M07051 and 61 and LSRE09.

Floor Plate a. Metal floor plate between ventilation unit .4g (given value of containment and clean-up demia pit, at personnel lock building).

area. Loss of M07070, M07064 and M07068.

Footnotes:

(')See Figure V-1 for location of block walls. (Note: block wall M100.PIS is in the fuel cask loading dock area znd, therefore, not shown in Figure V-1.)

(2)Feilure of these components leads to one of the three initiating events.

NOO383-2628A-BQ01-NLO4 u

79 TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Consequences of Failure and Failure Mode

b. Second level mezzanine floor plate in outer .35g (given value of service building) electrical penetration room. Loss of SV4980 and 84, LT3180 and 84 and PI367.

Inst ruments Instrumentation mounted to wall in rod drive .4g (given value of containment access area. Loss of containment level switches building).

LS3562, 3564 and 3565.

Jurction Boxes a. JB-21 mounted to screenhouse wall. .5g (given value of screenhouse).

Loss of PS789.

b. JB-UPS-A mounted on top of UPSA battery .35g (given value of service building).

charger cabinet. Loss of UPSA and LT3180 and 84, CV4180, PS789 ar.d SV4984.

c. JB-97 in sphere ventilation room (air shed). .35g (given value of service building).

Loss of M07072, Core Spray pump, M07066, M07030 and N07072.

d. Terminal box mounted to wall above the second .35g (given value of service building).

level in the outer electrical penetration room.

Loss of LS3550, LT3171 and 75, LS3562, 64 and 65, M07070 and 71, LSRE09, rod drive pumps, M07050, M07053 and 63, CV4180, SV4980, M07051 and 61, M07064, M07068.

Footnotes:

(1)See Figure V-1 for location of block walls. (Note: block wall M100.PIS is in the fuel cask loading dock area and therefore, not shown in Figure V-1.)

(2) Failure of these components leads to one of the three initiating events.

NUO383-2628A-BQ01-NLO4 s

l 80 TABLE V-1 (Continued)

Assumed Capacity (ZPGA)

Component Consequences of Failure and Failure Mode Vest Line Lube oil tank vent line in outer electrical pene- .35g (given value of service building).

tration room. 19ss of LT3184 and 80, SV4984, SV4980.

Metal Hatch Access hatch cover to the regen /non-regen heat .4g (given value of containment exchanger room. Loss of LT31/5. building).

Recirc Pump Valve Recirc pump suction valve operator, MO-N003A, .4g (given value of containment located over rod drive access area. Loss of building).

containment. Icvel switches LS3562, 64 ard 65.

Overhead Light Light suspended from ceiling in core spray .4g (given value of fuel cask pump room. Loss of M07056. loading dock).

Poison Tank Liquid poison storage tank at emerstency .4g (given value of containment condenser level. Loss of LS3550. building).

Pipe Support Pipe and conduit. support located at the north .4g (given value of containment side of the emergency condenser level. Loss of building).

LS3550.

Screen Safety screen mounted to floor, south of the .5g (given value of screenhouse).

diesel fire pumps. Loss of diesel fire pumps.

Loose Equipment a. Tools and equipment in rod drive access area. .4g (given value of containment Loss of containment level switches LS3562, building).

64 an3 65.

Footnotes:

(3)See Figure V-1 for location of block walls. (Note: block wall M100.PIS is in the fuel cask loading dock area and, therefore, not shown in Figure V-1.)

Failure of these components leads to one of the three initiating events.

NUO383-2628A-BQ01-NLO4

t 81 TABLE V-1 (Continued)

Assumed Capacity (ZPCA)

Component Consequences of Failure and Failure Mode

b. Cabineta, desks and equipment in Room 441, 0.4g (given value of containment Radiation Protection counting room in building).

containment. Loss of LT3175.

Steel ceiling Steel ceiling plate in electrical penetration .4g (given value of containment room, inside containment. Loss of cables coming building).

into containment.

Vent Unit Ventilation units across from the clean-up demin .4g (given value of containment access area. Loss of LT3184. building).

Vent Unit 2 Ventilation ducts in su corner of containment 4g (secured to Reactor building /

- loss of ECS circuits internal structure) /

Footnotes:

See Figure V-1 for location of block walls. (Note: block wall M100.PIS is in the fuel cask loading dock area end, therefore, not shown in Figure V-1.)

(2) Failure of these components leads to one of the three initiating events.

NUO383-2628A-BQ01-NLO4 L

l 42 1

FIGURE V-1 COLT!!Ci LINES 6

5)

=

. 1 1 Menuk '

_ . %g . .- . _ _

m 2 . .

g 0

5Q s

... . . . . _. . 2 .. -. .

- i b

up ,,

8 19 ohuks -

v g .

tt I

S 14 ns u 44 .

uMi T bdews ,

g 2 lo i n . Is a 4,; .9 15 11 C. o

.  ;. . 1.2 . IJ . , . . . .. . - - . .

PLl* VIN OF M50NRY WALLS l

NUo383-2628A-BQO1

__ _ i

83 VI. METHODOLOGY FOR IDENTIFICATION OF THE SEISMIC "WEAK-LINKS" AT BIG ROCK POINT Three transients were identified in Section II as being those transients which require sufficient systems and equipment to characterize adequately Big Rock Point operational response to a given seismic event. The list of systems associated with these three transients and the manner in which they interact was sufficiently complete to assure that all important plant system functions were identified regardless of the actual transient which might occur as a result of the earthquake. The three transients chosen were loss of offsite power; medium steam line break inside containment and an ATWS. In this section, the procedure by which the most fragile combinations of equipment given any of these transients can be identified will be described. These most fragile combinations of equipment are referred to as the "weakest-links" with respect to attaining safe shutdown following an earthquake at the Big Rock Point site.

The event trees developed for the purpose of this study were presented in Section IV and are sisilar to the event trees presented in the Big Rock Point PRA. The system functions important to safe shutdown of the re&ctor are identified in the event tree headings. As stated in Section IV, these headings differ slightly from the headings of the event trees presented in the PRA as, conservatively, they do not include those systems or functions which are not easily shown to be capable of surviving the earthquake. They also contain more detail than do the PRA event trees with respect to those functions which will most likely be required. Existing systems or sets of equipment were identified which will fulfill each function identified by the event tree headings. The logic by which each of these systems succeeds or fails also was extracted from the PRA in the form of fault trees. For any given sequence, a fault tree was applied to each heading in the sequence for which a system functionally failed. By combining the fault trees for each system failure and performing Boolear logic on this combination of trees, the dependencies between each of the systems in a sequence was identified and a listing of minimum combinations of all the failures which must occur to result in a particular reactor state was developed.

As an example, there exists in the loss-of-offsite-power tree (at the end of Branch Point 10) a Sequence PEmC which defines a given set of system failures required to lead to a plant state in which inadequate coro cooling occurs. The particular system functional failursa which were assumed to occur in this sequence are makeup to the emergency condenser (Em) and core spray failure (C). Given that emergency condenser makeup depends on the FPS for its water source (just as does the core spray) there are some obvious dependencies between these two systems (core spray piping being a specific example).

The fault tree logic for emergency condenser makeup and core spray are /

presented in Apendix A of this report. The definitions of the bottom /

events in these trees are presented at the end of that section. It can /

NUO383-2628A-BQ01-NLO4

84 be seen from the fault trees that failure of either yard piping (PPYARD) /

or fire piping inside the containment (PP02) was sufficient to satisfy /

the logic of the trees and fail both systems. These two trees in /

Appendix A were combined under an AND gate and Boolean logic was /

performed by use of the SETS code. The results of this exercise are /

presented in Table VI-1 of this section. This table contains a partial /

list of the combinations of all the failures (cut sets) which must /

occur to attain the plant state PEmC. It can be seen that the yard /

piping and fire piping inside containment do in fatt show up as single /

events leading to the failure of both of these eyst. ems, confirming the /

dependency. Effectively, identification of this dependency in this /

manner indicates that the occurrence of either of :hese single events /

following an earthquake by themselves are sufficient to lead to a plant /

stato in which inadequate core cooling occurs. This exercise has been /

completed for all the system failures which occur in each sequence of /

the event trees presented in Section IV. There exists a table of cut /

sets similar to that presented in Table VI-1 for each sequence. Tens /

of thousands cut sets exist for the three-event trees as a vnole with /

the size of the cut sets ranging from one to seven members. /

In examining the event trees in Section VII, it may be noted that the detail of the trees has been substantially simplified over what exists in the PRA. As an example, the RDS tree was revised to include only a single train of power supplies, sensors, actuation cabinets and depres-surization valves because all four trains are essentially identical to each other in terms of their function, location and structural features. In other words, if one train fails as a rssult of a seismic event this study assumes the likelihood of a similar failure in the other trains is quite high. Table VI-2 contains a list of components which were modularized in this manner. Components which may have a dissimilar seismic resistance (such as the two diesel generators) or have functional diasimilarities in the way they operate (such as the t

' fire pumps and their power supplies) were not combined. Passive components and structures normally unimportant during these transients but whose failures may be made important as a result of ground motion were added to the trees (such as masonry walls).

Additional modularization of bottom events was performed as neeled as /

the fault trees were combined and run through SETS. This modulariza- /

tion consisted of combining a set of independent bottom events beneath /

an OR gate into a single bottom event. Bottom events simplified in /

this manner were compressed everywherc that specific combination of events occurred in the trees being run. No bottom event was included in a compressed event identifier that occurred by itself elsewhere in the tree so as not to lose the ability to identify all dependencies between systems. The definitions of these compressed events were saved for later use in identifying the weakest-links in the seismic response of the plant.

Determination of the weakest-links after an earthquake requires knowledge of'the seismic strength of the component and the response of the structure to which the component is mounted to ground motion. Given hT0383-2628 A-By01 -N1,04

85 l i

l i these, a best estimate ground acceleration which will result in the failure of a given component located at a specific location in the i plant can be determined. Applying this acceleration to each component i within a cut set, one can then determine the acceleration at which all i members of a cut set will fail. This acceleration is the acceleration t at which the strongest component in the cut set fails and represents  ;

j the seismic resistance of that cut set. Those cut sets which are

( satisfied at the lowest ground acceleration are the seismic weak-links  !

at the Big Rock Point Plant.

y t

l l Section V presented a table of failure modes and effects on all components  !

for which a bottom event exists in the fault trees of Section VII. For each component a conservative ground acceleration was presented above ,

which this study assumes the component fails. Some components have not j been characterized in sufficient detail to estimate a gronnd acceleration ,

at which they will fail. These components have been assigned an  ;

arbitrary capacity of zero 3 This approach artificially raises the importanco of these components for seismic events and allows a relative l

determination of the value of evaluating these components further. l The acceleration at which the most fragile of tre weak-links is satisfied is representative of that size earthquake the plant can be expected to survive without sufficient seismically-induced failures to result in core damage. It is these weak-links at which further evaluations or plant modifications should be aimed if any are necessary. Evaluation and modification of components in the more seismically resistant cut sets produces little measurable berefit unless the weaker-links are also addressed.

