ML20035F288

Summary of 930401 Meeting w/ABB-C-E in Rockville,Md to Discuss Analyses of Accidents Presuming common-mode Failure of RPS & ESFAS

ML20035F288
Person / Time
Site:	05200002
Issue date:	04/15/1993
From:	Wambuch T Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9304210114
Download: ML20035F288 (38)

Text

-

Docket No.52-002 April 15, 1993 APPLICANT: ABB-Combustion Engineering, Inc. (ABB-CE)

PROJECT:

CE System 80+

1

SUBJECT:

MEETING

SUMMARY

FOR MEETING ON INSTRUMENTATION & CONTROL (I&C)

DIVERSITY AND COMMON-MODE FAILURE On April 1,1993, a meeting was held at the Nuclear Regulatory Commission (NRC) offices in Rockville, Maryland, between ABB-CE and the NRC staff to discuss the analyses of accidents, presuming a common-mode failure of the reactor protection system and the engineered safeguards features actuation system. The meeting attendees are listed in Enclosure 1.

The material presented by ABB-CE is provided in Enclosures 2 and 3.

i ABB-CE took credit for leak detection capability inside containment to exclude a main steamline break (MSLB) inside containment and break sizes in the i

primary coolant system (loss-of-coolant accident (LOCA)) greater than a double-ended break of a 12 inch line. The results of the eight accident analyses were presented.

The NRC staff commented that the core design used for these analyses must be identified as requiring staff approval if any changes are made. The staff stated that there must be a combined license action item (COLA) requiring the operator training program to include this common-mode failure event coincident t

with accidents. The staff asked whether ABB-CE was committing to 100-percent testing of the loop controller inputs. ABB-CE responded that they would consider this and inform the staff at a later date. The staff stated that the inspections, tests, analyses, and acceptance criteria for the reactor i

protection system, the engineered safety features actuation system, and the main control room must include the design features provided for this event.

The staff concluded by stating that it was considering the necessity for analyses of the large-break LOCA and the MSLB inside containment, and would inform ABB-CE later.

I (Original signed by)

Thomas V. Wambach, Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors 9304210114 930415 PDR ADOCK 05200002 and License Renewal A

PDR Office of Nuclear Reactor Regulation

Enclosures:

DISTRIBUTION w/o enclosures:

l As stated ACRS (11)

GGrant, 17G21 TMurley/FMiraglia l

JMoore, OGC CGoodman, 10D24 JBongarra, 10D24 cc w/ enclosures:

SBSun, BE23 MRubin, 8E23 RJones, 8E23 See next page GWest, 10D24 GThomas, 8E23 MWaterman, 8H3 1900M EKendrick, 8E23 AEl-Bassioni,10E4 EJordan, MNBB 3701

. DISTRIBUTION w/ enclosures:

Docket File PDST R/F DCrutchfield RBorchardt SMagruder PDR TEssig TWambach MFranovich PShea A

/

0FC:

LA:PDST:ADAR PM:

S A

SC:PDST:

AR TEssig/g-[v NAME.: PShea/$6 TWa ach:tz 04/g/93 04/$/93 04/p/93A Fo3 DATE:

t m n ey - er OFFICIAL RECORD COPY: MSM401.TVW bM iL v:fd,,a b,3i

$\\

ABB-Combustion Engineering, Inc.

Docket No.52-002 i

cc:

Mr. Regis A. Matzie, Vice President

)

Nuclear Systems Develupment I

Combustion Engineering, Inc.

I 1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 Mr. C. B. Brinkman, Acting Director Nuclear Systems Licensing Combustion Engineering, Inc.

1000 Prospect Hill Road Windsor, Connecticut 06095-0500 Mr. C. B. Brinkman, Manager Washington Nuclear Operations Combustion Engineering, Inc.

i 12300 Twinbrook Parkway, Suite 330 i

Rockville, Maryland 20852 Mr. Stan Ritterbusch Nuclear Systems Licensing Combustion Engineering, Inc.

1000 Prospect Hill Road Post Office Box 500 Windsor, Connecticut 06095-0500 Mr. Sterling Franks U. S. Department of Energy NE-42 Washington, D.C.

20585 t

Mr. Steve Goldberg Budget Examiner 725 17th Street, N.W.

Washington, D.C.

