ML033440488

Appendix R Regulatory Conference with Notes

ML033440488
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/10/2003
From:	- No Known Affiliation
To:	Office of Nuclear Reactor Regulation
References
FOIA/PA-2003-0358
Download: ML033440488 (91)

Text

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REGULATRY IICNEECE July 10, 2003 -A

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OPENING REMARKS. ) ..

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• -I Craig Anderson (.

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INTRODUCTION C'r 1

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Sherrie Cotton - .-

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Director, Nuclear Safety Assurance

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Opening Remarks Craig'Andersori

. I VP, -ANO-r I.-

Introduction Sherrie6Cotton

/I Director, NSA.,

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• A Risk Assessment Methodology Dale James '

Manager, EP&C 4, r

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Fire Modeling Bijan' Najafi ' ' / .

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II SAIC Analyst,. I.).

Break I . ,

I Probabilistic Safety Assessment Jessica Walker p ,

PSA Engineer' ,- F S

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I Overall Summary Joe Kowalewski I-

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Director, DE' - .

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Closing Remarks Craig Anderson VP, ANO,-- 'Z II .

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Risk Assessment Methodology

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Dale James Manager, Engineering Programs and  ;. \-  ;: .-

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Problem Statement I* . . Iv

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• NRC ConclUsions ..

-ANO's reliance on manual actions in lieu' of providing separation'design features is in violation of'Appendix R  ; .

-"'ANO's strategy for implementing manual actions ISr inadequate . /

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Risk Assessment Overview I..

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• NRC's preliminary SDP evaluation concluded'!

unacceptable (greater thangreen) increase in core M " I. VI I II-  :-I I .

damage frequency f I .

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, .,bs Key assumptions in NRC evaluations vs ANO's Wr preliminary assessment /* . Ic7 b I.

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- Heat release rate . II Ih*I I I'

- Human error probability ) . I ..

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• Subsequent'site,-specific in-depth assessment . .

- Results incorporated into 'Unit. 1 PS IAmodel, to derive ACDF C- . ..

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Risk Ilnforriied Stgrategyfor Zone 99M- K . 8'."

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_. . CircuitLAnalysis &

Fire Modeling 99M I~~ --. .,,

Simulator;Scpnario i - Development

}i .'1, L Location Evaluatibn

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Procedure -

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Risk Assesmet Compatr

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NRC . .Al k , .

• 4250 F cable failure

• 700dF cabIleI,.failure .,  :

temperature 1-1 .temperature B.. r .1 e,

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• Zone.wide prompt: Limited.time. . . ph:ase'd I'

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damage c amage I

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• Generic HRA

• Plant specific, HRA

- Based on zone wide - Scenario specific prompt damage actions evaluated k . t Included L@OP -N6 LOOP, ,

Greater than Green

• Green. finding findi~ng

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Bijan Najafi' C.

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SAIC Analyst" I

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4 .i LocatiodEvaluation ',

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In our analysis we will show that: .

- Damage to equipment and instruments needed for safe 4. I shutdown will be limited to portions of the. room ' . . \

- Failures will occur over a period of time,. and

- No credible.fire can be postulated that leads to zone-wide 1.

!darmage. ' .'. I .Oi

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Fire Modeling in the 4KV.Switgear Room 13.-

(Fire; Zone.99.-M)

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

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• Fire scenario selection , I! A/-

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• Fire characterization I .-

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Fire modeling, evaluation of the consequences and timing of the'fire scenarios,

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• Results and conclusions I .

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• l ~~~) 1w6%S-Unit 1 4KV Switchgear -Room (fire zone 99M)

EC204 C204 EC204 EC205' E E203 BE202 EB202 EB202 .

EC236 EC236

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A4 Switchgear

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1 Unit, I 4KV Switc'hgearom(ie. zone 99M)'

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transformer *-.

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Typical ANO switchgear cabinet wiring, control cubicle

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Fire Scenario Selection':,

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-I- -N -I- ro---VA UU Fire'scenarios define pot'ential'ranges of damageby.a fire ,I

- 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 I

- 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':.

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Three distinct fire scenario classifications:. -

-;el,.

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

• 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 damage (

to targets. anhd ignites exposed dables injthe vicinity ,. .' :

• The fire may continue in the switchgear and'grow within the,ighited. -I.

combustibles (e.g., cable trays) in the vicinity L.

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

- A transient fire that may spread into cable trays A transient fire between B55 and B56 was selected as the nrraximum expected scenario due to its' potential for extent and timing of damage *i .

Circuit 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.

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Eight fire scenarios selected.'represent credible fire risks for 99M - ..  ;; ..

Scenario a - Fire in A4 switchgear.

- Scenario 1b - 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.to this presentation -: ' .

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

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• ElectricalIcabinet fires . I r' -.

- The heat release rate-'data profile is

, HRRProflle ,

based on the b-6st available fire test data 120 -

• Sandia National Lab (NUREG/CR-4527, 100-,

87/88) and VTT (Valtion Teknillinen

'Tutkimuskeskus, 94/96) in Finland 80 _ ' I

• Same test used in the NRC SDP analysis 60 - . _

- The ANO HRR is based on the highest

• r40 ,,/,

peak of ST5 (unqualified, open 110 KBTU loading) and all qualified, vertical x20-,

I)

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'cabinets (excluding PCT6 and test 23.

with'1.5 MBTU loading).' 0. '5..

• 10. .15. 20:

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• 'the NRC HRR is based on test 23 .. . . . .

(qualified, open 1.47 MBTU loading) and . Tlnie [mln] , J,

• test 24,(unqualified, open, 1.44 MBTU) P. .

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- Time-to-peak is based on the average .I.

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- Tests are based on"control panels I I

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- The switchgear, MCC's and load \ /

certters; are enclosed with sealed  ;',,'

penetrations  ; IS I - .

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• Used for'scenarios 1a, 2- 5 r 4 4.'.4I,4.

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II Fire Scenario Selection: I.,-

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. . I 0 NRC and ANO SDP Analyses; . ef .I

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• NRC SDP firescenarios , , :  ? ../ .. ... *.~ -)  : ic

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- Based on fire size .