The fault trees for each of the system headings in the loss of offsite power, long-term cooling and medium steam-line break-event trees are represented in Section VII. The evaluation of the Big Rock Point Plant as it exists today using this methodology is presented in Section VIII.

Evaluation of potential modifications of the weak-links identified by this method is also presented in Section VIII.

NUO383-2628A-BQ01-NLO4

86 TABLE VI-1 LOSP2 =

PPYARD +

CRANE 75T +

PP02 +

SPEPRM +

M10018 +

TURBLDG +

RXBLDG +

PPHTC

• IST-1 +

TBERM

• IST-1 +

IST-8

• IST-9 +

PPO4W

• IST-9 +

PP03

• IST-9 +

M10019

• IST-9 + l M10014

• IST-9 +

M10017

• IST-8 +

M10017

• PPO4W t M10017

• PP03 + ,

M10019

• M10017 +

M10014

• M10017 +

t PPO4W

• VPI301 +

PP03

• VPI301 +

M10019

• VPI301 +

TBERM

• IST-3

• IST-4 +

TBERM

• UPSCHG

• IST-3 +

hT0383-2628A-BQ01-NLO4

l i

. 87 TABLE VI-1 (Continued)

TBERM

• JBUPSA

• IST-3 + {

TBERM

• M10010

• IST-3 +

TBERM

• M10006

• IST-3 +

TBERM

• M10003

• IST-3 +

TBERM

• PPHTC

• IST-3 +

PPHTC

• IST-2

• IST-3 +

M10019

• PPHTC

• IST-3.t (

M10014

• PPHTC

• IST-3 + l t

RE092ND

• IST-1

• IST-13 + i LSRE09

• IST-1

• IST-13 + f IST-1

• IST-10

• IST-11 +

i WALKWAY

• IST-1

• IST-10 +

l RE092ND

• IST-1

• IST-10 + j PSIG11

• IST-1

• IST-10 +

i LSRE09

• IST-1

• IST-10 +  !

VPI301

• M10020

• IST-8 + f M10014

• VPI301

• M10020 + l TBERM

• IST-3

• IST-5

• IST-6 +

TBERM

• M10017

• IST-3

• IST-6 +

TBERM

• SDGRAT

• IST-3

• IST-7 +

i TBERM

• SDCTUE

• IST-3

• IST-7 +

r TBERM

• CBSDG

• IST-3

• IST-7 +

TBERM

• SDGTR2

• IST-3

• IST-7 + f i

TBERM

• SDGTR1

• IST-3 IST-7 + r l

TBERM

• SDG

• IST-3

• IST-7 + {

WALKWAY

• IST-1

• IST-2

• IST-13 +  ;

NUO383-2628A-RQO1-NLO4 i

H 88 t

TABLE VI-1 (Continued)

PSIG11

• IST-1

• IST-la

• ICT-13 + ..

M10019

• RE092ND

• IST-3

• IST-4

• IST-13 +  !

M10019

• UPSCHG

• RE092ND

• IST-3

• IST-13 + l 4

M10019

• JUBUPSA

• RE092ND

• IST-3

• IST-13 + ,.

t M10019

• M10010

• RE092ND

• IST-3

• IST-13 + l 4

M10019

• M10008

• RE092ND

• IST-3

• IST-13 +

1 i

M10019

• M10006

• RE092ND

• IST-3

• IST-13 +

M10019

• M10003

• RE092ND

• IST-3

• IST-13 + I l

j M10014

• RE092ND

• IST-3

• IST-4

• IST-13 + l t

M10014

• UPSCHC

• RE092ND

• IST-3

• IST-13 v l

M10014

• JBUPSA

• RE092ND IST-3

• IST-13 +

M10014

• M10010

• RE092ND

• IST-3

• IST-13 +

1 l M10014

• M10008

• RE092ND

• IST-3

• IST-13 +  !

M10014

• M10006

• RE092ND

• IST-3

• IST-13 +  !

M10014

• M10003

• RE092ND

• IST-3

• IST-13 +

l RE092ND

• IST-2

• IST-3

• IST-4

• IST-13 +

a

UPSCE3

• RE092ND

• IST-2

• IST-3

• IST-13 +

l JBUPSA

• RE092ND

• IST-2

• IST-3

• IST-13 +

J l M10010

• RE092ND

• IST-2

• IST-3

• IST-13 +

} M10008

• RE092ND

• IST-2

• IST-3

• IST-13 +

M10006

• RE092ND

• IST-2

• IST-3

• IST-1* +

M10003

• RE092ND

• IST-2

• IST-3

• IST-13 +

l M10019

• LSRE09

• IST-3 IST-4

• IST-13 +

i d

M10019

• UPSCHG

• LSRE09

• IST-3

• IST-13 +

l M10019

• JbUPSA

• LSRE09

• IST-3

• IST-13 +

M10019

• M10010

• LSRE09

• IST-3

• IST-13 +

NUO383-2628A-BQ01-NLO4

89 TABLE VI-1 (Continued) l M10019

• M10008

• LSRE09

• IFi-3

• IST-13 +

l M10019

• M10006

• LSRE09

• IST-3

• IST-13 +

M10019

• M10003

• LSRE09

• IST-3

• IST-13 +

M10014

• LSRE09

• IST-3

• IST-4

• IST-13 +

M10014

• UPSCHG

• LSRE09

• IST-3

• IST-13 +

M10014

• JBUPSA

• LSRE09

• IST-3

• IST-13 +

M10014

• M10010

• LSRE09

• IST-3

• IST-13 +

M10014

• M10008

• LSRE09

• IST-3

• IST-13 +

M10014

• M10006

• LSRE09

• IST-3

• IST-13 +

M10014

• M10013

• LSRE09

• IST-3

• IST-13 +

LSRE09

• IST-2

• IST-3

• IST-4

• IST-13 +

UPSCHG

• LSRE09

• IST-2

• IST-3

• IST-13 +

JBUPSA

• LSRE09

• IST-2

• IST-3

• IST-13 +

M10010

• LSRE09

• IST-2

• IST-3

• IST-13 +

M10008

• LSRE09

• IST-2 ^ IST-3

• IST-13 +

M10006

• LSRE09

• IST-2

• IST-3

• IST-13 +

M10003

• LSRE09

• IST-2

• IST-3

• IST-13 +

IST-2

• IST-3

• IST-4

• IST-10

• IST-11 +

WALKWAY

• IST-2

• IST-3

• IST-4

• IST-10 +

RE092ND

• IST-2

• IST-3

• IST-4

• IST-10 +

PSIGil

• 1ST-2

• IST-3

• IST-4

• IST-10 +

LSRE09

• IST-2

• IST-3

• IST-4

• IST-10 +

UPSCHG

• IST-2

• IST-3

• IST-10

• IST-11 +

UPSCHG

• WALKWAY

• IST-2

• IST-3

• IST10 +

UPSCHG

• RE092ND

• IST-2

• IST-3

• IST10 +

UPSCHG

• PSIC11

• IST-2

• IST-3

• IST10 +

NUO383-2628A-BQ01-NLO4

' 90 ,

TABLE VI-1 (Continued)

. t UPSCHG

• LSRE09

• IST-2

• IST-3

• IST-10 +

JBUPSA

• IST-2

• IST-3

• IST-10

• IST-11 +

JBUPSA

• WALKWAY

• IST-2

• IST-3

• IST-10 +

JBUPSA

• RE092ND

• IST-2

• IST-3

• IST-10 + j JBUPSA

• PSIG11

• IST-2

• IST-3

• IST-10 +

JBUPSA

• LSRE09

• IST-2

• IST-3

• IST-10 +

2 M10010

• IST-2

• IST-3

• IST-10

• IST-11 +

i M10010

• WALKWAY

• ts"-2

• IST-3

• IST-10 +

M10010 * *!002:::

• ftT-2

• IST-3

• IST-10 +

M10010

• PSIG11

• IST-2

• IST-3

• IST-10 +

l M10079

• LSRE09

• IST-2

• IST-3

• IST-10 + ,

M10008

• IST-2

• IST-3

• IST-10

• IST-11 + >

t

M10007

• WALKWAY

• IST-2

• IST-3

• IST-10 +

t I I

M10008

• RE092ND

• IST-2

• IST-3

• IST-10 +

M10008

• PSIG11

• IST-2

• IST-3

• IST-10 +

i j M10008

• LSRE09

• IST-2

• IST-3

• IST-10 + [

I 3

M10006

• IST-2

• IST-3

• IST-10

• IST-11 + l r

! M10006

• WALKWAY

• IST-2

• IST-3

• IST-10 + t 1 I M10006

• RE092ND

• IST-2

• IST-3

• IST-10 + [

i, r M10006

• PSIG11

• IST-2

• IST 3

• IST-10 + f M10006

• LSRE09

• IST-2

• IST-3

• IST-10 + ,

I

, h10003

• IST-2

• IST-3

• IST-10

• IST-11 +

M10003

• VALKWAY

• IST-2

• IST-3

• IST-10 +  ;

i

M10003

• RE092ND

• IST-2

• IST-3

• IST-10 +

M10003

• PSIC11

• IST-2

• IST-3

• IST-10 +

i M10003

• LSRE09

• IST-2

• IST-3

• IST-10 +

NUO383-2628A-BQ01-NLO4

\

l I

• 91 I

TABLE VI-1 (Continued) i M10019

• IST-4

• IST-3

• IST-10

• IST-11 +

M10019

• WALKWAY

• IST-4

• IST-3

• IST-10 + l 8

l M10019

• RE092ND

• IST-4

• IST-3

• IST-10 + t l M10019

• PSIG11

• IST-4

• IST-3

• IST-10 +.

i 1 M10019

• LSRE09

• IST-4

• IST-3

• IST-10 +

M10019

• UPSCHG

• IST-3

• IST-10

• IST-11 +

M10019

• WALKWAY

• UPSCHG ' IST-3

• IST-10 +

M10019

• RE092ND

• UPSCHG

• IST-3

• IST-10 +

M10019

• PSIG11

• UPSCHG

• IST-3

• IST-10 +

M10019

• LSRE09

• UPSCHG

• IST-3

• IST-10 +

M10019

• JBUPSA

• IST-3

• IST-10

• IST-11 +

NUO383-2628A-BQO1-NLO4

92 TABLE VI-2 Major Components Modularized Due to Locational, Functional and Major Components System Structural Similarities Not Modularized Discussion Fire Protection MO 7070, MO 7071 (AC Core Spray Valves) Electric Fire Pump AC and DC core spray valves System MO 7051, MO 7061 (DC Core Spray Valves) Diesel Fire Pump were not modularized due to PSIG 11 A through H (Reactor Pressure) dependencies on different LSRE09 A through H (Reactor Level) power sources.

Fire pumps .not modularized due to dependencies on different power sources.

