ML033110418
| ML033110418 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/10/2003 |
| From: | Anderson C Entergy Operations |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2003-0358 | |
| Download: ML033110418 (91) | |
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ARKANSAS NUCLEAR ONE APPENDIXR REGULATORY CONFERENCE
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OPENING REMARKS
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Craig Anderson HI Vice President, ANO I"
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\. K IK INTRODUCTION I
I Sherrie Cotton Director, Nuclear Safety Assurance
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p i AGENDA 4 Opening Remarks Craig Anderson
- 1VP, ANO Introduction Sherrie Cotton Director, NSA Risk Assessment Methodology Dale James Manager, EP&C Fire Modeling Bijan Najafi SAIC Analyst Break
.'Probabilistic Safety Assessment Jessica Walker PSA Engineer; Overall Summary Joe Kowalewski Director, DE
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Closing Remarks* 'Craig Anderson \
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I f ,I a: Risk Assessment Methodolog
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I Dale James Manager, Engineering Programs and Components ... A I I
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p I Problem Statement 6
- N RC Conclusions
- ANO's reliance on manual actions in lieu of providing separation design features is in violation
.of Appendix R I
- ANO's strategy for implementing, manual actions is inadequate
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3' Risk Assessme ,ntOverview 7
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- NRC's preliminary SDP evaluation concluded unacceptable (greater than green) increase incore damage frequency .'.
- Key assumptions in NRC evaluations vs ANO's i
.p . preliminery.,assessment
- Heat release rate C-Hunman error probability
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Subsequent site-specific in-depth assessment
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- Results incorporated into Unit 1 PSA model to C
' derive ACDF r
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r Risk Informed Strategy for Zone 99M 8
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1 Circuit Analysis &
Simulator Scenario 24
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Fire Modeling go Location Evaluation
- Development
., 99M
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- a. . qI Procedure x,- *
- r Simulator Exercise Igo Development 99M , .
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I HRA Development
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T Total Unit Risk Ii
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! 2 Risk Assessment Comparison
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NRC :ANO 4250 F cable failure -7000 F cable failure temperature temperature Zone wide prompt Limited time phased damage ;I damage t .
Generic HRA
- Plant specific HRA
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, - Based on zone wide - Scenario specific operator prompt damage actions evaluated, Included LOOP % - No LOOP ¢
- Greater than Green
- Greeh finding I, fineding I
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1'0 I I FIRE MODELING I
Bijan- Najafi SAIC Analyst i
'Risk nformed Strategy for Zone 99M
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r Circuit Analysis &
Simulator Scenario 0
Location Evaluation Development AOLG
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Procedure Development l
99M _. ,~~~~~~~
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Su Mmary ~12 In our analysis we will show that:,
-Damage: to equipment and instruments needed for safe-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, damage
Fire Modeling in the 4KV Switchgear Room 13 (Fire Zone99-M)
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- Unit 1 4KV switchgear room (fire zone 99M)
- Fire scenario selection
- Fire characterization
- Fire modeling, evaluation Of the consequences and timing of the fire scenarios I :( A
- Results and conclusions ( '.k I.
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Unit 1 4KV Switchgear Room (fire zone 99M)
EC204 EC204 I~~~~~~~
EC204 I
,14 EB203 EB202 E8202 EB202 .
EC1530&
EC1504 V
Unit 1 4KV Switchgear Room (fire.zone 99M) 15 (J I (
.9v zj 99M - north view l
B55 MCC Si
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Y28 Inverter A6 I
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. ,, 99M - north view lj . ,~~~~~~~~~~~~~~~~~~~~~~~~
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. A4 Switthgear C : . ! At I I.
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Unit 1 4KV Switchgear Room (fire zone 99M)
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.N B6 Load center Dry-type transformer
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Typical ANO switchgear cabinet wiring, X control cubicle
I Fire Scenirio Selection:* 17 General Approach Fire scenarios define potential ranges of'damage by a fire
-' They define sequence, and timing of failures, ie., ,
'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 I. ./ '.
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Fire Scenario Selection: 18 General Approach I
- Three distinct fire scenario classifications:
- An electrical fire (non-energetic) in any of the electrical cabinets in the room
- Fire may spread in the cable trays, but requires considerable time
- Circuit damage/failures follow a time-phased sequence with first damage
- after 10 minutes
- AhiIh energy arcing fault switchgear fire that may initiate secondary fire!
' The event has an initial (immediate) pressure phase that causes damage to targets and ignites exposed cables in the vicinity
- The fire may continue in the switchgear and grow within the ignited combustibles (e.g., cable trays) in the vicinity
- ' There is an initial/immediate circuit damage/failure followed 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 maximum
,expected'scenario due to its potential for extent-and timing of damage Circuit damage/failures follow a time-phased sequence with first damage A after 10, minutes 2
Fire Scenprio Selection:
Scenarios Modeled in Zone 99M Eight fire scenarios selected represent credible fire risks for 99MV 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,,.,
stacktray
- Scenario 6b - Welding/cutting fire between B55 and B56 below tftree stack tray' Illustration of these scenarios is provided in the attachment to this presentation
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Fire Scenario Selection: 20 NRC and ANO SDP Analyses NRC SDP fire scenarios
- Based on fire size Total room heat-up and zone-wide damage Y
- Electrical cabinet and electrical equipment fires ANO SDP fire scenarios
- Based on source and target-set characteristics and configuration.
- Local. as well as zone-wide damage
- Electrical cabinets in zone
- High energy'arcing faults in the 4KV switchgear (a "beyond design basis event")
Transient fires including, hot work
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Fire Characterization 21 Electrical cabinet fires The heat release rate data profile is based on the best available fire test /
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Sandia National Lab (NUREG/CR-457, 87/88) and VTT (Valtion Teknillinen Tutkimuskeskus, 94/96) in Finland
,Same twst used in the NRC SDP analysis
- The ANO HRR is based on the highest peak of ST5 {unqualified, open 110 XKBTU loading) and all qualified, vertical cabinets (excluding PCT6 and test 21-with 1,5 M lBTUIoading)
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- The NRC HRR Isbased on test 23 C Int ialifia' nrnar 1 A7 MRTI I InnriinnI Andr4 test 24 (unqualified, open, 1.44'MBTU)
- Time-to-peak is based on the average
- Tests are based on control panels
- The ~witchgear, MCC's and load centers are enclosed with sealed .. I penetrations ( '
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- Used for scenarios 1a, 2 >5
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I-~~~~~~~~~~~~~~~~~2 Fire Characterization (cont.) 22 Cable trayfire heat release rate: Q0
=0.45 qbs N
- Widely used model from Society of Fire Protection Engineers handboo Used for scenarios that includ ignited cable trays Transientfirps: 150KW
- Typical refuse based on fire tests, at SNL/LLNL documented in EPRI Fire: PRA Guide
- Used for scenarios 6a and 6b,
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11 Fire Characterization (cont.) 23 High-energy Switchgear Arcing F I The darmage/ignition zone of the initial pr sure phase is derived fror'US nuclear experience (next slid (EPRI SU105928 Supp to N)
-EPRI Fire PRA Guide)
Ensuing electrical cabinet fires (the switchgear or others exposed tots arcing fault) follow the same behavior as the non-energeti electrical cabinet fires
- Potential ensuing cable fires'spread horizontally and spread faster 5 vertically through cable tray stacks
, Observations:l
- Experience of the US nuclear industry indicates that damaging/severe switchgear fires tend to be of the energetic arcing fault type Used in scenario 1'b I-.