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• Total room heat-up and zone-wide damage I

- Electrical cabinet and electrical equipment'fires I I I I.

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I I ANO SDP fire scenarios I ,:

- Based on source and target-set ch'aracteristics and' - . . I ..

configuration ' . ,

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• Local as-well as zone-wide damage

- Electrical cabinets;.in zone', 1.

- High energy arcing faults in the 4KV.switchgear.(a "bNey IOhd design basis event"), . .

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- Transient fires including hot. work.' .. .

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.5 Cable tray fire heat release rate: Q Ct"

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

Used for scenarios that include ignit~ed cable trays 5).

Transient fires: 150KWV -

- Typical refuse based on fire tests at SNL/LLNL documented in EPRI

,Fire PRA Guide " .-.

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• Used for s-cenarios 6a and 613 g S ,,  ; ' , . '

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Fire Characterization (cont.) ' . ..

High-energy Switchgear Arciig Fire.

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

• En'suing electrical cabinet fires (the switchgear or,others exposed to 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 arcing fault type , .,

• Used in scenario 1.b . . " ."

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Fire Characterization (cont) .) ,-

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ZOI of the High-energy Switchgear ArcingFire Top VieW

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Relative Door - I I . . Cable trays-are Height CDblb Ni Cab s CAb K offset as .,I indicated in top 1I - _ -_

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a-Firet Characterization: N e 1

25 NRC and ANO SDP Analyses II (

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Electrical cabinets . J I

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

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- ANO: 100KW, peaking in 12 minutes

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High energy fires in switchgear; I r,

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- NRC: Assumed covered by the range of HRR I

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- ANO: Empirical model based on 'experience' (previous slide),

O- I . . I damage/ignition within: five ft. .I

.1 Transient fires . .

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- NRC:.Out; of 1,scope .. I . . I

- ANO: 150KW, peaking in 1.0 minutes A'...

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Fire Modeling. .I ..II-. I

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_61 Model. for.Preed-iction of Fire Growth" . .

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I- I Hand calculations, used to calculate time to localized target

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- Target immersed in flame . .

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- Target in..the.fire plume I Target in-the, ceiling jet, and

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- Target'in the-flame radiation zone

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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. I $

widely used for the range of fire conditionsexpected in Zone 99M

• Cable fires: Used fire tests for both growth through stacks and horizontal cable tray. ) - .. j

- An empirical model used to determine the extent and timing of the spread through the stack (based. on SNL tests documented in NUREG/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). ' - ' I I

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- I I -I Fire Modeling:

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Target Da.nagel.gnon. . 3. 3

• 27)

Tar~get Damage/lgfiition N\ . I Targets (cables) are considered damaged or ignited.

when their surface temperature reaches 700OF I

-, Thermoset insulated cable predominantly used in'the plant as verified through original and current design and installation specifications ' 's

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

high risk zones ,. ' I I.

• This is the critical difference between the NRCC and ANb analysis as it relates to the extent and timing of firp damage.

- 425.F vs 700.F . . _

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- Critical to extent and timing of damage and fire growth .0,

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Fire Modeling: /

28 Target DamagelIgnition . .I

• .6 Assumed cables inside metal conduits daMragqd. at the same critical temperature, but will not contribute to ( I '

room heat-up -

N  :

High-ienergy arcing fire,' *1 I .

- 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 the short-lived (seconds) initial pressure spike. i '. I

• Spurious operation 'of damaged circuits were modeled.

In some cases, the likelihood of the spurious"actuation was obtained from the EPRI. Expert Elicitation r'ep~o't?' , I (EPRI 1006961) which' was estimated in part based -on.

the data from EPRI/NEI circuit..failure. characterization fire . te sts .. .*.

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. ~ ~~ . .

CFAST Results *I I Scenario lb, Open door .I/

.sI

(

s.

I .

1., -

(i.

-2500 ., 600?-

p1.,

I i F. .

I

- v.5 0 0/ ~, I.

LL .:

- 400 -,o '- .

,- .. I -

A-

-. 300 X 4),. ..

i2Q0 E . j I _

iW1 00

.I 0 I~ . .. I 4

. I . I, , .., .

I °r 2 4 t 6 8 10 12 f14.:

S.*

Time [Mini .,

I ._

I .

. I

,I, ... ,. I

- Calculated HRR m InpLut HRR - UL4Temperiatutie

'p. J. .

I I., ,.

e-

. f I.7

.1 . .

I' I.~~~~~~~~~~~~~~~~~ I... - , . .

I I

-I 2_

.~~ . . ~

Results (cont) . IL ~~~~I

.. I% ~~.1 - - .3

• -~~~_I A high-energy switchgear fire (scenario l b) is the maxi 3urn ex ected,.fire scenario - *I

- 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,

• The floor-based sources of fire in fire zohe 99M are electrical ca6inets and transients' . l - - i'

)t

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

I(.

- The floor-based fire intensity needed tb generate damaging (7000F) HGL'is-2MW. . . . .. . 1 .7 I r XA a a_ 0 _ * ..

- inone o Tne Tloor-oasea rires'are capaoie.ot sucn'intense neat, 11 I I-I

. I  :

Only cable fires are potentially capable of generating. such intennsity'if enough cable is involved. I . 1 -

- 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 the top 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 12f . - '.

- The temperature peaks at 5500-535 F 0 .

- The fire has to be below the settled smoke layer (4-5 ft) for the cable' firesto continue to grow - X I' I . .. II I. I . -

7.

I . . I . . I .

/ II

, II I - '. I t.

j

I 1, * .. /3 . . .. 7 I

. I . . .I I

-- N ) .I

/ T I R -o . If . 5,. , 1 r .

1 , I I I . . I Results (cont.) .. .~~~~~~~~~~~~~~~~ I-I . . . . .

I I

. 31 ..

I .:" I.V

,- , I t/v,' l-',

_i'v'

/%;

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"),

I ' ' ,~

I * .