UPS A separated from other UPS to account for EDG starting circuit dependency.

RDS UPS A-D Sensor Actuation Cabinets LT 3180 through 3183 (Drum Level)

LT 3184 through 3187 (Reactor Level)

>dS Timers PS 789 through 796 (Fire Pump Pressure)

LV 4180 through 4183 (Isolation Valves)

SV 4980 through 4983 (Isolation Valve Air Supply)

SV 4984 through 4987 (Depressurization Valves)

Enclosure Spray PS 636 A and B (Containment Pressure) NO 7064 (DC Encl Spray) Enclosure spray valves not PS 7064 A and B (Containment Pressure) MO 7068 (AC Enct Spray) modularized due to dependencies on different power sources.

Emergency MO 7053 and MO 7063 (DC Outlet Valves)

Condenser PSRE07 A through D NUO383-2628A-IV)01-NLO4

. . - - _ - - . - . -- - . - . . . . . = _ _ . - . . . _ . . - . __ . . _ . . .= _. . -.

93 TABLE VI-2 (Continued)

Major Components Modularized Due to Locational, Functional and Major Components Syston Structural Similarities Not Modularized Discussion Emergency Power Emergency Diesel Diesel generators not Generator modularized due to dis-Standby Diesel similarities in rearting Cenerator circuitry (one automatic, one nanual) and structures in which they are housed (one in the screenhouse, one on a truck bed).

Post Incident Core Spray Pumps Syster. IA and 2A Bus CRD Makeup Control Rod Drive Pumps IA and 2A Bus NUO383-2628A-BQ01-NLO4

94 VII. FAULT TREE LOG 70

SUMMARY

The fault tree logic used to identify the structural weak-links following a seismic event at the Big Rock Point Plsnt is contained in Appendix A. The trees are drawn for those systems identified as being dmportant during an earthquake in Section III of this report. A fault tree is available for each system heading located in the event trees of j Section IV. The bottom events are described in the following table and i

include all components and dependencies defined in the failure modes and effects table of Section V.

, The first four fault trees were used for determining cut sets leading to core damage during LOCAs and Primary System Isolation Failures (with MSIV fault tree). These four trees include Post-Incident System Core Spray, Reactor Depressurization System and Enclosure Spray and included all dependencies on ac and de power sources UPS and all shared systea i

components. The remaining trees were used to evaluate the loss of offsite power and long-term cooling event trees and have had the ac and UPS power dependencies removed as these power sources are included in independent event tree system headings. There are two emergency condenser makeup, emergency condenser valve, core spray, RDS and fire pump Sc.es to account for the availability or absence of emergency

power.

i i

l I

l l

i l

l l

NUO383-2628A-BQ01-NLO4

95 1

VIII. RESULTS AND CONCLUSIONS A. Results Table VIII-1 of this section contains the results of the evant /

tree / fault tree evaluations cerformed using the methodology described /

in Section VI. This list of cutsets represent the seismic weak-links /

at Big Rock Point. The cutsets presented in Table VIII-1 were limited /

to cutsets whose seismic capacities were determined to be less than /

0.30G. This cut off point was arbitrarily chosen as a point to limit /

printing output. /

In Table VIII-1, the cutsets are ranked from the most fragile to those /

vith the greatest seismic capacity. The first column in this table /

list the sequence designator used throughout the analysis to represent /

the sequence PEvR (Less of station power with Emergency Condenser Valve /

failure and failure of RDS). These designations are listed in Table /

VIII-2. The second column in Table VIII-l lists the cutset capacity, /

or ground acceleration at which core damage is assumed to occur due to /

the sequence of events represented in the cutset. The cutsets are /

grouped by initiating events, ie., Loss of Station Power (LOSP). LOCA /

and Long Term Cooling (LTC). Cutset component identifier definitions /

are listed in Table VIII-3. /

As stated in Section VI, the most fragile cutsets are those of which /

backf,s or further analysis vill be directed. To determine which /

component (s) of the cutsets our upgrading effort vill be directed at, a /

list was compiled of all of the basic events whose ground acceleration /

< vas less than 0.30G, that were present in the weak-links cutsets (Table /

VIII-4). It is from among these basic events that the components for /

upgrading vill be chosen. Cutset by cutset, the basic events were /

reviewed and a determination was made as to which component of the /

cutset would be easiest and most ecst effective to fix, that would /

raise the capacity of the cutset. Since the upgrading of the seismic /

capacity of a particular basic event often raised the capacity of /

several cutsats, there are not as many weak-links to upgrade as there /

are weak-link cutsets. Table VIII-5 groups the baste events chosen for / j upgrading by ground acceleration. We can see by the summarisation of / i failures why these basic events have such a pervasive effect on safety. /

Several methods are available for raising the assumed capacity of each /

of the cutsets in which the above mentioned weak-links are a part. /

Note that an upgrade of any one of the weak-link basic events in a / >

curset can raise the capacity of a cutset. This is because the capacity /

of the cutset is equivalent to the strongest component of the cutset. /

therefore it is not necessary to consider each weak-link in a given / )

cutset. Backfits may take a variety of forms ranging from simple / f

~

procedure revisions to major structural upgrading of plant s t ructure s / :

and equipment. /

/

hT0383-2628A-BQ01-NLO4

96 Proposed fixes for the identified weak-links are listed in Table /

VIII-6. These are merely recommendatiens to the plant staff and are /

not the only alternatives. /

AWS was not handled in the same manner as the other two initiating /

events, LOSP and LOCA because of the lack of fault tree and event tree /

~

, logic. Instead, a reactor internal evaluation was ceanpleted by /

Structural Mechanics Associates (SMA 13703.01 October, 1985) to essess /

the seismic resistance of equipment important to shutting down the /

plant. The analysis addressed two technical issues, 1) the ability of /

the plant to SCRAM and 2) the integrity of the reactor vessel supports. /

The results of the SMA reactor internal evaluation indicated that the /

core assembly would shift and preclude scram at 0.20G. The reactor /

, vessel support evaluation indicated the supports would fail at 0.21G. /

Since these items are greater than the 0.120 ground response spectrum, /

l it is not mandatory that we perform seismic upgrades. However, the /

Technical Review Group will consider any modifications useful in /

i raising the seismic capacity of the weak-links of Big hock Point. /

B. Conclusions Numerous backfits and evalua*, ions are suggested in the previous section /

as potential dispositions for raising the cutset capacities to levels /

4 at which they are no longer limiting. These dispositions vary in /

j complexity from reanalyzing to major structural reinforcement of large /

j itema such as masonry walls and emergency condenser supports. In /

i selecting the backfits to be upgraded, it is desirable to select those /

which are most pervasive in their effect on upgrading the seismic /

capacity of the plant and the least ccerly in terns of capital /

, expenditures and other resources. Cost / benefit analyses are useful in /

the weighing of one backfit against another during analyses such as

/

these. Unfortunately, the methodology used in this report does not /

j easily lend itself to performing detailed cost beneth evaluations. /

'l The Safe Shutdown Earthquake (SSE) was da; ermined to be 0.12G by Reg /

Guide 1.60. In order to establish an adequate margin of safety, the /

analysis was bound to 0.18G. The weak-links less than or equal to 0.12G /

(the SSE) were presented to the Technica'. Review Group (TRG) in the /

January 1988 meeting. These vesk-links are the Emergency Condenser /

supports and block walls M10018 and M10021. The TRG requested an /

investigation of the cost involved in upgrading these two weak-links. /

Proposed fixes for these weak-links will be imples.ented. Weak-links /

greater than 0.12G will be reviewed by the TRG and considered for /

upgrading provided they are not cost prohibitive after the items /

in the 0.12C category have been resolved. All cost beneficial weak- /

links greater than 0.12g will be upgraded. They will be reviewed by /

TRG in order of the weakest-links first. /

LC0383-2628A-BQ01-NLO4

97 i

By upgrading the weak-links identified as those that fail equal to or /

below the SSE, and proceeding beyond the SSE by 50% or until it /

becomes cost prohibitive, the conclusion is reached that Big Rock /

, Point is seismically resistant, with an adequate margin of safety, /

to earthquakes that are Ifkely to occur in this general local. /

I l

j I

a f

1

}

(

i NUO383-2628A-BQ01-h104

98 TABLE VIII.;

LOSP18 .00E+00 M07053

• PS6362ND +

LSOP27 .00E+00 PS6362ND

• ISOSCRAM +

LOSP25 .00E+00 M07070

• M07051

• ISOSCRAM +

LOSP17 .00F,+00 CRANE 3T

• M07051

• M07053 +

LOSP17 00E+00 M07070

• M07051

• M07053 +

LOSP26 .00E+00 CRANE 3T

• M07051

• IS0 SCRAM +

1. LOSP18 1.20E-01 ECSHELL

• RDSCV2ND +

LOSP17 1.20E-01 CRANE 3T

• M07051

• ECSHELL +

LOSP17 1.20E-01 M10021

• VPI301

• M10020

• ECSHELL +

LOSP17 1.20E-01 M10021

• VPI301

• M10020

• M07053 +

LOSP17 1.20E-01 M07070

• M07051

• ECSHELL +

i LOSP3 1.20E-01 VPI301

• M10020

• IST-7

• IST-9 +

f LOSP3 1.20E-01 VPI301

• M10020

• PS6362ND

• IST-7 4 LOSP18 1.20E-01 ECSHELL

• PS6362ND +

LOSP18 1.20E-01 M07053

• RD$CV2ND +

LOSP26 1.20E-01 M10021

• VPI301

• M10020

• ISOSCRAM +

LOSP27 1.20E-01 RDSCV2ND

• ISOSCRAM +

LOSP25 1.20E-01 ISOSCRAM

• VPI301

• M10020 +

LOSP16 1.20E-01 VPI301

• IST-1

• IST-6 +

f LOSP1 1.20E-01 VPI301

• M10020

• IST-8 +

f LOSP2 1.20E-01 VPI301

• M10020

• IST-8 +

LOSP2 1.30E-01 M10018 +

LOSP1 1.30E-01 M10018 +

LOSP16 1.30E-01 M10018 +

LOSP14 1.30E-01 M10018 +

LOSP19 1.30E-01 M10018 +

LOSP15 1.30E-01 M10018 +

1. Denotes sequencee new to this analysis.

NL'0383-2628A-BQ01-NLO4 1

9@

TABLE VIII-1 (Continued)

LOSP12 1.30E-01 M10018 +

LOSP6 1.30E-01 M10018 +

1. LOSP28 1.30E-01 IS0 SCRAM

• M10018 +

i LOSP26 1.30E-01 ISOSCRAM w M10018 +

1. LOSP25 1.30E-01 ISOSCRAM

• M10018 +

d LOSP27 1.30T-01 ISOSCRAM

• M10018 +

1. LOSF20 1.30E-01 ECSHELL

• M10018 +

i LOSP23 1.30E-01 ECSHELL

• M10018 +

1. LOSP23 1.30E-01 M07053

• M10018 +

1. LOSP20 1.30E-01 M07053

• M10018 +

1. LOSP31 1.30E-01 ISOSCRAM

• M10018 +

i LOSP6 1.30E-01 VP1301

• M10020

• IST-2

• IST-4 +

1. LOSP26 1.30E-01 M07070

• DC

• ISOSCRAM +

1. LOSP21 1.30E-01 ECSHELL

• M10018 +

2 i LOSP21 1.30E-01 M07053

• M10018 +

1. LOSP26 1.30E-01 CRANE 3T

• DC

• IS0 SCRAM +

LOSP4 1.30E-01 M10018 +

2 LOSP3 1.30E-01 M10018 +

! LOSP7 1.30E-01 M10018 + .