20503 Mr. Raymond Ng 1776 Eye Street, N.W.

Suite 300 Washington, D.C.

20006 Joseph R. Egan, Esquire Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W.

Washington, D.C.

20037-1128 w-t-- - - - '

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MEETING ATTENDEES April 1, 1993 ORGANIZATION AFFILIATION Tom Wambach NRC/NRR/ADAR C1 aire Goodman NRC/NRR/DRCH (HF)

James Bongarra NRC/NRR/DRCH (HF)

S. B. Sun NRC/NRR/SRXB l

Mark Rubin NRC/NRR/SRXB Robert Jones NRC/NRR/SRXB George Thomas NRC/NRR/SRXB Mike Waterman NRC/NRR/DRCH/HICB Edward Kendrick NRC/NRR/SRXB Adel El-Bassioni NRC/NRR/SPSB i

l Garmon West, Jr.

NRC/NRR/DRCH/HHFB Joe Rezemides ABB-CE l

Stan Ritterbusch ABB-CE l

Alfred Hyde ABB-CE I&C I

Fred Carpentino ABB-CE R. B. Fuld ABB-CE l

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4/01/93 9 EVENTS & ASSOCIATED RESULTS EVENT RESULTS LOSS OF FLOW RATE DNBR 2 1.24 SEIZED / SHEARED SHAFT DNBR 2 1.24 SGTR DNBR 2 1.24 LETDOWN LINE BREAK DNBR 2 1.24 CEA EJECTION DhBR 2 1.24 0

CLAD TEMP s 2200 F FUEL ENTHALPY s 280 CAL /GM 0FFSITE DOSES s 10CFR100 MSLB CLAD TEMP 5 2200 F 0

0FFSITE DOSES s 10CFR100 RCS PRESSURE s 3200 PSIA FWLB CONTAINMENT PRESSURE s-145 PSIA LOCA CLAD TEMP s 2200 F 0

0FFSITE DOSES 5 10CFR100

4/01/93 BEST ESTIMATE EVALUATION ASSUMPTIONS 1.

EVENT & CHF (N0 ADDED SINGLE FAILURES).

2.

NOMINAL OPERATING PARAMETERS.

i 3.

CONTROL SYSTEMS PROVIDE AUTO AND MANUAL MITIGATION / REC 0VERY ACTIONS.

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4.

ALTERNATE PROTECTION SYSTEM PROVIDES HPP Rx f

TRIP AND EFAS.

5.

LEAK DETECTION CAPABILITY LIMITS LOCA BREAK SIZE AND MSLB BREAK LOCATION.

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t 4/01/93 LOSS OF RCS FLOWRATE BEST-ESTIMATE DNBR ASSUMPTIONS:

PRESSURIZER PRESSURE 2250 PSIA 0

COLD LEG TEMP 556 F VESSEL FLOW RATE 461,200 GPM CORE POWER 3914 MWT ASI

- 0.07 FR 1.50 l

RESULTS:

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MINIMUM DNBR = 1.74 o

OPERATOR ACTION REQUIRED > 30 MINUTES l

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4/01/93 SINGLE RCP SHAFT BREAK / SEIZURE BEST-ESTIMATE DNBR ASSUMPTIONS:

PRESSURIZER PRESSURE 2250 PSIA 0

COLD LEG TEMP 556 F VESSEL FLOW RATE 461,200 GPM CORE POWER 3914 MWT ASI

- 0.07 FR 1.50 RESULTS:

0 NO Rx TRIP o

RCS FLOW RATE DROPS FROM 100% TO 74.8%

o MINIMUM DNBR = 1.41 i

o OPERATOR ACTION REQUIRED > 30 MINUTES I

l

4/01/93 SYST 800 SEtZED ROTOR SYST 80o SE! ZED ROTOR POWER SCS TEMPEMTUES i

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PRESSURIZER PRESSURE 2250 PSIA COLD LEG TEMPERATURE 556 F 0

VESSEL FLOW RATE 461,200 GPM CORE POWER 3914 MWr ASI

- 0.07 FR 1.5 RESULTS:

o MANUAL Rx TRIP AT 30 MINUTES o

MINIMUM DNBR = 1.35 l

r M

STEAts C,04ERATOR TUBE RUPTURE WWTH NO RPS ACTUATK)N

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4/01/93 LETDOWN LINE BREAK RESULTS ASSUMPTIONS:

