ML17334A539

Summary of 970923 Meeting W/Util Re Analysis of Recirculation Sump Inventory.List of Meeting Attendees & Meeting Handouts Encl

ML17334A539
Person / Time
Site:	Cook
Issue date:	11/05/1997
From:	John Hickman NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9711200355
Download: ML17334A539 (59)

Text

November 5, 1997 LICENSEE:

American Electric Power FACILITY:

Donald C. Cook Nuclear Plant, Units 1 and 2

SUMMARY

OF SEPTEMBER 23, 1997, MEETING ON ANALYSISOF RECIRCULATION SUMP INVENTORY On September 23, 1997, NRC staff members met in Rockville, Maryland, with representatives of American Electric Power (AEP). The purpose of the meeting was for AEP to discuss their analysis of recirculation sump inventory to address Item 1 from the September 19, 1997, confirmatory action letter. A list of the meeting participants is included as Attachment 1, and a copy of the meeting handouts is provided as Attachment 2.

The licensee opened the meeting with a basic discussion of the lower containment design and the problem involving the pipe annulus area which can accumulate up to 333,000 gallons of spray flow before overflowing into the active sump.

The discussion then involved the calculations performed to credit the ice melt contribution to the active sump inventory.

Analysis codes were chosen to minimize ice melt for conservatism.

Scoping runs were performed to determine which break size was limiting. The benchmarking of the code was also discussed.

The meeting concluded with the licensee's intent to submit a technical specification amendment which would require that the minimum ice necessary for sump recirculation inventory be maintained in the ice bed.

Further discussion on the benchmarking of the code was also anticipated.

Original signed by:

John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316 Attachments:

1. List of Meeting Participants 2.

Meeting Handouts cc w/atts:

See next page DISTRIBUTION: See next page lllllllllllllllllfllllllllllllllll

'OCUMENT NAME: G:iDCCOOKttc0092397.MTS To receive a copy of this document, Indicate In tho box: "C" = Copy without enctosures "E" "-Copy with enctosures "N" ~ No copy OFFICE NAME

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November 5, 1997 LICENSEE:

American Electric Power FACILITY:

Donald C. Cook Nuclear Plant, Units 1 and 2

SUMMARY

OF SEPTEMBER 23, 1997, MEETING ON ANALYSISOF RECIRCULATION SUMP INVENTORY On September 23, 1997, NRC staff members met in Rockville, Maryland, with representatives of American Electric Power (AEP). The purpose of the meeting was for AEP to discuss their analysis of recirculation sump inventory to address Item 1 from the September 19, 1997, confirmatory action letter. A list of the meeting participants is included as Attachment 1, and a copy of the meeting handouts is provided as Attachment 2.

The licensee opened the meeting with a basic discussion of the lower containment design and the problem involving the pipe annulus area which can accumulate up to 333,000 gallons of spray flow before overflowing into the active sump.

The discussion then involved the calculations performed to credit the ice melt contribution to the active sump inventory.

Analysis codes were chosen to minimize ice melt for conservatism.

Scoping runs were performed to determine which break size was limiting. The benchmarking of the code was also discussed.

The meeting concluded with the licensee's intent to submit a technical speciTication amendment which would require that the minimum ice necessary for sump recirculation inventory be maintained in the ice bed.

Further discussion on the benchmarking of the code was also anticipated.

Original signed by:

John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316 Attachments:

1. List of Meeting Participants

2. Meeting Handouts cc w/atts:

See next page DISTRIBUTION: See next page DOCUMENT NAME: G:IIDCCOOKic0092397.MTS To receive a copy of this document, Indicate In the bord "C" a Copy without enciosures "E" ~ Copy with enciosures "N" ~ No copy OFFICE NAME DATE PD3-3:PM E

JHickman 11/~97 PD3-3:LA E

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001

- November 5,

1997 LICENSEE:

American Electric Power FACILITY:

Donald C. Cook Nuclear Plant, Units 1 and 2

SUMMARY

OF SEPTEMBER 23, 1997, MEETING ON ANALYSISOF RECIRCULATION SUMP INVENTORY On September 23, 1997, NRC staff members met in Rockville, Maryland, with representatives of American Electric Power (AEP). The purpose of the meeting was for AEP to discuss their analysis of recirculation sump inventory to address Item 1 from the September 19, 1997, confirmatory action letter. A list of the meeting participants is included as Attachment 1, and a copy of the meeting handouts is provided as Attachment 2.

