ML20091D976

Forwards GE Responses to Resolution of Issues Related to Chapter 15 of Advanced Bwr,Draft Ser,Incorporating Responses to NRC Staff Comments and to Outstanding Issues 136 & 139

ML20091D976
Person / Time
Site:	05200001
Issue date:	04/06/1992
From:	Marriott P GENERAL ELECTRIC CO.
To:	Pierson R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
MFN-081-92, MFN-81-92, SLK-9249, NUDOCS 9204130227
Download: ML20091D976 (50)

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April 6,1992 MFN No. 081-92 Docket No. STN 52-001 SLK 9249 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20S55 Attention:

Robert C. Pierson, Director Standardization and Non Power Reactor Project Directorate

Subject:

GE Responses to the Resolution of Issues Related to Chapter 15 of AllWR DSER SECY 91355

Reference:

GE Responses Agenda Items 1,5,9 and 16 Discussed Durir; the GE\\NRC Reactor Systems Branch Meeting on November 20-21,1991, MFN No. 010-92 dated January 10,1992 Enclosed are thirty four (34) copies of the GE responses to NRC staff comments and to Outstanding Issues 136 and 139. In addition, the test has been revised to incorporate the response to Outstanding Issue 135 provided in the referenced letter.

It is intended that GE will amend the SSAR with this response in a future amendment.

Sincerely, f

-[,

fawn dy P.W. Marhihtt, Manager Regetatory and Analysis Services M/C 444, (408) 925-6948 cc: F.A. Ross (DOE)

N.D. Fletcher (DOE)

C. Poslusny, Jr.

(NRC)

O. Thomas (NRC)

R. C. Berglund -

(GE)

J. F. Ouirk (GE) 1001.1.0 bp I

'9204130227 920406 PDR ADOCK 05200001 PDR; A

ABWR Standard Plant arv c documentation. The limiting events which 15.0.43.3 Barrier Performance establish CPR operating limit:

The significant areas of interest for (1) Limiting Pressurization Esents: Inadvertent internal pressure damage are the high pressure closure of one turbine control valve, and portions of the reactor coolant pressure generator load rejection with all bypass boundary (the reactor vessel and the high vahe f ailure, pressure pipelines attached to the reactor vessel). The plant shall meet the criteria in Appendix 4B.

(2) Limiting Decrease la Cort Coolant Temperature Event 6 Lm J Tw i-m %.

15.0.4.5.4 Radlobgical Consequences Fuot u s.n.,, Wtvolk G.,%. M*o. m D < s-a.A For the core loading in Figure 4.31, the in this chapter the consequences of resulting initial core MCPR operating limit is radioactivity rel-

for the core loauing in 1.17. The operating limit based on the plant Figure 4.31 duri g the three types of events

loading pattern will be provided by the utilhy (a) incidents of moderate frequency (anticipated applicant referencing the ABWR design to the operational occurences); (b) infrequent USNRC for information, see Su'osection 15.0.5.2 incidents (abnormal operational occurences); and for interface requirement.

(c) limiting faults (design basis accidents),

are given. For all events whose consequences Results of the transient analyses for are limiting, a detailed quantitative evaluation individual plant reference core loading patterns is presented. For nonlimiting events, a will differ from the results shown in this qualitive evaluation is presented or results are chapter. However, the relative results between referenced from a more limiting or enveloping core associated events do not change. Therefore, case or event.

only the results of the identified limiting evente given in Tables 15.0 4 will be provided by 15.0.5 Interface Requirements the utility applicant referencing the ABWR design to the USNRC for information. See Subsection 15.0.5.1 Anticipated Operational Occurences 15.0.5.1.

(AOO) 15.0.4.5.1 Effect of Single Failures and The results of the events identified in Operator Errors Subsection 15.0.4.5 for plant core loading will be provided by the ultiity applicant referencing The effect of a single equipment failute or the ABWR design to the USNRC for information.

malfunction or operator error is provided in Appendix 15A.

15.0.5.2 Operating Umits 15.0.4.5.2 Analysis Uncertainties Tbe operating limit resutting from tbc analyses normally provided in this subsection The analysis uncetainties snect the criteria in will be provided by the ultiity applicant Appendix 4B.

referencing the ABWR design to the USNRC for information.

In Table 15.0 3. a summary of applicable accidents is provided. This table compare-GE 15.0.5.3 Design Basis Accidents calculated amount of failed fuel to that u:.. in worst case radiological calculatiuons for the Results of the design basis accidents core shown in Figure 4.3-1.

Radiological including radiological consequences will be calculations for a plant initial core will be provided by the ultiity applicant referencing provided by the utility to the USNRC for the ABWR design to the USNRC for information.

information. (See Subsection 15.0.5 for interf ace requirements).

Amendment 15 1504

l

-i

'.AB M unmora Standard Plant nov c Table 15.01 INPUT PARAMETERS AND INITIAL CONDITIONS FOR SYSTEM RESPONSE ANALYSIS TRANSIENTS (Continued)

29. S/R Valve Reclosure Setpoint Both Modes

(% of setpoint)

- Maximum Safety Lim *: (used in analysis) 98

- Minimum Operational Limit 93

30. High Hux Trip (% NBR)

Analysis Setpoint (125 x 1.02) 127.5

31. High Pressure Scram Setpoint (kg/cm g) 77.7

32. Vessellevel Trips (m above bottom of separ; tor skirt bottom)

Level 8 - (L8) (m) 1.73 Level 4 - (L4) (m) 1.08 level 3 -(L3) (m) 0.57 Lev-! 2 - (L2) (m)

-0.75

33. APRM Simulated Thermal Power Trip Scram % NBR Analyds Setpoint (115 x 1.02) 117.3 Time Constant (sec) 7

34. Reactor Internal Pump Trip Delay (sec) 0.16

35. Recirculation Pump Trip inertia Time Constant for Analysis (sec) '"

0.62

56. Total Stcamline Volume (m )

113.2

37. Set presjure of Recirculation pump trip (kg/cm g) 79.1 For trasents simulated on the ODYN model, this input is calculated by ODYN.

" EOEC = End of Equilibdum Cycle.

'" The is ertia time constant is defined by the expression:

K.

2/J n

, where t

= inertia time constant (sec);

t=

g T, J,

= pump motor inertia (kg-m);

n

= pump speed (rps);

g

= gravitational constant (m/sec7; and T,

= pump shaft torque (kg m)

Amendment 13 15A6 1

-{

l ABWR 234 imAB Standard Plant REv C Table 15.0 2 RESULTS SUM 51ARY OF SYSTEM RESPONSE ANALYSIS TRANSIENT EVENTS Mr.x Core No Mn Mn Average

,f Durr er.

Mn Mn Venet Steam Surface Vahts of Sub Neutror.

Dome Bottorn Une lleat Dus a

Freq. Fint Bla dent Sect a Figur2 Dua Preuvre Pressurt Preuvre

(% of in Cats-Blo=-

(seconds) 2 2

2 m

Q De sm et" M (b81CD g}

(W Cmg)

(Kr/Cm g) 33 y,5) ggy g;;g; ggg 15.1 Decrease in con coolant temperature 15.1.1 less of Feed-112.8 73.1 75.9 71.6 112.8 0.07 a

0 0

water beattrig 15,1 2 15.1 2 Runout of one 104 5 73 2 75 8 71.7 101 8 0 06 a

0 0

feeduter pum; 129.0 83.3 64 9 82.8 105.9 C.10 a+

10 6

15.1.2 15.1 3g "-.. _. :;-.

gu c.s,. us

. go. :w Do nd g,;q, 15.s.3 15.1-4 Opening of 102.1 73.1 73.6 71.6 100.0 a

0 0

one Bypass Vaht 15.1.3 15.1 5 Opening ot all 102.0 80.4 81.8 80.1 100 0 a+

0 0

Control and Bypass Vahts 15.1.4 Inadvertent open SEE TEX 1' ing of One SRV 15.1.6 Inadvertent RHR SEE

'mXT Shutdown Cc-ting 15 2 increau in Reactor Presure 1511

$5.21E C1caurt of One 2!ts" 75f p

g 109T 0.10 a

0 0

Turbine Control #di 15.I 7 ?, (,

}},';

o3 L Valvt 1511 15 2-2 Prn Regu-1543 15.8 87.4 85.1 103.0 N/A c

18 6

tator Dommcate Fail.

152.2 15 2-3 Generator Imd 148.1 83 2 84.7 82.7 100.2 0.06 a

10 5

%ecten, Bypaa on Frequency defmition is discussed in Subsection 13.04.1 Not limiting (See Subsection 15.04.5.)

a Moderate Frequency b

Infrequent c

Limiting Fault N/A Not applicable This event should be classified as a limitingfault. However, criteria for moderate frequer.: inciden:s are

+

conserva:ively applied.

15 0-7 Amendmerit 15

^ A.BWR m61mD

.[

Standard Plant Rev c Table 15.0 2 RESULTS

SUMMARY

OF SYSTEM RESPONSE ANALYSIS TRANSIENT EVENTS (Cont.)

Mu Con No.