Fire"Characterization (cont.)
ZOI o'f the 'High-energy Syit gear Arcing Fire V
W161111 CA)
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Relative Door Height I ,
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Fire Characterization:
N ~~~~~25 NRC and ANO SDP Analyses
- Electrical cabinets
-NRC:4 200-590OKW, peaking in 105 seconds /,
-ANO: 100KW, peaking in 12 minutes
- High energy fires inswitchgear
-NRC: Assumed covered by the range of HRR
- -ANO: Empiric'al model based on experience (previous slide),
damage/ignition within five ft.
- Transient fires
-NRC: Out~of scope ANO:A150KW, peaking in 10 minutes
Fire Modeling:
Model for Prediction of Fire G r
- Hand calculations used to calculate time t ocalize target damage -
- Target-immersed in flame
- Target in the fire plume Target in the ceiling jet, and*
-Target in the flame radiation zone
- CFAST used to calculate room temperature and target damageignition due to hot gas layer
- CFAST and simple correlations such as Heskestad, are validated and widely used for the range of fire conditions expected in zone 99M Cable fires: Used fire tests for both growth through stacks and horizontal cab!4 tray
- 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 for fire propagation along horizontal cable tray (EPRI NP 7332)
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'5 Fire Modeling:
Target Damragelignition Targets (cables) are considered damag r i wnhen their surface temperature reaches 700 F
- Therrhoset insulated cable predominantly us e plant verified through original and current design and installation specifications
- Thermoplastic insulated cables are not used in ANO Unit /
high risk zones, This is the critical difference between the NRC and AN analysis as it relates to the extent and timing of fire da
- 425F s 7000F i l
-' Critical to extent and timing of damage and fire growth
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i l Fire Modeling:
(I 28 Target Damage/lgnition
. Assumed cables inside metal conduits 4maged at the same critical temperature, but will notjontribute to room neat-up High-energy arcing fire
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- Assumed raceways and cabinets in th zone-of-influence are damaged with exposed cables (trays )ignited
- Assu nservative for the conuits (ifstainless or
'- anized steipes) where they are likely to survive the ort-liv'(seco s Spurious operation of damaged circuits were modeled.
In.some cases, the likelihood of the spurious attuatibn was obtained from, the EPRI Expert Elicitation report, (EPRI 1-006961 ).which was estimated in part based' on the. data from' EPR'I/NEI circuit failure characterization I.
fire tests.
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Results -i; i29
-CFAST Resuits
- Scenario b, Open door 2500-2000 -is '
- . 0 4500 V l ! n. _ _ _ _ _ _ _
Results (contf) 30 '
A high-energy switchgear fire (scenario b) is the maximum expected fire scenario
- Initial HE phase could lead to ignition of as much'as 12 linear ft of cable tray,
- After the initial HE pehase, ensuing cable fire may grow although at a very slow rate
- The floor-based sources of fire in fire zone 99M are electrical cabinets and transients i : I
- The likely'location of electrical cabinet fires (flame) is below 5ft off the floor onc e OLII breaker cubicle is open in the highenergy event .
- The floor-based fire intensity needed to generate damaging (700 0F) ,GL i -2M
_- None of the floor-based fires are capable of such intense heat
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Only cable fires are potentially capable of generating such intensity ifo e is involved
-' Cabre tray fires are elevated fires, (none of the cable trays in fire zone 99M are located below the 8ft door opening)
- Cable fires are expected to be in the smoke layer once thesmoke layer reaches the top of the door. Once in the smoke layer, intensity of the cable fire will be controlled by the oxygen availability, which is not enough to the combustion process With bn elevated cable fire that grows at -a rate of inearft/hr MO as input
' .- The oxygen depletion occurs very quickly, regardssoqr closed door.
- The cable fire does not grow beyond the initial 12 ft I - The temperature peaks at 500-535*F
<e~ The fire has to be below the settled smoke layer (4-5 ft) for the cable
- o grow I I
Results (cont-) 31
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The limiting fire scenario, one that can generate a damaging HGL, isnot credible '
- The non-suppression probability by the brigade for very long duration cable fires(100 minutes for the high-energy switchgear event) is 0.01 (per EPRI 'Fire PRA Guide) ,
- Fuel depletion, cables ignited earlier have burned out 7 Parts of the cable trays are coated with flamastics which both delays ignition and slows propagation of cable fires
- Continued growth of the non-piloted cable fire for a long time is not likely. (Tests reported in NUREG/CR-5387 state that cable fires,
. .spreading horizontally only as it progressed from level to level")
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- Maximum expected fire is a high-energy switchgear fire
- No credible fire reaches 7000F in this room (limiting fire scenario)'
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Results (cont.) ' 32 Comparison of,NRC and ANO Results Damage threshold
- - NRC: 425°F
- ANO: 7000..
- Heat release rate
- NRC: 500KW 500 fire peaking in,105 sec. 400
- ANO:. 100KW peaking 300 in 12 min (Scenario 1a) + cable fires and high energy fault in A4 0 switchgearandtcable fires (ScenarioIb) 0 250 500 750 0 High energy arcing fault - ANO1TenpScenaaOpenDoor in the 4KV switchgear \v-NRC: Not arialyzed tNRCULTenp ANO: Limiting scenario in terms -ANO-ULTefp, Scenario lb, Open D/ .
of its consequence, i.e., affected . D6.
circuits and timing I
- Neither analysis reaches 700 0 7F y
Frequency of Fire Scenarios 33 Fire risk = S (Scenario Frequency) x CCDP
- Scenario Frequency is derived from multiplication of:
- Generic fire frequency Based on the EPRI FIVE method (EPRI TR 105928 page 4-7)
- Severity Factors
- High energy weighting factor for the 4KV switchgear Based on operating experience (EPRI fire dat base)
- Prompt suppression of transient fires by plant personnel or fire watch
- .Manual suppression by fire brigade
- For scenarios that critical target is beyond plume, ceiling jet or flame radiation 2one,
- Next presentation discusses development of the CCDPj and' fire risk , ;
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Frequency of Fire Scenarios in Fire Zone 99M ANO SDP Analysis Results 0 . lI MFloor Raio of HE Pns y
'Gemeuic, oalf (IgnItion area ratio Beert event for a plant Pns by Sourcefreqency source F r severe personnel fin Results b" welghtIng fires awttchgear or fire brigade
__________________ feetof) fire watch la Fire Inthe A4 svitchgear. .
Nomnai valoes 100 KW fire - 1.50E2 2.50E41 5.88E-01 1.00E.0W 1.20E-01 2.50E15 1.OOE+Od 1.OOE+00 6.62E.05 lb High energy arcing fault any of the A4 sWtchgear breaker cubbhes 150E-02 2-QE41 _588E01 1.00E+00 1.20E-011 2.50E0-01 zI1.EO .0E+W 1.9fEf04
.2 Fire inth 655 MCC. Nominal 100 KW fire. Fires hI Inveter Y28 are bounded by his scenarlo.
_3 FbeIntheB58_MCC._._Naninal__ 1.50E-02 2:50E-01 . 5.88E-02 1.OOE+00 1.20E-01 IME400 1.OOE+0W 1.00E+00 2.65E-OS 3 -FIre Inthe 656 MCC. Nominal 100 KW fim 1.50E-02 2.50E-01 5.88E-02 1.00E+0 1.20E-01 1.00E+00 1.00E+00 1.WE+00 2.65E-05 4 Fire In the Y22 Inverter. Base case 100 KW fre. F~s Y24 a.nd Y25 are bounded by this scenario 1.50E-02 2.50E-01 5.88E-02 1.00E+00 1.20E-01 I 0E+0 t00E+00 5.00E-01 1.32E.OS 5 Flre nthe Load Center 8.