The licensee opened the meeting with a basic discussion of the lower containment design and the problem involving the pipe annulus area which can accumulate up to 333,000 gallons of spray flow before overflowing into the active sump.

The discussion then involved the calculations performed to credit the ice melt contribution to the active sump inventory.

Analysis codes were chosen to minimize ice melt for conservatism.

Scoping runs were performed to determine which break size was limiting. The benchmarking of the code was also discussed.

The meeting concluded with the licensee's intent to submit a technical specification amendment which would require that the minimum ice necessary for sump recirculation inventory be maintained in the ice bed.

Further discussion on the benchmarking of the code was also anticipated.

John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of'Nuclear Reactor Regulation Docket Nos. 50-315 and 50-316 Attachments:

1. List of Meeting Participants 2.

Meeting Handouts cc w/atts: See next page

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Indiana Hichigan Power Company Donald C.

Cook Nuclear Plant Units 1 and 2

cc:

Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, IL 60532-4351 Attorney General Department of Attorney General 525 West Ottawa Street

Lansing, HI 48913 Township Supervisor Lake Township Hall P.O.

Box 818

Bridgman, HI 49106 Al Blind, Site Vice President Donald C.

Cook Nuclear Plant 1 Cook Place

Bridgman, HI 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office 7700 Red Arrow Highway Stevensville, HI 49127 Gerald Charnoff, Esquire

Shaw, Pittman, Potts and Trowbridge 2300 N Street, NW.

Washington, DC 20037 Hayor, City of Bridgman P.O.

Box 366

Bridgman, HI 49106 Special Assistant to the Governor Room 1 - State Capitol

Lansing, HI 48909 Drinking Water and Radiological Protection Division Hichigan Depai tment of

, Environmental guality 3423 N. Hartin Luther King Jr Blvd P.O.

Box 30630, CPH Hailroom

Lansing, HI 48909-8130 Steve J.

Brewer Indiana Hichigan Power Company Nuclear Generation Group 500 Circle Drive

Buchanan, HI 49107 E.E. Fitzpatrick, Vice President Indiana Hichigan Power Company Nuclear Generation Group 500 Circle Drive

Buchanan, HI 49107

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DISTRIBUTION FOR

SUMMARY

OF SEPTEMBER 23, 1997, MEETING WITH AMERICAN ELECTRIC POWER Hard Co w/atts:

Docket File PD III-3 Reading PUBLIC GGrant, Rill OGC ACRS David A. Lochbaum, UCS E-mail w/att 1:

SCollins (SJC1)

FMiraglia (FJM)

GTracy (GMT)

RZimmerman (RPZ)

EAdensam (EGA1)

GMarcus (GHM)

TMartin (SLM3)

Richard Lobel (RML)

MEETING ATTENDEES NRC AND AMERICAN ELECTRIC POWER DISCUSSION OF ANALYSISOF RECIRCULATION SUMP INVENTORY SEPTEMBER 23, 1997 NRC John Hickman Richard Lobel American Electric Power Steve Brewer Jeb Kingseed Karl Toth Brenda Kovarik Raymond Sartor Robert Henry (Fauske 8 Associates, Inc.)

Michael Young (Westinghouse)

Gallo and Ross Joseph Gallo Attachment 1

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POST LOCA SUMP-LEVEL ISSUES Introduction (AEP)

Background (AEP)

III.

Calculations/Modeling (Fauske)

IV.

Conclusions (AEP)

Lower ontainment im i ie c ematic (333,000 gallons)

Inactive Sump (Pipe Annulus) 6'1 2'-0" (337,000 gal excl. cavity)

(276,000 gallons)

Active Sump 602'-10" (110,000 gallons)

(126,000 gal Reactor Cavity 598'-9"

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ontainment pray E

im i ie ow c ematic Containment Spray (CTS)

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"-"-'-"-RCS"-""'"-"'-'CCS via via Fan stairwell Accumulator Rooms Inactive Sump 0O (0

via Ice I-Refueling Melt Canal and Drains Condensed Steam Active Sump Liquid Break Flow Reactor Cavity

SCOPING CALCULATIONSOF THE COOK NUCLEAR PLANT POST-SBLOCA SUMP FILLEVALUATIONS PERFORMED WITH MAAP4 R. E. Henry G. T. Elicson C. E. Henry C. Y. Paik Fauske 8c Associates, Inc.