Ma Mu Avengt of Duration Ma Mu Vesel Sicam Sudace WMs of Sub Neutron Dome Bottom

!)nt Heat Dux A

Freq. Fint Blomhn Section Figure Flus Preuure Pressun Preuure (9r of in Ce te. Blo. Occonds) 2 2

2 G

Q Desemice 3.118 (Kr'Cm 1)

$&!fm g)

(Ar!Cm 8)

Imtit!)

23, Ed 6

/

1512 15.2 4 Generator ized 155.3

&41 85.8 83.6 100.5 0.07 a+

14 5

Rejection, Failun of One Dypasa VaM 13.2.2 15.2 5 Generator Lead 184 6 86.1 87.7 85.6 102.3 0.10 a+

18 6

Rejecten with failure of all Bypau Valves 1513 15.2 4 Turbine Tnp 122.1 83 0 64 6 82 6 100.0 0.05 a

10 5

Bypau-On 1513 15.2-7 Turtune Trip 131.9 64 1 85.6 83 4 100 0 0.05 a+

14 5

w/ Failure of One Dypau Vaht 15.2.3 15.2-8 Turbine Tnp 158.6 86.1 87.7 85.4 100.6 0.J8 a+

15 6

with fadun of all Bypau Whea 15.2 4 15.2-9 Instenent 102.1 84 6 86 4 84.1 100.1 a

18 5

MSlV Cicsun 15.2.5 15.2-10 Lm of 122.3 83.0 64 6 82.6 100.0 a

10 f

Condenser Vacuum 15 *.5 15.2-11 loss of AC 113.2 82.9 84 4 82.7 100.0 0.05 a

10 5

Power 1517 15.2 12 less of All 102.0 73.1 75.7 71.6 100.1 a

0 0

Feed =3ter Flow 1518 Feedwater Piping Break SEE

' EXT Frequency definition is discussed in Subsection 150,4.1 Not limiting (See Subsection 150.4.5) a Moderak Frequency b

infrequent c

Limi:ing Fault N/A Not applicable This event should be classified as an infrequent event or a limitingfault. However, criteria for

+

moderatefrequent incidents are conservatinly applied.

Amendment 15 15.0 8

ABWR 2mius Standard Plant Rrv c Table 15.0 2 RESULTS

SUMMARY

OF SYSTEM RESPONSE ANALYSIS TRANSIENT EVENTS iCont.)

g Mu Cort N:

M at.

M as.

Average of Duration Mas M az.

Venel Steam Surface Vahts of Sub Neutron Dome Bottom Linc llcat Dun A

Frtq. Fint Blo*4oun Section Dgvre Put Pnuure Preuvre Pnuurt

(% of in Cate. Dic=-

(unnds) 2 2

2 Q

d Demtion M Gs1Cm g)

(ht!Cm g]

OJt!Cm &)

lada!)

fl.E Em* h -

15 19 Failure of RHR SEE TITT Shutdown Coohng 15 3 Decrease in Reactor Ceolant System Flow Rate 15.3.1 15 3-1 Tnp of Thee 102.0 73.3 76 0 71.7 100.1 0.04 s

0 0

Reactor laternal Pump 15 3.1 153 2 Tnp of All 102.0 83.2 84.1 82.7 1003 c

Ructor hternal Pumps 15 3.2 15 3-3 Fast Runback 102.0 73.0 75.9 71.6 100 0 a

0 0

of One Reactor laternal Pump 15 3.2 15M Fast Runback 102.0 73.1 76 0 71.6 100.0 a+

0 0

of All Raa: tor Internal Pumps 1533 1535 Seizure o(One 102.0 73.1 75 9 71.6 100.0 c

0 0

Reactor latetus!

Pump 15 3.4 One Pump Shaft SEE TEXT Break 15.4 Ructrvity and Power Dutnbution Anomahes 1541.1 RWE Refueling SEE

'IT.XT Frequency definition is discussed in Subsection 15.0.4.1 Not limiting (See Subsection 15.0.4.5.)

CPR cnterion dm not apply. PCT <593.30C a

Moderate Frequency b

Infrequent c

Limiting Fault 1his event should be classified as a limitingfault. However, criteria for moderatefrequent incident.s are

+

conservatively applied.

15.0-9 Amendment 15

[ ;<'

YABWR uA6im4D -

g-.=

Standard Plant Rrv C Table 15.0 2 RESULTS

SUMMARY

OF SYSTEM RESPONSE ANALYSIS TRANSIENT EVENTS (Cont.)

Max Core No Max.

Max.

Average of.

Duration Max..

Max.

Vesel Stearn Surface -

Valves et Sub Neutron Dome Dottom -

the liest Mux

.A Freq. I~ttst - Blowdown 5:ction Figure Mux Preuure Preuvre Preuure

(% of in Cate-Dlow-(seconds) 2 2

2

.d M - - Desenpeien 3,. 'I!,B (b/Cmg)

(h/Cm 1)

(M/Cm E)-

Inl!.!!.D C'.lk 120.* IL3 -

/

154.12 RWF Startup SEE TEXT l' 4.2 RWE at Power SEE 1TXT 1543 Control Rod SEE

'1TXT Whsoperation 15AA AbnormalStartup -

SEE TEXT of One Reactor laternal Pamp 15.43 15 4-2 Fut Runout 89.A 71.1 72.3 70.6 116.1 0

0 of One Reactor Internal Pump 15A3 15.4-3 Fut Runout 135.0 72J 74.7 71 3 1683 a+

0 0

ot All Reactor Internal Pumps l 15A.7 Misplaced Dendic SEE TEXT Accident 1548.

8tod Ejection Accident SEE TEXT 154 9 -

Control Ec4 Drop Accident '

SEE, TEXT 15 3 Increase in Reactor

- Coolant laventory

~ 153 1 Inadvertent 102.0 73.1 75.6 71.6 100.0 a

0 0

l 153.l.

IIPCF Startup Frequency definition is ',.ussed in Subsection 15.0.4.1 Not limiting (See Subsection IS.O.4.S.)

Transients initiatedfrom lowpower.

a Moderate Frequency b

. infreque st c^

Limiting F:.att

+.

This event should be classified as a limitingfault. However, criteria for moderatefrequent incidents are conservatively applied.

i-i Amendment 17 15A10 4

_. ~,.....

. 1 1

ABWR nwwan Standard Plant nry c Table 15.0 3 SL31 MARY OF ACCIDENTS FAILED FUEL RODS GE NRC b'

CA14ULATED WORST-CASE SUtlSECTION tea uw f+(4 bowaada)

TITLE VALUE ASSUMI' TION I.D.

M t o. 2 */.

o r. z.I 15 3.1 Trip of All Reactor Internal Pumps None e < 0, 2. '/

J 1533 Seizure of one Reactor laternal Pump None None 7

15 3.4 Reactor laternal Pump Shaft 3reak None None 15.6.2 Instrument Line Break None None 15.6.4 Steam System Pipe Break Outside None None Containment 15.6.5 LOCA Withia RCPB None 100 %

15.6.6 Feedwater Line Break None None 15.7.1.1 Main Condenser Gas Treatment N/A N/A System Failure 15.7 3 Liquid Radwaste Tank failure N/A N/A 15.7.4 Fuel. Handling Accident

< 125 125 15.7.5 Cask Drop Accident None All Rods in Cask Table 15.0-4 CORE-WIDE TRANSIENT ANALYSIS RESULTS TO BE PROVIDED FOR DIFFERENT CORE DESIGN h1AX. CORE h1AX.

AVERAGE NEUTRON SURFACE FLUX llEAT FLUX L LTA TRANSIENT WNBR)

MNBR)

Cf.B FIGURE Closure of One Turbine Control Valve X

X X

X Load Rejection with all Bypass Vahes X

X X

X Failure o,._.,c,n.x

.m. "-

X X

X X

FuMA h //A Fai/a

- Mu:n pama4 l

t5 0-11 Amendment 15 l

.-AB W

=*

m swaxa

• Standard Plant uv A-.

4 SECTION 15,1 CONTENTS Seetion Title-Eage 15.1.1 -

Loss of Feednter Heatine 15.1 1 15.1 1 Identification of Causes and Frequency Classification 15.1 1 15.1.1.1.1 Identification of Causes 15.1 1 15.1.1.1.2 Frequency Classification 15.11 15.1.1.2

_ Sequence of Events and Systems Operations 15.1 1 15.1.1.2.1 Sequeace of Events 15.1-1 15.1.1.2.2 Systems Operation 15.1 1 15.1.1.3 Core and System Performance 15.1 2 15.1.1 3.1 Input Parameters and initial Conditions 15.12 15.1.1.3.2 Results 15.1 2 15.1.1.4 Barrier Performan'cc 15,1 2 15.1.1.5 Radiological Consequences 15.12 15.1.2-Feedwater Controller Failure-Maximum Demand 15.1 2 15.1.21 Identification of Causes and Frequency

, Classification -

15.1 2 15.1.2.1.1 Identification of Causes 15.12 15.1.2.1.2 Frequency Classification 15.1-3 15.1.2.1.2.1 Runout of One Feedwater Pump 15.13

@e,4o A*c8/w Alf%

A*w= bu(

15.1.2.1.2.2 noaum vi a wu. ---w.

i muy 15.1-3 15.1.2.2 '

Sequence of Events and Systems Operations 15.1-3 15.1 il

' ABM ux6iorwa o

Standard Plant L

PJV A SECTION 15.1 CONTENTS (Continued)

Section l'Illt E.aga 15.1.2.2.1 -

Sequence of Events 15.1-3 15.1.2.2.1.1 Runout of One Feedwater Pump 15.1 3

. 15.1.2.2.1.2 g

D"-^"'

' T : F:: F ':- P 15.13 Y

15.1.2.2.1 3 Identification of Operator Actions 15.1 3 yf x

f 15.1.2.2.1 3.1' Runout of One Feedwater Pump 15.13 Qy v.

