ML16148A416

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Summary of 801216 Meeting W/B&W Owners Group Re Abnormal Transient Operating Guidelines (Atog) Program & TMI Action Item II.K.3.30,small Break LOCA Models.List of Attendees & Viewgraphs Encl
ML16148A416
Person / Time
Site: Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  Duke Energy icon.png
Issue date: 01/30/1981
From: Throm E
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 8102270389
Download: ML16148A416 (30)
v • d • e


Text

Distribution Licensees:

Mr. W. Parker Mr. R. C. Arnold Mr. J. J. Mattimoe Mr. W. Cavanaugh us Mr. R. P. Crouse Mr. J. A. Hanc/

Docket Files -

50-269, 270, 289, 302, 312, 313, and 346 NRC PDR LPDR TERA - 8 NSIC ORB Reading HDenton ECase DEisenhut RPurple RTedesco TNovak GLainas RReid TIppolito SVarga DCrutchfield RAClark ORB Project Manager Licensing Assistant OELD AEOD - J. Heltemes IE -

3 SShowe (PWR)

Meeting Summary File -

ORB RFraley, ACRS (16)

Program Support Branch GZech JRoe NRC Participants DFRoss PSCheck TPSpeis RSB Reading SMacKay

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 JAN 30 1

Docket Nos.

50-269, 270, 289, 302, 312, 313 and 346 FACILITIES:

OCONEE 1, 2, and 3; THREE MILE ISLAND 1; RANCHO SECO 1; ARKANSAS NUCLEAR ONE, UNIT 1; DAVIS-BESSE 1; and CRYSTAL RIVER 3 LICENSEES:

BABCOCK AND WILCOX REACTOR OWNER GROUP - DUKE POWER COMPANY METROPOLITAN EDISON CO., SACRAMENTO MUNICIPAL UTILITY DISTRICT, ARKANSAS POWER AND LIGHT CO., TOLEDO EDISON CO.,

and FLORIDA POWER CORPORATION

SUBJECT:

SUMMARY

OF MEETING WITH THE B&W OWNERS GROUP CONCERNING THE ABNORMAL TRANSIENT OPERATING GUIDELINES (ATOG) PROGRAM AND TMI ACTION ITEM II.K.3.30 SMALL BREAK LOSS-OF-COOLANT ACCIDENT MODELS (DECEMBER 16, 1980)

The purpose of the meeting was to continue discussions on the ATOG program for B&W reactors and to provide the Owners Group with a list of.concerns in responding to the TMI Action Item II.K.3.30. A list of attendees is provided in Enclosure 1.

The meeting agenda is provided in Enclosure 2.

Summary on ATOG Program The staff stressed the point that the ANO-1 submittal needs to be suitable for a generic review of the ATOG program. In addition the TMI Action Plan scheduled is not changeable and the January 1, 1981 date for the formal submittal is to be met.

The PTRB is currently drafting a set of guidelines to address the procedures and methodology for the vendor programs in response to TMI Action Plan Item I.C.l and would like to incorporate the Owners Group ATOG program in the draft. A March 1 date is tentatively planned for the draft release.

The Owners Group has indicated that they would use 10CFR 50.59 to implement the ATOG program and are not particularly concerned as to what we actually review.

Questions generated on the ANO-1 submittal should be addressed to ANO-1 or to all the licensees.

The ANO-1 draft ATOG program will be formally submitted, with a schedule for completion of the final report, on January 1, 1981 as required.* It will be noted that the draft is very much preliminary and that there is no guarantee the sub stantial changes will not be made in the final report. In addition it will be noted that Inadequate Core Cooling will be added to the ATOG program.

A formal presentation on the ATOG program was made by J. J. Kelly of B&W. The slides are provided in Enclosure 3.

  • Each licensee will reference the ANO-1 submittal and supply a schedule.

JAN 3 0 1981 Summary for II.K.3.30 The staff identified the following nine areas of concern which should be addressed by the Owners Group in future submittals:

1. Need to verify the current non-condensible model and the conservatism of the condensation heat transfer rate in the steam generator.
2. Need to verify the non-equilibrium model and to justify that the amount of ECCS water injected i~s conservative.
3. Need to.discuss the pressurizer model and the effects of a non-equilibrium model.
4. Need to address the formation of a, steam bubble in thq hot leg "candy cane".

(Is it a real or calculated phenomenon?) Experimental verification believed necessary.

5. The staff indicated that a mechanistic model of the steam generator heat transfer should be developed. A best estimate or verified conservative model would be acceptable.
6. As part of the additional.systems verification needed, the following Semiscale and LOFT tests should be considered:

Semiscale S-07-10D, LOFT L3-1, L3-5 and L3-6.