Presented to The Nuclear Regulatory Commission September 23,- 1997 Rockville, Maryland

OUTLINE Introduction to the MAAP4 computer code Benchmarking calculations for this effort.

Comparison ofthe containment pressure and ice melt rate with LOTIC-3.

Comparison ofthe RCS break flowwith the NOTRUhP model.

Comparison with the Westinghouse ice condenser experiments.

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Important input parameters and sensitivity studies for the D.C. Cook nuclear plant evaluations.

BASIS FOR INVESTIGATINGA SPECTRUM OF LOCACONDITIONS A LO CA must be large enough for the containment sparys to be activated and needed over the long term.

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Ifthe LOCA is too large, the entire RCS willbe depressurized, LPI willbe initiated and the core will be cooled with cold water leaving the RCS break location. In this case, the containment sprays will only be required early in the accident.

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The sensitivity calculations show that the utilization ofcontainment sprays is the greatest for small LOCA conditions.

Hence, it is recommended that the NOTRUMP/LOTIC-3 calculations be utilized for a two-inch size LOCA. Moreover, the utilization of containment sprays is determined by the transient containment pressure including the sprays turning on ifthe pressure increases to 2.9 psig and turned offat 1.5 psig.

INTRODUCTION TO THE MAAP4 COMPUTER CODE

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MAAP4 is an integral plant representation and includes models for:

the reactor core response, the coolant system response, the containment response, the contributions of the emergency safeguard

features, and the response of adjacent plant building (auxiliary building, etc) where appropriate.

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As an integral system model, the focus of MAAP4 is on the total plant response to postulated accident conditions, with particular emphasis on accident management evaluations.

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As an integral system model, the MAAP4 focus is on the best-estimate evaluations for all phenomena evaluated.

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IV4MP4 Modular Accident Analysis Program

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MAAP4is a modular computer program written in fortran and is directed at evaluating the integral response ofthe RCS, containment and ESFs to a broad spectrum ofpossible accident conditions.

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The MPAP4 code is fast running (variable timestep) and has been developed and used for:

PWR NSSS (B&W,CE+ ~W designs 0

large dry containments, 0

subatmospheric containments, 0

ice condenser containments.

BWR NSSS (ABB + GE) designs 0

Mark I containments, 0

Mark IIcontainments, 0

Mark IIIcontainments.

CANDUNSSS designs 0

Ontario Hydro containment designs with the vacuum building, 0

AECL design with a separate containment for each reactor.-

VVERNSSS designs 0

reactor confinement including the bubble tower.

Fugen NSSS design 0

single containment with a suppression pool.

MAAP4 Modular Accident Analysis Program (Continued)

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MAAP4 modeling includes:

Response

to LOCA or inadequate cooling conditions.

Models for core degradation, core melt progression, debris quenching, etc.

necessary to evaluate severe accident conditions.

generalized containment model that promotes extensive containment nodalization ifdesired.

This generalized containment is used for all types of containment represented by the MAAP4 designs listed above.

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MAAP4 contains a dynamic benchmarking capability that enables the best-estimate models to be benchmarked with available experiments and experience.

These benchmarks can be easily repeated as the code evolves.

Steam Outfet (to turbine)

Steam Generator Steam Outfet (to turbine)

Feecfwator iniet (from condenaer)

Main Coolant Pump Feedwater tnlet (from cond.neer)

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Cold Leg Tubes Hot Leg Tubes 10 Hot 9 Leg Reactor Dome I

Hot Leg 3 Intermediate Leg Cold Leg 13 12 8 'Unbroken'oops Core Oowncomer Cold Leg 1 'Broken'oop

- (Nodailzatlon Same as Unbroken Loop)

Intermsdlate-Leg Figure 3-1 PWR primary system nodalization for Westinghouse 4-loop design.

