ML033580469

Appendix R Regulatory Conference with Comments

ML033580469
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: ML033580469 (10)

Text

T1 --

• (*;, '-..

' 1 S

1 I 7 1.

I 0 l. N

.I UJI 141' I'

ARKANSAS NUCLEAR ONE APPENDIXR

- REGULATORY CONFERENCE I.

I

(

July 10, 2003 k..

I, I I I .. -  %% I

-0e

. cb

4.'.

... ;tv.

l-..

- 1. ....

)

I

..T -

hA L

LV, 7 -

OPENING REMARKS

.1 J 1 I

II I I.

el

~ ~ I I I~ I ~

'I i

I, i

Craig Anderson

. Vice President, ANO I!

A/

. t

, "M

' im Risk Assessment Overview' i

i 7 4K.

C-v

• NRC's preliminary SDP evaluation concluded unacceptable (greater than green) increase in core damage frequency I .

• Key assumptions in.NRC evaluations vs ANO's

-pr.eli"minary assessment ' . *r 1

- Heat release rate

- Human error probability 10 -

S Subsequent site-specific in-depth assessment...

- Results incorporated into Unit 1 PSA model 'to derive ACDF I

.. \

L.  :

-I, I ,.-

e, CD-Unit 1 4KV Switchgear Room (fire zone 991 EC204 EC204 EC204 EC205  ;

EB203 EB202 EB202 E8202 E8203 EC236 EC236 , . v (enters)I J-.

L I'

I

.I 4 - ' -o itv-

,. ( ..

- eD f'ire (Characterization

• Electrical cabinet fires .I The-heat release rate data profile is' HRR Profile based on the best available fire test i-data 120 Sandia National Lab (NUREG/CR-45 27, 100 87/88) and VTT (Valtion Teknillinen Tutkimuskeskus, .94/96) in Finland 80

• Same test used in the NRC SDP analysis 60

- The ANO.. HRR is based on the highest 40 j

peak of.ST5 (unqualified, ope'n 110 20 KBTU loading) and all qualified, vertical cabinets (excluding PCT6 and test 23 0.

with 1.5 MBTU loading) 5 10;

• The NRC. HRR is based on test 23 (qualified, open 1.47 MBTU loading) and Time [min]'

test 24 (Unqualified, open, .1.44 MBTU)

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

- Tests are based on control'panels (I I I

- The switchgear, MCC's and load centers are enclosed with sealed penef rations I I I I.,

• . Used for-scenarios Ia, 2 - 5 C, rC

-I.

Results (cont.) 32 Comparison of NRC -and ANO Results lDamage threshold;

., Damage thResh:d 0

.Upper Layer Timperature in,99M-

-NRC: 425 F

- ANO: 7000F 800 Heat release rate 600

- NRC: 500KW _500 fire peaking in 105 sec. '-' 400 _ _-

- ANO: 1OOKW peaking -300 in 12 min (Scenario la) + cable 200 fires and high energy fault in A4l10 10 switchgear and cable fires (Scenario Ib) 0 250 500 - 750 1000 High energy arcing fault Time sec in the 4KV\ switchgearAN0-ULTenpScenario la Open Door

- NRC: Not analyzed . NRC-ULTerrp ANO: Limiting scenario, in terms UL Terrp Scenario>bOpenDoor.

of its consequence,.i.e., affected circuits and timing'  %

Neither analysis reaches 700 0 F I

I, Res-u Its: 34 Frequency of Fire Scenarios in Fire Zone,99M ANO SDP Analysis Results -

0 1 ~~~~~~~~~WF WIS lorRatio of HE Pnsb 2 ~~~~~~~~~~~~~~~~Generic (location igton area ratio Severity eetfra pat Psb IS <"Source Frequency weighting source (transient Factor severe e hie Results U) ~~~~~~~~~~~~~~~~weighting switchgear or fire brigade factor) factor) fires) fire wtch la Fire In the A4 swtchgear.

bNomnalvalue, I j0KWAr fire 11.50E-02 2.50E-01 5 88E-0 1.00E+00 1.20E-01 2.50E-01 1 O00 1.0OE.00 6.62E-05 lb High energy arcing fault in any of the A4 switchgear breaker cubicles

_______________ _ 1.1.50E42 2.50E4 1 588E-01 1.OOE+00 1.20E-01 7.50E-041 .000 1,E0 2 Fire in the B55 MCC. Nominai__

100KW fire. Fires in Inverler Y28 .

are bounded by this scenario.