O l'.1.2.2.13.2

. _. ' T ~ Tv.- mm Tmue 15.1 4 3

15.1.2.2.2 Systems Operation 15.1-4 4

- 'Q

. 15.1.2.2.2.1 Runout of One Feedwater Pump 15.1-4 1

15.1.2.2.2.2

_ - :' T- : _2...m i u2ue 15.1-4

\\

15.1.2 3 Core and Sptem Performance 15.14

- b.1.23.1 Input Parameters and Initial Conditions 15.14 LL.

15.1.2 3.2 Results -

15.1-4 15.1.2.3.2.1 unout of One Feedwater Pump 15.1-4 15.1.2 3.2.2 '

--.. m'. wu a suu =c1 1 usup 15.1-4 15.1.2.4 Barrier Performance 15.14

- 15.1.2.5 RadiologicalConsequences 15.15 15.13 Pnssun Regulator Failum--Onen 15.15 g".

15.1 3.1 Identification of Causes and Frequency Classification 15.15 15.1 3.1.1.

Identification of Causes 15.15 15.1 3.1.2 Frequency Classification 15.1 5 15.1 iii

ABWR numw Sinndard Plant krv c SECTION 15.1 CONTENTS (Continued)

Sect 10n Tillt Eggt 15.1.6.2.2 Systems Operation 15.1 8 15.1.6.3 Core and System Per'ormance 15.19 15.1.6.4 Barrier Performance 15.1 9 15.1.6.5 Radiological Consequences 15.19 TABLES Table Illlt Eage 15.1 1 Sequence of Events for less of Feedwater H:ating 15.1 10 l

0 Loss of 55.6 C Feedwater Heating 15.1 10 15.1-2

• r.V t o.

Len *4 l6 7*c Fu k k Hr*Jr.g a v.I-r o 15.1-3 Single Failure Modes for Digital Controls 15.1 11 15.1-4 Sequence of Events for Figure 15.12 15.1 12 15.1-5 Sequence of Events for Figure 15.13 15.112 15.1 6 Sequence of Events for Figure 15.14 15.1-13 15.1 7 Sequence of Events for Figure 15.15 15.1 13 15.1 8 Sequence of Events for inadvertent Safety / Relief Valve Opening 15.1-14 15.1 9 Sequence of Events for inadvertent RHR Shutdown Cooling Operations 15.1 14 15.1 vi Amendment 13

. ~

?

'ABWR'-

- muun.

Ei

. Standard Plant

_ - arv c

- SECTION 15.1-ILLUSTRATIONS :

Baure Iille Page r

1S11.

Simplified Block Diagram of Fault Tolerant Digital Controller System 15.115 15,1 2 Runout of One Feedwater Pump 15.1 16-Iw@ M+< IL Fs%r - %im WA..

15.1-3 T, i... " -;_

15.1 17 15.1 4 Inadvertent Opening of One Bypass Valve 15.1-18 15.1-5, -

Inadvertent Opening of All Control and Bypass Valves if.1 19 '

/

t i

i t

15.1 vii Amendment 18

ABM unmoru S.tandard Plant prv c 15,1 DECREASE IN REALTOR COOLANT C' #" M # #

TEMPERATURE Bec use this event is very slow, the operator 15.1.1 Loss of Feet. water Heating action will terminate this event. Therefore, the worst event is the loss of feedwater heating 15.1.1.1 Identification of Causes and resulting in a temperature difference just below Frequency Classification the AT setpoint. However, a loss of 55.60C l feedwater temperature is anclyzed to bound this 15.1.1.1.1 Identincation of Causes event.

A feedwater heater enn be lost in at least two 15.1.1.1.1 Frequency Classification ways:

The probability of this event is considered I

(1) steam extraction line to heater is closed; lov enough to warrant it being categorized as an infrequent incident. However, because of the or lack of a sufficient frequency data base, this (2) steam is bypassed around heater, transient disturbance is analyzed as an incident of moderate frequency.

The first case produces a gradual cooling of the feedwater. In the second case, the steam 15.1.1J Sequence of Events and Systems bypasses the heater and no beating of that Operation feedwater occurs. In either case, the reactor vessel receives cooler feedwater. The maximum 15.1.1.2.1 Sequence of kaents number of feedwater heaters which can be tripped or bypassed by a single event represents the most Table 15.11 lists the sequence of events

. severe transient for analysis considerations. for this transient.

bid '~gThis event has been conservatively estirnated to h

incur a loss of up to 55.60C of the feedwater 15.1.1.2.1.1 Identification of Operator heating capability of the plant and causes an Actions increase in core inlet subcooling. This increases core power due to the negative void Because no scram occurs during this event, no reactivity coefficient. However, the power immediate operator action is required, As soon increase is slow, as possible, the operator should verify that no operating limits are being exceeded. Also, the

, The feedwater control system (FWCS) inclubs a operator should determine the cause of failure

) logic intended to mitigate the consequences of a prior to returning the system to nortnal.

loss of feedwater heating capability. The system

- will be constantly monitoring the actual 15.1.1.2.2 Systems Operation feedwster temperature and cornparing it with a reference temperature. When a loss of feedwater in establishing the expected sequence of heating is detected (i.e., when the difference events and simulating the plant performance, it between the actual and reference temperatures was assumed that normal functioning occurred in exceeds a AT setpoint, which is currently set the plant instrumentation and controls, plant at 16.70C), the FWCS sends an alarm to the protection and reactor protection systems.

operator. The operator can then take actions to mitigate the event. This will avoid a scram and The high simulated thermal power trip (STPT) reduce the a CPR during the event. rl.a scram is the primary protection systesu trip in

.fa u, rip..) 4 4,, reg. # $ t mitigating the consequences of this event.

However, the power increase in this event is not

/t C 4 I J f., a d e, w.4 r e. Ag I 1.igh enough to initiate this scram. Operation (4.fcJt4 Q; 4 m. 4, i tg of engineered safeguard features (ESF) is not 4 h % % % N A.u. tt 4 4dvv f' A N m : d v w..s:;:al & 2 p.mndmut t$ M l' JY Md % **

1511 frrh r l % % M M J lits,

o 9

~

.i IW/.s (O T4 AcWA 4

L &

e cA %AJ M

.c e yi o,xvA.

%~

.4g~ y ~ =.J / d m I= n vf m-n< Am 9WC*C slo ew

( too ' F) hxdadn My.

Th & v.

.rydn.rk shan u.< (o.aso cavmon A Fce ed.I-/ %

t o. / - 3 M

& 's ng=O W.

L.j<d, +h PJ.WpA Qkwn+G pa 4 4 4</a, a h 4 R

/o. /- / 4 f.co L so'c'lr3*p3

, +ee w f ht.6 *c.

(fou'F) A y n &+ derf % +4 do-wa'd c.,stys.u 7.5 wMve_.

MN DAMAH Standard Plant uv c expected for this transient.

15.1.2 Feedwater Controller Failure--

Maximum Demand 15.1.13 Core and System Performance 15.1.2.1 Identification of Causes and 15.1.1 3.1 Input Parameters and Initial F'requency Classification Conditions 15.I.2.1.1 Identificttion of Causes The transient is simulated by programming a change in feedwater enthalpy corresponding to a This event is postulated on the basis of a 55.60C loss is feedwater heating. A. A ca.4 single failure of a control device, specifically w:/4 $la 4 T* r4t Ne % me 3 4 g 7,c, one which can directly cause an increase in

?

15.1.1.3.2 Results i ! O % %p,(,

coolant inventory by increasing the feedwater flow.

Because the power increase during this event is relatively slow, it can be treated as a quasi The ABWR feedwater control system uses a steady-state transient. The 3 D core simulator, triplicated digital control system instead of a has been used to evaluate this event for the single-channel analog system as used in current equilibrium cycle. The results are summarized in BWR designs (BWR 2-6). The digital systems TablGr l5.12W tr, l-m,

consist of a triplicated fault. tolerant digital controller, the operator control stations cud The MCPR response of this event is small due displays. The digital controller comins tivec to the mild thermal power increase with shifting parallel processing channels, each -. mining l axial shape. The worst a CPR response is 0.07.

the microprocessor based hardware and associated software necessary to perform all the co arol No scram is initiated in this event. The calculations. The operator interface provides l increased core inlet subcooling aids thermal inforaiation regarding system status and the margins. Nuclear system pressure does not change required control functions.