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e oker Level Conhul KEY CCW - Component Cockng WWr CSS - Conhhment Spay Sveieca CST - Condenaeka 8 tank EACW - EeeenQ Aaw Wcder FW - Feedwaier LPI - LowPreeara nleeoton MSIV-Main Sleem leoiedon Vetvee POAV - Power OpeccWd ReM Vetvea PZR - I'reearher M-Aeeldwet Heat Removal f08 - Reaabr Pcoleoiion Syetem RPV - Reader Freed Veeeel RNST WhirrSerrepe Trrer Sl-SRV -

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RRQ40007.CDR M4044 D.C. Cook MAAP4 10 Node/18 Junction Containment Model

MAAP MAAP Phenomenological Routines Experimental Data and Conditions BENCH BNCHHL BNCHSW CREEP DECOMP BNCHGE FLOEXP OUTPUT BNCHSG HLNC HEATUP BNCHCS GENERALIZED CONTAINMENT Strategy for Incorporating Dynamic Benchmarks Into the hhMP4 Code

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MAJOR BENCHMARKS MAAP4 Oyster Creek loss-of-feedwater (BWR)

Crystal River loss-of-feedwater and stuck open PORV (PWR)

Peach Bottom turbine trip tests (BWR)

Tokai-2 turbine trip (BWR)

Davis-Besse loss-of-feedwater (PWR),

Brown's Ferry fire (BWR)

Current Dynamic Benchmarks TMI-2 (RCS)

TMI-2 containment PHEBUS LOFT-FP-2 v'0-5 hrs.)

v'0-5 hrs,)

CORA (BWR & PWR)

CSTF W ice condenser tests Ice condenser DB calculation W SG tests Material creep ABCOVE aerosol tests ORNL fission product release tests DEMONA aerosol tests LACE ACE experiments BETA experiments

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BENCHMARKCALCULATIONS FOR THE D.C. COOK SBLOCA SUMP FILLEVALUATIONS

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Containment pressure and ice melt comparison with LOTIC-3.

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The break flow rate spectrum used in the MAAP4 scoping calculations compared with the NOTRUMP model.

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Comparison of the MAAP4 ice condenser model with the %estinghouse experiments.

WHAT IS EXPECTED FROM THE BENCHMARKCALCULATIONS

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Assure consistency between "best estimate" and "design basis" analyses.

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Assure consistency of the MAAP4 ice condenser model with the experimental basis large LOCA, medium LOCA, small LOCA.

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COMPARISON WITH THE LOTIC-3 RESULTS

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This comparison is an evaluation of the respective containment models.

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The boundary conditions for both evaluations are the mass and energy releases from a six-inch cold leg LOCA as calculated by NOTRUMP.

These calculations are available for the first 1500 seconds.

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Given the NOTRUMP mass and energy releases and the specification of the Cook containment, the resulting containment response for the two models can be compared.

DCCOOK 6

INCH LOCA MAAP4 LOTIG-3 0

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200 400 600 800 1000 1200 1400 1600 1800 TIME SEC

COMlPAjRISONOF NOTRU1MP AI%)IKAAP4BREAKFI OW RATES AT t=0 SECS BREAK SIZE (IN)

MAAP4

.BREAKFLOW RATES LBM/SEC NOTRUMP 5500 5000 550 1.5 313 1.0 139 0.5 35

BENCHMAPZQNG WITH THE WESTINGHOUSE ICE CONDENSER EXPERIMENTS

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The Westinghouse ice condenser experiments have been run for a variety of break sizes.

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These experiments were performed for a full scale segment of the ice condenser with an ice basket height that is three-fourths of the plant.

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The principal information from the experiments is the depressurization of the simulated RCS, the pressure in the containment lower compartment, the pressure in the containment upper compartment, the temperatures of gases exiting the top of the ice condenser, the drain temperature of water leaving the ice condenser and the approximate ice melted.

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It is important that the integral system model be consistent with the experiments since the ice melt rate is a major contribution to the integral containment response and is also an important component of the water inventory in the circulation sump.

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MAAP dynamic benchmarks are being performed for three different break sizes investigated in these experiments which are generally representative of a large LOCA (Test A), a medium size LOCA (Test C) aud a small LOCA (Test F).

These beuchmarits are performed using the dynamic benchmarking capability in the MAAP4 code.