PRESSURIZER PRESSURE 2250 PSIA 0

COLD LEG TEMPERATURE 556 F i

VESSEL FLOW RATE 461,200 GPM CORE POWER 3914 MWr j

ASI

- 0.07 FR 1.5 1

i RESULTS:

o THE CHAPTER 15 ANALYSIS ASSUMES THERE ARE NO PROTECTIVE ACTIONS FOR 30 MINUTES WITH MINIMUM DNBR RESULTS B0UNDED BY THE SGTR RESULTS HEREIN.

o A DIVERSE ' HARDWIRED' SWITCH TO ACCOMPLISH LETDOWN LINE ISOLATION IS l

BEING ADDED.

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4/01/93 CEA EJECTION BEST-ESTIMATE DNBR.

ASSUMPTIONS:

RCS PRESSURE 2250 PSIA COLD LEG TEMP 5560F VESSEL FLOW RATE 461,200 GPM CORE POWER 3914 IWr ASI

- 0.07 FR PRE-EJECTED

= 1'. 0 POST-EJECTED = 1.75 RESULTS:

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0 Rx TRIP AT 3.29 SECONDS (APS) o NO FUEL CENTER LINE MELTING (FIGURE ATTACHED) 0 o

CLADDING TEMP BELOW 660 F (FIGURE ATTACHED) o N0 FUEL FAILURE:

MINIMUM DNBR = 1.40 (FIGURE ATTACHED) o RADIALLY AVERAGED FUEL ENTHALPY < 100 CAL /GM o

RADIOLOGICAL-CONSEQUENCES B0UNDED BY LOCA o

OPERATOR ACTION REQUIRED > 30 MINUTES

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4/01/93 STEAM LINE BREAK BEST-ESTIMATE DilBR ASSUMPTIONS:

PRESSURIZER PRESSURE 2250 PSIA 0

COLD LEG TEMP 556 F VESSEL FLOW RATE 461,200 GPM CORE POWER 3914 MWr ASI

- 0.07 FR 1.50 RESULTS; o

Rx TRIP AT 17.3 MINUTES (APS) o PEAK RCS PRESSURE LESS THAN 2950 PSIA o

NO FUEL CENTER LINE MELTING (LESS THAN 45000 )

F o

CLAD TEMPERATURES LESS THAN 1150 F 0

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100 k

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400 800 1200 1600 SECONDS

4/01/93 i

MAIN STEAM LINE BREAK 0FFSITE RADIOLOGICAL DOSES ASSUMPTIONS:

o MAIN STEAM LINE BREAK -

OUTSIDE CONTAIIMENT o

MAIN STEAM LINE ISOLATION -

MANUALLY AT 30 MINUTES o

100% GAP FISSION PRODUCTS RELEASED o

NO SG SECONDARY SIDE DECONTAMINATION (BROKEN GENERATOR - INTACT GENERATOR:

30 MINUTES) o.

PRIMARY-TO-SECONDARY LEAKAGE 3

PER NUREG-0017 (REV 01)

.00625 GPM o

.EPRI URD CHI /O o

NUREG - 1465 SOURCE TERM l

RESULTS:

TWO HOURS 0 EAB 92.0. REM THYROID 3.0 REM WHOLE BODY EIGHT HOURS 0 LPZ 41.0 REM THYROID 0.3 REM WHOLE BODY

4/01/93 l'

l FEEDWATER LINE BREAK CONTAINMENT PRESSURE ASSUMPTIONS:

o FULL POWER INITIAL CONDITIONS o

MANUAL MAIN STEAM /FEEDWATER ISOLATION 0 30 MINUTES o

CREDIT FOR 2 CONTAINMENT SPRAY TRAINS AT 17 MINUTES o

.CONTINU0US ADDITION OF EMERGENCY J

FEEDWATER RESULTS:

o Rx TRIP ON APS HIGH PRESSURIZER PRESSURE o

PEAK CONTAINMENT PRESSURE = 74 PSIG (PCP = 94 PSIG IF SPRAYS AT 30 MINUTES) y w

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FEDMATER LINE BREAK AT FULL POWER 100 CONTAINMENT DESIGN PRESSURE = f(PSIC CONTAINNENT LEVEL C LIMIT = 130 PSIC l

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0 400 800 1200 1600 2000 TIME. SECONDS M a vi u ( { cr i'og 3OM M

c a y s f.

y i

4/01/93 LOCA BEST-ESTIMATE CORE INVENTORY ASSUMPTIONS:

BREAK SIZE BRANCH LINES i

BELOW 12" DIA.