2 significantly (less than 0.4 Kg/Cm ) and consequently, the reactor coolant pressure Redundant transmitters are provided for key boundary is not threatened.

process inputs, and input voting and validation are provided such that faults can be identified 15.1.1.4 'tiarrier Performance and isolated. Each system input is triplicated internally and sent to the three processing As noted previously the consequences of this ch t.nnels. (See Figure 15.1 1) The channels event do not result in any temperature or will produce the same output during normal pressure transient in excess of the criteria for operation. Interprocessor communication which the fuel, pressure vessel or containment provides self diagnostic capability. A two out-are designed; therefore, these barriers maintain of three voter compares the processor outputs to ]

their integrity and function as designed.

generate a validated output to the control actuator. A separate voter is provided for each 15.1.1.5 Radiological Conr*quences actuator. A *ringback' feature feeds back the final voter output to the processors. A voter Because this event does not result in any fuel failure will thereby be detected and alarmed.

failures or any release of prirnary coolant to In some cases a protection circuit will lock the either the secondary containment or to the actuator in*c its existing position promptly environment, there are no radiological after the ta..ure is detected.

consequences associated with this event.

Amendment l$

15 1-2 i

I

.SMN nA61oo.e Standard PlanL_

Rrv_c Table 15.13 lists the failure modes of a remaining feedwater pump will decrease to offset triplicated digital control system and outlines the increased flow of the failed pump. The the effects of each failure. Because of the effect on total flow to the vessel will not be triplicated architecture, it is possible to take significant. The worst additional single fail-one channel out of service for maintenance er ure would cause /A feedwater pumps to run out repair while the system is on line. Modes 2 and to their maximum capacity. However the proba-5 of Table 15.13 address a f ailure of a bility of this to occur is extremely low (less component while an associated redundant component than 7 x 10 5 failu per reactor year).

is out of service. This type of failure could potentially cause a system failure. However, the 15.1.2.1.2 Frey,ency Classification probability of a component failure during servi-cing of a counterpart component is considered to 15.1.2.1.2.1 Runout of One Feedwater Pump be so low that these failure modes will not be

- considered incidents of moderate frequency, but Although the frequency of occurrence for this will be considered limiting faults, event is less than once per 100 reactor years, this event is consersatively evaluated as an Adverse effects minimization is mentioned in incident of moderate frequency.

the effects of Mode 2. This feature sterns from F w -Aerce+ b % F A - A + -

.5e additional intelligence of the system 15.1.2.1.2.2 D e ^ r '" " 7 lba nci provided by the microrsrocessor. When possible, the system will be piogrammed to take action in The frequency of occurrence for this event is the event of some fa;iure which wi'.I reduce the estimated to be less than once per 10000 years, severity'of the transient. For example, if n e It should be classified as a limiting fault as total steam flow or total feedwater flow signals specified in Chapter 15 of Regulatory Guide failed, the feedwater control system will detect 1.70. Nonetheless,.. m i

m. m l

x this by the input reasonability checks and ; -.. - un uo ugnmcam nu pm vu -

automatically switch to one element mode (i.e.,

ci m.

GR umit, the criteria of moderate control by level feedback only). The level frequent incidents are conservatively applied to control would essentially be unaffected by this this event.

failure.

15.1.2.2 Sequence of Events and Systems The only credib'e single failures which would Operation icad to some adverse affect on the plant are Modes 6 and 7, r. failure of the output voter and 15.1.2.2.1 Sequence of Events a control actuator failure. Both of these failures would lead to a loss of control of only 15.1.2.2.1.1 Runout of One Feedwater Pump one actuator (i.e., only one feedwater pump with increasing flow). A voter failure is detected by With momentary increase in feedwater flow, the ringback feature. The FWCS will initiate a the water level rises and then settles back to lock-up of the actuator upon detection of the its normal level. Table 15.14 lists the failure. The probabilities of failure of the sequencing of events for Figure 15.1-2.

variety of control actuators are very low based p.u/ r./w c..,M il Fa:Im.4 -A;. %

on operating experience (less than 0.0088 15.1.2.2.1.2 " - J T2 F % Nm ha failures per reactor year). In the event of one pump run.out, the FWCS would then reduce the With excess feedwater flow, the water level demand to the remaining pump, thereby rises to the high level reference point, at automatically compensatin6 or the excessive flow which time the feedwater pumps and the main f

from the failed pump. Therefore, the worst turbine are tripped and a scram is initiated.

single failure in the feedwater control system Table 15.15 lists the sequence of events for causes a run-out of one feedwater pump to its Figure 15.1-3. The figure shows the changes in maximum capacity. However, the demand to the important variables during this transient.

Amendment t3 151 3

~---

ABWR mmn

- Signdard Plant pfv c required. As soon as possible, the operator assumed to be 75% of tated flow at the design should verify that no operating limits are being flow for du/ 74.9 kg/cm'g. The total feedwaterl pressure 9 execeded. Also, the operator should determine pumps runout is assumed to be 1309 the cause of failure prior to returning the of rate;d at the design pressure of 74.9 system to normal.

kg/cm'g.

PuAsh Cei,hol/n Fa,%.a.

15.1.2.2.13.2 ""- -"' -"u rm_. -- *- r 15.1.2.3.2 Results Mwsm 92 mA 4 The operator should:

15.1.2.3.2.1 Runout of One Feedwater Pump (1) observe that high feedwater pump trip has The simulated runout of one feedwater pump terminated the failure event; event is presented in Figure 15.12, When the increase of feedwater flow is sensed, the (2) switch the feedwater controller from auto to feedwater controller starts to command the manual control to try to regain a correct remaining feedwater pump to reduce its flow output signal; and immediately. The vessel water level increases slightly (about 6 inches) and then settles back (3) identify causes of the failure and report to its normal level. Tl;c vessel pressures only all key plant parameters during the event, increase about 0.1 kg/cm'. MCPR remains above the safety limit.

l 15.1.2.2.2 Systems Operation FwhA C+ NL F* N - A a.,

15.1.2.3.2.2 L...ous m s wo r ceu,,m....,

15.1.2.2.2.1 Runout of One Feedwater Pump D4 ed g

The simulated runout of Lwa feedwater pumps Runout of a single feedwater pump requires no iie is shown in Figure 15.13. The high l

protection system or safeguard system operation, water level turbine trip and feedwater pump trip

~

This analysis assumes normal functioning of plant are initiated at approximately 18 seconds.

Justrumentation and controls.

Scram occurs and limits the neutron flux peak F=JA c.., hells Fdfe and fuel thermal transient so that no fuel 15.1.2.2.2.2 'C C E *

-- t damage occurs. It is calculated that the MCPR r

Mu.%W P 4 6 * * *(

s right at the safety limit. Therefore, the To properly simulate the expected sequence of design limit for the moderate frequent incident events, the analysis of this event assumes normal is met. The turbine bypass system opens to functioning of plant instrumentation and limit peak pressure in the steamline near the 2

controls, plant protection and reactor protection SRVs to 82.8kg/cm g and the pressure at the systems. Imp'ortant system operational actions b < am of the vessel to about 84,9 kg/cm'g.

for this event are high level tripping of the main turbine and feedwater purnps, scram and The level will gradually drop to the Low recirculation pump trip (RPT) due to turbine Level reference point (Level 2), activating the trip, and low water level initiation of the RCIC system for long term level control.

reactor core isolation cooling (RCIC) system to maintain long. term water level control following The applicant will provide reanalysis of this tripping of feedwater pumps.

event for the specific core configuration.

15.1.23 Core and System Performance 15.1.2.4 Barrier Performance 15.1.23.1 Input Parameters and Initial As previoJsly noted the consequence of this Conditions event dor. not result in any temperature or pressur. transient in excess cf the criteria for The runout capacity of one feedwater pump is which the fuel, pressure vessel or containment are designed; therefore, these barriers maintain Amendment 15 151 4

'ABWR

=a i

Standard Plant prv c discharged into the supprenion pool. The sudden 15.1.6 Inadvertent 111111 Shutdown Cooling

^

increase in the rate of steam flow leaving the Operation reactor vessel (auses a mild depressuriration I.

t r mie nt.

15.1.6.1 Identification of Causes and F'requency Claulfication

'~

h j

Tbc SD&PCS senset the nuclear system preuure decrease and within a few seconds closes the

!!.1.6.1.1 Identification of Causes j turbin, control valves f ar enough to stabilire the reactor vessel pressure at a slightly lower At design power conditions, no conceivable l value and the reactor settles at nearly the malfunction in the 6butdown cooling system could initial power level. Thermal margins decrease cause a temperature reduction.

only slightly through the transient, and no fuel damage results from the transient. MCPR is In startup or cooldown operation, if the essentially unchanged at.d, therefore, the safety reactor were critical or near critical, a scry limit margin is unaffected and this event does slow increase in reactor power could result. A not have to be reanalyzed for specific core shutdown cooling malfunction leading to a configurations, moderate temperature decrease could result from n.pd @

misoperation of the cooling water controts for 15.1.4.4 Barrier Pstformance the RHR beat excbar.gers. 'ib resulting temperature decrease woul. ause a slow As presented previously, the transient insertion of positive rea,s. ;ty tuto the core, resulting from a stuck open relief valve is a If the operator did not act to control the power mild depressurization wnich is within the range level, a high neutron flux reactor scram would l of normal lot d following and therefore has no terminate the transient without violating fuel significant effect on RCPB and containment design thermal limits and without any measurable preuure limits.

increase in nuclear system pressure, 15.i.45 Radiologica! Consequences 15.1.fLil Frequency Claulf1 cation (j

ideW A4

'3ile the consequence of this event does not Because no single failure could cause this l

result in fuel failure, it does result in the event, it ' categorized as a limiting fault. //s+w.%

discharge of normal coolant activity to the cv.'iten fv m asfe,. A [va p.A.4 A c.J Q,

surpression pool vit SRV r.peration. Because this 15.1.6J Sequence of Dents and Systems 7 ea %

activity is contained in the primary containment, Operation i

'"l y

k99*

there will be no caposures to operating personnel. Because this event does not result in 15.1.6 2.1 Sequence of Events an uncontrolled,elease to the environment, the plant operator can choose to leave the activity A shutdown cooling malfunction leading to a bottled up in the containment or discharge it to moderator temperature decrease could result from the environment under controlled release misoperation of the cooling water controls for cocciticus, if purging of the containment 15 RHR heat exchangers. T%c resulting temperature chosen, the release will be in accordance with decrease causes a slow insertion of positive the established technical specifications; reactivity into the core. Scram nceurs before therefore this event, at the worst, would only any thermal limits are reached if tbc operator s

result in a small increase in the yearly dces not take action. The sequence of events integrated caposure leve..

for this event is shown in Table 15.19.