7. The overall thermal-hydraulic behavior of the core during uncovery should be verified against applicable experimental data, particularly the recent ORNL data.
8. The influence of metal heat on the system pressure response, particularly on the time of ECCS injection was identified as an area of concern and should be shown to be properly considered in the analysis models.
9. The break flow model needs to be confirmed. The use of combined models with various discharge coefficients applied to them needs to be compared to.

a best estimate model to demonstrate conservatisms.

A discussion of the 'impact of reactor coolant pump pperations was held. The staff identified their present thinking regarding acceptance criteria for allowing manual pump trip.

1. If the operator is allowed to trip the pumps manually, the criteria of 10CFR 50.46 must be met using an acceptable model (Appendix K model with a best estimate verification analysis of LOFT L3-6) assuming RCP trip at 10 minutes or at the specified criteria, whichever is longer.
2. If pump trip is required in less than 10 minutes, using the Appendix K analysis model, then a best estimate analysis with pump trip assumed at worst time into the event must result' in acceptable consequences.

A minimum acceptable time for operator action will have to be identified and justified.

3 JAN 3 0 1981 A schedule and work scope for the II.K.3.30 item is to be submitted by January 30, 1981.

Edward D. Throm Reactor Systems Branch Division of Systems Integration

Enclosures:

1. Attendance List
2. Meeting Agenda
3. ATOG Slides

ENCLOSURE 1 ATTENDEES:

NRC B & W TOLEDO EDISON E. D. THROM J. J. KELLY T. MYERS J. GUTTMANN R.C. JONES S. JAIN F. ODAR, RES R. A. TURNER D. BECKMAN J. J. CUDLIN B.W. SHERON H. BAILEY G. S. VISSING N. K. SAVANI G. MEZETIS R. PITMAN J. CLIFFORD GPU R. URBAN T.C. BRUUGHTON M. GOODMAN C.W. SMYTH AP & L DUKE POWER D. WILLIAMS R. L. GILL M. SMITH SMUD TVA D. WHITNEY F. A. KO0NTZ, JR R. A. DIETERICH CPC0 FLORIDA POWER CO.

W. J..HALL H. M. PERRY L.S. GIBSON A. F. FEGENDRE, JR.

WPPSS A. HOSLER

B&W OWNERS GROUP MEETING WITH NRC TMI-2 & ATOG SUBCOMMITTEES December 16, 1980 Tuesday, December 16, 1980 (P 110 -

8:30)

TMI Action Plant Item I.C.1 (Short Term Accident & Procedures Review)

1. Inadequate Core Cooling Guidelines
2. Transients and Accidents TMI Action Plan Item II.K.3.30 (SB LOCA Methods)
1. Introduction
2. Discussion of work to address NRC expectations in the following areas:

Areas of the model which will be upgraded Areas of the model which will.be further justified Test data which will be used

3. Owners Caucus
4. Scope and schedule for completion of the above.

ATOG Objective Simplify operator problem of identifying and treating abnormal transients t0

Closing the loop Designers Operators tna s

Trainers uideline procedure writers

Methodology Event trees Analysis Design basis/expected plant response Simulation Operator feedback

Event Trees Purpose Systematically determine various plant conditions which can evolve following a postulated initiating event Assumptions Initial conditions Equipment failures Operator action

TranJsients Selected for Guideline Preparation Increase in heat removal by secondary system

  • Small steam leaks
  • Excessive feedwater flow Decrease in heat removal by secondary system
  • Inadequate core cooling
  • Loss of coolant

Event tree bounding assumptions Initial conditions

-Power range.

-No equipment tagged out.

-Equilibrium core.

Equipment failures

-Consequential failures.

-Non-safety system components only.

-Active failures only.

Operator action

-Will be given the opportunity to act when required by procedure.

-His actions can be correct, incorrect or failure to act.

-Errors will not be random.

-Actions, including mistakes, will be complete within a system.

-No artificial time constraints.

Prior plant transient experience

Analysis

Purpose:

Realistically portray expected plant response Analyze: Design success path All single failure paths Discuss subsequent failures:

Verify LOCA paths covered in small break guidelines

Design Basis/Expected.