INTERHEOIATE DECK DOORS RECEIVER VESSEL LATTICE FRAMES I

I'CE BASKET BOILER VESSEL RUPTURE DISK ANO ORIFICE DIVIDER DECK DISCHARGE PIPE INLET DOORS TURNING VANES DISCHARGE NOZZLE Isometric View of Boiler and Receiver Vessels at the Waltz MillTest Facility

1400 1200

~ESTi IMGHOUSE ICE COINOENSEA EXPAIIIVIEINlTS DATA MAAP4 CO ce 1000 800 600 IK 400 T

ST F

200 0

200 400 6OO SOO 1000 TIME (SECONDS) 1200 1400 1600

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MAAP4 DATA

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CONCLUSIONS WITH RESPECT TO THE BENCHMABXINGACTIVITIES The comparison of the MAAP4 containment model and LOTIC-3 for the six-inch cold leg LOCA show good agreement between the ice melt rate and the transient containment pressure history with LOTIC-3 having a somewhat higher ice melt rate and higher containment pressure consistent with the design basis philosophy of the code.

~

The break flow rates considered by MWQ'4 are in good agreement with the flow rates from NOTRUMP.

Also, the spectrum of LOCAs considered in the MAAP4 analysis span those which are to be investigated by the DBA codes.

Hence, these provide the important insights related to where the DBA calculations should be performed.

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Comparisons of the MWQ'4 best-estimate model with the full scale experiments show a consistent response of the containment with the measured behavior.

This is true for both the containment pressure response and the ice melt conditions.

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The composite of these benchmarking activities shows a consistent representation of the containment response for the best-estimate scoping model (MWQ'4) and the design basis calculations (NOTRUMP and LOTIC-3).

Furthermore, the best-estimate model is consistent with the results of the large scale experiments used to characterize the response of an ice condenser containment to variety of LOCA conditions.

SENSITIVITYSTUDIES FOR THE COOK NUCLEARPLANT SUMP FILLEVALUATIONS

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The most important sensitivity calculation is to consider a variety ofbreak sizes. In this regard, the MAAP4sensitivity studies willinvestigate LOCAs from one-half inch to six inches in diameter.

This encompasses the entire small LOCA range.

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The particular issue of interest for this evaluation is the sump depth for the containment spray pumps.

Consequently, the duration ofthe containment sprays is an important variable in this evaluation.

Therefore, the variations in.plant parameters, within tech spec limits, willbe assessed to determine the influence that these could have on the use ofcontainment sprays.

Typically, the containment sprays are automatically initiated at 2.9 psig and are to be turned offby the operator when the pressure decreases to 1.5 psig. The influence of conditions whereby the sprays would be turned on at pressures less than 2.9 psig willbe evaluated through these sensitivity evaluations.

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Other plant parameters influence the mass ofair in containment, the condensing capability ofthe sprays, etc..

These willalso be investigated in these sensitivity calculations.

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DCCOOK 10 NODE 2'OLD LEG LOCA, COOLDOWN AT 30 F/HR SPRAYS OR AT I.S

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4-2 Table 4-1 Sensitivity Run No.

Sl S2 S3 S4 S5 S6 S7 Parameter Chan ed Core power = 3315 MWt.

Core power = 3250 MWt.

Core power = 3588 MWt.

Core power = 3425 MWt.

RWST temperature = 105 F.

RWST temperature = 70'F.

Containment gas temperature = 60'F.

Comments Licensing value for Unit 1.

Nominal reactor condition.

Licensing value for Unit 2.

'ominal operating power for Unit 2.

Nominal value.

Minimum tech spec value.

Lowest value to maximize the mass of air in containment.

S8 S9 S10 Sll Containment gas temperatures

- UC = 100'F,

- LC = 120'F,

-. DEC = 120'F.

Thermal conductivity of containment structural heat sinks decreased by a value of 1.4.

Thermal conductivity of containment structural heat sinks increased by a factor of 1.4.

Heat exchanger cooling rates set at minimum lake water temperature =

'5'F.

Minimize the influence of containment structural heat sinks.

Maximizes the influence of.

containment structural heat sinks.

Minimizes ice melt.

FAD,97-104; Rev. 0

0 CONCLUSIONS FROM THE PRELIMINARY SENSITIVnVSTUMKS

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Allofthe calculations considered for an ice mass of 2.37x10 lb show water levels in the lower compartment are sufficient to support the necessary containment spray pump operation.

~

The most limitingLOCAs appear to be the two-inch diameter range since these minimize the amount of injection to the RCS, and therefore the RCS cooldown, while still requiring the containment sprays to be activated.

However, even with these two-inch LOCAs there is no challenge to the containment spray pump operation for an ice mass of 2@3 7x 1 0 Ibme

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