15,1.5 SpNtrum of Steam System Piping 15.1.622 sptem operation Failures inside and Outside Containtnent in a PWR A shutdown cooling malfunction causing a moderator temperature decrease must be This event is not applicable to BWR plants. considered in all operating states. Howeser, U 1-8 Amendment t$

--.-.____.___._____m_

-rM Yh cLk e L ny oj.rja a,

+G..cs.,y n n Lp4 4 s r yns.

poi gn+ u n p.

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+G

r. /.Jup%

aus.

4 u.

mk a f.o.w 4 g9. y *c C /w 'F) a <. h a

Sm a n&ske&

L:&yle l.

ABWR nun,4n Sundard Plant niv c Table 15.1 1 SEQLTNCE OF EVEN~TS FOR EOSS OF i EEDWATER llEATING IISIE (tto nIAT(or 46 fc )

0 Initiate a 55.tPC, temperature reduction in the

~

feedwater splem 5

Initial effect of unhented feedwater starts to raise core pov tr level 1(0(est.)

Reactor variables settle into new steady state Table 15.12 LOSS OF !!.60C FEEDWATIR llEATING l

DOC' to EOC' Change in l

Core Power (9) 12.8 Change in MCPR 0.07 BOC = Beginning of Cycle 74,,4 Ju Hg &y EOC=

i 4,n,

,g 7 og DC $ EOC

$A w en te) 39 h

ck)L L o,o u u crit.

P Amendment 13 MIO

'I lABWR muun Standard Plant prv c Table 15.14 SEQUENCE OF EVENTS FOR FIGURE 15.12 IIA!E. hrs)

EVENTS 0

laitiate sinnulated runout of one feedwater pump (at system design pressure of 74.9 kg/cm*g the pump runout flow is 75% of rated feedwater Dow) l

- 0.1 Feedwater controller starts to reduce the feedwater Gow from the other feedwater pump 16.6 Vessel water level reaches its peak vahe and starts to return to its j

normal value

-60 (est.)

Vessel water level returns to its norrnal value.

Table 15.15 SEQUENCE OF EVENTS FOR FIGURE 15.13 TIME (seel EVENT

\\/

All

/

0 Initiate simulated runout of sep feedwater pumps (130% at system design pressure of 74.9 kg/cm'g on feedwater Dow) 1835 12 vesselicvel setpoint initiates trips rnain trebine and feedwater pumps.

1836 Reactor scram and trip of / Rips are actuated by stop valve position switches 18.5 Main turbine bypass v.; vet opened due to turb%e trip 20.1 SRVs open due to high pressure

> 25 SRVs close

> 40 (es,t.)

Water level dropgd to low water level setpoint (level 2)

> 70 (est.)

RCIC Dow into vessel (not simulated)

Amendment 15 13 1 I2 I

ABWR

uno, Standard Plant nrv c Table 15.16 SEQUENCE Of INEN~IS FOR IIGURE 15.1-4 H3EDtsj EVENTS 0

Simulate one bypass vahe to opeu

- 0.5 Pressure control system senses the decrease of reactor pressure and commands control vahes to close 5.0 Reactor settles at another steady state Table 15.17 SEQUENCE OF EVENTS FOR f IGURE 15.15 TIME (sec)

INENTS 0

Simulate all turbine contini valves and bypass vahes to open.

2.8 Turbine control valves wide open.

2.87 Vessel water level (12) trip initiates main turbine and feedwater own ne trips.

qwfs 2.9 Main turbine stop valves reach 85% open position and ini'iates reactor scram and trip of 4 RIPS.

2.97 Turbine stop valve closed.

17.2 Vessel water level reachu L2 setpoint. The remaining 6 RIPS are tripped. RCICis initiated.

36.2 cw turbine inlet pressure trip initiates main steamline isolation i

41.2 Main steam isolation valves closed. Bypass valves remain open, exhausting steam in steamlines downstream of isolation valves.

47.2 (est.)

RCIC flow enters vessel (not simulated).

i Amendment 13 1$ l 13 i

1

• l'

' MM n m man Standard Plant sum x i

i Table 15.18 SEQUENCE OF EVENTS FOR INADVERTENT SAFETY / RELIEF VALVE OPENING j

TIME Isec)

L1TEI O

Initiated opecirig of one SRV.

0.3 (est.)

Relief flow reaches full flow.

j 15 (est.)

System establishes new sieady. state operation.

7totest.)

s pan w p s I Wp :- % m.4e &sdi s,, - w ~ r.. e.. nr y.ss n n. w a.

11 C o (*.'* )

supuro,s yesI 4*, m L vuab. "Go.4, w A., %

1 i -4 4..,v 4 4 /.'C < A /.

Table 15.19 SEQUENCE OF EVENTS FOR INADVERTENT RHR SHUTDOWN COOLING OPERATION APPROXIMATE ELAPSED TIME EVENT 0

Reactor at states B or D (of Appendix 15A) when RHR shutdown cooling inadvertently activated.

0-10 min.

Slow rise in reactor power.

+ 10 min.

Operator may take action to limit power rise, flux scram will occur if no action is take.

l j '.

15.1 14 1

....,_.m...-_---.,

- - - -. _, - - - -, -. ~., _,,, -. _. - -,,

...,._.._,.-_..~n_

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-- - - i y

3 100

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s 73 __ 1 VESSLL PHES ntSE(PSn

/

[

,'\\

2 Siu titJE PHES nt3E (PSI) l i

+

O 1

3 ItNIBrJE PluS HISE (PSI) i

~y 1 7JEUTRON FLUX :

}

4 RE18EF VALVE FLOW (PCT) to 2 PEAK FUEL CENTER TEMP 5 BYPASSVALVE F1OW(PCT) e-

~>

g 50 3 AVE SURFACE HEAT FLUX I

25 --- G TLKtB STEAM FLOW (PCT)

I 4 FEEDWATER FLOW J

i

- ! 5 VESSEL STEAM FLOW

. f u e ty y 1@

f k

0 t-1*

-25' * '- -

O 5

to 15 20 0

5 to 15 20 TIME (sec)

TIME (sec)

I 1 LEVEL (INCH 41EF-SEP-SKIRT) 2 W R SENSED LEVEL (INCHES) t50 1

3 N R SENSED LEVEL (2JC3IES) 4 CORE INLET FLOW (PCT)

}.

5 PUMP FLOW 3 (PCT) k "8 S 9 5--

" 5 2

E d A 100 I

'M8 M

!N b

?

_t I

q O

1 1

i M

5 1 VOtD REACTMTY t

2 DOPP1LR REACTIVITY 50 m

3 SCHAM REACTIVITY t

V LL 4 TOTAL REACTIVITY 1

j LaaaIaaaa

+2'anaaIaam*

4 i

0 5

to 15 23 0

5 10 15 20 i

iIME (sec)

TIME (sec)

FUYL kWk St$w - Mg;g

" 2 0' "

y I.

Figura 15.1-3 K T ; C ".' T O ' " a c r m

^m m iajt,g, yg i

t t

1

}

l 1 NEUTRON f LUX 1 VESSEL PRES RISE (PSI)

N

[

~

2 PEAK FUEL CENifIl IEMP 2 STM t INE Piti S fltSE (PSI)

$C g

150

- 3 AVE SUHFACE lit Al I L UX 150

^

+-

3 TUltalNE I1tES FitSE (PSI) 3j 2

4 I EEDWATER FLOW

'Q 4 I4ELIEF VALVE FLOW (PCT)

DW G

5 VESSEL SIEAM i1.OW 5 BYPASS VALVE FLOW (PCT) cW O

I,

.i

, 6 TURU STE AM FLOW (PCT)

_5_

e h

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L h

_ (Q' G1 L N-o t

I 3

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t N

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\\

o 50 h

- 50 M

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g, O f

\\

1-71--- l\\

/

1 h= 2_ w

-300 **'-

O to 20 30 40 0

10 20 30 40 ilME (sec)

TIME (sec) 1 LEVEL (INCil REF-SEP SKIRT) 2 W H SENSED LEVEL (trJCHES)

\\

150 3 N H SENSED L EVEL (INCliES) 1 4 COHE INLET FLOW (PCT)

/

A

- I'[

5 PUMP FLOW 1 (PCI)

/

i

/

@[/

E E

o 100 s

N-g q

5 i

i~

1 VOID REACTIVITY l

M E8

-I 2 DOPf4 Ell REACilVITY g

3 SCilAM ilEACTIVIIL g

in 5

44 4 TOTAlilEACTIVITY s-V g

o

• • 41211 3

J t

-?

s u. La i_a4

?