• Maximum expected fire is a high-energy switchgear fire No credible fire reaches 7000F in this room (limiting fire '

scenario) ' . .. . (' - . ) -~~~~13*

(

V. I*~ ~~~~

. . I- ~ j. * '5. I ~ .

I I S I ,~~~~~~~~~~

.A r

(.

,,- .1

/ I~~~~~~~~~ ....

Results (cont.).. -- ... .j* .

• . 32, Comparison of NRC and ANO Results I I.  :  % I

-AI I

i f' I I . A I  ;. ,

. . .0-

. I *

• Damage threshold ..

." UpperLayerTemperaturein99M  :

- NRC': 4250F ..

ANO: 700°F -.

t 800

• Heat release rate 700 600 - -~ .. L,.

- NRC: 500KW 500 fire peaking in 105 sec'.

u-400 - .,

- ANO: IOOKW peaking, in 12'min (Scenario la) + cable ->

120O0 I.

fires and high energy fault in A4 switchgear and cable fires (Scenario lb) I, 0 250 500'. 7590 1000

• High energy arcing fault Ti'Je sc p --

in th~e 4KV-switchgear *' -.- ANO-ULTerpScenario.laopenDoor

- NRC: Not-analyzed ' _4-NRCULTe np '

- ANO: Limiting scenario in terms -- AN0JL Tenrp, Scenario lb, Open Door of its consequence, i.e, affected *-I '

circuits a'nd timing I

tC

• Neither analysis reaches 700°F .% .

, J I

I I I--

4

. r!

I .

. 11 Il; I .

I  %'.1. .I b'

'IL-Frequency of Fire Scenarios I . .

.33

..I

-/

. 1.I I ~~~

.. . I~~~.

• ir P rik k N'f (.1rPn'nqrin

-'~

D ' ^-'- ' ' - S ' W /

Frprui ipnrv) j - -J ' I ., Cn -

flP.

Scenario Frequency is derived from multiplication of:

. .S

- 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 switchgear -I11 Based on operating experience (EPRI fire data base) 1.I

- 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 NI,

)

Next presentation discusses development of the CCDP 1 and fire risk I I I .

I .

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Rc0hc1u

• \wsJ L4 Uhs cit111*, 'j . , , . * ' ,,

Frequency of Fire Scenarios in, Fire Zone 99M

' '" {' '~~~~~~~~~~~I( ' , ,-". ' '..' ' [

.1 \ .1  ?

4.

ANO SDP Analysis Results 0 *. ~~~~~~~~~~~WIFloor ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~We IoofH Pns,by U

~~.Sourc. ~~~~Generic Frequency (location weighting Pg ~ n source

-weighing area ratio everiy evn fra severe pln personnel Psb fire Results

. l Frequency l weightrng (transient U Factor o r g I C

________ _____________________________ ______factor) factor) Fires) fire ivatch I C.

1a .Fire in the A4 sv~tchgear.. _ I 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.O5 0 lb 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 Y28 J4

,.I, ':

are 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. Nominal 0.0 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 this scenario. 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-06 6a Transient fire in areas of the room .

where cable trays are exposed to .

a floor-based fire. Nominal Value of 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 I

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

Ventilation Subsystems . . / , . I - 4,4E06 t.;

' _ .____  ! ~ ~ ~~,4t.0.

Y * /

'I N I , . , . ., ,.. , i~ z~

N 1.'

- > ~~~~~~ ~ ~~~~~~~~~~~~~~~~~~~~~

'N'-

Fire Modeling Summary Maximum expected fire scenario in firezone 99M is a high , l 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

~

C4,~ ~ /oVtta , ,2~-3, wA~~~~LJ~~~c~~~eI~v~' ,-s.)... FOpAeGA apj rz (4 c' 6 i;26/ /' igH 74--l b , 7 Day .ac ~~~~~~~~* FA , i.

,E,- '-* ,,A 86zg  ? -  ; r $ a r

• k %5 . Ace-.~~~~~~~~~~~~~~~~~~~~~~~*

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. ~~~~~~~.

Probabilistic Safety Assessment .

I I )

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Jessica Walker I .. ~,9 .A ,

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11~~~~~~~~~~~~~~.. \ ' ..

'%-^.' .*I .

I C-PSA Engineer .

.. I.

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4. 'I.I I .. ,-, 4"

-I 44

4. 4

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Introduction I.

.37 '

I

• )

S -,

I'

• Key system failures in 99M .

I- -

C.

I -. I .

• Affected components due to cable failure S.

.. I

,: . . A. . .

• Key. operator action/response * -k *.*

K

• Simulator scenario/results-' ..s .

C.~~~~~~~~~~~~~~~~~~~~~~~~~~~~

I*** I(** . ,<..

• Operator action probabilities. .1 C-,

• K x, s

K A.. I;

• CCDP calculation t r

\

• Delta CDF determination N I~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ ,, A. -

• 1*

• ¢ a}

. I

\ \

\ - . In .

• SI.

• . *1

• l3/4 / . *1

, A..

i . , , . -.

Is,

• l #

N ,  :..

4 I I.,

C C

.1 I I

- . ')V.

I .

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Risk Info-rmed Strategy for Zone 99M k " . - -I- I I -\ 38-I 'i N

1.

Fire Modelin"g 99M l 4 J.'~~~. I *..

N . 4. )

. I -

. z ,' - ProcdclurE Simulator Exercise *1 I I ., .

If. I ..- Dev6lopme

. '\I - - c -

I I 1. .,  % 99M

• . . I I

I

.. I I I

'. "I I-I .,  % .I

.... J, I I / . . .  %

IF 11 . .

I , . I

. , , ,~~ (~~~~~~I . .).

J I. "% . I

• I I.,- .. I 4  % .

HRA Development .. ., .

nt -

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~~~~~~~~~~~~~~~

.. ", O' II~~~~

I"!~ .

'If, .

-I~~~~~~~~ I

• . x

.- i I II"

.I

'C

.1 0 .

PSA 11II I .\ I . .