LOSP33 1.30E-01 IS0 SCRAM

• UPS2ND +

f LOSP33 1.30E-01 IS0 SCRAM

• M10018 +

i LOSP29 1.30E-01 IS0 SCRAM

• M10018 +

f LOSP30 1.30E-01 ISOSCRAM

• M10018 +

LOSP11 1.30E-01 M10018 + r LOSP17 1.30E-01 M10018 +

LOSP10 1.30E-01 M10018 +

1. Denotes sequences new to this analysis.

I NL'0383-2628A-BQ01-NLO4

100 TABLE VIII-1 (Continued) f LOSP17 1.30E-01 CRANE 3T

• DC

• M07053 +

f LOSP17 1.30E-01 CRANE 3T

• DC

• ECSHELL +

1. LOSP17 1.30E-0; M07070

• DC

• M07053 +

f LOSP17 1.30E-01 M07070

• DC

• ECSHELL +

LOSP24 1.30E-01 ECSHELL

• UPS2ND +

LOSP24 1.30E-01 M10018 +

LOSP3 1.30E-01 M10018 +

LOSPS 1.30E-01 M10018 +

LOSP24 1.30E-01 M07053

• UPS2ND +

f LOSP10 1.30E-01 UPS2ND

• M10021

• VPI301

• M10020 +

f LOSP32 1.30E-01 IS0 SCRAM

• M10018 +

f LOSP20 1.30E-01 ECSHELL

• M10018 +

f LOSP20 1.30E-01 M07053

• M10018 +

f LOSP10 1.50E-91 UPS2ND

• PP06C

• VPI301

• M10020 +

f LOSP10 1.50E-01 UYS2ND

• Pt0$T

• VPI301

• M10020 +

f LOSP26 1.50E-01 PPO'T

• ISOSCRAM

• VPI301

• M10020 +

f LOSP26 1.50E-01 PP06C

• ISOSCRAM

• VPI301

• M10020 +

f LOSP17 1.50E-01 PP06C

• M10020

• VPI301

• ECSHELL +

f LOSP17 1.50E-01 PP06C

• M10020

• VPI301

• M07033 +

f LOSP17 1.50E-01 PP05T

• M10020

• VPI301

• ECSHELL +

f LOSP17 1.50E-01 PP05T

• M10020

• VPI301

• F07053 +

f LOSP17 1.60E-01 M10013

• M10020

• VPI301

• M07053 +

1. LOSP17 1.60E-01 M10013

• M10020

• VPI301

• ECSHELL +

f LOSP26 1.60E-01 M10013

• VPI301

• M10020

• 190 SCRAM +

LOSP24 1.60E-01 ECSHELL

• M10012 +

LOSP24 1.60E-01 M07053

• M10012 $

1. Denotes sequences new to this analysis.

NUO383-2628A-BQ01-NLO4

101 TABLE VIII-1 (Continued)

1. LOSP10 1.60E-01 M10012

• M16013

• VP1301

• M10020 +

f LOSP10 1.60E-01 UPS2ND

• M10013

• VPI301

• M10020 +

1. LOSP10 1.60E-01 M10012

• PP06C

• VPI301

• M10020 +

1. LOSP10 1.60E-01 M10012

• PP05T

• VPI301

• M10020 +

f LOSP14 1.60E-01 M10012

• UPS2ND

• ECSHFtt +

f LOSP14 1.60E-01 M10012

• UPS2ND

• LS3550 +

f LOSP14 1.60E-01 M10012

• UPS2ND

• PP06C

• VPI301

• M10020 +

r LOSP14 1.60E-01 M10012

• UPS2ND

• M10021

• VPI301

• M10020 +

1. LOSP14 1.60E-01 M10012

• UPS2ND

• PP0$T

• VPI301

• M10020 +

f LOSP14 1.60E-01 K10012

• UPS2ND

• M10013

• VPI301

• M10020 +

1. LOSP10 1.60E-01 M10012

• M10021

• M10020

• VPI301 +

1. LOSP10 2.00E-01 UPS,7ND

• PPYARD +

LOSP3 2.00P-01 PS6362ND

• IST-1

1. LOSP15 2.00E-01 M10012

• UPS2ND * 'PSWCW +

LOSP19 2.00E-01 PPSWCW

• ECSHELL +

LOSP19 2.00E-01 PPSWCW

• M07053 +

10SP3 2.00E-01 IST-1

• IST-9 +

LOSP26 2.00E-01 PPYARD

• 1SOSCRAM +

LOSP17 2.00E-01 PPYARD

• M07053 +

LDSP17 2.00E-01 PPYARD

• ECSHELL +

LOSP28 2.00E-01 ISOSCRAM

• PPSWCW +

• LOSP10 2.00E-01 M10012

• PPYARD +

LOSP11 2.00E-01 UPS2ND

• PPSWCW +

LOSP11 2.00E-01 M10012

• PPSWCW +

10SP4 2.00E-01 PPSWCW +

LOSP6 2.00E-01 IST-1

• IST-4

• IST-5 +

LOSP6 2.00E-01 PPYARD

• IST-4

• IST-5 +

1. Denotes sequences new to this analysis.

NL'0383-2628A-BQO1-NLO4

102 TABLE VIII-1 (Continued)

LOSP6 2.00E-01 VPI301

• M10020

• IST-1

• IST-4 +

LOSP6 2.00E-01 PPYARD

• VPI301

• M10020

• IST-4 +

f LOSP14 2.00E-01 M10012

• UPS2ND

• PPYARD +

1. LOSP1 2.00E-01 M10020

• VPI301

• PPYARD +

LOSP2 2.00E-01 PPYARS +

LOSP2 2.20E-01 CRANE 75T +

LOSP1 2.20E-01 CRANE 75T +

LOSP14 2.20E-01 M10012

• UPS2ND

• CRANE 757 +

f LOSP6 2.20E-01 CRANE 75T

• IST-4 +

LOSP17 2.20E-01 CRANE 75T

• ECSHELL +

LOSP16 2.20E-01 IST-1

• IST-5 +

LOSP17 2.261-01 CRANE 75T

• M07053 +

LOSP25 2.20E-01 IS0$ CRAM

• CRANE 75T +

LOSP10 2.20E-01 M10012

• CRANE 75T +

LOSP10 2. 20 F.-01 UPS2ND

• CRANE 75T +

LOSP26 2.20E-01 ISOSCRAM

• CRANE 75T +

f LOSPl? 2.50E-01 PPO4V

• VPI301

• ECSHELL +

f LOSP17 2.50E-01 PP03

• VPI301

• ECSHELL +

f LOSPl? 2.50E-01 PPO4V

• VP1301

• M07053 +

f LOSP17 2.50E-01 PP03

• VPI301

• M07053 +

LOSP19 2.50E-01 PPSCNH

• M07053 +

f LOSP15 2.50E-01 M10012

• UPS2ND

• PPSCNH +

1. LOSP15 2.50E-01 M10012

• UPS2ND

• PM07 +

f LOSP19 2.50E-01 PPSCH

• ECSHELL

• PM06 +

f LOSP19 2.50E-01 PM07

• ECSFELL

• PM06 +

f LOSP19 2.50E-01 PPSCNR

• ECSHELL +

1. Denotes sequences new to this analysis.

NUO383-2628A-BQ01-NLO4

103 TABLE VII2-1 (Continued)

1. LOSP19 2.50E-01 PPSCKH

• M07053

• PM06 +

LOSP19 2.50E-01 PM07

• M07053

• PM06 +

LOSP4 2.50E-01 PPSCNH +

LOSP29 2.50E-01 ISOSCRAM

• PPSCNH +

LOSP28 .1.50E-01 ISOSCRAM

• PM07

• PM06 +

LOSP4 2.50E-01 PM07

• PM06 +

LOSP10 2.50E-01 UPS2hD

• PP04W

• VPI301 +

LOSP10 2.50E-01 M10012

• PPO4W

• VPI301 +

LOSP10 2.50E-01 M10012

• PP03

• VPI301 +

LOSP10 2.50E-01 UPS2ND

• PP03

• VPI301 + ,

LOSP11 2.50E-01 M10012

• PPSCNH +

LOSP25 2.50E-01 ISOSCRAM

• VPI301

• PPO4W +

LOSP25 2.50E-01 IS0 SCRAM

• VPI301

• PP03 +

1. LOSP6 2.50E-01 PPO4W

• VPI301

• IST-4 +

f LOSP6 2.50E-01 PP03

• VPI301

• IST-4 + r i LOSP11 2.50E-01 UPS2ND

• PM07

• PM06 +

LOSP17 2.50E-01 CRANE 75T

• M07053 + s LOSP26 2.50E-01 PPO4W

• VPI301

• IS0 SCRAM +

4 LOSP26 2. 0E-01 PP03

• VPI301

• ISOSCRAM +

LOSP11 2.50E-01 UPS2ND

• PPSCNH +

LOSP11 2.50E-01 M10012

• PM07

• PM06 +

f LOSP11 2.50E-01 UPS2ND

• PPSCH

• PM06 +

LOSP3 2.50E-01 PP03

• VPI301

• IST-9 +

LOSP3 2.50E-01 PP03

• VPI301

• PS6362ND +

LOSP3 2.50E-01 PPO4W

• VPI301

• IST-9 +

LOSP3 2.50E-01 PPO4W

• VPI301

• PS6362ND +

1. Denotes sequences new to this analysis.

Nt'0383-2628A-BQ01-NLO4 L

t 106 TABLE VIII-1 (Continued)

LOSP19 2.50E-01 PPSHCN

• M07053 +

LOSP19 2.50E-01 PM07

• ECSHELL

• PM06 +

LOSP19 2.50E-01 PPSCNH

• ECSHELL +

f.0SP19 2.50E-01 PPSCNH

• M07053

• FM06 + .

i LOBP1 2.50E-01 PPO4W

• VPI301 +

LOSP1 2.50E-01 PP03

• VPI301 +

LOSP2 2.50E-01 PP03

• VPI301 +

LOSP2 2.50E-01 PPO4W

• VPI301 + r i LOSP3 3.00E-01 PP03

• IST-8

• IST-9 + '

