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July 10, 2003 -A

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OPENING REMARKS., ...

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Craig Anderson

Vice President, ANO.. ~ ~ ~~~~1 . A

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~~~~. e; INTRODUCTION

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. Ir I . .Sherrie Cotton -I .

Director, Nuclear Safety Assurance. -1I I I-

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AGENDA..I t-~~~~~~1

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Opening Remarksr

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Introduction

Craig'AndersoriVP, -ANO-

Sherrie6CottonDirector, NSA.,

Dale James 'Manager, EP&C

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Risk Assessment Methodology

Fire Modeling I .

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Bijan' Najafi ' 'SAIC Analyst,.

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Break I � . ,

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Probabilistic Safety Assessment

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Overall Summary

Jessica WalkerPSA Engineer'

Joe KowalewskiDirector, DE' -

Craig AndersonVP, ANO,--

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Closing Remarks,.

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Dale James

Manager, Engineering Programs and

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* NRC ConclUsions.I

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-ANO's reliance on manual actions in lieu' ofproviding separation'design features is in violationof'Appendix R ; .

-"'ANO's strategy for implementing manual actionsinadequate . /

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Risk Assessment Overview,; ...

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* NRC's preliminary SDP evaluation concluded'!unacceptable (greater thangreen) increase in core

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damage frequencyKey assumptions in NRC evaluationspreliminary assessment''

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in-depth assessment. .

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Risk Ilnforriied Stgrategy for Zone 99M- K . 8'."

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Fire ModelingMo4 it •OIA;d.

99M --I~~ . .,,Simulator;Scpnarioi - Development

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_ . . r CircuitLAnalysis &}i , .'1 L Location Evaluatibn

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Simulator-Exercise .4

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[, HRA DevelopmentI

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Risk Assesmet Compatr -eon.',. ~ ~ ~ ~ ~ ~ ~ f _ .) ..

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* 4250 F cable failure

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* 700dF cabI.temperature

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* Zone.wide prompt:damage

* Generic HRA- Based on zone wide

prompt damageIncluded L@OP

Greater than Green

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* Green. findingfindi~ng

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Bijan Najafi'i C.

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Risk informeddStrategy for Zone 99MC * * .- . .. 99M... 1. 1 .. I.1 .I

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Simulator Scenario! Development.,

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Circuit AnalVsis & sLocatiodEvaluation ',

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CrSimulator.,Ex ( '

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'HRA Development.

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Summary,

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In our analysis we will show that: .,

- Damage to equipment and instruments needed for safeshutdown will be limited to portions of the. room '

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- Failures will occur over a period of time,. and- No credible.fire can be postulated that leads to zone-wide

!darmage. '1.

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Fire Modeling in the 4KV.Switgear Room(Fire; Zone .99.-M)

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Unit. 1 4KV switchgear room (fire zone 99M)I~~~~~~~~~~~

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* Fire characterization I .-

Fire modeling, evaluation of the consequences andt. .-. 1.

timingI.

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of the'fire scenarios,. ;....

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Unit 1 4KV Switchgear -Room (fire zone 99M)EC204 C204 EC204 EC205'E E203 BE202 EB202 EB202 .EC236 EC236

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15',Unit 1 4KV Switchgear Room (fire.zone M)/ . . . . . .I . . - I -. . .5

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Unit, I 4KV Switc'hgearom(ie. zone 99M)'-\ . ~.(I

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Typical ANO switchgear cabinet wiring,control cubicle

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Fire Scenario Selection':,G2nera.I Annroach

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Fire'scenarios define pot'ential'ranges of damageby.a fire- They define sequence and timing of failures, i.e:,

equipment and, instruments .- Ensure that risk-significant'failure. sets are identified

Considerations for selection of fire scenarios .. '- Location of critical cables in the room- Potential characteristics. of the fire sources. located in

the zone, thermal and high energy --- Configuration of the combustibles in the room ... .

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Fire Scenario Selection':..~ ~ I

General ApproachI .C. * .

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Three distinct fire scenario classifications: . -

- An electrical fire (non-energetic) in any of the electrical cabinets inthheroom ' * , ' ,

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*Fire may spread in the cable trays, but requires considerable time* Circuit damage/failures follo\ a time-phased sequence with first damage

after 10 minutes . .- A high energytarcing fault switchgear fire that may. initiate secondary

fire, - .

* The event has an' initial (immediate) pressure phase that causes damageto targets. anhd ignites exposed dables injthe vicinity ,. .' :

* The fire may continue in the switchgear and'grow within the,ighited.combustibles (e.g., cable trays) in the vicinity L.

There is an initial/immediate circuit damage/failure fo~lowed by potentialtime-phased circuit damage/failures: *. ( '

- A transient fire that may spread into cable traysA transient fire between B55 and B56 was selected as the nrraximumexpected scenario due to its' potential for extent and timing of damageCircuit damage/failures follow a time-phased sequence'with first damage-after 10 minute '

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Fire Scenario Selectio'n: ' 19

Scenarios.-Modeled in Zone 99M.*~~~~~ ~ ~~~~~~~~~~~~ . * *

Eight fire scenarios selected.'represent credible fire risks for99M - .. ;; ..

Scenario a - Fire in A4 switchgear.- Scenario 1 b - High' energy fire in A4 switchgear .- Scenario 2 - Fire in the B55 motor control center- Scenario 3 - Fire in the B56 motor control center- Scenario 4 - Fire in the Y22 inverter '- Scenario 5- Fire'in the B6.load center .- Scenario 6a - Transient fire between B55 and B56 below three

stackA.tray '- Scenario 6b - Welding/cutting fire between B55 aind B56 below

three stack tray . . . .,.,

Illustration of these scenarios .is. provided in the' atta.ch'ment.tothis presentation -: ' .

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Fire C haracterization'. . I.1 *..> 2i,

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* ElectricalIcabinet fires- The heat release rate-'data profile is

based on the b-6st available fire testdata

* Sandia National Lab (NUREG/CR-4527,87/88) and VTT (Valtion Teknillinen

'Tutkimuskeskus, 94/96) in Finland* Same test used in the NRC SDP analysis

- The ANO HRR is based on the highestpeak of ST5 (unqualified, open 110KBTU loading) and all qualified, vertical'cabinets (excluding PCT6 and test 23.with'1.5 MBTU loading).'

* 'the NRC HRR is based on test 23(qualified, open 1.47 MBTU loading) and

*test 24,(unqualified, open, 1.44 MBTU)- Time-to-peak is based on the average- Tests are based on"control panels- The switchgear, MCC's and load \

certters; are enclosed with sealedpenetrations ;

* Used for'scenarios 1a, 2- 5

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Fire Scenario Selection:NRC and ANO SDP Analyses;

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* .~ -) :* NRC SDP firescenarios , , :- Based on fire size

* Total room heat-up and zone-wide damage- Electrical cabinet and electrical equipment'fires

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ANO SDP fire scenarios- Based on source and target-set ch'aracteristics and'

configuration ' . ,* Local as-well as zone-wide damage

- Electrical cabinets;.in zone',- High energy arcing faults in the 4KV.switchgear.(a "bNey

design basis event"), . . ." , ,- Transient fires including hot. work.' .. .

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Cable tray fire heat release rate: - 0.45 *qb' (-AQ Ct''" b I .. .

Widely used model fromr ',Society of Fire 'Protection Engineers (SFPE),handbobko'*.

Used for scenarios that include ignit~ed cable trays

Transient fires: 150KW V -

- Typical refuse based on fire tests at SNL/LLNL documented in EPRI,Fire PRA Guide " ' ' .-.

* Used for s-cenarios 6a and 613 ' . ' I -g S ,, ; ' , . ' -. ; . 'n i'~~~~~~~~~V 7. -.. ' , ' .

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Fire Characterization (cont.) . ..''High-energy Switchgear Arciig Fire.

The damage/ignition zodne of the initial pressure ,hase is derivedfrom US nu'clear experience (next slide) (EPRI SUI 05928 Supp toEPRI Fire PRA Guide)- . . '

* En'suing electrical cabinet fires (the switchgear or, others exposedto its arcing fault) follow th6 same behavior as the non-energetic ,electrical cabinet fires ' ' .

* Potential ensuing cable fires spread horizonta'ly and-spread faster. . ,vertically thIrough cable tray stacks' . '

* Observations: . . ('; . , , -

- Experience of the US nticlear industry indicates that..damaging/severe switchgear fires tend to be of the energetic arcingfault type , .,

* Used in scenario 1.b . . " ."

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Fire Characterization (cont) .) ,-ZOI of the High-energy Switchgear ArcingFire

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Firet Characterization:NRC and ANO SDP Analyses

Electrical cabinets

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- NRC: 200-500KW,'peaking in 105s'econqs .. I

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- ANO: 100KW, peaking in 12 minutes

High energy fires in switchgear;- NRC: Assumed covered by the range of HRR

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- ANO: Empirical model based on 'experience' (previousdamage/ignition within: five ft.

Transient fires . .

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- NRC:.Out of scope .. I; 1,- ANO: 150KW, peaking in 1.0 minutes

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Fire Modeling.Model. for.Preed-iction of Fire Growth"

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Hand calculations, used to calculate timedamage . ,

- Target immersed in flame- Target in..the.fire plume

Target in-the, ceiling jet, and -

- Target'in the-flame radiation zone -^

to localized target'; . . E -

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* CFAST used to' calculate room temperature and target;damage/ignition due to h6t gas layer ,,

- CFAST and'simple correlations such as Heskestad, are validated and.widely used for the range of fire conditionsexpected in Zone 99M

* Cable fires: Used fire tests for both growth through stacks andhorizontal cable tray. ) - ..