100KW nominal HRR. *1.502E-02 2.60E-01 5.88E 02 1.00E+00 1.20E-01 1.00E+W _1.00E+00 200E-01 5.2 E 6a TransIent fire Inareas of the room where cable trays are exposed to . ..
a "for-basedfire. NomInal Value of 150KW. 3.60102 2.O2E+/-
_E1.00E-01 1.00E+00 1.00E+00 1:00E+00 I00E-01 6.48EC05 eb Cblo fire caused by welding and cuttg In reas of th roon where cable trays are exposd to a foor..
. . tiased fire. Nominal Value of (I 1 501tW.
_____ 1.30E-4 2.OOE+
. 2.00E402 1. 1.00E+00 1.00E+00 5.OOE-02 1.00E+00 2.60E47 NRC SDP Analysis Results (May 15, 2003 Supplemental Letter-Page 25)
-S Source -Frequency
-Eletricai cabinets , 2.3E-04 1*..
Transfonfiers , 1.6E-05 Ventitation Subsystems 4.4E-06
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Fire, 9 Modeling ..Summary 35
- Maximum expected fire scenario in fire zone 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 secondary cable fires l l \ ,
- Credible firis will not result in a hot gas I yer (imiting fire scenario) in excess of the cable failure temper ture.
- Zone wide damage is.not credible,,
- Adequate margin V i
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Probabilistic Safety Assessment~~~~~~~~~~~~~~~~~~~~~~
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- r Introduction I 37 I I Key'system fiurds in99M Affected components due to cable failure Key operator actilon/respons , X.
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- Simultor scenaloesdut re I. I Operator action probabilities CQDP calculation /
- Delta CDF determination
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Risk Informed Strategy for Zone 99M 38 Fire Modeling 99M
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, S W~gWy S l C 9 ). , [ ¢ t ~~~~~Procedure EOP/AOP/Prefire Plans Simulator ExerciseDeveloment
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HRA Development PSA-
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Key Systems Affected in the Risk-Significance Determination (Fire Zone 99M)
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The following systems/trains are reCtae to fire induced power losses of A4 an
- One train and the swing pump of service water One'train and the swing pump of HPI (makeup) V f \
- The A4 associated diesel is no longer usable QC S
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Circuit Analysis 40
- Detailed circuit analysis performed on zone 99M
- Investigation of cables located in the trays and conduits I
associated with the target sets
- Analysis showed no loss of offsite power associated with zone 99M
- NRC evaluation did use loss of offsite power
- Analysis of associated failure modes for affected cables Failures unrelated to safe shutdown also examined to-provide accurate portrayal of the risk caused by the fire
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Systems Affected in the Risk-Significance Determination (Fire Zone 99M)
Scenario specific failures are based on cable location; subsets of the following are impacted for each scenario
- EFWflow control valves Loss of power to these valves will fail them open (desired state)
-Subsequent spurious operation not probable
,7 - DC control power to the A3 switchgear fails
- Breakers to remain static and require manual closure at the switchgear
- P-7B (motor driven EFW pump) suction valve could spuriously close Cable C in conduit; purious operation not probable but assumed in evaluation l~ ;Steam admission valves for P-7A (turbine-driven EFW pump) s'.
- Requires local action to start P-7A Aux-lube. oil pumps for the unaffected HPI train
- Requires local start of HPI pump when affected C-
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Operator Actions/Response in the Risk-Significance Determination (Fire Zone 99M) 42 A subset of the following operator actions are required in each 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 HP) A.
- Starting HPI fr make-up (long term action)
- May require, local start of pumps depending on fire scenario
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- Emergency diesel generatorrecoveries were not necessary
.. due to the lack of a LOOP event
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I~~~~~~~~~~~~~~~~~~~~~~~~~ I Previous,. ~~ ~~~ ,
Procedures .
vs ZoneSpecific I 43 Procedures
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- Previous procedures
- Combination of EOP/AOP/Pre-Fire plan
- Opporqunistic approach'
- Plant condition determines action, I
- New procedures Zone-specific fire procedures
! -Tactical approach
- Reduces impact and probability of spuric)us operations 7~ 'II
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Summary
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4i tg y:;pti .4{~-rF i - jr > a Q2 NewPmedur.e -l e
hprevous procedures discuss this ngreat Starting EFW P-7A detail. Spurious and false Indicators are not I manually and positioning mentioned which could delay operator Discussion'in hw procedure includes ndicators.
< , * ~~~~~~~~~~~~~~~~~functional associated Valves , ,, response.
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Lack of adequate and, correct indication s I,I I Cotrolin EFW (A or B) r~spoeue ics hslclo 'Ldirectlyfddiscussed procedure t Inithe inew'iii 2 Cotroiing nro2gcontrol FW{ArB}Prpoius romr action. discuss this local or procedures
, _ tsooverfill prevent , ; , which makes this action more likely in the new rocedure.
I L al closing pf bug A3 ibis actloo not explicitly discussed in the
. i N ssv tchgear for P-7B and nomnal operating procedures but is discussed The newprcedue'explicitly ddresses A (e.g., inrerter fires)' fi Altemate Shutdown.
IP locally closing these breakers.
I Discutsed In prevous procedures. The timing of this action depends on when letdown is The new procedure addresses the isolated.possIbility of starting the HPI pump locally.
Isolation of letdown to Inboth the pre ous and new procedures, this Inboth the previous and new procedures, 5 amid needing HPI action Is discussed and can be performed in this action Is discussed and can be (Makeup) sooner the control room. performed in the control room.
Inboth the previous and new procedures, this Inboth the previous and new procedures, 6 Switch t6 recirculation action Is discussed and can be performed In this action is discussed and can be ongter ,oog ,the control roomri. ' pertbmed in the control room.
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Simulator Scenario for Zone 99M 45 Fire, damage chosen to provide HRA information for multiple oprrato r actions Fire model beginning with an A4 switchgear fire
- Fire propagated throughout zone causing wide-spread cable damager
- Damage for scenario extends beyond credible fires Realistic control room communication challenges
- Fire brigade leader communication.
- Tirnelines based'on actual fire drill
< Included need to contact local fire department
- In plant auxiliary operator used for operator actions
- Radio and telephone communications used
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C-Simulator Scenario . I.I ,
for Zone 99M I
46
- ~.1 Simulator scenario failures included:
- Direct failuresi,
- . A4 switchgear (4KV) J B6 load center (480 VAG)
-. Included spurious operations y
- P-7BEFW suction valve closed at T=15
- ipcluded failed and incorrect indications
. Multiple panel indications failed (EFW, HPI, P )wer) (
Randomi annunciators spuriously alarmed
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Simulator Scenario for Zone 99M 47 I (,.!
- Three crews ran simulator with previous procedures,
. Two crews ran simulator with training on zone specific fire procedure One crew with each procedure contained operators in the plant. simulating local actions
- Controllers were present in the field to evaluate local manual'actions
- Time to locati'on Potential hazards i*
- Communication barriers
- Observers in the simulator to evaluate control room actions X S
- Including time to perform in control room actions
- Procedure usage
- Work practices due to loss of indications
Risk Informed Strategy for Zone 99M .
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If Simulator Scenario Development '1 e Circuit Analysis &
Location Evaluation I
99M II 91
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/* K Total Unit Risk 2-A.