1. LOSP3 3.00E-01 PP03

• PS6362KC

• IST-8 +

f LOSP3 3.00E-01 IST-7

• IST-S

• IST-9 +

f LOSP3 3.00E-01 PS6362ND

• IST-7

• IST-8 +

r v LOSP3 3.00E-01 PPO4W

• IST-9

• IST 8 +

1. LOSP3 3.00E-01 PPO4W

• PS6362ND

• IST-8 +

, LOCA I

CSLOCA .00E+00 M07070

• M07051 +

CSLOCA .00E+00 CRANE 3T

• M07051 +

~

RDSLOCA .00E +00 PS6362ND +

ESCLCA .00E+00 PS6362ND +

ESCLCA .00E+00 PS636

• PBSHIELD +

ESCLCA .00E+00 PS7064

• PBSHIELD +

ESCLCA .00E+00 CRANE 3T

• PBSHIELD +

i ESCLCA 1.10E+00 M10021 +

CSCLCA 1.10E+00 M10021 +

RDSLOCA 1.20E+00 ECSHELL + ,

CSLOCA 1.20E+00 ECSHELL +

1. Denotes sequences new to this analysis. ,

l i Nt0383-2628A-BQ01-NLO4

105 TABLE VIII-1 (Continued) l l PISLOCA 1.20E+00 ECSHELL +

PISLOCA 1.20Z+00 VPI301

• M10020 +

ESLOCA 1.20E+00 IC5 HELL +

ESLOCA 1.30E+00 UPS2ND +

ESLOCA 1.30E+00 M10018 +

PISLOCA 1.30E+00 M10018 +

CSLOCA 1.30E+00 M10018 +

l '

RDSLOCA 1.30E+00 UPS2ND +

l RDSLOCA 1.30E+00 M10018 +

l l CSLOCA 1.50E+00 PP06C +

l l CSLOCA 1.50E+00 PP05T +

ESLOCA 1.50E+00 PP05T +

i ESLOCA 1.50E+00 PP06C +

1 ESLOCA 1.60E+00 M10013 +

CSLOCA 1.60E+00 M10013 +

RDSLOCA 1.60E+00 M100UPS +

RDSLOCA 2.00E+00 PPSWCW 4 ESLOCA 2.00E+00 FPSWCW +

CSLOCA 2.00E+00 PPSWCW +

CSLOCA 2.00E+00 PriARD +

PISLOCA 2.00E+00 PPSWCW +

ESLOCA 2.00E+00 PPYARD +

ESLOCA 2.20E+00 CRANE 75T +

PISLOCA 2.20E+00 CRANE 75T +

CSLOCA 2.20E+00 CRANE 75T +

CSLOCA 2.50E+00 PPO4W +

1. Denotes sequences new to this analysis.

NUO383-2628A-BQO1-NLO4

[ i

(

I 106 TABLE VIII-1 (Continued) d CSLOCA 2.50E+00 PPSCNH +

CSLOCA 2.50E+00 PP03 +

CSLOCA 2.50E+00 PPSCNH

• PM06 +

CSLOCA 2.50E+00 PM07

• PM06 +

RDSLOCA 2.50E+00 PPSCHN +

RDSLCOA 2.50E+00  !'M07

• PM06 +

PISLOCA 2.50E+00 PPSCHN +

PISLOCA 2.50E+00 VPI301

• PP03 +

PISLOCA 2.50E+00- VPI301

• PPO4W +

PISLOCA 2.50E+00 PM07

• PM06 +

ESLOCA' 2.50E+00 PPSCNR +

ESLOCA 2.50E+00 PP03 +

ESLOCA 2.50E+00 PPO4W +

ESLOCA 2.50E+00 PM07

• PM06 +

ESLOCA 2.50E+00 PPSCN

• PM06 +

LONG TERM COOLING LTC10 1.30E-01 M10018

. LTC1 1.30E-01 M10018 LTC6 1.30E-01 M10018 LTC2 1.30E-01 M10018 LTC3 1.30E-01 M10018 LTCPUY 1.30E-01 M10018 LTC5 1.30E-01 M10018 LTC10 1.60E-01 UPS2ND

• M10012 +

LTC6 1.60E-01 UPS2ND

• M10012 +

LTC5 1.60E-01 UPS2ND

• M10012 +

1. Denotes sequences new to this analysis.

NUO383-2628A-BQ01-NLO4 t

107 TABLE VIII-1 (Continued)

LTC2 2.50E-01 PPCPD4

• PPYARD +

,LTC2. 2.50E-01 PPCRD4

• CRANE 75T +

LTC2 2.50E-01 CV4090

• CRANE 3T

• DC +

LTC2 2.50E-01 PPCRD3

• CRANE 3T

• DC +

LTC2 2.50E-01 PPCRD4

• M07070

• DC +

LTC2 2.50E-01 CV4090

• CRANE 3T

• M07051 +

LTC2 2.50E-01 PPCRD3

• CRANE 3T

• M07051 +

LTC2 2.50E-01 PPCRD3

• M07070

• M07031 +

LTC2 2.50E-01 CV4090

• PPO4W

• VPI301 +

LTC2 2.50E-01 PPCRD3

• PPO4W

• VPI301 +

LTC2 2.50E-01 CV4090

• PP03

• VPI301 +

LTC2 2.50E-01 PPCRD3

• PP03

• VPI301 +

LTC2 2.50E-01 PPCRD4

• PP06C

• VPI301

• M1009.0 +

LTC2 2.50E-01 CV4090

• PPOST

• VPI301

• M10020 +

LTC2 2.50E-01 PPCRD3

• PPOST

• VPI301

• M10020 +

LTC2 2.50E-01 PPCRD4

• M10021

• VPI301

• M10020 +

LTC2 2.50E-01 CV4090

• M10013

• VPI301

• M10020 +

LTC2 2.50E-01 PPCRD3

• M10013

• VPI301

• M10020 +

LTC2 2.50E-01 CV4090

• CRANE 75T +

LTC2 2.50E-01 CV4090

• PPYARD +

LTC2 2.50E-01 PPCRD3

• CRANE 75T +

LTC2 2.50E-01 PPCRD3

• PPYARD +

LTC2 2.50E-01 PPCRD4

• PP03

• VPI301 +

l LTC2 2.50E-01 PPCRD3

• M07070

• DC +

l LTC2 2.50E-01 CV4090

• M07070

• DC +

LTC2 2.50E-01 PPCRP4

• CRANE 75T

• M07051 +

1. Denotes sequences new to this analysis.

NUO383-2628A-BQ01-NLO4 L

108 TABLE VIII-1 (Continued)

'LTC2 2.50E-01 PE.dD4

• PPO4W

• VPI301 +

LTC2 2.50E-01' CV4090

• M07070

• N07051 +

LTC2 2.50E-01 PPCRD3

• M07070

• M07051 +

LTC2 2.50E-01 PPCRD4

• CRANE 3T

• DC +

LTC2 2.50E-01 CV4090

• M10021

• VPI301

• M10020 +

LTC2 2.50E-01 CV4090

• PP06C

• VPI301

• M10020 +

LTC2 2.50E-01 PPCRD3

• M10021

• VPI301

• M10020 +

LTC2 2.50E-01 PPCRD4

• PP05T

• VPI301

• M10020 +

LTC2 2.50E-01 PPCRD4

• M10012

• VPI301

• M10020 +

LTC2 2.50E-01 PPCRD3'* PP06C

• VPI301

• M10020 +

LTC1 2.50E-01 CRANE 75T

• PPCRD4 +

LTC1 2.50E-01 VPI301

• M10020

• PPCRD4 +

LTC1 2.50E-01 PPO4W

• VPI301

• CV4090 +

LTC1 2.50E-01 PPO4W

• VPI301

• PPCRD3 +

LTC1 2.50E-01 PP03

• VPI301

• PPCRD4 +

LTC1 2.50E-01 CRANE 75T

• PPCRD3 +

LTC1 2.50E-01 CRANE 75T

• CV4090 +

LTC1 2.50E-01 PPO4W

• VPI301

• PPCRD4 +

LTC1 2.50E-01 M10020

• VPI301

• CV4090 +

LTC1 2.50E-01 PP03

• VPI301

• CV4090 +

LTC1 2.50E-01 M10020

• VPI301

• PPCRD3 +

LTC1 2.50E-01 PP03

• VPI301

• PPCRD3 +

LTC3 2.50E-01 CV4090

• PS6362ND +

LTC3 2.50E-01 PPCRD4

• RDSCV2ND +

LTC3 2.50E-01 PPCRD3

• RDSCV2ND +

LTC3 2.50E-01 CV4090

• RDSCV2ND +

(Denotes sequences new to this analysis.

NUO383-2628A-BQ01-NLO4

l 109 7; TABLE VIII-l (Continued)-

LTC3 2.50E-01 PPCRD3

• PS6362ND +

LTC3 2.50E-01 PPCRD4

• PS6362ND +

LTCPUY 2.50E-01 CV4090

• UPS2ND +

LTCPUY 2.50E-01 PPCRD3

• UPS2ND +

LTCPUY 2.50E-01 CV4090

• M10012 +

LTCPUY 2.50E-01 PPCRD3

• M10012 +

LTCPUY 2.50E-01 PPCRD4

• M10012 +

LTCPUY 2.50E-01 PPCRD4

• UPS2ND +'

1 l

l i

l l

1. Denotes sequences new to this analysis.

NUO383-2628A-BQ01-NLO4

4 110 TABLE VIII-2 SEQUENCE DESCRIPTOR DESIGNATORS DESIGNATOR SEQUENCE DESCRIPTOR LOCA PISLOCA LLp CSLOCA LC RDSLOCA LR ESLOCA LE LOSP (LTC) PL LOSP1 PEmLp LOSP2 PEaC LOSP3 . PEAR LOSP4 PF LOSPS (LTC) PQL ,

LOSP6 PQEmLp LOSP7 PQEmC LOSP8 PQF LOSP9 PUL LOSP10 PUEa LOSP11 PUF LOSP12 PQEmR LOSP13 (LTC) PUQL  !