C. .. ..

I ~~~ (I . .N, ~~'I1 N J I 00 I ..1 I .I Ii b

I,

-N I v I.

'Cf Total Unit Risk i.

) Ad s-.

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.. I I

e I ..

/ .. . '_. I Z . -. i./

.:,. .7'.

. , . k

(  ;',

k,

.1 .

. ,f C. I Key.Systems-Affected in the Risk- . . .-. .39. . .

I Significance Determi~nation (Fire'Zone. 99M) .

I3 . .

1 . \.

N.*

,. I .

..... .. . . . ,`I j I-

I*

.I I,;

The following systems/trains. are directly faiied due to I-fire induced poWer losses.of A4 and B6 ._

~

Is,

! NI .: .

- One train and.the swing pump of service water .

.d I.

.. I

.I I. . .

'I

-,-One train and the swing' pump of HPI (makeup) .

I j/

't - . ,,

K'

-. The-A4 associated diesel is.no longer usabld'.

I . .*

-I.

J .r /.. I. , , - .;

., I.

Ic ,

1, . 1

. I . . I I,. I. I . , .,,/

- . R.

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. ~~~~~~~~~~~~~

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I I . I I .I Circuit Analysis .!

I .

I

" - I .-) .

.5' 40 I.:

I I *

('V.

• I II I Detailed. circuit 'analysis peiformed on zone .99M 'S

,I

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

.5 I I I

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

I x .

. I-.  ;- i-

- NRC evaluation did use loss of offsite power

• Analysis of associated failure modes for affected

~~~,I ~ ~~~~.A ~ ~ ~~~~~~. cables 4, I

. .1

.I.

I .

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

, ~~~~~~~~,-

.. c

/ .

.II

., I Ir I .  :..

, ,I I I ' I I -

-6 Systems Affected in the Risk-Significaipcer.  ; 41.

Determination (Fire Zone 99M) v

~~~~~~~~~* r -

Scenario specific failures are based on cable focation; .r 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, - -.

I I

- DC, control power to the A3 switchgear fails I

. . . C t

• Breakers to remain static and require manual closure at the switcahgear' .I

- P-7B (motor driven EFW'pump) suction valve could spuriously close I

.1 ,

,I

. I I I

• Cable in conduit; spurious operation not probable but assunied in evaluation

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

... Requires local action to start P-7A, ' ,-

- Aux-lube oil pumps for the unaffected HPI train e-.

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

, .t IC

. f I  ; I I - .

I ~ ~ ~~~'

11

,t.,

• * , J } _ _: ._~~~~~~~j J .

Operator Actions/Response in the Risk-Significance ..

Determination (Fire Zone. 99M) -2 l

• subset of the following operator actions are rewquired in each A

scenario ; - .

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.'

.%J C

- Stafting HPI for-make-up (long term action)  ;

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

. . \~~~~~~~~~~~~~~~~~~~~~~9

/ . , *

• Emergency diesel generator rdcoveries were not ne'cessar due to the lackof a LOQP event -. .r ,.

, ~~~~~~~~~~~. . .. A ..

~~~~~~~~~~~~.

9 , x . . - . . *

, ~ ~~~~~~~~~~~~~~~~~~~~~~~~~~~~ , . M ..

I .'

. V i, ' ' ' .' ' ' ,* ., ' ' 9_.

I -- ' .9/.'

f J B).

,. -1

p.

II I ).  :*

Previous Procedures vs Zone Specific .I 43

/

Procedures.

.j\N, -

.I Provini i . -nrn-I I .V W V-J J 1M VL4 IJ  %..,1%AL41 '-'I

- Combination of EOP/AOP/Pre-Fire plan

. L

- Opportunistic approach :j,

- Plant condition determines .action

• I N

New procedures 11 I .

- Zone-specific fire procedures - 1.1 , . . .

- Tactical approach

- Reduces impact and probability of spurious operations t-.

fi

. . *C C

• I r _,

. ,  ; , .,, ' *' ,'1.- p .3 S~  :. . '.3a-

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A.

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Summary of Procedural I ..

Guidance . ¶1* -

1. -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 includes ... I..

1 manually and positioning mentioned which could delay operator indicators. v.

- . , . ~~~~~~~~~~~~functional associated valves response. '_. Or} * /

Controlling EFW (A or B) Previous 2 to prevent overfill control room action.

soca o prodredscss Lack of adequate and'borrect indication,is directly discussed in the new procedure" which makes {his actionmore likely ir).the

. I -

I.

I.

____________________________________ ,newprocedure. , -

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 -

I .~ 'I1 4 Starting PID Makeup of this action depends on when letdown is The new procedure addresses the 4,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. /1,

. . -_I~~r.

.X

-J I

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, ,' - I \ . I I

.I * -1 . , *:

... I.

t) / . I'

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• i V.'. I,.. . . -:.

.-_ I Simulator Scenario for Zone 99M I I I ( I j.

*I 45

.i *I, _; 1.

U

.- .. . I.

N ..E

-N I,. ji' Fire damage chosen to provide HRA information for multiple ' .: I.

operator actions" .

.,- Fire model.beginning with an A4 switchgear fire II Firehlon g ingwie-predanb

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

/ damage', .' . .4.

- Damage for scenario exte'nds beyond credible.fires'

r. - . ~~~~~~~lef~1. a I s

• reaiistic control room communication cnaiienges

• f -

-, Fire brigad6 leader.communicati'on ' , * -V...

-.4 4.

Timelines based on actual fire drill .

• V

• I nrIi i nrlarP4rifs tn a, A I( 1 r r n n f q; I n nn f;rn i.I% %%a ILIa'.JL IJ%.JUI A II ranrnk I

rfmtnnf%

LI II 1 II L , .A  ; * ** U

- In plant auxiliary operator used f6r operator.actions - I -

,1 I

• Radio 6hdtelephone communications used V. \

I

-.. 5. I.,. 4 V q.. lzf ' v

.. I I . .

. eiI

~~~~~

~

I

, .. I

.}

/

I . .

I.

1*

'. I..