Simulator Scenario Results 49 Simulator runs using previous EOP/AOP/ Pre-Fire plan approach and unrehearsed crews (3)
EOP approach for plant trip provided adequate initial core cooling Pre-fire plan used. tp show faulty indications and possible local actions
'Crews responded appropriately and in a timely manner
-. Plant maintained at a safe stable. state Simulatorruns with crews trained on new tactical 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 Crew performance using either previous or new procedures met Appendix R requirements for achieving safeshutdown
Risk
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Informed Strategy for Zone 99M . 50 Circuit Analysis &
Simulator Scenaric Location Evaluation, Fire Modeling 99M , Development
- 1. I 99M j
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{EOP/AOP/Prefire Plans Simulator Exercise Procedure Development 99M
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- HRA methods for quantification demonstrate there is an
.impactof fire on reliability of human actions PreviQusivs new procedures for shutdown
- Previous procedures use an opportunistic approach to control, where crews respond to cues and symptoms by selecting EOPs for that condition with the aid of pre-fire plans New procedures assist crew to; respond using a more tactical
'control process Use of either approach demonstrated
- Identifying symptom or cue will generate appropriate response for either procedure
- Ability to recover from spurious actuations Enhanced in new procedures I'~~~~~~~~~~~iI
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Method for Updating HRA Assessments to Account for Fire 52 The current Unit 1 model for human recovery actions in internal events PRA is based on a time reliability curve' HRA accounts for operational context by adjusting parameters such as:
Rule-based vs knowledge-based behavior
- No burden vs burden
'Other'performance influencing factors
- In,current assessment, effects of fire are not addressed nor are model parameters available; therefore, a different adjustment
,method was identified
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. . . . .~~~~~1~
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N EPRI Calculator
- Industry sponsored method provides a process for book keeping HRA evaluations
- Includes several methods for quantification
- Industry and NRC sponsored
- Generic data quantitatively differentiate human error probabilities (HER's) for key characteristics of procedures and man machine interface
- HRA analyst judgment is still required
. ,~~~~~~~~~~~~~~~~~~~~~~N x
. ,~~~~~~~.
Evaluation of Fire Impact on Probability 54 Seven cognitive assessments on differences in procedures
- Availability of information
-> Failure of attention
- Misread/miscommunication data
- Information misleading
- Skip a step in procedure
- Misinterpret instruction
- Misinterpret decision logic Probability of execution also calculated for fires based on inputs in the HRA calculator' Sf .5 .. ,
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Summary of AHEP Applications Due to Fire 55
. I -
Case' Event ID Basic Event Description A Pcop A Pexe A HEPgre I FIREOLDP Actions are carried out within the 9.8E-03 7.50E-04 1.1 E-02 control room - previous.
2 FIRENEWP Actions are carried out within the 2.6E-03 6.10E-04 3.2E-03 control room - new 3 99-MFIRECR Realistic fire in 99M decisions in 9.8E-03 2.OOP-02 3.OE-02 99;MslRE9'
, . control room with local manual
. .. . ~~~~actions; 99-MFIRECRE' . Realistic fire in 99M early control 4.7E-03 4.3E-04 5.1E-03
____room ____ actions , ._. _
_5 99!MFIRELOCAL Local actions taken by field operators 1.5E-02 2.6E-02 4.1 E-02 6 NotFeasible . ' 1 1 1 7 i No Change) 0 0 al
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' Comparing HFEs from PRA baseline with HFEs i 99M fire Fire in 99M .increases.human failure event (HFE) for typical feasible actions over initial intemal evepts PR from zero to a value in range of 3E-3 to 4E-2 for various scerarios and conditions If action is not feasible, then HFE assessment is set at 1.0
- Very small difference in impact of previous versus new procedures Comparison of previous and new procedures on the HFEs for fire Impact In 99M
~~~~~~~~~~~41 prrrviouWs rocedlTure:IIlHFl )L_
t . _ _ _ _ . _ _ _ _ ~~~~~~~~~~~~~~~~~. I I I I __._ ___
_ Kombined Fire Value usir (0.0001 0.001 0.01 0.1 '
Value of current HFEe In the Bn PRA model
Human Error Probability Comparison 57
- NRC approach assumes zone wide damage at time zero'
- NRC approach included loss of offsite power Operator Action NRC Value No NRC Value ANO Value ANO Value
.Procedure W/Procedure Previous Newiv Establish EFW 1 0.6 0.11 0.098 (A3 local start) ,
Establish EFW ' 1 0.6 0.038 0.026 (Control EFW).
Establish Feed 0.75 0.55 : 0.008 0.008 Bleed Estaplish Feed 0.75 0.55 0.11 . 0.098
& Bleed
/ (A3 Local Start) -.
Secure Diesel 0.75 0.55 Not needed due to no loss of offsite I- . . with no Service -power Water - .
Risk Informed Strategy for Zone 99M 58 Ir Fire Modeling. 99M H, _ .
Simulator Scenario Development Circuit Analysis &
IrLocation Evaluation 99M
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_-P Simulator Exercise D 'evelopment
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CCDP Determination for Zone 99M 59
- Eight fire scenarios in zone 99M quantified
- Current Unit 1 PSA model used Fire modeling targets used to determine failed components 94* Spurious operation probabilities used in high-energy electrical fault scenario lb
- All other Scenarios conservatively assume the spurious operation will occur.
All components failed together (conservative)
'Timings only used to disallow spurious operation of components whose control cable would be lost after power loss
., /~~~~~~~~~~~~~~
> . .. .~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
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SDP Process Review 60
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- Created, eight fire scenarios Used, fire modeling/characterization
- Determined falures for each scenario
/~~~~~~~~~~~~~~~~~~~~~~~~~~~
Used simulator exercises and industry experts to determine reliability of necessary operator actions Combined interaction into plant specific PSA model
- Calculated, change in risk between the previous and new procedures\ . <.
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. ~~~~~~~ ~ ~~~~~ ~~ ~ ~ ~~~~~ ~ ~ ~~~~~~~~~~~~~~~~~~~~~~~~~~~- .
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Fire Risk in Zone 99M j 61 I . I
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ILe.l8E U.~~~~~0 to Flm in the B55 MCC. Nom l ,
Fire Inthe A4 switcCNgearn ,
a Nominal value, 100 KW fire 6.62E-05 3.12E-04 1.06E-04 2.06E-08 1.37E-08 6.98E-09 High energy 100arcing fault Firesin i-^N anypof Infre.