LOSP14 PUQEa -

LOSP15 PUQF LOSP16 PEvLp LOSP17 PEvC I

LOSP18 PEvR ,

LOSP19 PEvF LOSP20 PEvQLp LOSP21 PEvQC LOSP22 PEvQR LTC LOSP23 PEvQF LTC1 PYLp LOSP24 PEvU LTC2 PYC LOSP25 PILp LTC3 PYR LOSP26 PIC LTC4 PQLp LOSP27 PIR LTC5 PQC LOSP28 PIF LTC6 PQR LOSP29 PIQLp LTCPUY PUY(3) ,

LOSP30 PIQC LTC10 PQUY ,

LOSP31 PIQR LOSP32 PIQF LOSP33 PIU NUO383-2628A-bQ01-NLO4 y-,r-,.-, y..- - . .- - . _ . # -%.,- . - ,,.,~%,-. _~-,,_.,.,,,---r,-- __--,_p-.__,,_,_

, - . , _ - - -,, .e.- -- -

111 TABLE VIII-3 BOTTOM EVENT DEFINTION LIST NAME GROUND ACCEL DESCRIPTION 31802ND 0.40 VENT DUCT AND RCW PIPING 31842ND 0.40 SD MIRROR & EMERG LIGHT ACSHTCH 0.50 CCW PUMP ACCESS HATCH IN SCREENHOUSE ROOF AIRDUCT 0.40 AIRDUCT IN CORE SPRAY TEST TANK AREA ATRDUCT1 0.40 AIRDUCT IN CORE SPRAY TEST TANK AREA ASDBAT 0.60 ASDB BATTERIES ASDBBE 0.60 ASD BUILDING FAILURE ASDBLDG 0.60 ASDB STRUCTURE ASDCHG 0.60 ASDB BATTERY CHARGER BATCHG 0.35 DC 3ATTERY CHARGER BEAM 1T 0.40 1 TON BEAM ON ECS LEVEL BUS 1A 0.35 AC BUS 1A BUS 1Y 0.32 PANEL 1Y BUS 2% 0.35 AC BUS 2A BUS 2B 0.35 BUS 2B C05 0.35 AUTO THROWOVER PANEL CAC61 0.35 RDS ACTUATION CAB CBIA2A 0.35 CIRCUIT BREAKER 1A-2A

CBIA2B 0.35 CIRCUIT BREAKER 1A-2B CB1Y 0.32 CIRCUIT BREAKER 1Y CB1YLI 0.32 CIRCUIT BREAKER RX LEVEL INDICATOR CB1YPB 0.32 CIRCUIT BREAKER RDS PUSH BUTTON CB2A2B 0.35 2A-2B TIE BREAKER CB2B 0.35 CIRCUIT BREAKER 2B CB3550 0.35 CIRCUIT BREAKER LS-3550 CB7050 0.35 MSIV CIRCUIT BREAKER CB7051 0.35 CIRCUIT BREAKER M07051 & 61 CB7053 0.35 CIRCUIT BREAKER MO-7053 & 63 i CB7054 0.35 CIRCUIT BREAKER MO-7054 CB7064 0.35 CIRCUIT BREAKER M07064 CB7066 0.35 CIRCUIT BREAKER MO-7066 CB7068 0.35 CIRCUIT BREAKER MO-7068 CB7070 0.35 CIRC 1'IT BREAKER MO-7070 i

CB7072 0.35 CIRCUIT BREAKER MO-7072 CB7080 0.35 CIRCUIT BREAKER MO-7080 CB721D33 0.35 CIRCUIT BREAKER LS-3550 CBEDG 0.35 EDG CIRCUIT BREAKER CBPM02 0.35 CIRCUIT BREAKER CORE SPRAY PUMPS CBPM04 0.35 CIRCUIT BREAKERS CRD PUMPS CBPM04A 0.35 CIRCUIT BREAKERS CRD PUMPS NUO383-2628A-BQ01-NLO4 L

112 TABLE VIII-3 (Continued)

CBPM05 0.35 EFP CIRCUIT BREAKER CBPM06 0.35 CIRCUIT BREAKERS ETP LBSDG 0.35 CIRCUIT BREAKERS SDG CDST 0.95 CONDENSATE STORAGE TANK C05 0.35 AUTO THROWOVER PANEL COMPEQP 0.35 EQUIP IN COMPUTER RM COMPRMC 0.35 COMPUTER RM CEILING COMPWL 0.35 COMPUTER ROOM WALLS COOLUNIT 0.35 HEATING / COOLING UNIT NEAR 1Y CRANE 25T 0.32 25T CRANE CPANE2T 0.50 2T CRANE CRANE 3T 0.00 3T CRAhT CRANE 75T 0.22 75 T CRANE FAILS CRDFIL 0.40 CRD FILTERS CRDPD 0.40 CRD PULSE DAMPENER CRTRWL 0.35 COMP RM/ TELEPHONE WALL CV4090 0.25 CRD PUMP SUCTION VLV CV4180 0.40 RDS ISOLATION VLV CV-4180-83 DC 0.13 DC POWER GATE DCBAT 0.35 DC BATTERIES DCBUS 0.35 DC BUS D01 DCLTS 0.32 LIGHTS ABOVE STATION BATTERIES ECSHELL 0.12 EMERG. CONDENSER SHELL FAILS EDG 0.50 EDG EDGBAT 0.50 ED3 BATTERIES EDGBE 0.50 EDG FAILURE EDGBLDG 0.50 EDG BUILDING STRUCTURE EDGFUE 0.99 EDG FUEL EDGFUSES 0.99 EDG UNERGROUND FUEL TNK EDGTRN 0.99 OPERATOR ER00R TO TRANSFER EDG TO ASDB EDGTRNSFR 0.40 TRS-1442 AND TRS-1401 EMLTEYS 0.50 EMERG LIGHT & EYE WASH STATION EQUIPLCK 0.40 EQUIPMENT LOCK STRUCTURE FDSCCOMP 0.35 FILE DRAWER & STORAGE IN COMPUTER RM FLRGRAT 0.40 FLOOR GRATING OVER ROD DRIVE ACCESS FLRGRAT1 0.40 FLOOR GRATING OVER ROD DRIVF ACESS FLRPLT 0.40 FLOOR GRATES, DEMIN AREA FPPINH 0.35 FIRE PUMP INHIBITOR HAND SWITCH GRASA 0.40 WALKWAT GRATING AT SD ACCESS AREA GRAT2 0.40 FLOOR GRATING IN CONT. ELECTRICAL PENETRATION RM GRATPER 0.40 WALKWAT GRATING NEAR PERSONNEL LOCK HS7053 0.35 ECS OUTLET VLV PAND SWITCH HS7066 0.35 HAND SWITCH FOR EC OUTLET VALVES HS7068 0.35 HAND SWITCH MO-7068 HS7080 0.35 HAND SWITCH FOR MO-7080 11S7902 0.35 ECS MAKE-UP VLV HSPM02 0.35 CORE SPRAY PLHP CONTROL SWITCH HSPM04 0.35 CRD PUMPS HAND SWITCH HSPM07 0.50 HAND SWITCl! FOR FIRE PLHPS NUO383-2628A-BQ01-NLO4

l 113 TABLE VIII-3 (Continued)

HSRDS 0.35 RDS CONTROL SWITCHES HSSDL 0.35 RDS HAND SWITCH DRUM & RX LEVEL HSVEC1 0.40 HAND SWITCH FOR ECS MAKEUP VLV HX01 0.40 CORE SPRAY HX HYPOThK 0.50 HYPOCHLORITE TNK INST 1 0.40 INSTURMENTS IN ROD DRIVE ACCESS IS0 SCRAM 0.00 MSIV CONTROL SIWTCH ISOVLV3 0.99 RDS ISOLATION VALVES JB21 0.50 JUNCTION BOX 21 IN SCREENHOUSE JBUPSA 0.35 JUNCTION BOX UPSA CABINET LEQUIP 0.40 LOOSE EQUIPMENT IN ROD DRIVE ACCESS LERE08 0.40 STEAM DRUM LEVEL ELEMENT LI3380 0.35 SD LEVEL INDICATION LIIA40 0.35 RDS RX WATER LEVEL INDICATOR C40 LIRE 19 0.35 SD LEVEL INDICATION LOCAHOSE 0.99 FIRE HOSE TO CORE SPRAY HEAT EXCHANGER LOTVL 0.35 LUBE OIL TANK VENT LINE LPNLAS -0.35 LOCAL PANEL AND SFHERE VENT ISOLATION VLV IN AIR LS3550 0.12 ECS SHELL LEVEL SWITCH LS3562 0.00 LEVEL SWITCH LS3564 0.00 LEVEL SWITCH LSRE09 0.40 RX LEVEL SWITCHES RE09 LT3171 0.40 CONTAINMENT LEVEL TRANSMITTER LT3175 0.40 CONTAINMENT LEVEL TRANSMITTER LT3180 0.40 RDS RX LEVEL TRANSMITTER LT3184 0.40 SD LEVEL TRANSMITTER LTIA39 0.40 RVG RX WATER LEVEL TRANSMITTER LTRE20 0.40 SD LEVEL TRANSMITTER

, M10001 0.33 BLOCK WALL

{ M10003 0.63 BLOCK WALL M10004 0.53 BLOCK WALL M10005 0.53 BLOCK WALL M10006 0.33 BLOCK WALL M10008 0.31 BLOCK WALL M10009 0.31 BLOCK WALL M10010 0.33 BLOCK WALL M10012 0.37 BLOCK WALL j M10013 0.16 BLOCK WALL M10014 0.53 BLOCK WALL M10016 0.53 BLOCK WALL j M10017 0.99 BLOCK WALL M10018 0.13 BLOCK WALL M10019 0.53 BLOCK WALL M10020 0.12 BLOCK WALL M10021 0.11 BLOCK WALL M10022 0.30 BLOCK WALL M100PIS 0.40 BLOCK WALL MODULE M100UPS 0.16 COMBINATION BLOCK WALLS I METHAT 0.40 ACCESS HATCH TO RECEN-NON-REGEN RM i

NUO383-2628A-BQ01-NLO4 i

114 TABLE VIII-3 (Continued)

M07050 0.40 MSIV M07050BE 0.40 MSIV M07051 0.00 CORE SPRAY VLV M07051 & 61 K07053 0.00 ECS VLV M07053 &63 M07054BE 0.40 MO-7054 M07064 0.40 ENCLOSURE SPRAY VLV N07064BE 0.40 ENCLOSURE SPRAY VLV M07066 0.40 FIRE WATER TO CORE SPRAY HX VLV M07066BE 0.40 FIRE WATER TO CORE SPRAY HX VLV M07068 0.40 ENCLOSURE SPRAY 5.'LV M07068BE 0.40 ENCLOSURE SPRAY VLV N07070 0.00 CORE SPRAY VLV M07070 671 M07070BE 0,00 CORE SPRAY VLV M07070 & 71 M07072 0.40 FIRE WATER VLV AROUND PIS M07072BE 0.40 FIRE WATER VLV AROUND PIS M07080 0.40 BYPASS VLV FOR M07066 MON 003A 0.40 RECIRC PP SUCTION VLV MANUAL OPERATOR OVLEYW 0.50 OVERHEAD LIGHT & EYE WASH NEAR FIRE BAT OVRHDLT1 0.40 OVERHEAD LIGHT IN CORE SPRAY HX RM P7 BAT 0.50 DFP BATTERIES P7FUE 0.99 DFP UNDERGROUND FUEL P7 FUEL 0.99 DTP UNDERGROUND FUEL PBRDS 0.35 RDS PUSH BUTTONS PBSC'H 0.25 FIRE PMP PUSH BUUTON, SCREEN HOUSE PBSHIELD 0.00 LEAD SHIELD PT-174 PI367 0.35 PRESSURE IND PIIA07 0.40 RX PRESSURE INDICATION PIIA07BE 0.40 RX PRESSURE INDICATION PM02 0.40 CORE SPRAY PUMPS PM04 0.40 CONTROL ROD DRIVE PUMPS