.' A ~. .- < I Simulator Scenario for Zone 99M

. I

• tI .

-46 1

4 .

Simulator scenario failutes includbd:

.1

.J

('I

- Directfailures -4 I

• A4 switchgear (4KV)

• B6, load center (480. VAC) ,

• EFW flow control valve power failure

• HPI auX-lube oil pump power fairure

- Included spurious operations *'

I II 000,

• P-7B EFW suction valve closed at T-1 5 7 . .

. I I. ,

- Included failed and incorrect indications A

<- . i e., I. .

. 1.1 , ... iI Multiple panel indications failed (EFW, HPI, P ower) > 11 . . , .

I .

• Random annunciators spuriously alarmed, .

. . * . . . .~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~A

/'

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e.

A-.

I I

'-,I.

't

I . . .I I

.I I . IL iI I jI , . I

. I - I - 11 ,

.. . II

1. " . .-

. . ~~~~~~~

I Simulator I Scenari6 for Zone 99Mf

• I r1 4-.- *i 47 ) N-I/

. I (i I

, *(

I I I * ) .. .I I

.I I

• Three crews ran simulator with previous, procedures.

\* I IA

• Two crews ran simulator with training on zone specific.fire procedure .*. . , -' 'CA-'

• I.-. 1 I Qr -

• One crew With each. procedure. contained operators n,the plant simulating' local actions - ' /

7*

Controll'ers were present in,the field to evaluate local manual actibns' ' . " I I ...

• Time toJocation. . ' ,

. I' I .

• Potential hazards I.I

, p.

• Communication barriers I -

I - 1-I

\** I*

I. I.

• Observers in the simulator to evaluate control room actions I

• Including time to perform in.control room actions -, j Procedure usage. \ r ..

• Work practices due to loss of indications a I'~~~~~~~~~~~~

• ,. .'~~~~~~.

v

t. . . ,..

I.

i

.J~~~~~~~~~~,

J . -.

I. I for ZoneII 99M" I%,

Risk Informe'd Strategy I >e

. I... J 41 .. . ..

X. ,

.. . I

.. Circuit Analysis & ; . I.

I. Location Evaluation-I 99M h..

m

. I

,.-,)

I

.) N

_-A IL

.. t-~~~~~I

. . I\ ...

' ' i

I I

I, i e Simulator Scenario Results , f Simulator runs using previous ESPIAOP/ Pre-Fire plan 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 .1:

- Crews responded appropriately and, in a timely-manner

- Plant maintained at a safe stable state

• Simulator ruhs with crews trained on new. tactical I /

procedure 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 4,*

I

• Crew performance using either previous or new' procedures met Appendix R requirements for 1 achieving safe shutdown .

. 11 I . II

I .

.I I I 1, I

• 4 Risk Informed Strategy for Zone 99M. * ' 5O~~~~~~

L I ,

I I I I Simulator Scenario circuit Analy~is & I Fire Medelinj-99M F.oato Evaluation.

Development 99M /

4,~~~~~~~~~~~~~~ f: . T . I .1 I ..

94 1%

III .

-N Procedure. I '

e .

Simulator Exercise LEO P/AOP/Prefire Plans] -

I i % ),

I-NDevelopjbent 0eW 1 M. " t.

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/.. .,I HRA methods for-quantification demonstrate there is an impact of fire on reliability of human actions t

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Previous vs' new procedures for shutdown C

- 'Previous procedures use an opportunistic approach to control,, e

' . .4 where crews respond to cues and symptoms by selecting EOPs for that condition with the'aid of pre-fire plans i:

4 New procedures assist crew to. respond using 'a more tacti(.-.al,- .

control process " ' '

v.,

• Use. of dither approach,'demonstrated ' '

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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

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Method for Updating HRA Assessments to ... 5 52 A .......

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I The current Unit 1 model for human recovery actions. in-internal ( ..

events PRA is based on a time reliability curve .'

.I HRA~accounts for operational context by adjusting. param'eters.

such as: , , , ., , , ,, , , . , , , .

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-. Rule-based vs knowledge-based behavior

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- No burdenys burden I . I -

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- Other performance influencing factors,

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,Incurrent assessment, effects of fire are not addressed 'nr ayre.

model param'eters available; therefore, a different ajdjustment- . ,

method 'iXas.identified' ',

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• EPRI HRA calculator used to assess differences of fire . ..

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• inciuaes several metnoas tor quantmication

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- Inuustry. ana NRu sponsorea. I I .

t Generic data quantitatively differentiate' human error,

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Evaluation ofFire Impact on, Probability: Il II 5

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Seven cognitive assessments, on differences in procedures,..

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- Availability of information . ,. ,, I.

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Summary of AHEP Applications Due to-Fire . I .

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.. i Case Event ID Basic Event Description A Pcoa A Pexe A HEPf~re ,I I 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-03 control room - new t

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

• 3 MFIRECR Realistic fire In 99M decisions in 9.8E-03 2.OOE-02 3.?E-02 control 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-02 6 Not Feasible - -1 1 1 7 No Change .  : .. . 0 0 .0. .

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S Comparing HFEs from PRA baseline with HFEsin99M fire I ..

r. .r- . N Fire in 99M increases human failure event (HFE) for typical feasible actions over .,

initial internal events PRA from zero to a value in range..of ,E-3 ,to 4E-2 for various *56-scenari6s 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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Value of current HFES In the Base PRA model ,a, I. A ., .:I. .. .

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- I Human Error Proba'bility, Comparison..

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• NRC approachI assumes zone -

wide damage at time zero, I

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• NRC approach included I ..

loss of offsite powerI Operator Actidn NRC Value No NRC Value . ANO Value . ANO Value

-, Procedure W/Procedure Previous New

.. i Establish EFW . .0.6. 0.11 0.,098 i " .

(A3 local start)

. I I

Establish EFW 1 0.6 . 0.038 0.026 I.

I .

(Control EFV) . .

Establish Feed 0.75  ; 0.55 0.008 / 0.008

&Bleed ' . . .