- An empirical model used to determine the extent and timing of thespread through the stack (based. on SNL tests documented inNUREG/CR5384 and in the EPRI Fire PRA Guide'

- Ten linear ft/hr is the generally used available model.f fire'-propagation along horizontal cable tray (EPRI NP 7332). ' - '

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U~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~Fire Modeling:Target Da.nagel.gnon. .Tar~get Damage/lgfiition

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Targets (cables) are considered damaged or ignited.when their surface temperature reaches 700OF

-, Thermoset insulated cable predominantly used in' the plant asverified through original and current design and installationspecifications ' 's

- Thermoplastic' insulated. cables are not used in ANO Unit 1',high risk zones ,. ''

* This is the critical difference between the NRGC and ANbanalysis as it relates to the extent and timing of firp damage.

- 425.F vs 700.F . . _

- Critical to extent and timing of damage and fire growth

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Assumed cables inside metal conduits daMragqd. atsame critical temperature, but will not contribute toroom heat-up -

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High-ienergy arcing fire,'- Assumed raceways and cabinets in the zone-of-in,'fluence are

damaged with exposed cables (trays) ignited- Assumption conservative for the conduits (if stainless or

galvanized steel pipes) where they are likely to survive theshort-lived (seconds) initial pressure spike. i '. I

* Spurious operation 'of damaged circuits were modeled.In some cases, the likelihood of the spurious"actuationwas obtained from the EPRI. Expert Elicitation r'ep~o't?'(EPRI 1006961) which' was estimated in part based -on.the data from EPRI/NEI circuit..failure. characterizationfire . te sts .. .*.

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A high-energy switchgear fire (scenario l b) is the maxi 3urn ex ected,.fire scenario- Initial HE -phase could lead to ignition' of as much as 3 f cable tray- After the initial HE phase, ensuing cable fire may grow although at a very slow.rate,

* -~~~_I

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* The floor-based sources of fire in fire zohe 99M are electrical ca6inets andtransients' . -l - i'

- The likely location of electrical cabinetfires (flame) is below 5ft off the floor.once thebreaker cubicle is open-'in the high-energy event ' .

- The floor-based fire intensity needed tb generate damaging (7000F) HGL'is-2MW. .I r XA a a_ 0 _ * ..

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- inone o Tne Tloor-oasea rires'are capaoie.ot sucn'intense neat, 11 I I-Only cable fires are potentially capable of generating. such intennsity'if enough cableis involved.

- Cable'tray fires are elevated fires (none o6the cable trays in fire zoih 99M are.located below the 8ft doior opening) '

- Cable fires are expected to be irn the smoke layer once the smrok6 layer reaches thetop of the door. Once in the smoke layer, intensity of the, cable.fire will be controlled'by the oxygeriavailability,'which is not enough to sustain the combustion process

* With an elevated cable fire that grows at a rate of 10 linear fWhr as input.'- Th en depletion occurs very quicklv. re ardless 'of operi-or. closedddoor

The cable fire does not grow beyond the initial 12 f . - '.- The temperature peaks at 5500-535 0F .- The fire has to be below the settled smoke layer (4-5 ft) for the cable' firesto continue

to grow - X

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The limiting firo scenario,. one th't can generate a.drmaging'HGL, ;is not 6redible ' v -v - A,. -

The non-supprqssion probability by the brigade for very long duration.cable fires (100. minutes for the high-energy switchgear event); is501.01 -(per EPRI Fire PRA Guide) Y . ,, , , .

- Fuel depleti'on; cable ignited earlier have burned out ,- Parts of the cable trays are coated with flamrastics which both delays.

ignition and slows propagation of cable fires ,- Continued growth of the non-pildted cable fire for a long time is not

likely. (Tests reported in NUREG/CR-5387 state that bable fires.spreading horizontally only as it progressed from level .t'0.!evel"),

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* Maximum expected fire is a high-energy switchgear fire

No credible fire reaches 7000F in this room (limiting firescenario) ' . .. . (' -

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Results (cont.).. --Comparison of NRC and ANO Results

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* Damage threshold- NRC': 4250F ..

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- ANO: 700°F

* Heat release rate

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- NRC: 500KWfire peaking in 105 sec'.

- ANO: IOOKW peaking,in 12'min (Scenario la) + cablefires and high energy fault in A4switchgear and cable fires(Scenario lb)

* High energy arcing faultin th~e 4KV-switchgear

- NRC: Not-analyzed- ANO: Limiting scenario in terms

of its consequence, i.e, affectedcircuits a'nd timing

* Neither analysis reaches 700°F

." UpperLayerTemperaturein99M :

800700600 - -~500

u- 400 -120O0 ->

0 250 500'. 7590 1000Ti'Je sc --p

*' -.- ANO-ULTerpScenario.laopenDoor

' _4-NRC ULTe np '

-- AN0JL Tenrp, Scenario lb, Open Door

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Scenario Frequency is derived from multiplication of:- Generic fire frequency. ,

* Based on the EPRI FIVE metho'd (EPRI TR 105928 page 4-7)- Severity Factors,

* Based on tyoe and location of fire (EPRI TR 105928) !.

High energy weighting factor for the64KV switchgearBased on operating experience (EPRI fire data base)

- Prompt suppression of transient fires by plant personnel or fire W"Aatcc'* Based on operating experience (EPRI TR 105928, Appendix K)

- Manualsuppression by fire brigade -' i* For scenarios that critical target is beyond plume, ceiling jet or flame,

radiation zone

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ANO SDP Analysis Results4. -

0 ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~We IoofH Pns,by*. ~~~~~~~~~~~WIFloor~~.Sourc. ~~~~Generic (location Pg ~ n area ratio everiy evn fra pln PsbU Frequency weighting - source severe personnel fire Results. | Frequency | weightrng weighing (transient Factor U o r g

________ _____________________________ ______factor) factor) Fires) fire ivatch1 a .Fire in the A4 sv~tchgear.. _

Nominal valuA. 100 KW fire 1.50E-02 2.5E-01 5.88E-01 _1 .E+00 1:20E-O1 2.50E.41 1.00E+00 1.00E+0 e.62E.O5lb High energy arcing fault In any of .

the A4 Owitchgear breaker cubicles ..- . 1.50E-0; __________________ 2.50E-01 5.88E4O1 .00E+O0 JE.1 7.50E41 AQooEv0o 1j EQ9 1,gE.04

2 Fire In the B55 MCC. Nominal . .100 KW lire. Fires In Inverter Y28are bounded by this scenario.._____.__ 1.50E-02 2.50E-41 5.88E-02 1.00EW00 1.20E-01 1.OOE+00 1.00E+00 1.OOE+00 2.65E405

3 Fire in the B56 MCC. Nominal0.0 KWfire - 1.50E42 2.50E-01 5.88E-02 1.00E+00 1.20E-1 1.00E+00 1.00E+00 1.00E+00 2.65E-05

4 Fire In the Y22 Inverter. Base . . ..case. 100 KWVire. Fires in Y24:and Y 25 are bounded by thisscenario. 1.50E42 2 588E-02 1.00E0 1.20E-01 1.00E+00 1.00Ek+W 5.ooE-oi 1.32E4S5

5 Fire In the Load Center 6... . .* -

__ _ Qotnlmnua!1RR. 1 .50E402 5.88E-02 1.00E4 W 1.20E-01 1.00E*00 1.00E+00 . 2.00E-o1 5.29E-066a Transient fire in areas of the room .

where cable trays are exposed to .a floor-based fire. Nominal Valueof 150KW. 360E-02 2.00E+00 1.80E2 1.OOE01 .OOE+00 1.00E00W 5.00E-41 11OOE+00 6.48E45

6b Cable fire caused by welding and .cuttln In areas of the room where .cable trays are exposed to a Ploor: .

based fire. Nominal Value 6f_15OKW. 1.30E-3 2 ooE+00 2.00E402 1.OOE-4 1.OOE0 1 .OOE+00 5.oOE-02 11.00E.+ 2.6tE.07

NRC SDP Analysis Results (May 15,.2003 Supplermhental.Letter Page 25).Source : r , Frequency

Electrical cabinets ' . . , 2.3E-04

Transformers , - , , . .1.6E-05';-

Ventilation Subsystems . . / , . I - 4,4E06 t.;' _ .____ ! ~ ~ ~~,4t.0.

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Fire Modeling Summary

Maximum expected fire scenario in firezone 99M is a high , |energy switchgear fire -.

- Immediate damage caused by high energy event will be ^ -

limited to portions of the room- Followed by time delayed failures caused, by secdndary cable

fires ,

Credible fires will not result in a hot gas layer (limiting fire ;scenario) in excess of the cable failure temperature

- Zone wide damage is not credible 's ;- Adequate margin

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Introduction

* Key system failures in 99M .

* Affected components due to cable failure

* Key. operator action/response

* Simulator scenario/results-'C.~~~~~~~~~~~~~~~~~~~~~~~~~~~~

* Operator action probabilities.

* CCDP calculation t

* Delta CDF determinationI~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

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Risk Info-rmed Strategy for Zone 99M'i_

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Fire Modelin"g 99M l

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Key.Systems-Affected in the Risk- ...Significance Determi~nation (Fire'Zone. 99M)

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The following systems/trains. are directly faiied due tofire induced poWer losses.of A4 and B6

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- One train and.the swing pump of service water-,- One train and the swing' pump of HPI (makeup)-. The-A4 associated diesel is. no longer usabld'.

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Detailed. circuit 'analysis peiformed on zone .99M 'S

* Investi'gation'of cables'located in the trays-and conduits,associated with the taraet sets '

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-4~ Analysis showed no loss of offsite power associated,.withzone 99M ' . . . -.