Inverte Y28 . ,
the A4 switchgear breaker lb cubice. I .99E-04 I1.28E-03 9.01 E-04 2.564E-07 I1.791E-01t 7.55E-08 k Fire in the B5 MCC. Nominal 100 KW fire. Fires In Inverter Y28 2 lare bounded this byscenario. 2.65E-05 2.78E-04 79E-04 7.35E-09.. 4.74E-09 2.61 E-09 Fire Lthe B56 MCC. Nominal 3 100KW fire. 2.65E-05 2.78E-04 1.79E-04 7.35E-09 4.74E-09 2.61E-09 Fire in the Y22 Inverter. Base.
ce, freKW fire. Fires InY24 an, w aY2 are bounded y by this 4 scenario. 6.32E-05 3.98E-05 3.86E-05 5.27E-10 5.10E1-10 760E-1 Fire InLoad Center B6.
j5 100KW nominal HRR. 5.29E-06 3.02E-02 1.88E-02 1.60E-07 9.93E-08 6.07E-08 Transient fire Inareas of the room where cable trays are exposed to afloor-based fire. Nominal value 6a of 150KW. 6.48E-05 3.24E-03 2.12E-03 2.10E-07 1.37E-071 7.25E-08 Cable fire caused by welding and cutting in areas of the room where
- ~cable trays are exposed to a floor-based fire. Nominal value of 6b 150KW. 2.60E-071 3.24E-03 2.12E-031 8.41E-10 5.50E-10 2.91 E-1 0 99M 6.61 E-07 4.39E-07 2.21E-071
l
'.hCr.iticalRisk I nputs 62 Time-phased fire induced failures are a critical element
- Realistic.assessment of fire progression, failures in 0 - 60 minutes (targets of the high-energy switchgear damage immediate, the rest time-dependent failures are from ensuing cable fire)
- Critical cable insulation used is thermoset
- 70 0 °F cable damage temperature
'Operator action probabilities
- New procedures offer slight HEP improvement over previous procedures
- Human reliability analysis: CCDP indicates that impact of.
AHEP is measurable but small
Risk lnformed Strategy for Zone 99M 63 4e Fire Modeling 99M Simulator Scenario Development L Circuit Analysis &
Location Evaluation
.999M 1i 4 2
.. ~~ ~ ~ ~~~~~-P Procedure
[EOP/AOP/PrefirePlans ] Simulator Exercise Development I~~~~~~~~~~~~I 99M 1 I C
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01
... IL 11*1 I , 11 PSA A - .
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Total Unit Risk 64
- Focus on zones that have delta risk due to the difference in manual actions between two types of procedures
- Qualitative review of zones where manual actions are utilized
- Alternate shutdown zones screened
- Zones with automatic suppression screened
- Agrees with NRC SDP approach - Suppression provides at least one
- 1 order of magnitude in risk results and provides time for operator actions to be performed 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 operator actions,
- Similar to NRC results, the following zones remain:
-1 OON
- 104S :/
L
Total Unit Risk 65
, The assessment of fire risk in 99.M was extrapolated to two other Unit "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)
- Unit 1 A3 4KV switchgear room (OON)
- Similar to 99M in combustibles and fire sources
- Considerably less redundant train cable routed through zone
- Unit 1 electrical equipment room (104S)
& Lack of highs energy switchgear
- Considerably tess redundant train cable routed through zone Each zope is bounded by the results of 99M Conservative estimated fire risk (ACDF) for this condition Unit ,< 6. 07/yr I
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Unit 2 Risk 66 The four Unit 2 zones identified as risk significant by NRC wore qualitatively evaluated
- Each was assessed with walkdown and examination of the
.. t potential fire scenarios threatening the other train raceways
,(e.g., red train raceway in a green train room)
I
, . Conclusion,
- The four Unit 2 zones contain similar characteristics to the Unit 1 zone.s .1
- Two switchgear rooms
'Two rooms containing MCCs similar to 104S
, . . . N .~~~~~~~~~~~~~~~~~~~~~-
- " - The results from 99M bound these zones
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Summary of Risk Assessment 67
- ANO risk assessment conclUded that:,-,
Realistic fires will not achieve whole-zone damage as originally assumed in NRC evaluation'
- Realistic fires will result in time-phased damage of cables
- Manual actions required to achievesafe shutdown for a fire in zone'99M are. redible Simulator scenarios validated that operators could achieve safe shutdown.
- Met Appendix R requirements for achieving safe shutdown Conclusion& r.
~~.- Delta C.DF - '
Unit 1< 6_6-07/yr .
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.1 I OVERALL
SUMMARY
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Joe Kowalewski i e
.Director, Design Engineering, N
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Summary.;
., 69
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Detailed analysis of zone 99M
- Credible fires result in time-phased failures without zone-wide
'-damage (7000 F damage temperature for thermos'et cables) i- Detailedwcircu'it 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
,.I..
I -
- Risk assessment of zone 99M conservative with respect to other
.I zones-.
- Conservative estimate of total unit L\CDF is < 6.6E-07/yr The significance of the use of manual actions to achieve safe shutdown has very low safety significance and should be characterized as GREEN
.1
- 1. I
I
. Overall S-ummary (cont.) 70
- Defense in depth strategy to prevent and mitigate fires
- Explicit control.of combustibles
- Fire brigade effectiveness
- Rrimarily rely on barriers or physical separation for equipment required for safe. shutdown /
- Fire detection and suppression
- Limited use of manual actions utilized for Appendix R compliance
- Actions taken to further reduce risk
- Validated circuit'analysis
- Feasibility evaluation of manual actions (IE 71111.05)
- New procedures developed to enhance operator response
- Fire detection reliability improved iANO can successfully achieve safe shutdown inthe event of a fire Jin cany zone
71 (I
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CLOSING REMARKS 6
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SCIENCE SAIC APPLICATIONS INTERNATIONAL CORP.
BIJAN NAJAFI, P.E.
MANAGER, FIRE PROTECTION SECTION EDUCATION:
University of Washington: M-S., Nuclear Engineerir, 1979 Shiraz University B.S., Electrical Engineering, 1976 -
Registered Professional Mechanical Engineer, State of California
SUMMARY
OF CURRENT POSITION:
Mr. Najafi is the Manager of the Fire Protection Section at SAIC responsible for overseeing a program that includes domestic and international nuclear utilities, DOE facilities and commercial/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 (EPR) fire risk analysis and fire protection program. Over the past decade he has been instruiental in development of the EPRI fire risk technology currently in use in the U.S. nuclear power industry. This technology has also been used internationally in Europe and parts of Asia and South America. Mr. Najafi has conducted trainting courses in U.S. and Europe on Fire Technology, most recently a series of Fre Modeling courses for nuclear power plant fire protection engineers.
Mr. Najafi is an active member 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 Ameiican Nuclear Society's committee for the development of the Fire PRA Standards -
- Participating-member of variou s taskforces at Nuclear Energy Institute including the circuit failures issues taskforce in the development of the NEI-00-01, "Guidance for Post-fire Safe Shutdown. Analysis."
Invited panelist at the NRC-Industry re-induced circuttilures workshop onFebruary 19, 2003.
_
- Member of the NRC-Industry team for the revision of fip protection Signifcance Determination Process (SDP) - -
Member of the NRC's International Collaborative Project to Evaluate Fire Models foiNuclear Power 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 "isk Assessmet Guidelines," in progress. - - - '
-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~I
,~~~ ~ ~ ~~~ -
SCIENCE' SAIC APPLICATioNs INTERNATIONAL CORP.
EMPLOYMENT HISTORY:
Mr. Najafi is the Manager of the Fire Protection Program at SAIC responsible for overseeing a business area that includes domestic and international nuclear utilities, DOE facilities-and commercial/indusbial facilities.
He is one of the principal investigitors for Electric Power Research Institute (EPRI) fire risk analysis and fire protection projects. These projects included development of EPRI's Fire.PRA Implementaion Guideand Fire-Induced Vulnerability Evaluation (FIVE) methodclogy and application of these technologies to US nuclear power plant support Over the past decade Mr. Najafi has been instrumental ir development of the fire research program at EPRI to support nuclear power industry move ,towards a Risk-Informed/Performance-Based (RI/PB) fire protection rule. Under this program data and methods are being developed a Snore 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 Evaluaing Cable Wrap Fire Barrier Performance."