O 10 20 30 40 0

to 20 30 40 TIME (sec)

TIME (sec)

}

k

}-

wi m m l

~

Figure 15.1-5 OPENING OF ALL CONTROQBYPASS VALVES

ABWR moon Standard Plant arv A 3

t; -

SECTION 15.2 CONTENTS (Continued)

I Sedlon

'Bl}g Egge

r. t. I.r.1 fan n %I6 De au e es fs.L.a Radiolog& ctmical Consequences sj c u L /tu % I v 4 2 15.2.1.5 15 2 Wa4 jr.s.t.r. f 15.2.2 Generator Land Relection 15.22 15.2.2.1 Identification of Causes and Frequency Clasufication 15.23 15.2.2.1.1-Identifin. tion of Causes 15.2 3 15.2.2.1.2 Frequency Classification 15.2 3 1512.1.2.1 Genetator lead Rejection 13.23 15.2.2.1.2.2 Generator Load Rejection with Failure of One Bypass Valve 15.2 3 15.2.2.1.2.3 Generator lead Rejection with Failure of All -

Bypass Valves 15.2 3.

15.2.2.2 Sequence of Events and Systerns Operations 15.2 3 15.2.2.2.1 Sequence of Events 15.23 15.2.2.2.1.1 Generator Load I.yection-Turbine Control Valve Fast Closure 15.2 3 15.2.2.2.1.2 Generator Load Rejection with Failure of One Bypass Valve 15.2 3 15.2.2.2.1.3

- Generator Imad Rejection with Failure of All Bypa$$ Valves 15.2 3 15.2.2.2.1.4

! der..*fication of Operator Actions 15.23 15.2.2.2.2 Systems Operation 1124 15.2.2.2.2.1.

Generator Load Rejection with Bypass 15.2-4 15.2.2.2.2.2 Generator Load Rejection with Failure of One Bypasa Valve 15.2 !

l>

l 15.2-iii V.

1 1

ABWR memo

.4 Standard Plant piw 4 SECTION 15.2 CONTENTS (Continued)

SffilQ.11 lillt EAgt l

15.2.7.2.2 Sptems Operation 15.2 16 13.2.7.3 Core and Sptem Performance 15.2 16 15 2.7.4 Barrier Performance 15.2 16 15.2.7.5 Radiological Consequences 15.2 16 i

15.2.8.

Feedwater 1Jae BM 15.2 16

- 15 2.9 Eallutt.nLRiiRikvidemLCARllag 15.2 16 15J.10 Refertasta 15.2 17 TABLES Table 11 tic East 1

15.2 1s Sequence of Events for Figure 15.21%

15.2 18 ITvlb h

u 4 19 4 + % It.s -!6-15.2 2 Se uence of Events for Figure 15.2 2 15.2 18-15.2 3 Sequence of Events for Figure 15.2 3 15.2 19 15.24 Sequence of Ew:nts for Figure 15.2 4 15.2 19 13.2 5 Sequence of Events for Figure 15.2 5 15.2 20 15.2 4 Sequence of Events for Figure 15.24 15.2 20 15.27 Sequence of Events for Figure 15.2 7 15.2 21 15.2-8

.- Sequence of Events for Figure 15.2-8 15.2 21 15.2-9 Sequence of Events for Figure 154 9 15.2 22 15.2 10 Post Transient Release Rate to the Containment with Suppression Pool Cleanup 15.2 23

-15.2 11 Activity Released to the Environment 15.2 24 w

15.2x m.

_ _ _ _ _ _ =... - _.. _ _... _., _ _ _ _ - _ _.. ~..

.2.._,..

. ABWR mwmn Standard Plant nrv c

+:

SECTION 15.2 TAllLES (Continued)

Inhlt IMt Part 152 12 Estimated Doses and Atmospheric Dispersion Factors 15.2 24 15.2 13 Typical Rates of Decay for Condenser Vacuum 15.2 25 15.2 14 Sequence of Events for Figure 15.210 15.2 25 15.2 15 Trip Signals Auociated with the lou of Condenser Vacuum 15.2 26 15.2 It; Sequence of Event: for Figure 15.211 15.2 27 152 17 Sequence of Events for Figure 15.212 15.2 28 i

ILLUSTRATIONS Figure 11tle Eage 15.2 li.

F9t Closure of One Turbine Control valve 15.2 30 t r.s -I k cle a etsj u sf su n s,u kl.,( yk 15.2 2 Preuure Regulator Downscale Failure 15.2 31 152-3 Generator Load Rejestion with Byy.ss 15.2 32 15.2-4 lead Rejection with One Bypau Valve Failure 15.2 33 15.25 load Rejection with All B)pau Failure 15.2 34 l 15.2 6 Turbine Trip with Bypass 15.2 35 15.2 7 Turbine Trip with One Eypau Valve Failure 15.2 36 15.2 8 Turbine Trip with All Bypass Failure 15.2 37 15.2 9 MSiv Closure Direct scram 15.2 38 15.2 10 lou of Condenser Vacuum 15.2 39 15.2 11 Lou of AC Power 15.2-4u 15.2 ti Amendment 15

i

'ABWR u-

.(

SIAndard Plant ruv c 15.2 INCREASE IN REACTOR PRESSURE control processors, called

• pressure regulator downscale f ailure.' However, the probability of 15.2.1 Pressure Regulator Failure Closed this event to occur is extremely low (less than 7s10** failure per reactor year), and hence 15.2.1.1 Identification of Causes and frequency the event is considered as a limiting fault.

Classification 15.2.1.1.2 Frtquency Classification l

15.2.1.1.1 Identification of Causes 15J.l.l.2.1 Inadvertent Closure of One Turbine The ABWR steam bypass and pressure control Control Yahe system (SB&PCS) uses a triplicated digital control system, instead of an analog system as This event is conservatively treated as a used in BWR/2 through BWR/6. The SB&PCS controls moderate frequency event, although the turbine control valves and turbine bypass valves voter / actuator fallere rate is very low (0.0088 to raaintain reactos pressure. As presented in failure per reactor year).

Subsection 15.1.2.1.1, no credible single failure in the control system will result in a minimum 15.2.1.1.2.2 Pressurt Regulator Downscale demand to all turbine control valves and bypass Failure l valves. A voter or actuator failure may result in an inadvertent closure of one turbine control The probability of occurence of this event is valve or one turbine bypass valvt if it is open calculated to be less than 7x10 5 per year as at the time of failute. In this case, the SB&PCS shown in Appendix 15D. This event is treated as will sense the pressure change and command the a limiting fault.

remaining contro' valves or bypass valves, if t

needed, to open, and thereby automatically 15.2.1.2 Sequence of Events and System mitigate the transient and try to maintain Operstlori reactor power and pressure.

15 2.1.2.1 Inadsertent Closure of One Turbine Because turbine bypass valves are normally Control Valve 7

closed during normal full power operation, it is assumed for purposes of this transient analysis Postulating a cluator f ailure of the that a single failure causes a single turbine SB&PCS as presented la Subsection 15.2.1.1.1 control valve to fail closed. Should this event will c.use one turbine control valve to close.

occur at full ;ower, the opening of remaining The pressure will increase, because the renctor control valves may not be sufficient to maintain is still generating the initial steam flow. The the reactor pressure, depending on the turbine SB&PCS will open the remaining control valves l

design. Neutron flux will increase due to void and some bypass valves. This sequence of esents l

collapse resulting from the pressure increase. A is listed in Table 15.24for Figure 15.2.la der a. Prf reactor scram will be initiated when the high clo w W,'n n a,f4 rr,i -la f.s N w #r.r 4 flux scram setpoint is exceeded.

15.1.1J.1.2 Pressure Regulator Downscale

,4 l

No single failure will cause the SB&PCS to issue erroneously a minimum demand to all turbine Table 15.2 2 lists the sequence of events l

control valves and bypass valves. However, as for Figure 15.2 2.

discussed in Subsection 15.1.2.1.1, multiple failures might cause the SB&PCS to fail and 15.2.1.2.lJ Identification of Operator erroneously issue a minimum demand. Should this Actions occur, it would cause full closure of turbine l

controls valvet well as an inhibit of steam The operator should:

bypass flow and wereby increase reactor power and pressure. When this occurs, reactor scram (1) monitor that all rods are in; will be initiated when the high reactor flux scram setpoint is reached. This event is (2) monitor reactor water level and pressure; analyzed here as the simultaneous failure of two Ame ndme s. 13 1521 l

l

' AB M uxmma Standard Plant uv c p,y r.4 6-va ve es pesented in Figure 15.2.lA.The f

turbine auxiliaries);

analysis assumes that about 85% of rated steam flow can pass through the remaining three (4) observe that the reactor pressure relief turbine control valves, valves open at their setpoint; Neutron flux increases rapidly because of the (5) monitor reactor water level and continue void reduction caused by the pressure increase.

cooldown per the normal procedure; and When the sensed neutron flux reaches the high neutron Hus scruru setpoint, a reactor scram is (6) complete the scram report and initiate a initiated. The neutron flux increase is limited /0A6 maintenance survey of pressure regulator [To4M % NBR by the reactor sc'am, Peak fuel _g before reactor testart, g surface heat uux does not exceed FM% of its initial value. MCPR for this transient is still 15.2.1.2.2 Sptems Operation above the safety MCPR limitj Therefore, thel design basis is satisfico, c"'-"!::-gogg c--

15.2.1.2.2.1 Inadvertent Closure of One Turbine

"^

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h Control Valve 15.2.1.L2 Pressure Regulator Downicale Normal plant instrumentation and control sre Failure assumed to function. This event takes credit for high neutron flux scram to shut down the reactor.