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- NRC evaluation did use loss of offsite power . I-. ;- i -

* Analysis of associated failure modes for affected. - .. ~~~ ~ ~~~~ ~ ~ ~~~~~~. ,I .A cables4, I

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* Failures unrelated' to safe shutdown also examined/toprovide accurate portrayal of the risk'caused by th'efire

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Systems Affected in the Risk-Significa ipcer.Determination (Fire Zone 99M) v

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Scenario specific failures are based on cable focation;subsets of the following are impacted fo'r eah 'ch'enario

- EFW flow control valves ..* Loss of power t6 these' valves will fail them open (desired state)

Subsequent spuriobs operation riot problable, -

- DC, control power to the A3 switchgear fails* Breakers to remain static and require manual closure at the switcahgear'

- P-7B (motor driven EFW'pump) suction valve could spuriously close* Cable in conduit; spurious operation not probable but assunied in

evaluation ''

- Steam admission valves for P-7i (turbine-driven EFW pump) ..

... Requires local action to start P-7A, ' ,-- Aux-lube oil pumps for the unaffected HPI train

Requires local start of HPI pump when affected, .. . .

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Operator Actions/Response in the Risk-SignificanceDetermination (Fire Zone. 99M) -2 l

* A subset of the following operator actions are rewquired in eachscenario ; - .

Starting turbine-driven EFW pump P-7A manually and the;positioning of its associated valves ; - .':*

- Controlling EFW (A or-B) flow to prevent overfill )

- Local closure of A3 switchgear breakers for P-7B and HPI A.'- Stafting HPI for-make-up (long term action) ;

May require local start of pumps depending on fire scenario . .

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*Emergency diesel generator rdcoveries were not ne'cessardue to the lackof a LOQP event -. ,. .r

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Previous Procedures vs Zone SpecificProcedures. - , -

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- Combination of EOP/AOP/Pre-Fire plan- Opportunistic approach

- Plant condition determines .action

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New procedures- Zone-specific fire procedures -- Tactical approach- Reduces impact and probability of spurious operations

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Summary of Procedural Guidance .I ..

# -ii . y.Key-Action' :.- ,,P ous Pcedures'!.Sf.5 '_The previous procedures discuss this in great , --

Starting EFW P-7A detail. SpUrious and false indicators are not Discussion in new procedure includes1 manually and positioning mentioned which could delay operator- . , . ~~~~~~~~~~~~functional indicators. v.

associated valves response. '_.

prodredscss soca o Lack of adequate and'borrect indication,isControlling EFW (A or B) Previous directly discussed in the new procedure"

2 to prevent overfill control room action. which makes {his actionmore likely ir).the

____________________________________ ,new procedure. -,Local closing of bus A3 This action hot explicitly discussed in the P '

3 switchgear for P-7B and normal operating procedures but is discussed The new procedure~explicitiy addreses,HPI A (e.g., inverter fires) in Alternate Shutdown. , locally closlngthese breakers. v

Discussed in previous procedures. .,e timing -

4 Starting PID Makeup of this action depends on when letdown is The new procedure addresses the4,Starting isolated. possibility of starting theHPI pump locally.$

Isolation of letdown.to In both the previous and new procedures; this In both the previous and new, procedures,''5 avoid needing HPI action is discussed and can be perforirned in this action is discussed and can be

(Makeup) sooner , the control room. perfbrn-ed in the control room.

Switch to recirculation In both the previous and new procedures, this In both the previous and new procedures,6 action ls'discussed and can be performed in this action is discussed and can, be.

long-term cooling, the control room. performed in the control rgorn.

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Fire damage chosen to provide HRA information for multiple ''operator actions" .

.,- Fire model.beginning with an A4 switchgear fireFirehlo n g ingwie-predanb

- Fire p.ropaga'ted throughout zone causing.wide-spre~ad cable,/ damage', .' .

- Damage for scenario exte'nds beyond credible.fires'- . ~~~~~~~lef~1. I a s

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* reaiistic control room communication cnaiienges-, Fire brigad6 leader.communicati'on ' ,

Timelines based on actual fire drill .

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- In plant auxiliary operator used f6r operator.actions* Radio 6hdtelephone communications used

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Simulator scenario failutes includbd:- Directfailures

* A4 switchgear (4KV)* B6, load center (480. VAC) ,* EFW flow control valve power failure*; HPI auX-lube oil pump power fairure

- Included spurious operations *'

* P-7B EFW suction valve closed at T-1 5- Included failed and incorrect indications

Multiple panel indications failed (EFW, HPI, P* Random annunciators spuriously alarmed, .

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Simulator Scenari6 for Zone 99MfI4-.- *i

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* Three crews ran simulator with previous, procedures.* Two crews ran simulator with training on zone specific.fire

procedure .*. . , -' 'CA-' *

* One crew With each. procedure. contained operators n, theplant simulating' local actions - ' /

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Controll'ers were present in, the field to evaluate localmanual actibns' . " '

* Time toJocation. . ' ,

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* Potential hazards* Communication barriers

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* Observers in the simulator to evaluate control room* Including time to perform in. control room actions

Procedure usage.* Work practices due to loss of indications a

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Risk Informe'd Strategy for Zone 99M" I . .... I >e II

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Simulator Scenario Results

Simulator runs using previous ESPIAOP/ P re-Fireplan approach and unrehearsed crews (3):. .

- EOP.approach for plant trip provided adequate initial core cooling- Pre-fire plan used to show-faulty indications and possible local'.

actions- Crews responded appropriately and, in a timely-manner- Plant maintained at a safe stable state

* Simulator ruhs with crews trained on new. tacticalprocedure approach (2) /

- 'EOP for plant trip still used until fire confirmed- Using new procedures', crews directly implemented local'control of'

core cooling , ,- Plant maintained-at a safe stable state

* Crew performance using either previous or new'procedures met Appendix R requirements forachieving safe shutdown

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Risk Informed Strategy for Zone 99M. * ' 5O~~~~~~~jI ,

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Fire Medelinj-99M

LEO P/AOP/Prefire Plans] -

Simulator ScenarioDevelopment

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L circuit Analy~is &F.oato Evaluation.

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Simulator Exercise 94f: . I .1T .

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Manual Actions are Reliable ,. It! 51I . . . .I . . .1

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HRA methods for-quantification demonstrate there is animpact of fire on reliability of human actions

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Previous vs' new procedures for shutdown I4C

- 'Previous procedures use an opportunistic approach to control,, ewhere crews respond to cues and symptoms by selectingEOPs for that condition with the'aid of pre-fire plansNew procedures assist crew to. respond using 'a more tacti(control process " ' ' * '.

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*Use. of dither approach,'demonstrated ' '''' 4. 7 I

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- Identifying'symptom orx cue will generate appropriate response-'for either procedure ' .

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- Ability to recover from spurious actuations .* Enhanced in) new procedures , I

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Method for Updating HRA Assessments toA ....... s s - ___ _ ................ I . . . .. ..

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The current Unit 1 model for human recovery actions. in-events PRA is based on a time reliability curve .'

internal( .. -;.

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HRA~accounts for operational context by adjusting. param'eters.such as: ., , , , , , . , ,, , , ,

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- Other performance influencing factors,

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,In current assessment, effects of fire are not addressed 'nr ayre.model param'eters available; therefore, a different ajdjustment-method 'iXas. identified' ',

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* inciuaes several metnoas tor quantmicationI - -1. . i-.C- . �- - -1 K I rl%.41% - - .- - - - - -1

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Evaluation ofFire Impact on, Probability:.,. , ; . . .. Do ~~~~~~~~~~~~~~(.4

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Seven cognitive assessments, on differences in procedures,..- Availability of information . ,. ,,- Failure of attention' .. , .

- MisreadfmTiscomrnunication data .- Information' misleading . ,. -. .- Skip.a step in procedure- Misinterpret instruction . .

- Mi'sinterpret decision logic.. - . ..

* Probability of execution also calculated for fires based-o6n inpu,the HRA calculator x -..

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Case Event ID Basic Event Description A Pcoa A Pexe A HEPf~reI FIREOLDP Actions are carried out within the*. 9.8E-03 .7.750E-04 1.1E-02

control, room.- prevdiaus , .

*2 FIRENEWP -Actions are carried out within the , 2.6E-03 6.1OE-04 3.2E-03control room - new t

\ , . . \ s................ . : t ... . . ' *

3 -99-MFIRECR Realistic fire In 99M decisions in 9.8E-03 2.OOE-02 3.?E-02control room with local manual .. E...actions . , . ..

4 99-MFIRECRE Realistic fire in 99M early control 4.7E-03 , 4.3E-04 5.1 E-03._______ ._________________ room actions . ._._.

5 99.MFIRELOCAL . Localactions taken by field operators 1.5E-02 2.6E-02 4.1E-026 Not Feasible - -1 1 17 No Change . : 0 .. . 0 .0.

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Comparing HFEs from PRA baseline withHFEsin99M fire r. . N .r-

Fire in 99M increases human failure event (HFE) for typical feasible actions overinitial internal events PRA from zero to a value in range..of ,E-3 ,to 4E-2 for variousscenari6s and conditions. - , . -

* If action is not feasible, then IH1FE assessment is set at 1.0

* Very small difference in impact of previous versps new procedures.

Comparison of previous and new procedures on the HFEs for fire Impact In 99M

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Human Error Proba'bility, Comparison..I I~~~~ I

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* NRC approach assumes zone wide damage at time zero,I -

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* NRC approach included loss of offsite powerII ..