RCS PRESSURE 2250 PSIA 0

COLD LEG TEMP 556 F CORE POWER 3914 MWT MODERATOR REACTIVITY BEST ESTIMATE HPSIs NONE SITS 4 TANKS 0 600 PSIA CHARGING FLOWRATE 150 GPM 7

RESULTS:

j o

NO Rx TRIP l-o 6" PSV LINE O TOP 0F PRESSURIZER -

4 MINIMAL CORE UNC0VERY 0 23 MINUTES

)

(FIGURES ATTACHED) o 3" SPRAY LINE O TOP 0F COLD LEG -

NO CORE UNC0VERY FOR 30 MINUTES i

f

04/01/93 l

1 l

l l

l 0.15 FT2 BREAK IN TOP OF PRESSURIZER

~

)

l NO RCP TRIP i

l I

i l

1 l

I i

l e

l-4/1/93 m.m.

• M ESUM

• DOC 4 e CORE CORE Dese tA 2trao L

5 i

\\

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-so, E

Y

~

l

\\

l M

$2 i

4:33 l

Ch.

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4tD aEn tam 3 geIB 223 e

e eco eco -

5200

140, 2000 5

T 4 er$CC 1

ge4(ggggc I

i b

(

l I

i i

[

i 5

ibOW RATE PAiw s row uTE PATM 33 CX)RE -

BREAK arsao t.7:ma 4

3i500 m

w E

N I.

3 5

O e

5 3

r3 31530

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• **'E * *"2" CHARCINC so 300 25 23e 20 3

aus E.

3 3

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f ese

=

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1-4/01/93 i

LOCA i

0FFSITE RADIOLOGICAL DOSES f

i ASSUMPTIONS:

i

(

i o

LOCA INSIDE CONTAINMENT o

CONTAINMENT ISOLATION & SPRAY i

MANUALLY AT 30 MINUTES o

ANNULUS VENTILATION SYSTEM STARTED l

MANUALLY IN 60 MINUTES j

o 100% GAP FISSION PRODUCTS RELEASED -

j NO FUEL FAILURE FOR 30 MINUTES o

PRIMARY-TO-SECONDARY LEAKAGE i

PER HUREG-0017 (REV 01)

.00625 GPM i

j o

CONTAINMENT LEAKAGE 0.5% PER DAY l

0 INITIAL PRIMARY COOLANT I0 DINE 4

l CONCENTRATION = 1 xCI/GM-o EPRI URD CHI /O

't o

NUREG-1465 SOURCE TERM i

i RESULTS:

1 l

TWO HOURS 9 EAB

[

297.0 REM THYROID 3.0 REM WHOLE BODY j

30 DAYS 9 LPZ-46.0 REM THYROID i.

0.7

{

REM WHOLE BODY

SYSTEM 80+

NUPLEX 80+ DIVERSITY AND REDUNDANCY EMERGENCY MAIN CONTROL ROOM b

OPERATIONS D

F# *:lLITY CRTs 0

l 0-CHN SUP RT SHUTDO 1 OOM p

p g

CONTROL ROOM OFFICES SAFETY (Design Type 1) l NON-SAFETY (Design Type 2)

Em n

j(

(

(

I

(

L

L o

:E PR E ING INDICAT10 AND SYSTEM O>

DIAS-P ALARM SYSTEM l

Em

t. - - - - - -l RSP O Z

l

?

-O PLANT ESF COMPONENT PROCESS POWER

+

CONTROL SYSTEM l

COMPONENT CONTROL HgM PROTECTION l

g l

CONTROL SYSTEM,9 f

SYSTEM Op2 SYSTEM

^

a a

szg a

g ug a

a a

i i

y O

1 2

/

i Q Og N

REMOTE Z

U MULTIPLEXORS d

y U\\ ~ ~ N]

D. e5 m i

y PLANT COMPONENTS & SENSORS v,

-1 co W

?