As part of this process of continuous enhancement of technology, Mr. Najafi is currently the principal technical manager of a joint project between EPRI and USNRC office of Research for development of the next generation of Fire Risk 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-informed rule in fire protection. Mr. Najafi is the key in providing goals and directions to this prograr that includes the developmett of the first documented methodology for assessment. of fire risk during low power and shutdown modes of operation.
Between 1991 and 1997, Mr. Najafi managed Fire PSA projects at over eighteen (18) U.S. nuclear 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 GEs ABWR/SBWR Levd PRA, Comanche Peak Level I/II PRA support, Project Engineer (Technical Project Manager) for the Turkey Point Nuclear Power Plant (PWR-) Units 3 and 4 Level 2 PRA 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 (USI) A-46, "Seismic Qualification of Equipment in Operating Plants," for the Onah Public Power District staff.
During 1987-1988, he was the manager of a project toupdate the PRA for the Indian Point Unif 3,plant 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 program 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 Susquehaniia Steam Electric Station (BWJ.).
During 1985, Mr. Najafi was one of the pringipal authors of a PRA study for the Peach Bottom plant (BWR) as part of the NUREG-1150 program for Sandia National Laboratories. He was primarily responsible fortthe 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). his effort involved development of
-10g l ..
SCIENCE SAIC APPLICATIONS INTERNATIONAL CORP.
a computer-based probabilistic framework to integrate the vast body of knowledge that exists regarding LMFBR core disruptive accidents and their inherent uncertainty. The methodology not only estimates the uncertainties, but also can display the nature and extent to which the state of knowledge (or lack of knowledge) contributes to them. The potential application of the methodology to the PWR steam explosion events in large, dty containment was investigated The results of this study were published in the Nuclear Science and Engineering Journal.
Over the period 1982-1984, Mr. Najafi was the principal investigator of several system safety studies on the Clinch River Breeder Reactor Plant (LMFBR) that were presented to the Advisory Committee on reactor safeguards as part of a technical assistance effort for the NRC staff. Tbis effort covered a variety of limited-'
scope studies for both systems and consequence evaluations, including radioactivity release irequencies, 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 NRC to ensure that the LWR licensing requirements and associated Regulatory Guides that are applicable to LMFBR's are being applied to CRBRP.
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 (lMYBR) Nuclear Power Station in Kalkar, 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 construction.
Mr. Najafi has been one. of the principal participants in the risk reduction program conducted by the Nuclear Safety Analysis Center to investigate the PRA methodology for estimating incremental changes in plant reliability-and risk due to modifications. The methodology was validated using VEPCo'9 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 sudy. He was also the Task Manager and conducted the probabilistic analysis part of the accident evaluation chapter for the Seabrook Nuclear Power Plant (PWR-M) 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. bySATC. A steady-'
state model of the Morgantown plant using the PEPSE computer code was developed covering the boiler, turbine and balance of plant systems. 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 h as 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 Seisiic Safety Margin Research Program (SSMRP)for the Lawrence Livermore National Laboratories, where he dvooped the models for Emergency Core Cooling System and Residual Heat-Removal System for tlie Zion Nuclear Power Station Unit 1 WR-SQ. Later he developed a fault tree model for the auxiliary feeolwater system for San Onofre Nuclear GeneratingStation Unit 1 (PWR-W) to predict the systems reliability'under seismic loading. -
'\
SCIENCE S C APPLICATIONS INTERNATIONAL CORP.
SELECTED PUBLICATIONS: -
- 1. "Fire Modeling Guide for Nuclear Power Plant Applications," EPRI 1002981, August 2002.
- 2. "Fire Events Database for U.S. Nuclear Power Plants: Fire Initiation and Trends", EPRI 1003111, December 2001.
- 3. "A Pilot Plant Evaluation Using NFPA-805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants", EPRI 1001442, May 2001
- 4. "NFPA 805: Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants", National Fire Protection Association,2001 Edition (Contributing Author)
- 5. "Fire Barer Penetration Seal Handbook," EPRI 100196, July 2000
- 6. "SFPE Engineering Guide to Performance-Based Fire Protection", Society of Fire Protection Engineers, First Edition 2000 (Contributing Author)
- 7. "planning for Risk-Informed/Performance-Based Fire Protection at Nuclear Power Plants", EPRI TR-108799, December 1997
- 8. "Reducing Operations'and Maintenance Costs of Nuclear Power Plant Fire Protection Programs", EPRI TR-107337, December 1996
- 9. "Methods for Evaluating Cable Wrap Fire Barrier Performance", EPRI TR-106714, August 1996
- 10. "Fire Ignition Frequency Model at Shutdown for U.S. Nuclear Power Plants", EPRI TR-105929, December 1995
- 11. "Fire Probabilistic Risk Assessment Implementation Guide", EPRI TR-105928, December 1995 (and Supplement EPRI SU-105928)
- 12. "Fire-Induce Vulnerability Evaluation (FIVE) Software", EPRI AP-100530, February 1994
- 13. "Automatic and Manual Suppression Reliability Data for Nuclear Power Plant Fire Risk Analyses".
NSAC-179L, February 1994
- 14. "Fire Risk Analysis Code. FRANC", EPRI AP-103733,January 1994
- 15. "Fire-Induced Vulnerability Evaluation (FIVE)"' EPRI TR-100370, May 1992 (Contributing Author)
- 16. "Fire Events Database for US. Nuclear Power Plants", NSAC-178L, December 1991
- 17. 'Reference Plant Accident Sequence Iik1lihood haactezation: Peach Bottom, Unit 2," NTJREG/CR-4550, Volume 3. (With Alan Kolaczkowski, et at)
- 18. "An Assessment of Steam Explosions Induced Containment Failure," NUREG/CR-5030, February 1989 and Nuclear Science and Engineering Pecember 1987. ; -
- 19. "On the Probabilistic Aspects of a-Mode Containment Failure," T.G. 1heofanous, B. Najafi and E.
Rumble, Nuclear Science and Engineering, November 1985. - . -
20.- Incorporation of Phenomenological Uncertainties in Safetj Analysis Application to LMFBR Core Disruptive Accident Energetics," Proceedings of ANS/ENS International Topical Meeting on
- Probabilistic safety Methods and Applications, VoL 1, San Francisco, CA, February 1983.
21,,. "SSMRP, Phase I, Systems AnAlysis," NUREG/CR-2015, November 1981 (withJ.E. Wels, et al)-
- Data Systems
~' -& Solutions G. WILLIAMHANNAMAN, PHD EDUCATION:
PhD, Nuclear Engineering, Iowa State University, 1974 MS, Nuclear Engineering, Iowa State University, 1971 BS, Electrical Engineering, Iowa State University, 1965 WORK
SUMMARY
-Dr. Hannaman is a Professional Engineer with over 25 years of progressive consulting experience in solng 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) techniques 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. Supporting elements include data collection from training simulators, database development and integrating the rsults into risk and reliability studies to identify management priorities for enhanced design, operation, and maintenance.
PROFESSIONAL EXPERIENCE:
1999 to present, Senior Staff Engineer, Data Systems-& Solutions 1988 to 1999 Senior Staff Engineer, Science Applications International Corporation Recent Projects
- Support for EPRI projects in the 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 opertors.
o Support development of a procedure for addressing HRA in fire PSAs o Upgraded Monte Carlo Simulation software (STEIN) for effluating the impact of NDE measures on structural integrity o Developed template for performing Human Reliability Analysis - lesson plans
~o Support project on methods for evaluation of organlzatioial 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/ niclar waste mixtures for .
DOE/Bechtel.