Plant Response

  • Communication between designer and operator
  • Supports operator action portion of guidelines
  • Written for operator understanding j

Training Simulator

  • Test various methods of approach to guidelines 8 Verify final product 0 Train operator

Operator. Feedback o Detailed review of event trees o Input to guideline format O Plant walk through a Training

2600 2400 POST TRIP SWINDOW 2200 1.1TI SU1BCOOLED 2000 REGION SUPERHEAT REGION 1800 1600 C

1400 c

1200 STEAM PRESSURE END POINT-POST TRIP WITH FORCED 1300 toCIRCULATION (THOT COLO) AND FOR Li.NATURAL CIRCULATION (TCOLO) 2 800 NORMAL OPERATING POINT-POWER 600 OPERATION (THOT)

END POINT-POST TRIP WITH NATURAL SUSCOOLEO MARGIN 400 CIRCULATION (THOT)

LINE 400 450 500 550 OO 650 7

Reactor Coolant And Steam Outlet Temperature-F

2600 POST TRIP 2400 WINDOW 2200

[IZ]

2000 I

SUBC00LEO

$REGIEON SUPERHEAT

~1800 -REGION 41~

16000 1400 END POINT-POST TRIP WITH FORCED CIRCULATION (THOT U 1200 STA P TCOtO) AND FOR NATURAL STEAM PRESSURE ct CIRCULATION (TCOLO) 1000 NORMAL OPERATING POINT-POWER OPERATION (THOT.

SATURATION 1 END POINT-.POST TRIP WITH 00-I NATURAL CIRCULATION (THOT)

SUBCOOLED MARGIN LINE S450 500 550 600 650 700 Reactor Coolant and Steam Outlet Temperature. F

2600 2400 POST TRIP WI NOOW 2200 2000 SUBCOOLED RECICN SUPERHEAT 1800 -

REGION 1600 7-3 1400 m

END POINT-POST TRIP WITH FORCED CIRCULATION (THOT 1200 -

STEAM PRESSURE A TcoLo) AND FOR NATURAL LIMIT CIRCULATION (Tcoto) 1000 COLD HNORMAL OPERATING POINT-POWER OPERATION (THOT 800 00 SATURATION END POINT-POST TRIP WITH oo NATURAL CIRCULAIsoN (TO)

SUBCOOLED 400 MARGIN LINE no 450 500 550 600 650 100 Reactor Coolant adl Steam Outlet Temperature, F INADEQUATE SUBCOOLING MARGIN

2600 POST TRIP 2400 W INDOW 2200 IA cl.2000 MA 0 20110SUBOOLED.

REGIOOD SUPERHEAT.

REGION-REGION 1800 1600 40 -END POINT-POST TRIP WITH FORCED CIRCULATION (Tit0T COLD AND FOR ATURAL

% 1200 STEAM PRESSURE AND (oTU LIMIT

~1000 p__

__WE 00 NORMAL OPERATING POINT-POWER OPERATION (Tit0T) t 800 SATURATION 1END POthT-POST TRIP WITH F)

NATURAL CIRCULATION (TtOT) 600V1N SUBCOOLED 400 MARGIN LINE I460005 400450 500 55060 neactor Coolant and Steam Outlet Temperature. F LOSS. OF PRIMARY TO SECONDARY HEAT TRANSFER

2600 POST TRIP 2400 Wi NDOW 2200

[

UUL ca 2000 1

SUBCOOLED SUPERHEAT REG ION REGION S 1800 1600 j 1400 -END POINT-POST TRIP WITH FORCED CIRCULATION (THOT n1200 S1 P

) AND FOR NATURAL c

noCCOLD STAM PRESSURE LIMIT

-OR14AL OPERATING POINT-POWER OPER.ATION (TilOT)

SATURATION

-i END POINT-POST TRIP WITH NATURAL C-IRCULATION (THjOT) 100 SUBCOOLED 400 MARGIN LINE 4 b 00 0 0 550 600 650 10 Rt.ctot Coulant afnt Steitm Outlet Tenueralire. F EXCESSIVE PRIMARY TO SECONDARY HEAT TRANSFER

0 TO 5 MINUTES 2600 POST TRIP 2400 ~WINDOW O

2200

. 2000 SUBCCOLED I /2 REGION SUPERHEAT 3

18000 REGION 2

1600 3

1400 (01 5

END POINT-POST TRIP WITH FORCED CIRCULATION (T cu1200 o

STEAM PRESSURE A NCO D)

No FOR NATURAL LIMIT CIRCULATION (ToLD) o 1000

-3 NORMAL OPERATING POINT-POWER OPERATION (THOT S800 SATURATION END POINT-POST TRiP WITH 600 1.1 NATURAL CIRCULATION (TIIOT 400 SUDCOOLED MARGIN LINE 0