Establish Feed 0.75 0.55: 0.11 . 0A098

& Bleed I (A3 Local Start) . .

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Secure Diesel . 0.75 0.55 Not needed due to no loss of offsite I

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with no'Service power, I I .

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I n S Risk Informe d Strategy for Zone. 99M

i58

. . ' Circuit Analysis &

L Fire Modeling 99 M - .

Simulator Scenari Development 1 -

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Locatior i' Evaluation

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CCDP Determination for Zon'e 9.9M.), 595, ;

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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-energyelectrical faultscqenario1b . ... . . - Y

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

occur'. .I

• All components,.failed. together (conservative) .,

Timinrgs opily used to disallow spurious, of K ';

puiosoperatibr o qompqnents whose control cable would be lost after power los  ;

• HRA values for the previous and new procedures used to recover the- 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 industry-experts to'deterrrine reliability of -necessary -operator.actions I'

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

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

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. I . I I 6 Fire Risk in Zone 99M if -

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Fire in the A4 switchgear. .

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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 . ' . , '.

-.4 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-08 .

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v'  ! ,. S ' ,.... . .... . . ), , N - I

) .. . . I Fire in the B55 MCC. Nominal , . . 11 100 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-09 Fire 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 - *. - .

.,1

. Fire in the Y22 Inverter. Base.

case, 100 KWfire. Fires in Y24 .

.I I

• and Y25 are bounded by this

4. scenario. . 1.32E-05 .98E-05 3.86E-05 -5.27E-10 5.10E-iO 1.60E-11 I

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Fire 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-08 Transient fire in areas of the room . f I.

where cable trays are exposed to , -.

I -

. 7-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 I Cable fire caused by welding and * ,

I

. cutting in areas of the roomh where . .

.. 11 cable trays are exposed to a floor- .

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based fire. Nominal value of .. .1 6b 150KW. 2.60E-07 3.24E-03 2.12E-03 8.41E-10 5.50i_-10 2.91 E-10 I ~.-..

99M

• 6.61 E-07 4.39E-07 -2'.21E-071I

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

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Realistic assessment of fire progression, failures in 0 - 6OM minutes (targets of the high-energy switchgear damnage,'

immediate,'the rest time-dependent failures are from ensuing-cable fire) , * ',

I-Critical c"able insulation used is ,thermoset ...

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-7n~R11_ 1___%1_ _ I._ L . '. ^_ .

- -UUt cauie. uamage temperature I

. N Operator-action.probabilitiees

.~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~ .1 . .

I.

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- New procedures offer slight HEP imrprovement over prqvious nrn-'rtl

- Ia- ,  % el .

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- Human reliability'analysis: :CCDP indicates that'impact of Ii AHEP I immeasurable but small I' I . i S.

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Risk.Informed Strategy for Zn 9 9 M I1 7 Risk I 'nformed% 1 Strategy. for Zone 99,M .. . I J, .. ;... ,"

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63'.. _.

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I-1-1 .. CircuiitAnalysis &.

Simulator Scenario ,. <-Location Evaluation Fire Modeling 99NI - 10 4 >,

• Development-

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[EOQP/AOP/Prefire Plans]7 I Development ,.

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Total UnitRisk . ..

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oc'us~~A h-e tht_ on. ,o 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. ' . 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 to be performedd ' .

- Zones'with MFW unaffected screened '

MFW greatly extends time needed for ,EFW local actions I

- Zones with one complete train of core cooling unaffected screened .

.t

• Control room operation of equipment removes impact of'local operator . .

actions ' ( S -. ..

-. I .. I.,

• Similar to NRC results, the following zones remain; I .'

- 1OON . . .

$,.- I ,_ . IeI

- 104S' ' * . .

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-65

/  : X The assessment of fire risk in 99M was extrapolated to two other 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, (1ON)',v

, . 4...)

9. I Similar to 99M in combustibles and fire sources

' Considerably less. redundant train cable' routed through zone. V.

- Unit ' electrical equipment room (104S)

, Lack of high energy-switchgear / . .

• Considerably less redundant train cable routed. through zone )

. .. I

• Each zone is. bounded by the results of 99M

- Conservative estimated fire risk (ACDF) for this condition

• I

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

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Unit 2 Risk .. I I

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• 'The four Unit 2 zones identified as risk significant by NRC were ,qualitatively evaluated, ' /

(.5. I

• Each was assessed with'walkdown and examination of the potential fire scenarios threatening theother train raceways.

(e.g., red train raceway in a green train room)

• Conclusion I.I.

(I

-The four, Unit 2 zones contain similar characteristics to the Unit 1 zones . - r .

• 4. -

'I,

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• Two switchgear rooms

.- .-  ? .. %VI f

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^Two rooms containing MCCs similar to, 104S

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- The. results from 99M. bound. these 1zones II, 4

1. 1.

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. I Summiary

... I of .Risk Assessment I I

. ; I .

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': 67 '-

I

• ANO risk assessment concluded that: V I.

5,. ._

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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 " -I zone 99M are credible , . ,.I r, Simulator scenarios validated that operators could achieve safe

' _ -- .. S '.

shutdown.

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

• Concl lusion I . . .. , .

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• Unit.1 < 6.6E-O7/yr

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-' I OVERAIk SUMMA.sY IOVERALL

SUMMARY

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I. . I Joe Kowalewski . ,.

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,Director, Design Enginberin g...

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N Overall Summary  % I

• 1'

• i. \JJV Daealsfo I ,

- I.)

• Detailed analysis of ;zone 99M

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

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

- Simulator scenarios provided realistic data for assessment- of operator 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. safe shutdown has very low safety significance and should be-characterized as-GREEN. fl '_ yg 4

.1 e A?~~~~~~~~~~~~~~~~~~~~~~~~

. 1 . .

( .N t I

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) 7 Overall. Summary;,(cont.)

I6 * '

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j. I 1 70.

I' I,

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• ANO fire. protection program

- Defense in depth strategy to prevent and miitigate fires" '

- Explicit control.of combustibles Fi~rebrigaae efectiveness A1

- ' / I I

• 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' ' .'.