Nuplex 80+TM Defense-in-Depth (Current)

Success Path Critical Function Non-Safety Safety Manual Reactivity Control Rod Control, CVCS (Boration)

Safety injection System, Reactor Trip Reactor Trip Breakers Breakers Vital Auxiliaries AC Main Transformer, Alternate A/C Emergency Diesels, Aux Transformers DC Station Battery Station Battery RCS Inventory Control CVCS (Charging / Letdown)

Safety injection System Safety injection System RCS Pressure Control Heaters / Spray, Safety injection System, Safety injection CVCS (Charging)

Safety Depressurization System System Core Heat Removal Forced Circulation Natural Circulation RCS Heat Removal Main Feed Emergency Feed, Emergency Feed Shutdown Cooling &

Safety injection Safety injection System Containment isolation Control Valves isolation Valves isolation Valves Containment Environment Fan Coolers, H, Ignitors Containment Spray, Cont. Spray Recombiners Radiation Emission Monitor and Copntrol Radiation isolation of Release Isolation Valves Release Paths Paths Ak ER ED M EDED

TABLE 2-2 KEY INDICATORS OF CRITICAL FUNCTION STATUS j

i DISPLAYED CONTINUOUSLY VIA DIAS-P

.l i

Sensed Parameters:

i RCS Pressure Coolant Temperature (Hot)

. Coolant Temperature (Cold)

Containment Pressure (Wide Range) j l

Cortainment Pressure (Narrow Range)

Steam Generator Pressure i

Steam Generator Level (Wide Range) i Pressurizer Level Neutron Flux Power Level (Safety Channels j

Reactor Cavity Level RCS Radiation Level l

Containment Area Radiation Containment Hydrogen Concentration j

Containment Isolation Valve Position Emergency Feedwater Storage Tank Level i

Calculated by PAMI Computer:

Co. - [c Jt Temperatures Reactar Vessel Coolant level RCS Subcooling 1

__=

.u i

FIGURE 73-25

• P DIVERSE MANUAL ESF ACTUATION DivCR50 MANUAL Csr Actuar:0N INTERFACE TO ESF COMPONENTS Dtv A Dtv C O SI O CS O 4 Crt @

KCYe O tre @

LC - Loop Controtter MCC - Motor Control Center O $G1 Msiv CI - Containment Air Purge Volve Isotetton LI - Letdown Isolotton O SG2 Mstv

@. Monual Switches O Cl/Lt MCP CONTROLS RSP CONTPOLS HCP CONTROLS R$P CONTROLS

& INDICA 110NS

& INDICATIONS t INDICAf tDNS t INDICAT10N3 Esr-CC3 DivistCN A E$r-CCS DIVI 5!&l C I

I I

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DPERATOR RESPONSE TINES IN NINUkES FDLLOVING A CEA EJECTION 0

2 4

6 8

10 12 14 16 18 20 I

I I

I I

I I

I I

I l

J k Reactoe trip Alarmed octuated Closure of contoenment indica tica*-

ole pur0e isototion volves.

Contoinnent Sprey

- Hi Pwe Jk Jb Actuoted Pre-P SIAS Pressure Pzr-L tep 8 Steps CS gy_gn Step le REACTIVITY CONTROL INVENTDRY CONTAINMENT ENVIRONMENT i

i g

l DP-I N I I

DP-1 ver Fles Rx trip I!

Jg gg gg y

I I

\\ Completes verificatior DP-2 ottemp DP-1 ID's OP-1 correlates DPS Manuel e nted to Initiate of Steps 17 - 20.

need For with DIAS-P for Frze I Ch8P9'"9 L CIAS vio PPS.

Actuates reactor trip Level.

te tdo" containment sprey SUPV-1 deternmes Manuel ci sure or cont, oir purge via HV switches.

oP'I '"'i'OI'8 globol DIAS-N problem volves Instinted via HV switches.

"0"u'l I"D-and that DPS & DIAS-P OP-1 ID's reactivity display problem &

reports to SUPV-l.

EFV Mo'" fe'd b

SUPV-1 evotuotes with' Step 2 & 3 Ha'a i

DP-1 & essigns EPG VERIFY T/G Steps i

jg Steam Step 2 to DP-2.