A pressure regulator downscale failure is d/*yg simulated at 102% NBR power as shown in Figure Niter a closure of one turbine cor trol valve, 15.2 2.

the b flow rate that can be transmitted through the remaining three turbine control Ntr'e. Hux increases rapidly because of the valves depends upon the turbine configuration. void reouction caused by the pressure increase.

For plants with full arc turbine admission, the When the sensed neutron flux reaches the high steam flow through the remaining three turbine neutron flux scram setpoint, a reactor scram is control valves is at least 95% of rated steam initiated. The neutron flux increase is limited flow. On the other hand, this capacity drops to to 155% NBR by the reactor scram, Peak fud about 85% of rated steam flow for plants with surface heat flux does not exceed 103% of its

_ partial arc turbine admission. Therefore, this initial value. It i; estimated less than 0.2%

transient is less severe for plants with full arc of rods will get.into transition boiling, o

ne adraission.

'Ij cr In this analyg,1M cases Therefore, the design limit for the limiting g

witb partial. arc turbine admission k analyzed to fault event is met.

g cover all pleew pe AM efMA7 m/A '.

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r g 15J.l.4 Barrier Performance 15.2.1.2.2.2 Pressure ulator Downscale Failure 15.2.1.4.1 Inadvertent closure of One Turbine Control Valve Analysis of this event assumes normal

-functioning of plant instrumentation and Peal pressure at the SR valves reaches 74.5 controls, and plant protection and reactor pro.

kg/cm)g. The peak fessel bottom pressure tection systems. Specifically, this event takes reaches 78.2 kg/cm g, below the transient cre:'!! for high neutren flux scram to shut down pressure limit of 96.7 kg/cm*g.

the reactor. High system pressure is limited by the pressure relief valve system operation.

15.2.1.4.2 Prr.:.sure Regulator Downscale Failure 15.2.13 Cne and System Perfe.rmance Pealj pressure at the SRVs reaches 85.1 kg/cm g.

Th e pe a k ""'I 'I"* " P""

2 15J.lJ.! Inadvertent Clos re of One Turbine reaches 87.4 kg/cm g at the bottom of the Control Valve vessel below the nuclear barrier pressure f*f,,I,.s limiI.

A simulated c osure of one turbine control Amendment 15 1322

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-Subsection 15.2.4.5 cover abe consequences of 15.2.2.1.2.3 Generstor teed Rejection with this event.

Failure of All Bypass Valtes 15.2.2 Generator lead Rejection Frequency: <3.6x10 / plant year f

4 15.2.2.1 Identification of Causes and frequency Frequency Basis: Thorough scarch of domestic Classification plant operating records have revealed three instances of bypass failure during 628 bypass 15.2.2.1.1 Identification of Causes system operatiens This gives a probability of bypass failure of 0.0048. Combining the actual Fast closure of the turbine control valves frequency of a generatnr load 'cjection with the (TCV) is initiated whenever electrical grid failure rate of bypass yields.1 frequency of a i

dis'vrbances occur which result in significant generator load rejection with oypass failure of loss of electrical load on the generator. The 0.0036 event / plant year. With the triplicated turbine control valves are required to close as fault tolerant design used in ABWR, this failure rapidly as possible to prevent excessive frequency is lowered by at least a factor of

. overspeed of the turbine. generator (T.G) rotor.

100. Therefore, this event should be classified

= Closure of the main turbine control valves will as a liroiting fault, however, criteria for cause a sudden reduction in steam flow which moderate frequent incidents are conservatively results in an increase in system pressure and applied.

reactor shutdown.

15.2.2.2 Sequence of Events and System

- After sensing a significant loss of electrical Operation load on the generator, the tuibine control valves are commanded to close rapidly._ At the same 15.2.2.2.1 Sequence of Esents time, the turbine bypass valves are signaled to open in the ' fast

• opening mode by the Steam 15.2.2.2.1.1 Generator land Rejectiou-Turbine Bypass and Pressure Control System (SB&PCS), Control Valve Fast closure which uses a triplicated digital controller. As presented in Subsection !$ 1.2.1.1, no single A loss of generator electrical lead from high failure can cause all turbine bypass valves fall power conditions produces the sequence of events to open on demand. The worst single failure can listed in Table 15.2 3.

-only cause one turbine bypass vaive fail to open on demand. Therefore, the probability of this to 15.2.2.2.1.2 Generator lead Rejection with occur is very low (less than one failure every 11 Failure of 0ne Bypass Valse year). Therefore, generitor load rejection with failure of one turbine bypass valve is considered A loss of generator _ electrical load from an infrequent event; while generator load high power conditions with failure of one bypass rejection with failure of all turbine bypass valve produces the sequence of events listed in valves is a limiting fault '

Table 15.2 4.

15.2.2.1.2 Frequency Classincatiou 15.2.2.2.1J Generator land Rejection with Failurt of All Bypass Valves 15.2.2.1.2.1 Generator Imad Rejection A loss'of generator electrical load at high This event is categorized as en incident of power with failure of all bypass valves produces l moderate frequency, the sequence of events listed in Table 15.2 5.

15.2.2.1.2.2 Generator Load Rejection with 15.2.2.2.1.4 Identlacation of Operator Failure of One Basss Valve Actions

- This event should be categorized as an The operator should:

infrequent event.- However,' criteria for moderate frequent incidents are conservatively applied.

(1) verify proper bypass valve performance; l

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Simulate one main turbine control valve to,close l

0 Failed turbine control valve starts to close

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Neutron flux reaches high flux scram setpoint and initiates a reactor scram Fd Turbine bypass valves start to open.

.ef fI Water level reaches level 3 setpoint. Four RIPS are tripped.

Table 15.2 2 SEQUENCE OF EVENTS FOR FIGURE 15.2 2 TlME.hed 10tTAT 0

Simulate rero steam flow demand to main turbine and bypass valves.

O Turbine control valves start to close.

1.0 Neutron flux reaches high flux scam setpo;nt anu initiaies a reactor scram.

2.4 Four RIPS are tripped due to high dome pressure.

2.6 Safety / relief vMyes open due to high pressure.

8.9 Safety / relief vahes close.

9.4 Group 1 safety / relief valves open again to relieve decay heat 9.8 Group 2 safety / relief valves open again to relieve decay heat.

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SECTION 15.3 CONTENTS (Continued)

Secthn Htle Ease 153.13 11 Trip of Three Reactor laternal Pumps 15 3-2 153.13.2.2 Trip of All Reactor Internal Pumps 153 2 j

15 3.1.4 Barrier Performance 153 3 15 3.1.4.1 Trip of Three Reactor Internal Pumps 153 3

. 15 3.1.4.2 Trip of All Reactor laternal Pumps 153 3 15 3.1.5 -

Radjetogical Consequences 15 3-3 I r.1.s. S*. l w ;y a + r w as h % 4s.s J fa,,. y 15 3.2 Recirculation flow Control Failure-Dggranian Flow 153 3

'15 3.2.1 Identification of Causes and Frequency Classification 153 3 15311.1 Identification of Causes 153 3 i

15 3.2.1.2 Frcquency Classification 15 3-3 15311.2.1 Fast Runback of One Reactor Internal Pumps 15 3-3 15 3.2.1.2.2 Fast Runback of All Reactor Internal Pumps 13 3-3 15 3.2.2 Sequence of Events and Systems Operations 15 3-3 153.211 Sequence of Events 15 3-3 l.

15 3.2.2.1.1 Fast Run.back of One Reactor laternal Pumps 15 3-3 15 3.2.2.1.2 Fast Aunback ci all Reactor Internal Pumps 1534 153.2.2.13 Identification of Operator Actions 15 3-4

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153 12.13.1 Fast Closure of One Main Recirculation Valve 15 3-4

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.153.2.2.13.2 Fast Closure of Two Nu.: Recirculation Valves 15 3-4 15312.2 Systems Operation 15 3-4 i-l I $.l.f. D Y'j%f k

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SECTION 15.3 TABLES Iahlt.