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Operator Actidn NRC Value No NRC Value . ANO Value . ANO Value-, Procedure W/Procedure Previous New

Establish EFW . .0.6. 0.11 0.,098

(A3 local start)

Establish EFW 1 0.6 . 0.038 0.026

(Control EFV) . .

Establish Feed 0.75 ; 0.55 0.008 / 0.008

&Bleed . . .'

Establish Feed 0.75 0.55: 0.11 . 0A098& Bleed

(A3 Local Start) ..

Secure Diesel . 0.75 0.55 Not needed due to no loss of offsitewith no'Service power,

Water . . . .. . - *

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Risk Informe d Strategy for Zone. 99M..,

;i58

L Fire Modeling 99 M - .Simulator Scenari

Development

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Simulator Exercise, 1 I I Pro(Devel

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Analysis &i' Evaluation

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CCDP Determination for Zon'e 9.9M.),. P ~~~ ~~. , z ., *.- . - t '

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Eight fire scenarios in zone 99M quantified .,

- Current Unit 1 PSA model used - .: :

* Fire modeling targets used to determine failed components.-

* Spurious operation probabilities used in high-energyelectricalfaultscqenario1b . ... . . -

- All other scenarios c~onsEivatively assume the spurious operation will.occur'.

* All components,.failed. together (conservative) ., K ';Timinrgs opily used to disallow spurious, of

puiosoperatibr o qompqnentswhose control cable would be lost after power los ;

* HRA values for the previous and new procedures used to recoverthe- baseline CCDP values for 99M . . . .. . ' " -

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SDP Process Review. i :* '' 60

*Created. eight fire scenarios . .

*Used fire modeling /characterization.-Determined failures'for' each scenario .*

*Used 'simulator exercises and in dustry-experts to'deterrrinereliability of -necessary -operator. actions I'

*Combined interaction -into -plant specific PSA mode'l ,'

-Calculated change i'n risk between the pre-vious and newprocedures 4.

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Fire Risk in Zone 99M*1 Ic

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Fire in the A4 switchgear. .la Nominal value, 100 KW fire 6.62E-05 3.12E-04 2.06E-04 '.06E-08 1.37E-08 6.98E-09

High energy arcing fault in any of . ' . , '.the A4 switchgear breaker . . . C.

lb cubicles. 1.99E-04 1.28E-03 9.01 E-04 2.54E-07 1.79E-07 '7.55E-08v ' ! ,. S ' ,.... . .... . . ),

Fire in the B55 MCC. Nominal ,100 KW-fire. Fires in Inverter Y28

2 are bounded by this scenario.' 2.65E-05 2.78E-04 1.79E-04 7.35E-09 4.74E-09 2.61 E-09Fire In the B56 MCC. Nominal.

3 100 KW fire. 2.65E-05 2.78E-04 1.79E-04 7.35E-09 '4.74E-09 '2.61E-O9. Fire in the Y22 Inverter. Base.

case, 100 KWfire. Fires in Y24 .* and Y25 are bounded by this

4. scenario. . 1.32E-05 .98E-05 3.86E-05 -5.27E-10 5.10E-iO 1.60E-11Fire In Load Center B6. /

5 100KW nominal HRR. 5.29E-06 3.02E-02 1.88E-02 1.60E-07 9.93E-08 -6.07t-08Transient fire in areas of the room .

where cable trays are exposed to , -.

a floor-based fire. Nominal value . *6a of 150KW. 6.48E-05 3.24E-03 2.12E-03 2.10OE-07 1.37E-07 7.25E-08

Cable fire caused by welding and * ,. cutting in areas of the roomh where . .

cable trays are exposed to a floor- I . . ' .Ibased fire. Nominal value of ..

6b 150KW. 2.60E-07 3.24E-03 2.12E-03 8.41E-10 5.50i_-10 2.91 E-10

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Time-phased fire-induced failures are aP~~~~s; 1 ,; . .. .. .. . ... ~~~~~~~~~~~~~~~~~~~~~ -

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Realistic assessment of fire progression, failures in 0 - 6OMm inutes (targets of the high-energy switchgear damnage,'immediate,'the rest time-dependent failures are from ensuing-cable fire) * , ',

Critical c"able insulation used is ,thermoset ...-7n~R 11_ 1__ _% 1_ _ I._ L . . ^_ .'

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- -UUt cauie. uamage temperature I

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Operator-action.probabilitiees I..~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ I .1 . .

- New procedures offer slight HEP imrprovement over prqviousnrn-'rtl Ia , - - % el .

VI- I ULI ~ .

- Human reliability'analysis: :CCDP indicates that'impact ofI AHEP im measurable but small I ' I

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I1 7Risk.Informed Strategy for Zn 9 9 M

Risk I 'nformed% Strategy. for Zone 99,M J, .. ;... ,"N1

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. , 63'.._ . _ .I . . . .

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L Fire Modeling 99NI- �1-1� � r,

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Simulator Scenario*Development-

.. CircuiitAnalysis &.4 > ,. , <-Location Evaluation

99M, ''

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[EOQP/AOP/Prefire Plans]7 S

-'Procedure . ,

I I Development ,.~ -. *, .99M JI 0

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oc'us~~A on. ,o h-e tht_* Focus on zonres that have delta risk due to the"diff rencein.manual'actibns between two types of procedures

* Qualitative, review of zones where manual actions are,utilized , " I. '

- Alternate shutdown zones screened- Zones with automatic suppression screened - :

Agrees with NRC SDP approach - Suppressiorn provides at least ohe' "order of,magnitude, in risk results and provides time for operator actipons tobe performed d ' .

- Zones'with MFW unaffected screened '

MFW greatly extends time needed for ,EFW local actions- Zones with one complete train of core cooling unaffected screened

* Control room operation of equipment removes impact of'local operatoractions ( S -. '

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* Similar to NRC results, the following zones remain;- 1OON . . . ''

- 104S' ' 5.. . . 55..

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The assessment of fire risk in 99M was extrapolated to twoother Unit 1-fire zones:

- Each'was assessed with walkdown and examination of the.''potential fire scenarios threatening the other train raceways..(e.g.- red'train raceway in a -green train room) ' .

- Onit 1A3 4KV switchgear room, (1 ON)',vSimilar to 99M in combustibles and fire sources

' Considerably less. redundant train cable' routed through zone.- Unit ' electrical equipment room (1 04S)

, Lack of high energy-switchgear / . .* Considerably less redundant train cable routed. through zone )

* Each zone is. bounded by the results of 99M- Conservative estimated fire risk (ACDF) for this condition

.* Unit 1< 6.6EP07/yr , '*~~ ~ ~~ ~~ ,,,-

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Unit 2 Risk. .

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* 'The four Unit 2 zones identified as risk significant by NRCwere ,qualitatively evaluated, ' /

* Each was assessed with'walkdown and examination of thepotential fire scenarios threatening theother train raceways.(e.g., red train raceway in a green train room)

(.5. I

* Conclusion I.I.(I

-The four, Unit 2 zones contain similar characteristics to the1 zones . - r .

Unit;5.

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* Two switchgear rooms .- .-, ? .. %VI

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n . .5

.

* 4. -

..

^Two rooms containing MCCs similar to, 104S

- The. results from 99M. bound. these1zones.. ~ ~~~~~~~~~~ . . -

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Summiary of Risk Assessment... I . I

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* ANO risk assessment concluded that: V

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- Realistic fires will not achieve whole-zone damage as,originally assumed in NRC evaluation . ;. ;

- Realistic fires will result in time-phased damage.of cables5 '.'- Manual actions resquired to achieve safe. shutdown for.a fire in "

zone 99M are credible , .

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Simulator scenarios validated that operators could achieve safe' _ -- .. S '.

shutdown.- Met Appendix R.requirements for achievingsafe.shutdown.;

lusion .I . .. , .

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OVERAIk SUMMA.sY

IOVERALL SUMMARY1:

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* Detailed analysis of ;zone 99M

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- Credible fires result in time-phased failures without-zone-widedamage (700 0F damage temperature for thermoset cables) -

- Detailed circuit analysis indicates there is. not a loss of offsite powerfrom any-fire scenario

- Simulator scenarios provided realistic data for assessment- ofoperator reliability in the use of previous and new procedures'.

- ACDF for 99M is 2.2E-07/yr . -

Total Unit Risk- Two additional zones considered risk significant for Unit 1 '-- Risk assessment of zone 99M conservative with respect to other

zones ' .- Conservative estimate of total unit AGDF is < 6.6E-07/yr

* The significance of the use of manual. actions torachieve. safeshutdown has very low safety significance and should be-characterized as-GREEN. fl ' _ yg 4

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- Explicit control.of combustibles- Fi~rebrigaae efectiveness '

* Primarily rely'on bar iers o'r physical separation for equipment'.;* ~required for saf~e shutdpown .'- ,; ''. . \........................

7 - 'Fire detection. and.suppression ...

- =mited use of manual action utilized for Appendix R. compliance

Actions taken to further reduce risk' ' .'.

- 'Validated circuit analysis-- Feasibility evaluationof manual actiors (IE 71-111.Q5).- New procedures developed to enhance operator response.- Fire. detection reliability improved ..

* ANO can' successfully achieve safe shutdown in the. vernt. of a firein any zone

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SAIC.-. I

SCIENCE - .