. PM04A 0.40 CONTROL ROD DRIVE PUMPS 1 PM05 0.25 EFP I

PM06 0.25 EFP i PM07 0.25 DFP PM7 BAT I).50 DFP BATTERIES PNC09 0.50 DFP CONTROL PANEL PNL20 0.40 PANEL C20

'NL30

/ 0.40 PANEL C30 i PNL3Y 0.35 PANEL 3Y PNLC09 0.50 DFP CONTROL PANEL t PNLC17 0.50 EFP CONTROL PANEL PNLC18 0.50 PANEL C18 PNLC30 0.40 PANEL C30 PNLD01 0.35 PANEL D01 PNLD02 0.35 DC DISTR. BUS D02 PNLD10 0.35 DC PANEL D10

PNLD2D 0.60 ASDB DC DISTR. PANEL 2D POSTNK 0.40 POISON TANK PP01 0.40 FIRE PIPING IN CORE SPRAY PUMP PJi l

l NUO383-2628A-BQ01-NLO4

115 TABLE VIII-3 (Continued)

PP02 0.40 CORE SPRAY PIPING INSIDE CONTAINMENT } AILS PP03 0.25 WELDED FIRE PIPING FAILS PPO4W 0.25 WELDED CORE SPRAY PIPING PPOST 0.15 THREADED FIRE PIPING TURB. BLDG PP06C 0.15 VICTROLIC COUPLING-FIRE PIPING PPCRD1 0.40 CRD PIPE FROM PUMPS TO NC-18 PPCRD2 0.40 CRD PIPE FROM CONTAIhHENT TO PLHP SUCTION PPCRD3 0.25 CRD PIPE IN PIPE TUNNEL PPCRD4 0.25 CRD PIPE IN CONDENSATE PUMP ROOM PPCRD5 0.99 CRD PIPE UNDERGROUND FROM CDST PPHTC 0.32 HEATING PIPING, STATION POWER RM PPSCN 0.25 SCREENHOUSE PIPING PPSCNH 0.25 SCREENHOUSE FIRE PIPING PPSUP1 0.40 PIPE SUPPORTS, N SIDE OF ECS LEVEL PPSWCW 0.20 CIRC WATER PIPING PPYARD- 0.20 CAST IRON UNDERGOUND FIRE PIPING PS612 0.50 PRESSURE SWITCH PS615 0.50 PS-615 EFP DISCHARGE PRESSURE PS632ND 0.00 VENT DUCT & PIPING MONITOR NEAR PRESS. SWITCH PS636 0.00 ENCLOSURE SPRAY VC,V PRESSURE SWITCH PS6362ND 0.00 VENT DECT&ILRT PI?ING NEAR PRESS. SWITCH PS7054 0.00 PRESSURE SWITCHES PS7064 0.00 ENCLOSURE SPRAY VIV PRESSURE SWITCHES PS789 0.50 FIRE PUMP DISCHARGE PRESS SWITCH PSID28 0.40 RX HIGH PRESSURE SWITCH PSIG11 0.40 RX HIGH PRESSURE SWITCH PSRE07 0.40 ECS OUTLEY VLV CONTROL SWITCH PT174 0.35 PRESSURE TRANSMITTER RAC11 0.35 RDS ACTUATION AND SENSOR CABINET RDSCV2ND 0.12 ECS SHELL AND 2 T09 WINCH RDSHOIST 0.40 RDS HOIST ON ECS LEVEL RDSLTLS 0.40 LT3180 AND LT3184 RDSPIPE 0.40 RDS PIPING OUTSIDE SD RE092ND 0.40 RCW ASDB SPF PIPIN3 BY RE09 SWITCH RE092ND 0.40 RCW ABD SPF PIPING IN RE09 AREA R00702ND 0.35 I & C TRANSFORMER AND AC BUS 1 62 R0702ND 0.35 I&C TRANSFORMER AC BUS 1&2 R07064 0.99 RELAY CONTACTS MO-1054 R07070 0.35 RELAY CONTACTS FOR M07070 & 71 R00M441 0.40 ROOM 441, MISC. EQ'JIPMENT RXBLDG 0.40 RX BUILDING FAILS RXCLCHX 0.40 RX COOLING WATER EC SCHDW 0.50 SCREENHOUSE WALKWAY SCHTLY 0.50 2T TROLLEY IN SCREENHOUSE SCNHBLG 0.50 SCREEhTOUSE STRUCTURE SCNHSBLD 0.50 SCREENHOUSE STRUCTURE SchTSBLG 0.50 SCREENHOUSE STRUCTURE SCREEN 0.32 SCREEN IN FRONT OF TOOLCRIB SCRN 0.50 SCRENN BEHIND DFP hT0383-2628A-PQ01-NLO4

116 TABLE VIII-3 (Continued)

SDG 0.00 STANDBY DIESEL GEliERATOR SDGBAT 0.00 STANDBY DIESEL GEN BATTERIES SDGFUE 0.00 STANDBY DIESEL GEN FUEL TANK SDGTR1 0.15 SDG TRANFORMER #1 SDGTR2 0.15 SDG TRANSORMER #2 SDLVL 0.35 LIRE 19 AND LTRE20 AND LERE08 SLMEPRM 0.35 2ND LEVEL MAZZANINE, OUTER CABLE SPEPRM 0.40 STEEL CEILING-INNER ELECTR. PENETRATION SV4894 0.40 SOLENOID FOR CV-4090 SV4947 0.40 ECS MAKEUP VLV SV498' O.40 RDS ISOLATION VLV 4980-83 SV4984 0.40 RDS DEPRESS VLV TBERM 0.35 TERMINAL BOX ELECTRICAL PEN. RM TESTCAB 0.35 TEST CABINET IN COMPUTER RM TOOLCRIB 0.00 TOOL CABINETS IN TOOL CRIB TRANSTCH 0.50 CRD PP #1 DISC. SWITCH 1441 TURBLDG 0.32 TURBINE BUILDI!"G FAILS UPS 0.35 UPS ban ERIES UPS2ND 0.13 VENT, DUCTS, LIGHTS, UPS UPSA 0.35 UPSA BATTERIES UPSABE 0.35 UPSA FAILURE UPSBAT 0.35 UPS BATTERIES UPSBE 0.35 UPS FAILURE UPSCHG 0.35 UPS BATTERY CHARGER UVRE2B 0.50 UNDERVOLTAGE RELAY ON 2B VENT 2 0.40 SHPERE HEATING & COOLING UNIT DUCTS VENTUNT 0.40 VENTILATION CABINETS W OF PERSONNEL LOCK VPI301 0.00 CORE SPRAY CHECK VLV 301 & 302 WALKWAY 0.40 WALKVAY GRATING TO ELECT SC 4TRATION RM XLEVEL 0.40 CONTAINMENT LEVEL TRANSMITTERS NUO383-2628A-BQO1-NLO4

117 TABLE VIII-4 COMPONENTS OF CUTSETS THAT ARE LESS *HAN 0.30G PPYARD 0.20 IST-1 LOSP6 PPSWCW 0.20 LS3550 0.12 CRANE 757 0.22 PPO4W 0.25 IST-7 LOSP3 PP03 0.25- M10013 0.16 PPSCNH 0.25 M10021 0.11 PM07 0.25 PPOST 0.15 PM06 0.25 PP06C 0.15 PPCRD4 0.25 CV4090 0.25 IST-9 LOSP3 PPCRD3 0.25 RDSCV2ND 0.12 ECSHELL 0.12 RDSCV2ND 0.12 IST-5 LOSP16 CRANE 3T 0.00 TRNYPIPL 0.20 l M07051 0.00 CRANE 75T 0.22 -

M10021 0.11

- VP1301 0.00 IST-1 LOSP16 M10020 0.12 M07053 0.00 M07053 0.00 ECSHELL 0.12 M07070 0.00 PS636ND 0.00 IST-2 LOSP6 l IS0 SCRAM 0.00 PP06C 0.15 i M10018 0.13 M10013 0.16 DC 0.13 M10021 0.11  ;

UPS2ND 0.13 PPOST 0.15 PP06C 0.15 PP05C 0.15 M10012 0.16 IST-4 LOSP6 M10013 0.16 UPS2ND 0.13 PPSWCW 0.20 M10012 0.16 PBSHIELD 0.00 IST-1 LOSP3 4 CRANE 75T 0.22 PPYARD 0.20 i IST-6 LOSP16 PP03 0.25 IST-8 LOSP1 PPO4W 0.25 PP06C 0.15 M10020 0.12 M10013 0.16 M10021 0.11 PPOST 0.15 i

i I

NUO383-2628A-BQ01-NLO4

i 118 TABLE VIII-5 WEAK-LINFS WEAK-LINKS AT 0.11G M10021 This block wall is located between the machine shop and the laydown area in the turbine building. The 2400 V switchgear cable and the fire piping penetrate this block wall. Failure of this wall causes failure of the Core Spray and Enclosure Spray systems during a LOCA only because of lack of time to isolate the turbine building fire piping and circumvent the fire system pipe break. ,

(See Figure V-1 for location of this block wall)

WEAK-LINKS AT 0.12G ECSKELL The emergency condenser shell support failure causes the failure of the emergency condenser. This failure not only causes loss of the primary heat removal source during a loss of station power, but it also could cause a loss of the KDS valves, and loss of the fire water supply to the cota spray system either through fire water make-up line to EC rupture or the EC snell falling off the ECS level and rupturing the surrounding enclosure spray anJ core spray piping.

M10018 This block wall is the west wall of the station power room. Failure of this wall takes out the 2400V bus, Bus IA, 2A and 2B. Because of the large amount of dependence in all systems on these power supplies, failure of this block wall can fail multiple safety systems simultaneously.

(See Figure V-1 for location of this block wall).

WEAK-LINKS AT 3.13G M10020 This block wall is the south wall of the lube oil tank room. Failure of this wall will damage or fail the fire water supply to containment. This particular segment of piping can be isolated from the fire system by manual closure of VFP-29 and 30. ,

Closure of these valves eliminates flow diversion of the fire system and itakes other paths of fire water available, given time is available as in loss of station power sequesees where there is a need for NUO383-2628A-BQ01-NLO4

I i

119 TABLE VIII-5 (Continued) core spray, enclosure spray or emergency condenser make-up.

(See Figure V-1 for location of this block wall).

OR In order for fire water supply to core spray and VPI301 enclosure spray to fail from a pipe rupture in the lube oil tanf room, VPI-301 and VPI-302, the core spray check valves, must fail to-hold, thereby preventing an alternate path of fire water supply to the containment from flowing out of these paths.