I.

- 'Validated circuit analysis-

- Feasibility evaluationof manual actiors (IE 71-111.Q5).

- New procedures developed to enhance operator response. I

- Fire.detection reliability improved ..

• ANO can' successfully achieve safe shutdown in the. vernt. of a fire .- I in any zone . . I I

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CLOSING REMARKS

.. I II . .-

. ~ ~~~~~~~~~~ I I. .. .. . .

. . . .. . 1 . . -\

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-"I~~~~-

I', ... .- 7 i

I, .

Craig Anderson . I VP, ANO

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ADDITIONAL INFORMATION I..'

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. . .1 SCIENCE - .

SAIC .-. I APPLICATIONS INTERNATIONAL CORP. -

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, I I BIJAN NAJAFI, P.E. - ' / s- J MANAGER, FIRE PROTECTION SECTIONI I .  :.,

EDUCATION: * '

University of W;shington. MS., NucleaiEngineering, 1!i9 Shiraz University.. ' B.S., Electrical Engineering, 1976

.RegifteredcProfessional Mecharical Engijner, State of Califprnia I

SUMMARY

OF CURRENT POSITION:' . - I I .

• ., ,:.. *.*I . J,j--. * -.- )

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

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 assessment and 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 fire protection program. Over the past decade he has been instrumentil in development of the EPRI fire risk technology currently in use in the U.S. nucear power industry. This technology has also been used internationally in Europe and parts of Asia and South America. Mr. Najafi haslconducted training courses in U.S. and Europe on Fire Technology, ;nost recently a series of Fire Modeling courses for nuclear power plant fire 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 Fire Protection for Nuclear Facilities (801/805) /

• .. Principal member of the American Nuclear Society's committee for the development of the Fire PRA Standards

• Participating member of various taskforces at Nuclear Energy Institute including the circuit failures is'sues taskforce 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 Determination Process (SDP) _

Member of the NRC's "International Collaborative Project to Evaluate Fire Models for NuclearNP6wer Plant Applications." - -

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

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

%, N. l

.iSAIC * -

SCIENCE 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 fire protection 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 the fire 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,.

"WrapFire 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 directions to 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 in response to NRC's Individual Plant Examination for External Events (1PEEE) as well as Dodewaard Plant in the 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), and Systems Analysis Task Leader for the River Bend Station (BWR) Level 1 PRA. He' also served as an instructor in a-course on Seismic PRA and Unresolved Safety Issue (US) A-46, "Seismic Qualification of Equipment 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 and perform a SAIC/Utility-conducted Level I PRA. for a BWR-4 plant (confidential client). Mr. Najafi was involved in the N-Reactor Safety, and Reliability Evaluation 'rogram as the task leader responsible for analyzing 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 Diesel Generator 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 for assessment 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 Regarding LMFm 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 explosion events 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 reactor safeguards 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 frequencies unprotected reactivity insertion accidents, reliability-analysis of the Decay Heat.Removal System, and Core Disruptive Accident Energetics.. He :was also involved in review of the CRBRP Reliability Assurance program for the ARC to ensure that the LWR licensing requirernents and associated Regulatory'Guides that are 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

• systemsanalysis part of the probabilistic risk analysis study for the SNR-300 (LM;FBR) Nuclear Power Station in'<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 West Germany, including continuation of the SNR-300 constructibtf.

• Mr.Najafi has be'en one of the principal participants in thertisk reduction program conducted by the Nuclear Safety Analysis Center to investigate-the PRAniiithodology'for estimating incremental changes in plant reliability and risk due to modifications. The methodology was validated using VEPCo's Surry (PWR-W, with several shared systems) plant by estimating the incremental change in system reliability and plant safety as the result of the modification in system design and operation and sPecifications implemented since the original WASH-1400 study. ,He was also the Task Manager and conducted the probabilistic analysis part of the accident evaluation chapter for the Seab'rook Nuclear Power Plant (PWR-Y Environmental Report This study 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 project was 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 Safety Margin Research Program (SSMRP) for the Lawrence Livermore Natioial Laboratories, where he developed the models for Emergency Core Cooling System and Residual Heat Removal System for the Zion Nuclear Power Station Uiit'l (PWR-Z. Latei he developed a fault tree model for the auxiliary feedciater system for San Onofre Nuclear Generating Station Unit 1 (PWR-Y to predict the systems reliability under seismic loading.

• - ~~~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.2000

6. `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-107337, 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 1991 PantAccden-Seuene Lkelhood Characterization: .Peach Bottom, Unit 2," NUREG1R

17. Refrene

• 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 Core Disruptive 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 'PMD -

- . - 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 ofprogressive 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 using reliability and probabilistic risk assessment (PRA) lechniques during the design process and on operating plants. Developed and applied human reliability assessment (HRA) methods to consider the impact of operator interactions before and -during accident conditions. Sutporting elements include data collection from training simulators, database development and integrating the results into risk and reliability studies to identify management priorities for enhanced design, operation, and maintenance.

PROFESSIONAL EXPERIENCE:.

_ l 1999 to present, Senior Staff Engineer, Data Systems & Solutions l 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 PSAs o Upgraded Monte Carlo Simulation software (STEIN) for evaluating the impact of NDE measures on

' structural integrity o 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 GeneratorAssessment 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 (10CFR50) electronic database containing data from tests on pulled SG tubes to A joint venture between Rolls-Royce and SAIC

-  %.,G.William Hannaman Page 2of77, - -

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 between

i. _. -_ -' 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 current control 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 Training Course on Prevention and Management ofAccidents at Nuclear Power-Plantss

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

' Systematic Human Action Reliability Procedure (SHARP), presented examples and information at EPRI'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 andRisk InfonredApplications.

• 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 in severe accidents for use in evaluating therisk of operating steam generators with degraded tubes.

• Supported Entergy (ANO2) and SCE (SONGS) in evaluating human reliability during severe accidents 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 the NRC.

' 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 Sandia National Labs project to comply with DOE orders 5480.23, 5481.IB, and standards 1027-92 and 3009-94.