Illic hat 153 1 Sequence of Events for Figure 1531 153-8 i

153 2

~ Sequence of Events for I1gure 153 2 153 8 I

153 3-Sequence of Events for Figure 153 3 153 9 i

15 3-4 Sequet.cc of Events for Figure 1534 153 9 e

153 5

- Sequence of Event for Figure 153 5 15 3-9' ILLUSTR.ATIONS Bgun Illit hat 153 1 Three Pump Trip 153 10 153 2 All Pump Trip 153 11 ItJ~sA Cishly f%ps.,& nlweig hil (% f rip 7*

153 3 Fast Runback of One RIP 15 3-12 1534 Fast Runback of All RIPS 153 13 153 5

. One RIP Seizure 153 14 l

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i ABWR uuimu S111ndard Pitint Riv q 15J.. 2 Sequence of Esents and Systems system or safeguard system operation. This analysis assumes normal functioning of plant l Optration instrumentation and controls.

!!.3.1.2.1 Sequence of Esents 15.3.1.2.2.2 Trip of All Reactor laternal 15J.l.2.1.1 Trip of1hree Reactor laternal Pumps Pumps Analysis of this event assumes normal func.

Table 15.31 lists the sequence of events for Joning of olent instrumentation and controls, Figure 15.31.

sud pla.

ction and reactor protection systems.

15 3.1.2.1.2 Trip of All Reactor luternal Pumps if a trip of all RIPS is caused by multiple Table 15.3 2 lists the sequence of events for failures in an electrical power supply to the Figure 15.3 2.

RIPS, a reactor scram will be initiated at time 0 due to load rejection or turbine trip at time 15.3.1.2.1.3 Identification of 0perator Actions 0.

For other cauvs a reactor scram will be initiated upon the condition olVrapid core (Ew?

15.3.1.2.1.3.1 Trip of Three Rractor Internal ioastdown. liigh system pressure is limited by Pumps the pressure relief valve system operation.

Because no scram occurs for trip of three h'%#'bl'd N

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RIPS, no immediate operator action is required.

Qu q g,,

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hQ (d) that no operating limits are being exceeded. Tbc

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orcrator should also determine the cause of 15.3.1J Core and Systern Performance failure prior to returning the systera to normal operation.

15.3.1.3.1 Input Parameters and initial Condillons 153.1.2.lJ.2 Trip of All Reactor Internal Pump motors and pump rotors are simulated Pumps with minimum specified rotating inertlas. The nuclear conditions for the beginning of life The operator should ascertain that the reactor (BOC) are used to provide conservatise bounding I scram is initiated. If the main turbine and analysis.

] feedwater pumps are tripped resulting from reactor water level swell, the operator should 15.3.1.3.2 Results regain control of reactor water level through RCIC operation, monitoring reactor water lesel 15 3.1.3.2.1 Trip of nrre Reactor laternal and pressure after shutdown. When both reactor Pumpa pressure and level are under control, the operator should secure RCIC as necessary. The Figure 15.31 shows the results of losing operator should also determine the cause of the three RIPS. MCPR remains above the safety trip prior to returning the system to normal limit; thus, the fuel thermal limits are not operation.

violattd. During this transient, level swell is not sufficient to cause turbine trip and scram.

!!J.1.2.2 Systems Operation Therefore, this event does not have to be reanalyzed for specific core configurations.

15 3.1.2.2.1 Trip of There Reactor laternal Pumps 15 3.1.3.2.2 Trip of All Reactor Internal INmps Tripping of three RIPS requires no protection Amendment 15 15 M

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tN N ZW1%All Standard Plant nu Figure 15.3 2 graphically shows this event all ten reactor internal pumps (RIPS) at the with the minimum specified rotating inertia for same speed. As presented in Subsection the RIPS. The vessel water lesel swell due to 15.1.2.1.1, no credible single f ailure in the rapid flow coastdown is expected to reach the control system will result in a rninimum demand high level trip, thereby tripping the main to all klPs. A voter or actuator f ailure may turbine and feed pumps. Subsequent events, such result in an inadvertent runback of one RIP at as initiation of the RCIC system occurring late its maximum drive speed (-40G/sec.). In this in this event, have no significant effect en the case, the RFCS will sense the core flow change results. The peak tlad tem-rature during this and command the remaining RIPS to increase event is calculated to F

.s than 6000C, speeds and thereby automatically mitigate the which is below the applicat imit of 12000C.

transient and maintain the core flow.

15.3.1.4 Barrier Performanc(ew C-.

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r As presented in Subsection 15.1.2.1.1, multiple f ailures in the control system might 153.1.4.1 Trip of urce Rractor Internal Pumps cause the RFCS to erroneously issue a minimum demand to all RIPS. Should this occur, all RIPS The results shown in Figure 15.31 indicate could : educe speed simultaneously. Each RIP that peak pressures stay well below the 96.7 drive has a speed limiter which limits the 2

Kg/cm g limit allowco h the applicable code.

maximur, speed change rate to 5%/sec. However, Therefore, the barrier pressurc boundary is not the probability of this esent occurririg is low threatened.

(le s t, t h a n 7 x 10 5 f ailures per reactor year); and hence, the event should be considered 15.3.1.4.2 Trip of All Reactor internal Purnps as a limiting f ault. However, criteria for moderate frequent incidents are conservatively The results shown in Figure 15.3.2 indicate applied.

that peak pressures stay well below the limit allowed by the applicable code. Therefore, the 15.3.2.1.2 Frequency Classification barrier pressure boundary is not threatened.

15 3.2.1.2.1 Fast Runback of One Reactor 15.3.1.5 Radigtpgical Consequences internal Pump

% r4t -+ F Toy J.a IU7aiercai pumps oue to a lost, M The failure rate of a voter or an actuator is power supply is considered extremely unlikely to about 0.0088 f ailures per reactor year.

result in perforation of fuel under nditions o However, it is analyzed as an incident of boiling i sition. The ase of fission moderate frequency.

products woul ver, r uch less than the assumed in i oss Coolant Accident for an 153.2.1.22 Fast Runback oi All Reactor event qual probabi.

Therefore, th Internal Pumps o

radio ogical exposures noted in section 15.6.i This event should be classified as a limiting

_com the consequences of this event.

fault event. However, criteria for moderate 15.3.2 RecirculhGon Flow Control frequent incidents are conservatively applied.

Failure Decreasing Flow 153.2.2 Sequence of E5ents and Systems 15.3.2.1 Identification of Causes and Operation Frequency Classification 15.3.2.2.1 Sequence of Esents 15.3.2.1.1 Identification of Causes 15.3.2.2.1.1 Fast Runback of One Reactor The recirculation flow control system (RFCS)

Intnnal Pump uses a triplicated, f ault tolerant digital control system, instead of an analog system as Table 15.3 3 lists the sequence of events for used in BWR 2 through BWR 6. The RFCS controls Figure 15.3 3.

Amendment 17 1513

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....ABM _ l , Standard Plant _ usacorn g wv c t A 15.8 ANTICIPATED TRANSIENTS WITH-eliminating the scram discharge volume OUT SCRAM (mechanical common mode potential failure) and by having an electric motor run in diverse from 15.8.1 Requirements the hydraulic scram feature. SRP 15.8 requires a automatic recirculation This latter feature allows rod run in if pump trip (RPT) and emergency procedures for scram air header pressure is not exhausted ATWS, This SRP has been somewhat superseded by because of a postulated common mode electrical the issuance of 10CFR50.62, which requires the failure and simutaneous fe!!ure of the ARI BWR to have automatic RPT, an alternate rod system, and therefore satisfies the intent insertion (ARI) system and an automatic standby required by 10CFR50.62. Thus, the design does liquid contro system (SLCS) with a minimum flow not need an SLCS to respond to an ATWS capacity and boron content equivalent to 86 gpm threatening event. of 13 weight percent rodium pentaborate solution, The SLCS is required by 10CFR50 Appendix A 15.8.2 - Plant Capabilities Criterion and is described in Section 9. Becaase the new drive design climinates the For, ATWS prevention /mitigatior for ABWR, the previous common mode failure potential and followmg are provided: because of the very low probability of simultaneous modem 'iilure of a large number of- ' a. An ARI system that utilizes sensors and drives, a fanure to ;chieve shutdown is deemed logic which are diverse and independent incredible, h/W4A. e.a. fem *N,id* A " of the reactor protection system, 'E. .1. . _ :. _ ; - !!_! ' i b, Electricalinsertion of FMCRDs that also e C.- . Z. Supporting analysis is utilize sensors and logic which are documented in Appendir 15E. diverse and independent of the reactor protection nstem,

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,w(,,/ > capacity 4 W//N .(t r.aJs'r 4. 04 * ~ A ru s. D A k. %,f % & The ABWR has the ATWS-RPT feature which prevents reactor vessel oserpressure and possible 2 short term fuel damage for the most limiting Spi;p 4 f ocp/ ft h, postulated ATWS events. The design details of this system are given in Section 7.7. Emergency procedures for ATWS are described in Chapter 18. Thus, the SRP 15.8 is satisfied. The ATWS rul-e of 10CFR50.62 was written as hardware specific, rather than functionally, because it clearly reflected the BWR use of locking-piston control rod drives. The ABWR however, uses a fine motion control rod drive (FMCRD) design with both hydraulic and electric means to achieve shutdown. This drive design is described in detail in Section 4.6. The use of this design eliminates the common mode failure potentials of the existing locking piston LRD by Amendment 15 15.8-1 1}}