TRIP Steps 4 - 7 10-16 i

CDRE & RCS HEAT REMDV R Isotated i

f VITAL AC L DC 3

3 I oP-e l+

i i

i i

a Il Il 11 il 11 ii l I a

a 05-2 ID's display 7

/

/

/

problem & reports Complete verirication to SUPV-l.

OP-2 ottempts ID,s or Steps 10 - 16.

to wilote Monvet Aware oF o puss:bte SUPV-1 & DP-2 ESFAS via PPS.

Stops EFV.

octuation DIAS-N problem SUPV-1 diagnose DIAS-N

%n assigns Step 8 to DP-1.

display problem.

Actuates EFV '

feed dwred via HV switches, pumps.

s wt t the s.

Jk SUPV-2 & DP-3 Step 9 SI octuated 4

I oP-a F+

lRCSPRESSURE l

ortve in MCR il 11 II a

DP-3 ottempts f f

1 Actuates Verifies $1 L to wtiete

/ ID's /

Sg via HV RCS Pressure SIAS via E5r.

no St.

,,,t che s.

Controt.

lSUPV-2 1

11 11 11 11 i

SUPV-2 assesses criticot

/

/

s Directs operators function status & determines Assigns OP-3

\\ ssigns DP-1 the following to be in Jeopordy /

to actuate SIAS.

A to use HV swithes to codinue es backup.

Inventory Controt Assions OP-2 to initiote ERAS with steps Pressure Control and MSIS and stop main feed 17 - 20.

RCS Heat Removet Core Heat Removat i

7

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DPERATOR RESPONSE TIMES IN MINUTES FDLLOVING A MAIN STEAM LINE BREAK 0

2 4

6 8

10 12 14 16 18 20 1

I I

I I

l i

I i

l l

d k Re*Clo" i"'D Aloemed

'Ciu**'d mdicotions'

- He Pwe

- Low SG-P

- Low Pre-P 51AS Pressure D8 PZR empty Steps l

17-20 g

, Step li REACTIVITY CONTROL INVENTORY CONTAINMENT ENVIRONMENT H

g g

I I

OP-l vertfles Rx trip 33 3

i I

gg t

11 11 11 1

[

DP-1 ID's OP-1 correlates DPS Manual con of 3~

or charging &

0P-1 completes steps.

P eacto trip Level letdown.

SUPV-1 determines OP-t tnettetes globot DIAS-N problem monuot trip, and that DPS & DIAS-P DP-l ID's reactivity display peoblem &

reports to SUPV-l, try Mein feed c stedJk J 8W SUPV-t evoluotes with Step 2 & 3 Mo8n OP-1 & essigns EPG VERIrY T/G steps jg Steam Step 2 to DP-2.

TRIP Steps 4 - 7 10-16 Isototed 3

VITAL AC t DC gCORE & RCS HEAT REMOVAL I oP-e l+

3, i

11 11 it it it gg g g

=

OP-2 1D's disptoy 7

f

/

[y-1.

OP-2 ottempts ID's or Steps 10 - 16.

Complete verirication to initiate Monuot E}y*

Steps Aware of a possible SUPV-1 & OP-2 EsrAs via Pps.

octuotson DIAS-N probten TUPV-1 diagnose DI AS-N main assigns Step 9 to OP-1.

disploy probten.

Actuates EFV feed h dwired vio HV switches.

pumps.

s witches.

l 4

h SI

$UPV-2 t OP-3 Step 9 octuated b

lOP-3 h lRCSPRESSURE orive in MCR 8

Il ti 11 OP-3 ottenpts f f

ID,s /

Actuates Verifies 31 &

]

to initiote

/

Si vio HV RCS Pressure SIAS vio E$r, "O II-s wit che s.

Controt.

l lSUPV-2 %

i 11 11 11 it i

SUPV-2 ossesses criticot

/

\\

Directs operators function status & determsnes Assigns OP-3

% ssigns OP-1 A

to use HV switbes the follow'ng to be in Jeopordy, to actuote SIAS.

Inventory Controt Assigns CP-2 to initiate EFAS with steps i

Pressure Control and M515 ond stop main feed 17 - 20.

RCS Heat P=movat Core Neot Removol t[

CMF-AP!b n

-