• Applied'methodology on two facilities (Rocket launch and Accelerator). Results support safety documentation 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 facility operations.

ReliabilityDatabasedevelopment

• 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 I . f ' -

G.,William H~naman  %~~, - - 1 - -

np~rf I 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 capacity factor reports working with Stone and Webster Availability Assessxiient team. This involved building_

reliability block'diagrams for various systems to evaluate reliability and n'is -'--

.Oversightprojects .

.- *} 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 analysis of spray leaks during tank transfer operations, and evaluation of two differeiit pump system operating lifetimes 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 and supervision of analysts and (4) guidebook development such as PRA procedures guide and HRX4 guidelines 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 tube rupture 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 ' ' '~

S.

• Performed probabilistic safety analysis, reliability and availability assessments and evaluations on all of GAs operating and proposed plant designs.

• Developed and operated a computerized data base system of component and system reliability measures to analyze Fort St. Vrain availability experience as a way of improving new designs, including the Gas Turbine-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 system hazard auialysis reliability allocations, reliability predictions, -availability, and maintainability quantification.

• Provided training seminars on probabilistic risk assessment for PRA practitioner training and for shift technical advisors.

• Performed system reliability analysis to support qualification of reactor protection, control, heat removal, 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.

e I

I

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.  : . N

G William Hannamai Page4 of 7

• Established -and- maintained the component ,and system reliability data bank supporting "the; quantification 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

-* Obtained licenses for reactor operator and senior reactor bperator through the' NRC op a university training reactorwith over 100 startups and shutdowns( -'

Taught lab courses and helped prepare and present.training course for Duane Arnold Energy Center

• ~~operators in support of NUS trainhi&~ . .--

r .

.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 inrepaii plants for over 10,000 unique power system equipment failures. L U

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

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

• While in Westinghouse's Graduate Student Program performed rotating assignments in manufacturing facilities for transforimers and apparatus repair.

COMPUTER PROFICIENCY:

LanguagefTools: Microsoft Office Software, Math software, Monte Carlo Simulation, CAEIA HardwareSystems: PC, and Mac OperatingSysiems: Windows 95,2000, XT; OS8, and DOS I AMISCELLANEOUS '

ProfessionalAssociationsandMemberships: -

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

San Diego Section chairman 1979 San Diego Section executive committee, various years is -

• Technical program chairman for Embedded topical meeting on Advanced Nuclear Installation Safety, 2000, C f Assistant 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 and reliability responsible for standards on reliability methods. 2000 -2003 SC-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 Analysis Executive committee of the Southern California Chapter in 1989.

Organized and chaired technical session at PSAM 11

a _ Abwam-Hannaman G. - . * '  ;  ;

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 reactor safety 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 .CMand ATM levels and Toastmiaster of the year for Area,17 District 5.199912000 - 7

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- Reliability Assurance, UCLA 1975 *  :  :

• - Global Business ManagementUniversity ofPhoenix 1998 ' -

Reports, Hannaman, G. W. and I. B. Wall,,"Lesson Plans for HumaniReliability Assessments in PSAs," EPRI 1003329 June 2002. - .

Hannaman G. W (DS&S),V. Durbec aid C. Bauby (EdF), "Feasibiity Study for the Integratiod of EDFVs models 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 Qiality by Analysis of Uncertainty and Reliability in the HLW Process Control System," DataSystems & Solutions Report to Bechtel National Inc. May 2001.

Hannaman G. W., and. S. A. Fleger, Evaluation of HCR Methodology Implementation in PSA and Control Room Human Factors Review for Jose Cabrera Nuclear Power Plant, EPRI, Palo Alto, CA, April 2000,000000000001000028.

Hannaman, G. W., B. W. Johnson, Maureen K. Coveney, "Methodology For Steam Generator Condition Monitoring and Operational Assessment, Applying Monte Carlo Simulation," SAIC-97/1078, Science Applications International Corporation, San Diego, CA Dec 1998.

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

.Hannaman -G. W., M. Lloyd, B. Putney, G. Klopp, B. Johnson, A. Farruk, E. Fuller, and G. Pod "PSA Support 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, for Westinghouse Hanford Company Richland, Washington, Nov. 1995. -

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

Hannaman, G.- W. "Transforming PRA Results bino Performnance-based Criteria for PWR steam generator 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 I131 Release 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 about Human 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 Assessments Performed 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 r analysis," IAEA-TECDOC-592 Internationil Atomic Energy Agency Vienna Austria, April 1991.

.SHARPI ' A Revised Systeinatic Human Action Reliability Procedre;;(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 Production Designs," 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 Energy Agency, Vienna, 1989.

• Hannaman G. W. F. S. Dombek and Y. D. Lukic, Evailation of Key Human Interaction Postulated for EDF 1300 Mw(e) Nuclear Plants STGR and -SBLOCA Accident Sequences. .EDF Project, NUS Report 5105, 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 Generating Station ProbabilisticFRisk Assessment. NUS-4887, August 1986.

.Hannaman G. W. AJ. Spurgin and Y. D. Lukic, Quantification of A3 and H2 Procedures for a Standard 900 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.

I 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-Maintainability Plan 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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XPennarnti. 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 Challenge to SteamrGenerator Thbes During HypotheticalSevere Accidents," in-ANS Operating Reactor Safety Topical MeetingOct 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

• Riskduring 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 and Tools",)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

- Maintenance". F&th IEEE Conference on Human Factors and Power Plants. Monterey CA. June 1988.

and Hannaman G. W., G.R. Crane and D.H. Vrorledge "Application of a Human Reliability Model to Operator Respon'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 Probabilistic Safety 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 Chemical Process Industry" Published in Chemical Engineering Progress, January 1987.

Hannaman G. W:,The Role of Frameworks, Data; and Judgment in Human Reliability Analysis", Nuclear Science 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 Nuclear Society, 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 Cooling Reliability".]EEE Transaction of Reliability, Special Issue on Nuclear System Safety and Reliability, R-25 Number 3, August 75.