APPLICATIONSINTERNATIONAL CORP. -

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I ~ : . . - . , -1. .. , - W , , ,<BIJAN NAJAFI, P.E. - ' / s-MANAGER, FIRE PROTECTION SECTIONI

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EDUCATION: * '

University of W;shington. MS., NucleaiEngineering, 1!i9Shiraz University.. ' B.S., Electrical Engineering, 1976

.RegifteredcProfessional Mecharical Engijner, State of Califprnia

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I

SUMMARY OF CURRENT POSITION: ' . - I* .. *.* ., ,: I . J,j--. * -.-

Mr. Naja4 is thelManager of the Fire Protection Section at SAIC responsible for overseeing a program thatincludes domestic and international nuclear utilitiesIDOE facilities Andcomnmetcial/industrial facilities.

I .

)

EXPERIENCE:

Mr. Najafi is a nuclear engineer with over 23 years of experience, emphasizing Reliability, Risk Assessment,Fire Protection and Systems Analysis. His background includes development of methods for risk assessmentand fire protection as well as application of these techniques in solving plant-specific problems.

Mr. Najafi is the SAIC Manager for Electric Power Research Institute (EPRI) fire risk analysis and fireprotection program. Over the past decade he has been instrumentil in development of the EPRI fire risktechnology currently in use in the U.S. nucear power industry. This technology has also been usedinternationally in Europe and parts of Asia and South America. Mr. Najafi haslconducted training courses inU.S. and Europe on Fire Technology, ;nost recently a series of Fire Modeling courses for nuclear power plantfire protection engineers.

Mr. Najafi is an active memiber of the fire protection community. His contributions include:

* Principal member of the National Fire Protection Association (NFPA) Technical Committee on FireProtection for Nuclear Facilities (801/805) /

*.. Principal member of the American Nuclear Society's committee for the development of the Fire PRAStandards

* Participating member of various taskforces at Nuclear Energy Institute including the circuit failures is'suestaskforce in the development of the NEI-00-01, "Guidance for Post-fire Safe Shutdown Analysis."Invited panelist at the NRC-Industry fire-induced circuit failures workshop on February 19, 2003.

* Member of the NRC-Industry team for the revision, of fire protection Significance DeterminationProcess (SDP) _Member of the NRC's "International Collaborative Project to Evaluate Fire Models for NuclearNP6werPlant Applications." - -

* Member of the Society of Fire Protection Engineers (SFPE) task Groups for development of the, "SFPEEngineering Guide to Performance-Based Fire Protection," completed in 2000 and "Risk Assessment

..-Guidelines," in progress. - - x

% , N. l

* - SCIENCE .- . -.i SAIC - APPLICATIONS.INTEANATIONAL CORP. . .

a - '~~~~~~~~7

EMPLQYMENT HISTORY: ; - '

-' Mr.Najafi is the Manager.of ihe Fire Protection Program at SAICsesponsible for overseeing a businessarea..that includes domestic and international nuclear tilities, DOE facilities and commekcial/industrial facilities.

* . He is one otthe principal iniestigators for Electric PowerResearch Institute (EPRI) fire risk analysis and fireprotection proj&ts. These projects ipiuided development of'EPRI's Fire PRA.-Implementation Guide and

*Fire-Induced'Vulneiiability Evaluation 'FIVE) methodology and.application of'these technologies to US ,.nuclear power plant support Over the.past.decade Mr. Najafi has been instrumental indevel6pment of thefire research .program at EPRI to support nuclear power .industry move towards a Risk-

* -. . Informed/Pofornance-Based (RI-PB)ufire protectibniule. Under this program data and methods are being.. developed a more engineering-based (as opposed to prescriptive-based)'approach to fire protection. Several

methods where also developed to demonstrate use of the technology, such as "Methods for Evaluating Cable,."Wrap Fire Barrier Performance."

As part of this process ,of continuous dnlancement of technology, Mr. Najafi is ,currintly the principal." technical manager of a joint project between.EPRI and USNRC office of Research for development of the

next generation of Fire lRisk Analysis Methods that can support the fire protection industry in RI/PB rule.This is a ground braking exercise.in cooperative research between EPRI and NRC and key to improving the

* environment for risk-inf6rmed rule in-fire protection: Mr, Najafi is the key in providing goals and directionsto this program that includes the development of the first documented methodology for assessment of fire-.risk during low power and shutdown modes of operation.

Between 1991 and 1997, Mr:Najafi managedire PSA projects at over eighteen (18) U.S. ifuclear plants inresponse to NRC's Individual Plant Examination for External Events (1PEEE) as well as Dodewaard Plant inthe Netherlands. The experience .was part of the .process to'improve the Fire PSA data and methods-developed by EPRI (with Mr. Najafi as the Project Manager).

Between 1988 and 1993, Mr. Najafi served as SAIC Project Manager for GE's ABWR/SBWR Leidel 1 PRA,-Comanche Peak Level I/II PRA support, Project Engineer (Iechnical Project Manager) for the Turkey Point '.

- Nuclear Power Plant (PWR-Y 'Units -3 and 4 Level 2PRA with external events (excluding seismic), andSystems Analysis Task Leader for the River Bend Station (BWR) Level 1 PRA. He' also served as aninstructor in a-course on Seismic PRA and Unresolved Safety Issue (US) A-46, "Seismic Qualification ofEquipment in Operating Plants,- for thle Omaha Public Power District staff. ,.

During 1987-1988, he was the managerof a project to update the PRA for the Indian Point Unit 3 planit andperform a SAIC/Utility-conducted Level I PRA. for a BWR-4 plant (confidential client). Mr. Najafi wasinvolved in the N-Reactor Safety, and Reliability Evaluation 'rogram as the task leader responsible foranalyzing the Confinement, Reactor Trip, HVAC, and Emergency Core Cooling Systems.

Mr. Najafi was one of the principal authors of the Reliability-Centered Maintenance studies for the DieselGenerator Systems at the Catawba (PWR-, and Palo Verde (PWR-CE) Nuclear Power Plants, and-the River'Water Makeup System for the Susquehanna Steam Electric Station (BWR).

During 1985, Mr. Najafi was one of the principal authors dfa PRA study for the Peach Bottom plant (BWR)as part of the NUREG-1 150 program for Sandia National Laboratories. He was primarily responsible for the'modeling of the plant Safety Support Systems including Electric Power and Service Water Systems.

During 1985 and 1986, Mr. Najafi directed an NRC-sponsored work to develop a methodology forassessment of uncertainties in the phenomenological events (back-end). This effort involved development of

SCIENCE - -

SAI C APPLICATIONS ;'INTERNATIONAL CORP.

I , .~ ;, ,w, , .i

a cormputer-based probabilistic framework to integrate the vast body pf knowledge that existi RegardingLMFm R core disruptive accidents and their inherent uncertainty. The methodology not only estimates the-uncertainties; but also can display the nature and extent io which the state of knowledge (or lack of'.;knowledge) con~ibutes:to them.1 The potential application of the methodology to the PWR steam explosionevents in a largE, dry containment was investigated. .The results of this study wdre published in the Nuclear'Science and Engineetingjournal ' 4.''

; k ,._- .- . v- v

Over the period 1982-1984, Mr. Najafi was the principal investigator of several system safety studies on the'Clinch River Breeder'Reactor Plant alMFBR) (that were presented to the Advisory Committee on reactorsafeguards as part of a technical assistance effort for the NRC sfaff.- This effort c6vered a variety of limited-;scope studies forb'oth systems and consequence evaluations, including-radioactivity release frequenciesunprotected reactivity insertion accidents, reliability-analysis of the Decay Heat.Removal System, and CoreDisruptive Accident Energetics.. He :was also involved in review of the CRBRP Reliability Assuranceprogram for the ARC to ensure that the LWR licensing requirernents and associated Regulatory'Guides thatare applicable to LMFBR's are being applied to CRBRPI ,

During 1980-1981, Mr. Najafi acted as the task manager for the SAIC team to perform the probabilistic*systems analysis part of the probabilistic risk analysis study for the SNR-300 (LM;FBR) Nuclear Power Stationin'<Kilkar, West Germany. The objective of this two-year project was to provide safety-oriented information -

to a special commission of the German Parliament that was"considering appropriate energy policies for WestGermany, including continuation of the SNR-300 constructibtf.

*Mr. Najafi has be'en one of the principal participants in thertisk reduction program conducted by the NuclearSafety Analysis Center to investigate-the PRAniiithodology'for estimating incremental changes in plantreliability and risk due to modifications. The methodology was validated using VEPCo's Surry (PWR-W, withseveral shared systems) plant by estimating the incremental change in system reliability and plant safety as theresult of the modification in system design and operation and sPecifications implemented since the originalWASH-1400 study. ,He was also the Task Manager and conducted the probabilistic analysis part of theaccident evaluation chapter for the Seab'rook Nuclear Power Plant (PWR-Y Environmental Report Thisstudy was prepared for Yankee Atomic Electric Company in support of the Seabrook Station licensing.

Mr. Najafi was the principal investigator of a Heat Rate Improvement Study performed by SAIC. A steady-state model of the Morgantown plant using the PEPSE computer code was developied covering the boiler,turbine and balance of plant systens. A limited sensitivity analysis was performed to investigate the,sensitivity of plant heat rate to different plant operational conditions. The long-term objective of this projectwas to provide optimum operating strategies to be used as part of a plant performance monitoring system.

On several occasions Mr. Najafi has served as a lecturer for the reliability and safety analysis courses"conducted by Argonne National Laboratories on the application of probabilistic techniques for accident.sequence quantification in nuclear power plants.

Joining SAIC in 1979, Mr. Najafi participated in the system model development as part of the Seismic SafetyMargin Research Program (SSMRP) for the Lawrence Livermore Natioial Laboratories, where he developedthe models for Emergency Core Cooling System and Residual Heat Removal System for the Zion NuclearPower Station Uiit'l (PWR-Z. Latei he developed a fault tree model for the auxiliary feedciater system forSan Onofre Nuclear Generating Station Unit 1 (PWR-Y to predict the systems reliability under seismicloading.