Therefore M10020 vill only cause a failure if VPI-301 and 302 fail simultaneously with M10020. Therefore we can fix either M10020 or VPI-301 and VPI-302. A deterministic analyses of these check valves was never performed, therefore we assumed that they would fail a 0.00.

UPS2ND Failure of the vent ducts and lights in the UPS room to stay secured are assumed to fail fail the UPS batteries, and therefore the RDS system.

WEAK-LINKS AT 0.15G PP06C Failure of the Vicerolic coupled fire piping in the Turbine Building will cause an inability to supply water to containment through the Turbine Building.

This piping failure causes a problem in a LOCA situation in as not enough time is available to isolate the turbine building fire water supply and circumvent the break to supply water to containment.

PPOST Failure of the Threaded fire piping in the Turbine Building will cause an inability to supply water to containment through the Turbine Building.

This piping failure causes a problem in a TOCA situation in as not enough time is available to isolate the turbine building fire water supply and circut ant the break to supply water to containment.

WEAK-LINKS AT 0.16G M10013 This block wall is located along the east wall of the stock room. Failure of this block wall may damage or fail the UPS batteries or the fire piping.

This appears only in the LOCA event trees as there is NUO383-2628A-BQ01-NLO4

I l

120 TABLE VIII-5 (Continued) not time to isolate a turbine building fire piping break during a LOCA.

(See Figure V-1 for location of this block wall).

M100UPS M100UPS is a combination of block walls, the most limiting of which is M10013. It's effects due to failure, is explained above. All of the other walls that makeup M100UPS have ground accelerations greater than 0.3g.

WEAK-LINKS AT 0.20G PPYARD Failure of the fire piping in the yard enn inhibit water supply to the turbine building or the containment for core spray, enclosure spray or use for emergency condenser makeup. No credit is taken for bypassing the yard loop via the LOCA hose '

in the case of a LOCA in as there is not enough time available.

PPSWCW Loss of the circulationg water flow due to the expansion joint rupturing causes less of the main condenser heat removal and loss of the fire pumps due to flooding. This catastrophic loss of fire water leads to core damage in any event tree.

ATWS Shearing of the alignment pins in the core support [

plate causes lateral movement of the support plate, preventing rod insertion.

NUO383-2628A-BQO1-NLO4

'131 TABLE VIII-6

,si OPTIONAL FIXES WEAK-LINKS AT 0.11G M10021 1. Structurally reinforce the block wall.

2. Reassess the seismic fragility of the block walls using nonlinear dynamic analysis. (Computech Engineering Services proposal. February 10,1986).

OR 3. Relocate the affected equipment and cables to a more suitable location.

WEAK-LINKS AT 0.12G ECSHELL 1. Improve emergency condenser supports possibly by attaching the other 3 gussets to the floor pads.

M10018 1. Secure the wall to the building structure.

2. Structurally reinforce the wall.

3. Reassess the seismic fragility of the block wall using nonlinear dynamic analysis. (Computech Engineering Services proposal, February 10, 1986).

OR 4. Relocate the affected equipment and cables to a more suitable location.

WEAK ?. INKS AT 0.13G

!!10020 1. Securs wall to building structure.

2. Etru,turally reinforce the tiock wall).

3. Reassess the seismic fragility of the block walls uJing nonijnest dynamic analysis. (Computech Engineering Services proposal, February 10, 1986).

OR 4 Relocate the affected equipment and cables to a more suitabic location.

NUO383-2628A-BQ01-NLO4

123 TABLE VIII-6 (Continued)

VPI301 1. Analyze valves (VPI-301 and 302).

UPS2ND 1. Place cage like structure over batteries to prevent damage.

2. Place cage like structure over vent ducts and lights to prevent damage to the batteries.

WEAK-LINKS AT 0.1,5G PP06C 1. Replace threaded joints and victrolic couplings

& with welded joints.

PP05T

2. Current analysis techniques has shown that piping is much more resilient than previously thought.

Reanalyze.

WEAK-LINKS AT 0.16G M10013 1. Secure wall to the building structure.

2. Structurally reinforce the block wall.

3. Reassess the seismic fragility of the block walls using nonlinear dynamic analysis. (Computech Engineering Services proposal, February 10, 1986).

4 Relocate the affected equipment and cable to a more suitable locations.

M100UPS Fixing M10013 will take care of M100UPS.

WEAK-LINKS AT 0.20G PPYARD 1. Replace yard piping with non-cast iron piping.

2. Install a redundant fire system for the yard pipe section.

PPSWCW 1. Provide an.slysis.

ATWS 1. Install larger diameter core support plate alignment pins and bolts.

NUO383-2628A-BQ01-NLO4

APPENDIX A FAULT TREES NL'03e3-2628A-BQ01-NLO4

m, , -

ENCLOSURE SPRAY FAILURE - LOCA FAULT TREES 17 Pages NUO383-2628A-BQ01-Nt04

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e APPENDIX B SEISMIC FRAGILITY OF BIG ROCK POINT CORE ASSEMBLY AND REACTOR VESSEL SUPPORTS NUO383-2628A-BQ01-NLO4

ADDENDUM EVALUATION OF ALTERNATE SHUTDOWN BUILDING MODIFICATION NUO383-2628A-BQ01-NLO4

EVALUATION OF ALTERNATE SHUTDOWN BUILDING MODIFICATION OMISSION /

During the review process, it was determined that certain pieces of equipment /

associated with the Alternate Shutdown Building Modification had not been /

included in the analysis. This omission has been evaluated and it has been /

determined that there is no need to reanalyze the fault trees and the event /

trees. /

The Equipment that was not included in the analyses is an follows: /

1. Alternate Shutdown Building and Equipment Located Within /

This building was designed and constructed to be seismically resistant. /

The limiting groury seceleration has been determined to be 0.6G a great /

deal higher than the SSE of 0.12G, therefore, this omission has no /

impact on seismic resistance. /

2. Underground Conduit from the Alternate Shutdown Building to the /

Equipment Lock /

This conduit is buried 6 feet underground and was designed and installed /

to meet seismic requirements, therefore, this omission has no impact on /

seismic resistance. /

3. Conduit from Equipment Lock Area to Equipment /

The conduit runs around the perimeter of containment and is in no danger /

of degradation from falling objects, therefore, this omission has no /

impact on seismic resistance. /

i 4. Block Wall M10023 /

Conduits from the Alternate Shutdown Building to the control room /

penetrate this wall. No seismic resistance analysis has been performed /

on this wall so the most limiting block wall ground acceleration has /

been assigned (0.12G). /

Of the above omissions, only block wall M10023 is considered a threat to /

the structural integrity of the equipment in question. /

In all fault trees associated with the omission (MSIV EC-VALVES, /

EC-MAKE-UP), block wall M10023 is a single failure. That is, it alone can /

fail the above mentioned equipment. Therefore, replacement of the above /

systems' contribution in the existing cutsets with M10023 produces a list of /

cutsets that were omitted because of the error. /

i 4

NUO383-2628A-BQ01-NLO4

Below is a list of new cutsets developed by replacing the above failures. /

LOSP18 1.20E-01 M10023

• RDSCV2ND /

LOSP17 1.10E-01 M10023

• CRA':E3T

• N07051 /

LOSP17,26 1.20E-01 M10023

• M10021

• VP1301

• M10020 /

LOSP17,26 1.10E-01 M10023

• M07070

• N07051 /

LOSP3 1.10E-01 M10023

• IST-7

• IST-9 /

LOSP3 1.10E-01 M10023

• PS6362ND

• IST-7 /

LOSP18 1.10E-01 M10023

• PS6302ND /

LOSP18.27 1.10E-01 M10023

• RDSCV2ND /

LOSP26 1.20E-01 M10023

• VP1301

• M10020

• M10021 /

LOSP25 1.20E-01 M10023

• VP1301

• M10020 / '

LOSP16 1.20E-01 M10023

• VP1031

• IST-6 /

LOSP1 1.10E-01 M10023

• IST-8 /  ;

1.10E-01 /

LOSP17 M10023

• CRANE 3T

• N07051 LOSP2 1.10E-01 M10023

• IST-8 /

M10023 is the limiting component in a few of the above cutsets. Therefore. /

it should be considered as a weak-link in the seismic issue and should be /

addressed. /a analysis was never performed on this block wall because / >

important equipment was never affected by its failure until the Alternate /

Shutdown Modification was installed. By conservative assumption, the lowest /

ground acceleration determined by analysis for all other block walls at Big /

Rock Point was assigned to this wall (0.11G). Since this analysis shows the /

block wall's importance, M10023 will be included in the block wall /

reanalysis, associated with the proposed fixes for block walls M10018 /  ;

and M10021. / '

i P

i r

r I

[

l NUO383 2628A-BQ01-NLO4 4

Below is a list of new cutsets developed by replacing the above failures. /

LOSP18 1.20E-01 M10023

• RDSCV2ND /

LOSP17 1.10E-01 M10023

• CRANE 3T

• M07051 /

LOSP17,26 1.20E-01 M10023

• M10021

• VP1301

• M10020 /

LOSP17,26 1.10F-01 M10023

• N07070

• N07051 /

LOSP3 1.10E-01 M10023

• IST-7

• IST-9 /

LOSP3 1.10E-01 M10023

• PS6362ND

• IST-7 /

LOSP18 1.10E-01 M10023

• PS6362ND /

LOSP18.27 1.10E-01 M10023

• RDSCV2ND /

LOSP26 1.20E-01 M10023

• VP1301

• M10020

• M10021 /

LOSP25 1.20E-01 M10023

• VP1301

• M10020 /

LOSP16 1.20E-01 M10023

• VP1031

• IST-6 /

LOSP1 1.10E-01 M10023

• IST-8 /

LOSP17 1.10E-01 M10023

• CRANE 3T

• M07051 /

LOSP2 1.10E-01 M10023

• IST-8 /

M10023 is the limiting component in a few of the above cutsets. Therefore, /

it should be considered as a weak-link in the seismic issue and should be /

addressed. An analysis was never performed on this block wall because /

important equipment was never affected by its failure until the Alternate /

Shutdown Modification was installed. By conservative assumption, the lowest /

ground acceleration determined by analysis for all other block walls at Big /

Rock Point was assigned to this wall (0.11G). Since this analysis shows the /

block wall's importance, M10023 will be included in the block wall /

reanalysis, associated with the proposed fixes for block walls M10018 /

4 and M10021. /

NUO383-2628A-BQ01-NLO4

s SPA 13703.01 J

l

. SEISMIC FRAGILITY OF BIG ROCK POINT CORE ASSEMBLY AND REACTOR VESSEL SUPPORTS

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prepared for l

CONSUMERS POWER COMPANY Jackson, Michigan l October, 1985 1

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SMA 13703.01 4

I SE!SMIC FRAGILITY OF BIG ROCK POINT CORE ASSEMBLY AND REACTOR VESSEL SUPPORTS

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I by R. D. Campbell R. W. Warriner R. Sues prepared for CONSUMERS POWER COMPANY Jackson, Michigan October, 1985 I

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