11.50E-02? 2.50E-01 5 88E-02 1.OOE400 1.20E-01 10.ooE00 11.008+00 1.OOE800 2.65E405 3 Fire In the B56 MCC. Nominal

________ _0OKW fire 1.50E-02 2.50E-01 5.88E-02 1OOE+00 1.20E-01 1.OOE+00 . 1.00E+00 11.00rE400 _2.65E-05 4 Fire In the Y22 Inverter. Base .

case.rtOO KW fire. Fires in Y24 . .

and Y 25 are bounded by this scenario. 1.50E-02 2.50E-01 5 88E02 1OOE+00 1,20E.01 l.0Ed, 1.OOE*00 5.OOE-4

• 1.32E-05 5 Fire in the Load Center 86.

1WK om -n"Ln RR 1 50E-02 2.50E-01 _ 5 88E- 10 0 1E-01 1.00E00 1O.OEE0 2.00E41 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. 3 60E-02 2 OOE*00 11.80E-02 11.00E-01 1.00E+00 100+.00 5.00E-01 11.00E+00 6.i8E-05 6b Cable fire caused by welding and ..

cutting in areas of the room where I cable trays are exposed to a foor-based fire. Nominal Value of O1iKW. 1.30E-03 2OOE+00 2OOE-02 1 OOE-01 OOE800 1.00800 5.002-0 1.008+00 2.608-07 II

. I NRC SDP Analysis Results (May 15, 2003 Supplemental Letter Page 25)

? I Source . ' Frequency Electrical cabinets 2.3E-04

.1 Transformers .1.6E05 II Ventilation Subsystems. . 4.4E-06

(. . t --

(

Key Systems Affected in the Risk- I 39 Significance Determination (FireZone 99M) I

., I The following systems/trains are directly failed due to' fire induced power losses of A4 and B6 .

t . I

/*.

I

- One train and the swing pump of service water'

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

- The A4 associated diesel is no longer usa I ..

• - i

.i I pp- -

k I (

-I:

-. I-I II A.

A f J.

(. ,A ,

I.

II Human Error Probability I~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

. C Comparison 57

't

• NRC approach assumes zone wide damage at time zero .

f .

• NRC approach included loss of offsite power

-. . 1 U p U9

..I..

Operat& Actiton I

I

)

Establish EFW I

(A3 local start) I 'l Establish, EFW (Control. EFW,)

4.

I Establish Feed N.

& Bleed Establish 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 with no Service power Water" I

.! 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 iad international nuclear utilities, DOEfacilities and commercial/industrial facilities..

He is one of the principal investigators for Electric Power Res~earch Institute (EPRI)'fire risk analysis and fire protection projects. These projects included development of EPRI's Fire PRA Implementation'Guide and Fire-Induced Vulnerability Evaluation (FIVE) methodology and application of these technologies to US

• nuclear power plant support Over the past decade Mr. Najafi has been instrumental in 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;nore'engineering-based (as opposed to prescriptive-based) approach to fire protection. Several methods where also developed to demonstrate use of the technology, such'as "Methods for Evaluating Cable Wrap Fire Barrier Performance."

As part'of this .process of continuous enhancerment of technology, Mr. Najafi is currently the princpl technical anager of a oint roect between EPRI and USNRC office of Research for development of the next generaion of Fire Risk Analysis Ileiho shia G p e protection imdustry 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 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 managed Fire PSA projects at over eighteen (18) U.S. nuclear plants in response to NRC's Individual Plant Examination for External Events (IPEEE) as well as Dodewaard Plant in the Netherlands. The experience was part of the process to improve the Fie PSA data and methods developed 1y EPRI (with Mr. Najafi as the Project Manager).

Between 1988 and 1993, Mr. Najafi served as SAIC Project'Manager for GE's ABWR/SBKrR Level 1 PRA, Comanche Peak Level I/II PRA support, Project Engineer (Technical Project Manager) for the Turkey Point Nuclear Power Plant (PWR-YZ) Units 3 and 4 Level 2 PRA xith 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) A46, "Seismic Qualification of Equipment in Operating Plants," for the Omaha Public Power District staff.

During 1987-1988, he was the manager of a project to update the PRA for the Indian Point Unit 3 plant and perform a SAIC/Utility-conducted Level 1 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 Emnergency 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-I 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 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 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