* - ~~~SCIENCE', .

* - I~~~~NTERNATIONAL CORP. -r

SELECTED PUBLICATfONS: '*

. 1.. "Fire MIodeling Glide. for Nu~clea~r Power Plant Applications," EPRI 1002981, August 2002:

2.*"Fire Events Database for U.S. Nuclear Power Plants: Fiie Initiation and Trends", EPRI 10031 11,* /1 ~December.2001. -

I. "'A Pilot Plant Evaluation Us ngNFPA.'805; Perforrnance-Bas'~d Standard for F~ire Protection for Light

Water Rea ctor Electric 'Generating Pl.is" PI004 ay~2001

C' 'NFPA 805: Perfbrxnan ce-Based Standard for Fire Prdtectilon for Light Water lReactor Electric \* ~~Ginerating Plints"'?Nation-A Fire Protectibn Assiociatibon, 2001 Edition (Contributing Author)

- 5.' "'Fire Barria Penetration Seal Handbook,"EPRI 100196, July.20006. `SFPE Enginering Guide to Periormance-Based Fire Prote ton", Society of Fire Protection Engin.eers,

.First Edition 2000 (Contribtn Athor;* ~7,, "Planbing for Risk-Ifre/ friac-ased Fire'~rotection at Nuclear Po'wer Plants", EPRI TR-.5* ~~- 108799, December 1997

* 8. "Reducing Operations and Maintenance Costs of NucleAr Power Plant Fire Protection Programs", EPRI

TIR-10 7337, Decexinber 1996

9. "Methods for Evaluatinig Cable Wrap Fire Barrier Performance", EP~RI TR-106714, August 1996

1 0. "Fire Ignition Frequency Model at Shutdown for U.S. Nuclear Power Plants", EPRI TR-105929,Decemiber 1995'

1 1. "Fire Probabilistic Risk Assessment Implementation Guide", EPRVITR-105928, December 1995 (and* Supplement EPRI SU-405928)

12. "Fire-Induce Vulnerability Evaluation (FIVE) Softwire", EPRI AP-100530, February 1994

13. "Automatic and Manual Suppression Reliability Data for.Nuclear Power Plant Fire hisk Analyses',NSAC-1791, February 1994

* ~~14. "lPire Risk Analysis Code, FRANC", EPRI AP-1 03733, January 1994,

15.' "Fire-Induced Vulnerabiliy Evaluation (FIVE)", EPRI TR-100370, May 199~ (Contributing Author)

16. "Fire Events Database'for U.S. Nuclear Power Plants'" NSAC-178L~, December 199117. Refrene PantAccden-Seuene Lkelhood Characterization: .Peach Bottom, Unit 2," NUREG1R

*4550,1Volume 3. (With Alan KolaczkowskK,'etat)-

:1 8. "An Assessmefit of Steam Explosions Induced Containment Failure," NUREG/CR-5030, February 1989* ~~~and Nuclear Scien~e and Engineering, December 1987.

19. "On the Probabilistic Apcsof cL-Mode Containment Failure,`,T.G. -Theofanous, B. Najafi and E.* . ~~Rumble, Nuclear Sciince and Enieeig November 1985.

* ~~20. "Incorporation-of Phenomenological Uncertainties in Safety Analysis -Application to LMiFBR CoreDisruptive Accidernf Energetics," Proceedings of ANS/ENS International Topical Meeting-on

k Probabilistic Safety Methods and Applications,Vol. 1, San Francisco, CA, February 1983.

21. "SSNIRP, Phase I, Systems Analysis," NUREG/CR-2015, November 1981 (withj.E. Wells, et al.).

tt)~~~~~~~~~~. e . - I

'- it ' r l i . ..................... 9 _ - .. 9

!ata Systems i_&Solutions -

.; - . -. I .. - _.

G WILIAM IINAMAN D 'PM -

:- -. EDUCATION:x' *PhD, uclear EngineeringIowa StateUniversity, 1974

MS, NucearEngineering,lwa State Uiveisity, 197' ''

BS, Electrical EgineeringowaState University, 1965 - * is,

* WORK SUMMARY:

Dr. Hannaman is a Professional Engineer with over 25 years of progressive consulting experience in soiving* electrical and nuclear engineering problems for a wide range of nuclear reactor types, process plants and

-'.industrial facilities. .Applied educational background and experience to, resolve technical issues usingreliability and probabilistic risk assessment (PRA) lechniques during the design process and on operatingplants. Developed and applied human reliability assessment (HRA) methods to consider the impact ofoperator interactions before and -during accident conditions. Sutporting elements include data collectionfrom training simulators, database development and integrating the results into risk and reliability studies toidentify management priorities for enhanced design, operation, and maintenance.

PROFESSIONAL EXPERIENCE:.

_ | 1999 to present, Senior Staff Engineer, Data Systems & Solutions |I 1988'to 1999 Senior Staff Engineer, Science Applications Internati6nal Corporation

Recent Projects s ' -* Support for EPRI projects inlhe following areas: -

- o Development of simplified trip monitor for use in generation risk modeling of nuclear power plants.o Support Probabilistic Risk Assessment Evaluation of Spent Fuel Dry Storage Bolted Cask Designs in

* the area of initiating events, HRA and data evaluations.o Write guideline for efficiently developing derate and trip monitors for use by control room operators.o Support development of a procedure for addressing HRA in fire PSAso Upgraded Monte Carlo Simulation software (STEIN) for evaluating the impact of NDE measures on

' structural integrityo Developed template for performing Human Reliability Analysis - lesson plans

- o Suppoit project on methods for evaluation of organizational factors-* \ . * Independent safety reviewer for CANDU plant PSA in Romania. - Peer review of PSA modeling

results to recommend changes and upgrades. Also supported HRA training and applications.'. Developed uncertainty analysis tools for predicting the quality of glass/ nuclear waste mixtures for

DOE/Bechtel. . --

Steam Generator Assessment Softwdre development-.* Evaluate primary safety valve reliability under severe accident conditions given a leaking SG tube.* Compare EDF COMPRIS software code with STEIN to identify areas for enhancement in addressing

PWSCC through the use non-destructive (NDE) test measures.* Product manager for establishing EPRI web site for SG SGDSM for maintaining and updating a

quality assured (lOCFR50) electronic database containing data from tests on pulled SG tubes to

A joint venture between Rolls-Royce and SAIC

- %.,G. William HannamanPage 2of7 -7, -

support burst and leak rate correlations. The secure web site developed under an .ISO-9000 and -a1OCFR50 approved quality program supports data searches.''

* ' * Product -manager for development of the STEIN Monte Carlo co4e for use in evaluating Steam. i. ' *x' Generator ODSCC NDE results to predict operational assessment'and conditiob monitoring criteria.

_ - * Devefoped methodology using Montetarlo Siiiulation bf uncertainties for assessing margin betweeni. _. -_ -' an allowed 1131 dose and a predicted accidental release from degraded steam generator tubes;.

Human Relfability Assessmenti* Planned and documented human reliability assessments (HRAi) for four utilities as part of their IPEs

-*Developed and delivered a weekl6ng training course on HRA to Eletronuclear in Brazil.* Supported update-of VC Summer 1PEEE fires assessment as BRA task leader under SAIC and VCS,

quality assurance programs. Evaluated risk of using fire-einergency procedures for the currentcontrol room configuration. HRA methodology used NUREG/CR 4772,. &1278 and EPRI-TR-

-: n 100259.". Contributor to development of ASME PSA standard HRA hnd data sections. _ -

* Instructor on the subject of human' reliability for Argonne National Labs Inter-regional TrainingCourse on Prevention and Management ofAccidents at Nuclear Power-Plantss

* Managed 3-year project to extract data from events to enhance human reliability for activities duringless than full power operation. Reviewed the operator event data collection pro-grams, updated the.

' Systematic Human Action Reliability Procedure (SHARP), presented examples and information atEPRI's human reliability assessment-workshop, and applied SHARPI on specific accident sequences(e.g., Interfacing System Loss of Coolant Accidents).

* -Developed procedures, guidelines and project-instructions for performing HRA in two PRAs.* -Supported use of control room training simulators in'HRA studies for six utilities including Hope.- Creek and Laguna Verde.PRA and Risk Infonred Applications .* For Enteigy Operations, assisting in update of.PRAs-for ANO-2 (accident sequence overview), and

Waterford nuclear power plants (ISLOCA and ATWS support).* Applied time dependent integration of system recovery assumptions and human reliability models

with thermal hydraulic transient output to produce estimates of large early release frequencies insevere accidents for use in evaluating therisk of operating steam generators with degraded tubes.

* Supported Entergy (ANO2) and SCE (SONGS) in evaluating human reliability during severeaccidents to support risk informed evaluation of steam generator tube integrity including review of

-SAMGs, EOPs, plant interfaces, and simulator training. Presentations on results were given to theNRC.

' Performed multi-compartment fire risk analysis in support of the IPEEE at Quad Cities.* For CEGA contributed to guidelines for PRA application during the NPR-MHTGR design process.

Provide mini PRA study for the Environrmental Impact Statement for the NPR-MHTGR.Risk management* Supported development of methodology for blending risk-informed PSA with deterministic rules to,.

-demonstrate compliance with NRC's regulations governing steam generator operation. _* Developed qualitative risk assessment methodology and delivered training course on qualitative .

safety assessments including consideration of HRA for non-reactor facilities as part of a SandiaNational Labs project to comply with DOE orders 5480.23, 5481.IB, and standards 1027-92 and3009-94.