I I

I I

I 400 450 500 550 600 650 100 Reactor Coolant and Steam Outlet Temperature. F 7T -2 Acc 1DEv7

5 TO 8 MINUTES 2600 POST TRIP 2400 WINDOW 2200 2000 SUBCOOLED 13 REGION SUPERHEAT 160 1800

-REGION 1600 I.-

5 1400 6

END POINT-POST TRIP WITH S1SA SFORCED CIRCULATION (T N1200 NOT STEAM PRESSURE T

TcoLe)

AND FOR NATURAL 000LIMIT CIRCULATION (TCOLD) 1000 NORMAL OPERATING POINT-POWER OPERATION (TilOT 000 SATURATION END POINT-POST TRIP WITl 600 I

NATUtAL CIRCULATION (TT)

SU BCOL0 E

400 MARGIN LINE U E I

I I

I 400 450 500 550 600 650 700 Reactor Coulant and Steam Outlet Temperature, F

/

1 2

A ccE,-

8 TO 15 MINUTES 2600 POST TRIP 2400 WINDOW 2200 01 2000 SUBCOOLf0 V)

REGION SUPERHEAT S1800 REGIONEGION ck 1600 10 1400 END POINT-POST TRIP WITH FORCED CIRCULATION (THOT 120015A AND FOR NATURAL STEAM PRESSURE 1

TCOLD n

onM~n L

CIRCULATION (TCOLD) 1000 NORMAL OPERATING POINT-POWER OPERATION (THOT)

S800 SATURATION I kise POINT-POST TRIP WITH 600 NATURAL CIRCULATION (THOT)

SUBCOOLED 400 MARGIN LINE 0:

_1 IW

~

4G0 450 500 550 600 650 100

-Reucter Coiolint ana Stean Outlet Temperture. F cr -a2 A d l 15 TO 20 MINUTES 2600.

POST TRIP 2400 WINDOW 2200 2000 SUOCOOLED REGION SUPERHEAT

~1800 REGION 1600 U1400 -cr3 END POINT-POST TRIP WITN FORCED CIRCULATION (THOT STEAM PRESSURE 15 C

TCOLD) AND FOR NATURAL LIMIT CIRCULATION (TCOLD) 0 CA 1000 0--

-3

  • NORMAL OPERATING POINT-POWER OPERATION (THO SATURATION END POINT-POST TRIP WITH 600 r

NATURAL CIRCULATION (THOT)

SUBCOOLED MARGIN LINE Ul I

II 458 500 550 600 650 100 Reactoi Coolant ina Steam Outlet Temperiture. F 71-2 4CCIlb vN-

Part I. Organization SECTION J. Immediate actions SECTION II. Vital system status verification SECTION Ill.

A. Treatment of lack of adequate subcooling margin B. Treatment of lack of primary to secondary heat transfer C. Treatment of too much primary to secondary heat transfer D. Follow up actions for OTSG Tube rupture COOLDOWN PROCEDURES

  • Normal
  • Solid water cooldown

Part II. Organization VOLUME 1: Fundamentals of reactor control for abnormal transients A. Heat transfer a

B. Use of P-T diagram C. Abnormal transient diagnosis and' mitigation D. Backup cooling methods E. Best methods of equipment operation F. Stability determination VOLUME 2: Appendices - Selected transients A. Excessive feedwater B. Loss of feedwater C. Steam generator tube rupture D. Loss of off-site power E. Small ste-am line break F. LOCA

2600 2400

-POST TRIP WINDOW

-~

2200 r--

cu 2000 SUBCOOLED 1800 REGION 1300) 5 42 10000 3

1400 1200 STEAM PRESSURE 1000C LIIT 00 4

000 SUBCOOLED 400 MARGIN LINE 406 450 500 550 600 650 700 Reactor COGlant and Steam Outlet Temperature, F 500 700 900 1100

RPS P-T Trip Envelope 2300 IORMAL THOT"COLD POWER 20PERATION WINDOW 2100 1900 PSIG 1700 I

I I

I I

520 540 560 580 600 620 RPS Flux Offset Trip Envelope 125 NORMALIZED RC FLOW-4 PUMP OPERATION 00 4 PUMP LIMITS 75 3 PUMP LIMITS 50 25

-40

--20 0

+20

+40

+60

ATOG Full Scale Display 2500 2400 TUBE /SHELL 000 POST-TRIP WINDOW 2000 1600 PSIG RB RB PRESS TEMP 1200 C

+15 300

+10 250 800

+5 200

-0 150 400 0

5 _J.

100 Rp "PS" Q~PSIG o F 0

200 300 400 500 600 OF

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