SteamGeneratorAssessmentSofiwaredevelopment '
- 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 conning data~from tests on pulled SG tubes to.
Ajoint venture between Rolls-Royce and SNJC
G. Kiiam Hannaman Page 2 of 7 support burst and leak rate correlations. The secure web site developed under an ISO-9000 and I0CFR5O approved quality program supports data searches.
- Product manager for' development of the STEIN Monte Carlo code for use in evaluating Steam Generator ODSCC NDE results to predict operational assessment and condition monitoring criteria.
- Developed methodology using Monte Carlo Simulation of uncertainties for assessing margin between an allowed 1131 dose and a predicted accidental release from degraded steam generator tubes.
Human ReliabilityAssessments
- Planned and documented human reliability assessments (HRAs) for four utilities as part of their IPEs.
- Developed and delivered a weeklong training course-on HRA to Eletronuclear in Brazil.
- Supported update of-VC Summer IPEEE fires assessment as HRA task leader under SAIC and VCS quality assurance programs. Evaluated risk of using fire emergency procedures for the current control room configuration. HRA methodology used NUREG/CR 4772, &1278 and EPRI-TR-100259.
- Instructor on the subject of human reliability for Argonne National Labs Inter-regional Training Course on Prevention and Management of Accidents at Nuclear Power Plants
- eManaged 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 programs, 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).
- Supported use of control room training simulators in URA studies for six utilities including Hope Creek and Laguna Verde.
PRA andRiskInformed Applications
- For Entergy 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 the risk 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 Environmental Impact Statement forthe NPR-MHTGR Risk management
- Supported developient of methodology for blending' risk-informed PSA with deterministic rules to demonstrate compliance with NRCs regulations governing steam generator operatibn. -
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- Developed qualitative risk assessment methodology nd 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, 5481.1B, and'standards 1027-92 and 3009-94.
- Applied methodology on two facilities (Rocket launch and Accelerator). esults support safety documentation suitable for a facility safety analysis reportii 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 usi'g test demands and-low conditions to improve risk assessment results.
G. William Hannaman Page3 of 7 Ram analysis Supported the MHTGR conceptual design through incorporation of applicable operational experience, development of technical position papers to demonstrate that lessons learned from previous operating experience were considered in the advanced design, ind updated safety, availability, and plant capacity factor reports working with Stone and Webster Availability Assessment tearn; This involved building reliability block diagrams for various systems to evaluate reliability and risk.
Oversightprojects
- Served as secretary on senior review committee to evaluate selection criteria for the NPR-MHTGR containment.
- Project manager for independent reviews of PSAIHRA and human factors for Union Fensoa on a Spanish Reactor to identify cost effective risk reduction upgrades for control room interface
- Review of a spent fuel processing design for a DOE site.
- Performed review of human reliability assessments in the IPEs,
- Performed independent safety reviews of safety analysis reports and risk assessments including analysis of spray leaks during tank transfer operations, and evaluation of two different 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 performing in US and internationally (e.g., EdF).
- Transferred technology via: (1) seminars, (2) reviews of PRAs and HRAs, (3) HRA task definition and supervision of analysts and (4) guidebook development such as PRA procedures guide and HRA guidelines for specific 'projects (5) performing benchmark comparisons, (6) performing analysis, (7) reviewing work, (8) 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.
i
- Detailed safety reviews of design concepts such as thedvanced modular gas turbine reactor.
1974to 1981 StaffEngineerGeneralAtomics-,
- Performed probabilistic safetyanalysis, 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 iystem 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 fbo Chemical and Process System An-ysis Group on a 6man-year effort to collct date and develop-reliability evaluation methods including eliability block diagrams for process system hazard analysis reliability- allocations, reliability predictions, availability, and mitainability 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, ontrol, heat renmoval, 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.
f -
G. Wiflam Hannaman Page 4 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 models and common-cause failure models.'
1970 to 1974, Graduate Assistant and Senior Reactor Operator, Iowa State Universityl
- Obtained licenses for reactor operator and senior reactor operator through the NRC on a university training reactor, with 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 training.
1965 to 1970, Supervisor, Westinghouse Electric Corporation-Apparatus Repair Division
- Planned repairs and directed maintenance crews on chemicaL utility and industrial sites and in repair plants for over 10,000 unique power system equipment failures.
- Designed and implemented an I&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 tsting electrical and mechanical apparatus including electric motors, breakers, controls, transformers, generators, turbines compressors, magnets etc.
- While in Westinghouse's Giaduate Student Program performed rotating assignments in manufacturing facilities for transformers and apparatus repair.
COMUTER PROFICILENCY:;
Language/Tools:Microsoft Office Software, Math software, Monte Carlo Sirmulation, CAFTA HardwareSystems: PC, and Mac OperatingSystems: Windows 95, 2000, XT; OS8, and DOS MISCELLANEOUS ProfessionalAssociationsand Memberships:
State of California - Professional Nuclear Engineering Registration NU 1948 Since 1982 Member American Nuclear Society - -
San DiegoSection chainnan 1979 '
'San Diego Section executive committee, various years Technical program chairman for 9mbeddM topical neeting on Advanced Nuclear Installation Safety, 2000,
- Assistant Technical Program Chairman for Risk Management -Expanding Horizons 1992.
Human Factors Division, Executive Committee, 1987.
Safety Division Program Committee 2000. - /
Organized and chaired numerous technical sessiorA'for ANS. -
Paper reviewer for Nuclear Technology Member of Institute of Electrical and Electronics Epgineers Corresponding member of the Nuclear Eiineering Subcommittee SC-5 on human factors and reliability responsible for standards on reliability methods. 2000 -2003 SC-5 Committee member on Reliability 1976 to 1980, '
SC-7 Cdnunfttee member on Human Performance 1984-1986.
Organized aid chaired technical sessions at an EE meeting Society for Risk Analysis Executive committee of the Southern California Chapter in 1989.
Organized and chaired technical session at PSAM II.
G. Wliam Hannaman Page 5 of 7 Patents, Selected Publications, and Awards:
- Elected to Sigma XI, the research honor society in 1973 Elected to National Acadeny of Sciences 6-member panel on cooperation with USSR on reactor safety to identify needs and means for enhancing reactor safety. 1987
- 'Elected to Strathmore's Who's Who 1996X03
- Outstanding technical paper awards in ANS Meetings (e.g., ANS Midwest student conference 1974 and ANS summer meeting Human Factors Division 85, 88, and 93).
- Toastmaster CTM and ATM levels and Toastnaster of the year for Area 17 District 5 1999-2000
- Academic credit for Reliability Assurance, UCLA 1975 Global Business Management, University of Phoenix 1998 Reports Haniiaman, G. W. and L B. Wall, "Lesson Plans for Human Reliability Assessments in PSAs," EPRI 1003329 June 2002.
Hannanan G. W (DS&S), Y. Durbec and C. Bauby (EdF), "Feasibility Study for the Integration of EDF's models for PWSCC into EPRI's STEIN code," Joint EDF and DS&S Report to EPRI, May 19,2002.