* Applied'methodology on two facilities (Rocket launch and Accelerator). Results support safetydocumentation suitable for a facility safety analysis report in a risk-based format.

* For DOE used PRA and HRA methods to support reviews of DOE reactor projects and facilityoperations.

Reliability Database development* Establish a reliability and safety database for use during the MHTGR design process.* Developed data based mechanical reliability models for safety relief valves using test demands and flow

conditions to improve risk assessment results.

.I. . . .1

_;- .., I

G.,William H~namanI np~rf

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%~~, - - 1 - -

I: : 0S* ' -. - '- ; * -

Ram analysis 'Supported the MHTGR conceptual design through incorpoiation of applicable operational experience,

- development of technical position papers'to demonstrate that lessons leamed fro'm previous operating'experience were considered in the advanced design, and updated safety, availability, and plant capacityfactor reports working with Stone and Webster Availability Assessxiient team. This involved building_reliability block'diagrams for various systems to evaluate reliability and n'is -'--

.Oversight projects ..- *} Served as secretary on- senior review committee to evaluate selection-criteria- for th6 NPR-MIHTGR

cdntainment .* Project manager for independent reviews of PSAfHRA and human factors for Union Fensea cm a

_Spanish Reactor to identify cost'effective riskreductioi upgrades for control roomM interface. Review of a spent fuel processing design for a DOE sie. ',

Performed rdvfew of human reliability assessments in the lIPEs, .',

- Performed independent safety reviews of safety analysis reports and risk assessments including analysisof spray leaks during tank transfer operations, and evaluation of two differeiit pump system operatinglifetimes for Westinghouse'Hanford using FMECAs, fault trees,,aging models and data evaluations.

* Performed independent review of INEEL's ISLOCA methodology.

1981 to 1988, Senior Executive Engineer, NUS Corporation .

* Principle Investigator for. EPRI projects included development of a human reliability analysis- framework, (SHARP), human cognitive reliability (HCR) models, and international HRA benchmark

projects.* Project leader for integration 'of HRA models'to support simulator training, and model verification

studies involving collection of data at control room simulators (e.g., for boiling water reactors (BWR's)at ComEd,'PP&L, and PE). Supported use of simulator data gathering for verification of BWR EOPs.

* 'Technology transfer of HRA/PRA methods to clients performingin US and internationally (e.g., EdF).Transferred technology via: (I) serninars, (2) reviews of PRAs and HRAs, (3) HRA task definition andsupervision of analysts and (4) guidebook development such as PRA procedures guide and HRX4guidelines for sliecific projects (5) performning benchmark comparisons, (6) p'erforming analysis, (7)reviewing work, J8) planning risk related projects, and (9) recommending programs.-Reviewed use of the newly designed symptom based procedures in response to steam generator tuberupture and small break LOCAs to identify key operator actions.

* Probabilistic risk accident analysis of fires for the Limerick BWR.* Detailed safety reviews of design concepts such as the advanced modular gas turbine reactor.

1974 to 1981 Staff Engineer, General Atomics '~ ' '

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* Performed probabilistic safety analysis, reliability and availability assessments and evaluations on all ofGAs operating and proposed plant designs.

* Developed and operated a computerized data base system of component and system reliability measuresto analyze Fort St. Vrain availability experience as a way of improving new designs, including the GasTurbine-HTGR, steamer, fusion designs and others.

* Lead engineer for Chemical and Process System Analysis Group on a 6-man-year effort to collect data,and develop reliability evaluation methods including reliability block diagrams for process systemhazard auialysis reliability allocations, reliability predictions, -availability, and maintainabilityquantification.

* Provided training seminars on probabilistic risk assessment for PRA practitioner training and for shifttechnical advisors.

* Performed system reliability analysis to support qualification of reactor protection, control, heatremoval, main power systems, circulators and support systems for the large HTGR.

* Team member and key author of the PRA study known as the Accident Initiation Progression Analysis.

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: G William HannamaiI. ' . I -

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* Established -and- maintained the component ,and system reliability data bankquantification of event- tree/fault-tree scenario frequencies and uncertainties.

* Developed and applied probabilistic operator niodels and common-cause failure models.

11970 to'l 974, Graduate Assistant and Senior Reactor Operator, Iowa State University

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. : . N

supporting "the;

-* Obtained licenses for reactor operator and senior reactor bperator through the' NRC op a universitytraining reactorwith over 100 startups and shutdowns( -'Taught lab courses and helped prepare and present.training course for Duane Arnold Energy Center

r* ~~operators in support of NUS trainhi&~ . .-- .

.X

I 1 1965 to 1970, Supervisor, Westinghouse Electric Corporation Apoaratus Repair Division - . .

* Planned repairs and directed maintenance crews on chemical, utility and industrial 'sites and in repaiiplants for over 10,000 unique power system equipment failures. L U

* Designed and implemented an l&C temperature protection system for large electrical motors, anddesign of a transformer oil storage and transfer system. -,

* Developed procedures, criteria, 'and equipment for testing, welding,-and evaluating insulation andmechanical structures for serviceability and, if needed on the basis of predicted failures, appliedmethods for repairing, balancing and testing electrical and mechanical apparatus including electricmotors, breakers, controls, transformers, generators, turbines compressors,-,magnets etc.

* While in Westinghouse's Graduate Student Program performed rotating assignments in manufacturingfacilities for transforimers and apparatus repair.

COMPUTER PROFICIENCY:

LanguagefTools: Microsoft Office Software, Math software, Monte Carlo Simulation, CAEIA

Hardware Systems: PC, and Mac

Operating Sysiems: Windows 95,2000, XT; OS8, and DOSI

A MISCELLANEOUS 'ProfessionalAssociations andMemberships: -

State of California - Professional Nuclear Engineering Registration NU 1948 Since 1982Member American Nuclear Society -.

San Diego Section chairman 1979San Diego Section executive committee, various years is - *Technical program chairman for Embedded topical meeting on Advanced Nuclear InstallationSafety, 2000, C fAssistant Technical Program Chairman for Risk Management -Expanding Horizons 1992.Human Factors Division, ExecutivelCommittee, 1987..Safety Division Program Committee 2000.Organized and chaired numerous technical sessions for ANS.Paper reviewer for Nuclear Technology ' '

Member of Institute of Electrical and Electronics Engineers .

Corresponding member of the Nuclear Engineering Subcommittee SC-5 on human factors andreliability responsible for standards on reliability methods. 2000 -2003SC-5 Comrnittee member on Reliability 1976 to 1980,SC-7 Committee member on Human Performance 1984-1986.Organized and chaired technical sessions at an IEEE meeting

Society for Risk AnalysisExecutive committee of the Southern California Chapter in 1989.Organized and chaired technical session at PSAM 11

a _ G. Abwam-Hannaman - . * ' ; ;Page5of7 .7- A' -,

Patents, Selected Publications, and Awards: E '-* Elected to Sigrna A, the-research honer Society in 1973

- . -. .. -.. * Elected to National Academy of Sciences 6-member panel on cooperati6fwith USSR on reactorsafety to identify needs and means for enhancing reactor safety.1987 - -

: - --Elected to-Strathmore~s Whors Who 1996-03 - -

* ' Outstaniding technical paper awards in ANS Meetings (e.g.,ANS Midvest student conference 1974- .< !, ; ->, *and ANS summer meeting Human Factors Division 85, 88; and 93). .

- '-. Toastmaster .CM and ATM levels and Toastmiaster of the year for Area,17 District 5.199912000 - 7

'. .*Acadefmiic credit for *- ;,- Reliability Assurance, UCLA 1975 * : :

* - Global Business ManagementUniversity of Phoenix 1998 ' -Reports,Hannaman, G. W. and I. B. Wall,,"Lesson Plans for HumaniReliability Assessments in PSAs," EPRI1003329 June 2002. - .

Hannaman G. W (DS&S),V. Durbec aid C. Bauby (EdF), "Feasibiity Study for the Integratiod of EDFVsmodels for PWSCC into EPRI's STEIN code," Joint EDF and DS&S Report to EPRI, May 19,-2002.

Mickey M. B., G. W. Hannaman, B. W. Johnson, K M. Btemen, 'Verification ofIHL'W Product Qialityby Analysis of Uncertainty and Reliability in the HLW Process Control System," DataSystems & SolutionsReport to Bechtel National Inc. May 2001.

Hannaman G. W., and. S. A. Fleger, Evaluation of HCR Methodology Implementation in PSA andControl Room Human Factors Review for Jose Cabrera Nuclear Power Plant, EPRI, Palo Alto, CA, April2000,000000000001000028.Hannaman, G. W., B. W. Johnson, Maureen K. Coveney, "Methodology For Steam Generator ConditionMonitoring and Operational Assessment, Applying Monte Carlo Simulation," SAIC-97/1078, ScienceApplications International Corporation, San Diego, CA Dec 1998.

E. Fuller, E. Rumblej G. W. Hannaman, and M Kenton, "Risk Assessment Methodology f6r Complyingwith NRC Regulations on Steam-Generator Tube Integrity: Diablo Canyon as an Example Plant" LREPRI 550-7, Sept. 1997.

.Hannaman -G. W., M. Lloyd, B. Putney, G. Klopp, B. Johnson, A. Farruk, E. Fuller, and G. Pod "PSASupport For Steam Generator Degradation Specific Management" SAIC-1326, EPRI 550-7, March 1996.