Mickey M.B.,G. We-Hannaman, B. W. Johnson, K. M. Batemen," Verification of IHLW Product Quality by Analysis of Uncertainty and Reliability in the HLW Process Control System," Data Systems & Solutions Report to Bechtel National Inc. May 2001,-
. Hannaman G. W., and S. A, Fleger, Evaluation of HCR Methodology Implementation in PSA and Control Rdom Human Factors Review for Jose Cabrera Nuclear Power Plant, EPRI, Palo Alto, CAApril 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. Rumble, G. W. Hannamian, and M Kenton, "Risk Assessment Methodology for 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, C 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. Hannaan, "241-Y-101 Mixer Pump Lifetime Expectancy, for
-Westinghouse Hanford Company Richland, Washington, Nov. 1995. - '
Mahn-J. A., G. W. Ijannainan and P. M. Kryska, -Qualitative Methods for Assessing Risk," Sandia National Laboratories, SAND95 -0320, Albuquerque New Mexio May 1995. '
Hannann, G. W. Transforming, PRA Results into Performance-based Criteria Tor PWR steam generator Inspecfions and Management" White paper on EPRI project 550-07, March 1995 Otis, M. D. D. A. Bradley and G. W. Hannanian, Technical Basis for-Considering Uncertainties in 1131 '
Release and Dose Limits for a Postulated Accident EPRI TR-1 03878. EPRI, Palo Alto, CA March l996. -
Hannaman G. -W., W. Parkinson, and C. Donahue, Lessons Learnied from Documented Events about Human Reliability during Less Than-Full Power Operations, EPRI report TR-104783, Sept. 1993.
Hannaman . W. C. G. Donahoe and E. M. Dougherty, Insights from Human Reibility Assessments Perforied during Less Than Full Power Operations, EPRI, report SAIC-92/0056, SAIC San' Diego CA, March 1992.
-- -NSAC l54I5SLOCA Evaluation Guidelines," HRA methodology, EPRI, Palo Alto, CA Sept. 1991.
Hannaman G. W. and . Forester Analysis of Initiation of Boron Injection in Response to an ATWS, SAIC-91/1132 SAIC Report forTask 2 of Gulf States Utilities River Bend project, April 22, 1991.
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G. Wilam Hannaman Page 6 of 7 Hannaman G. W. and R J.. Budnitz, "Case Study on the use of PSA methods: Human Reliability analysis," LAEA-TECDOC-592 International Atomic Energy Agency Vienna Austriaj April 1991.
SHARPI - A Revised Systematic Humans Action Reliability Procedure, (with G. Parry, A. Spurgin and D. Wakefield), EPRI NP-7183-M, December 1990.
Contributor to Operator Reliability Experiments Using Power Plant Simulators, EPRI NP-6937 Volumes 1, 2, and 3, July 1990.
Hannaman G. W. Application of SHARPI to Interfacing System Loss of Coolant Accidents (ISLOCAs),
SAIC-90-1351, Science Applications International Corporation Report on EPRI Project 3206-14, September 19, 1990.
P. Lobner, L. Goldman, G. W. Hannaman and S. Langer Preliminary 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. Hannaman and J. Young, "Perspective on Application of Probabilistic Modeling Techniques to th Heavy Water 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.). LAEA-TECDOC-499, Technical Document issued by the International Atomic Energy Agency, Vienna, 1989.
Hannaman G. W. F. S. Dombek and Y. D. Lukic, Evaluation 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, December 1986.
Incorporation of Transient Response Implementation Plan Procedures into the Limerick Generating Station Probabilistic Risk 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 453 1;October 1984.
Hannaman W., and A.' Spurgin Systematic Huinan Action ReliabilityProcedure (SHARP) EPRI NP-3583 June 1984.
Review of the Sizewell Probabilistic Safety Study," (with S. Levine ) NUS-3446, April 1983-'
Hannaman G. W., W. Breher, R. Cantrell, and H. Hopkins, Reliability, Availability, and Maintainability Plan for the Solvcnt Refined Coal Demonstration Plant, V I & II. GA-C-16372, Solvent kefined Coal Int., Inc., July 1981 .- 'a
Hannaman G. W,\et7i1. Safety Program Plan - Sunmmary. USDOE Report GA-C-16244, Volumes II, and I performed for Solvent Refined Coal International, Inc., January 1981, App. June 198 1.
Hannanan G. W., et a. HTGR-RPR Capacity Factor. Estimate. GA-A-16242, General Atomic Co.,
January 1981.
Hannaman G. W. GCR-Data Bank Status 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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G. William Hannaman. /
Page 7 of 7 Papers Hannaman. G. W., "Safety Valve Reliability Assessments for PSAs," PSA 2002, ANS Probabilistic Safety Topical Meeting, Detroit Oct. 2002.
Johnson, B, G. W. Hannanan, and M A.Stutzke, "Operating Reactor Safety, Regulation and the Real World,"
in ANS Proceedings Operating Reactor Safety Topical Meeting, Oct. 1 -14,1998.
Fuller, Ed, E. Runble, G. W. Hannaman, and M. A. Kenton, " Assessment of Risks from Thermal Challenge to Steam Generator Tubes During Hypothetical Severe Accidents," in ANS Operating Reactor Safety Topical-Meeting Oct. 11-14, 1998.
Mahn J. A., G.W.-Hannaman, P. M. Kryika, "Qualitative Methods for Assessing Risk" 1995 ASME conf, 1995.
Hannaman, G. W. and Avtar Singh, "Human Reliability Database for In-Plant Application Of Industry Experience," PSAM IL 1993 Hannaman G. W. 'and A Singh "Assessments and Applications to Enhance Human Reliability and Reduce Risk during Less Than Full Power Operations" of EPRI,,ANS Risk Management embedded topical,June 1992.
Hannaman G. W. "Human Reliability Methods for Enhancing Performance," in Risk Management Expanding Horizons Henisphere Publishing, New York, 1991.
Hannaman G. W. and D.H. Worledge, "Some Developments in Human Reliability Analysis Approaches and Tools', Reliability Engineering and System Safety, Elsevier Publishers Ltd. England, V22 pg 235-256, 1988.
Hannaman G. W., F. S. Dombek, B. Y. 0. Lydell, and Y. D. Lukic, "Using Risk Analysis to Improve Testing and Maintenance". Forth EEE Conference on Human Factors and Power Plants. Monterey CA June 1988.
Hannaman G. W., G.R. Crane and D.H. Worledge "Application of a Human Reliability Model to Operator Response 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 Assessment and Risk Management PSA'87, Zurich, Switzerland, September 1987. a Hannaman G. W. et. al. "Applications of Human Reliability Models to Structure Measurements of Human-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. A Hanaman G. W. "The Role of Frameworks, Data, and Judgment in Human Reliability Analysis", Nuclear Science and Engineering. North Holland Publishing Comnpany, NEDEA 98L93, May 1986.
Crane G. and G. W. Hannaman,"Realistic Operator Response Measurements: Inputs to La Salle PRA", V 5, International Topical Meeting-on Nuclear Reactor Safety No. 700106, ANSi La Grange Park] L, Feb. 1986.)
"Synthesis of Experiefice Data for Risk Assessnent d Design Irnprovementof-Gas-Cooled Reactors"
.(with A.P. Kelly), Pro&e#ings of -Probabilistic 'Analysis of Nuclear Reactor Safety, American Nuclear Society, IL, May 1978. -
Probabilistic Risk Assessment of HTGR's (with Fleming, Houghton, and Joksimovich), Reliability Engmeering, Applied Science Publisher Ltd., England (1981) pp. 17-25.
Treatment of Operator Actions in the HTGR Risk Assessment Study, GA-A-1 5499, Winter ANS Dec.1979
- 1eming KN. and G. W. Hannanal "'Common Cause Failure Analyses in the Podication of Core Cooling .*
Reliability" EEE Transaction of leliability,.Special Issue on-Nuclear System Safety and Reliability, R-25 Number 3, August 75.-
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