A Dabiri, F. Johansen, B. Johnson, and B. Hannaman, "241-Y-101 Mixer Purni Lifetime Expectancy, forWestinghouse Hanford Company Richland, Washington, Nov. 1995. -

Mahn J. A., G. W. Hannaman and P. M. Kryska, "Qualitative Methods for Assessing Risk,' SandiaNational Laboratories, SAND95 -0320, Albilquerque, New Mexico, May 1995.

Hannaman, G.- W. "Transforming PRA Results bino Performnance-based Criteria for PWR steamgenerator Inspections and Management" White paper on EPRI project 550-07, March 1995.

Otis, M. D. D. A. Bradley and G. W. Hannamnan, Technical Basis for Considering Uncertainties in I131Release and Dose Limits for a Postulated Accident. EPRI TR-1 03878.-EPRI, Palo Alto, CA March 1996.Hannaman G. W., W. Parkinson, ajnd C. Donahue, Lessons Learned from Documented Events aboutHuman Reliability during Less Than Full Power Operations, EPRI report TR-104783, Sept. 1993.

Hannaman G. W. C. G.-Donahoe and E. M. Dougherty, Insights from Human Reliability AssessmentsPerformed during Less Than Full Power Operations, EPRI report SAIC-92/1056, SAIC San Diego CA,-March 1992.

NSAC 154 "ISLOCA Evaluation Guidelines," HRA methodology, EPRI, Palo Alto, CA Sept. 1991.

Hannaman G. W. and J. Forester Analysis of Initiation of Boron Injection in Response to an ATWS,SAIC-91/1132 SAIC Report for Task 2 of Gulf States Utilities River Bend project, April 22, 1991.

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Hannaman G. W. and IC J. Budnitz, "Case Study on the use of PSA methods: Human Reliability ranalysis," IAEA-TECDOC-592 Internationil Atomic Energy Agency Vienna Austria, April 1991.

.SHARPI ' A Revised Systeinatic Human Action Reliability Proced re;;(w'th G. Parry, A. Spurgin and'D. Wakefield), EPRINP-.7183-M,Decemberl190. - -.

Contributor to Operator Reliability ExperimnentsJsing Power P1ant Simulators',EPRI NP-6937 Volumes..1, 2, and 3, Julyl-990. ' '*, s

Hannaman t3. W. Applicatin of SHARPI to Interfacing System Loss of Coolant Accidents (ISLOCAs),.SAIC-90-1351, Sclence Applications Internitional Corporation Report oA-'EPRI Project 3206-14,September 19, 1990. ; . ' * ' . '

P-Lbbner, L., Goldman, oG. XW. Hannaxfiaand S. Langer. Preliminey Risk Assessment of-the NPR-MHTGR, App. B, Generic Reactor. Plant Description and Source Terms,'Environmental Impact'Statement, EG&G-NPR-8522, June 1989. .

Atefi, B., M. Drouin, W. Hannamdn and J. Young, "Perspective on Applicationof Probabilistic-*Modeling Techniques to the Heavy Watei and modular High Temperature Gas-Cooled New ProductionDesigns," SAIC-89/1146, McLean, VA Sept. 29, .1989. .

Models and Data Requirements for.Human Reliability Analysis (with A. D. Swain,.G. Mancini, L.Lederman, et al.). IAEA-TECDOC-499, Technical Documentissued by the Iritemational Atomic EnergyAgency, Vienna, 1989. *

Hannaman G. W. F. S. Dombek and Y. D. Lukic, Evailation of Key Human Interaction Postulated forEDF 1300 Mw(e) Nuclear Plants STGR and -SBLOCA Accident Sequences. .EDF Project, NUS Report5105, May 1988.

Probabilistic Safety Study Caorso NPP, (co-author) ENEL DCO 401.V40.VR.001, NUS-4954, Nov.1986.La'Salle Human Reliability Measurements Program Data Analysis. Prepared for ComEd, NUS-4965, DecembeF 1986.

Incorporation of Transient Response Implementation Plan Procedures into the Limerick GeneratingStation ProbabilisticFRisk Assessment. NUS-4887, August 1986.

.Hannaman G. W. AJ. Spurgin and Y. D. Lukic, Quantification of A3 and H2 Procedures for a Standard900 Mwe PWR Plant . Prepared for CEA, NUS Report 4935, August 1986.

Hannaman G-W. A.J. Spurgin and Y. D. Lukic, Human Cognitive Reliability Model'for PRA Analyses.EPRI Project 2170-3, NUS Report 4531, October 1984.

Hannamnan W., and A. Spurgin Systematic Human Action Reliability Procedure (SHARP) EPRI NP-3583 June.1984.

Review of the Sizewell B Probabilistic Safety Study." (with S. Ilevine ) NUS-3446, April 1983

Hannaman G. W., W. Brieher,;R.Cantrell, and H. Hopkins,'Reliability,-Availability, and-MaintainabilityPlan for the Solvent Refined Coal Demonstration Plant, V I & II. GA-C-16372, Solvent Refined Coal r

Int., Inc., July 1981.

Hannaman G: W., et. al. Safety Program Plan - Summary. .USDOE Report GA-C-1 6244, Volumes II, and,I performed for Solvent Refined Coal International, Inc., January 1981, App. June 1981.

Hannaman G. W., et. al. HTGR-RPR Capacity Factor Estimate. GA-A-16242, -General Atomic Co.,January 1981:

Hannaman G. W. GCR-Data Bank Stafus Report, US DOE Report GA-A-14839, General Atomic Co,July 1978.

HTGR Accident Initiation and Progression Analysis Phase II, (co-authored with K. N. Fleming, et al.).USDOE Report, GA-A-15000, General Atomic Co., April 1978.

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,Papers * -

X Pennarnti. G. W., "Safety Valve Reliability Assessments for PSAi," PSA 2002, ANS Piobabilistic Safety; ; , Topical Meeting, Detr9it Oct,.2002. -, ..

Johnson, B, G. W. Hannanan, and M.,A. Stitzke, "Oprating Reactor Safety, Regulation and the Real World,"inANS Proceedings Operating Reactor SafetyTopical Meeting, Oct. 11-14,1998. . '

^ - Fuller% Ed, E. Rumble, G. W. Hannamanad M. A.Kentoh,!'Assessmnit of Risks from Thermal Challengeto SteamrGenerator Thbes During HypotheticalSevere Accidents," in-ANS Operating Reactor Safety TopicalMeetingOct 11-14,1998.'

Mahn J. A., G.W. Hannarnan; P. M. Kyska, "Qualitative Methods for-Assessing Risk' 1995 ASME conf ,, . ~1995. n

-Haniaman, G. W. and Avtar Singh, "Human Reliability Database for In-Plant Application Of Industry -X ..Experience," PSAM II,1993 v _ - . .

Hinnaman, G.(W. and A: Singh "Assessments and Applications fo Enhafice'Human Reliability and Reduce*Risk during Less Than Full Power Operations" of EPRI, AN9 Rislk Management embedded topical, June 1992-. -

a . Hannaman G. W. "Human Reliability.Methodbfor Enhancing Performance," in Risk Management Expanding. -Horizois Hemisphere Publishing, New York, 1991. -. -

Hannaman G. W. and D.H. Worledge, "SomeiDevelopments in Human ReliabilityfAnalysis Approaches andTools",)Reliability Engineering aiid System Safety, Elsevier Publishersttd. England, V22 pg 235-256,1988.

- 'Hanna.maik G; W., F. S. Dombek, B. Y.- . Lydell, andY. D. Lukic, "Using Risk Analysis to Improve Testing; -and Maintenance". F&th IEEE Conference on Human Factors and Power Plants. Monterey CA. June 1988.

Hannaman G. W., G.R. Crane and D.H. Vrorledge "Application of a Human Reliability Model to OperatorRespon'se Measurements" in PSA and Risk Management PSA'87, Zurich, Switzerland, September'1987.

* . Hannaman G. W., F.S. Dombek and P: Moieni "A PRA-Based Human Reliability Catalog", for ProbabilisticSafety Assessmeit and Risk Management PSA'87, Zurich, Switzerland, September 1987. .

. -̂ HannamanhG. W. et. al. "Applications of Human Reliability Models to Structure Measurements of Human-a Performance in Simulations" Job Performance Measurement Technologies Conference, DOD, Wash., D.C.

3/87.

Guymer.P., G. Kaiser, T. McKelvey'and G. W. Hannaman "Probabilistic Risk Assessment in the ChemicalProcess Industry" Published in Chemical Engineering Progress, January 1987.

Hannaman G. W:,The Role of Frameworks, Data; and Judgment in Human Reliability Analysis", NuclearScience and Engineering. North Holland Publishing Company, NEDEA 98L93, May 1986. ,

Crane G. and G. W. Hannaman,"Real1stit Operator Response Measurembnts: Inputs to La Salle PRA", V 5,*\*.' \ International Topical Meeting on Nuclear Reactor Safety No. 70016 6, ANS, La Grange Park, IL, Feb. 1986.

"Synthesis of Experience Data for Rihk Assessment and Design Improvement of Gas-Cooled Reactors"(with A.P. Kely): Proceedings of Probabilkstic Analysis of Nuclear Reactor Safety, American NuclearSociety, IL, May 1978.:

" Probabilistic Risk Assessment -of HTGR's" (with Fleming, Houghton, and Joksimovich), Reliability; .- Engineering, Applied Science Publishers, Ltd., England (1981) pp. 17-25.

Treatment of Operatbr Actions in the HTGR Risk Assessment Study, GA-A-1 5499, Winter ANS Dec.1979

Fleming KN. and G. W. Hannaman "Common Caise. Failure Analyses in the Predication of Core CoolingReliability".]EEE Transaction of Reliability, Special Issue on Nuclear System Safety and Reliability, R-25Number 3, August 75.