ML20092F460

From kanterella
Jump to navigation Jump to search
Forwards Nonproprietary Portion of GE Response to Agenda Items Discussed During 911120-21 Meeting W/Nrc Reactor Sys Branch Re Standby Liquid Control Sys Instrumentation & Controls
ML20092F460
Person / Time
Site: 05000605
Issue date: 02/14/1992
From: Marriott P
GENERAL ELECTRIC CO.
To: Pierson R
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
References
EEN-9224, MFN-038-92, MFN-38-92, NUDOCS 9202190343
Download: ML20092F460 (28)
v • d • e


Text

_ _

('

' f_

GE Nudidat faergy '

.x-([

OP:p?w [.

.5

, n

-usem n u m cawe February 14,' 1992 '

_ MFN No.038 92 Docket No. STN 50-605 -

.i EEN 9224

' Document Control Desk i

-U.S. Nuclear Regulatory Commission t

? Washington, D.C. 20555 -

Attention:'

Robert C. Pierson, Director

Standardization and Non Power Reactor Project Directorate t

Subject:

. GE Response to Agenda items Discussed During the GE/NRCE

' Reactor Systems Hranch Meeting on November 20-21,1991

Reference:

. OE Response to' Agenda Items Discussed During the GE/NRC

~,

Reactor Systems Branch Meeting on November 20 21, 1991,.

- Proprietary Information, Dated February 14,1991, MFN No. 039 92 Enclosed are thirty-four (34) copies of the GE response to the subject discussion items, numbered :

3,6,7 and 17(

! A portion 'of Item 3 contains proprietary information and is submitted under separate cover (see reference)..

-li is intended that GE will amend the SSAR, where appropriate, with these responses in a future amendment,~

' Sincerrly.

P.W. M iott, Manager L

Regulatory and Analysis Services l}

M/C 382,;(408) 925-6948 -

' cc: : F. A. Ross

. (DOE) -

- N. D, Fletcher; (DOE)

C. Poslusny, Jr. :

. (NRC)

' G. Thomas.

' (NRC)

- R. C. Berglund (GE)-

i J. F. Quirk '

(GE) -

JN E

i 9202190343 920214 g

PDR-ADOCK 05000605 A.

PDR l'

r-RESPONSE TO AGENDA ITEM 3 The automatic start text changes are provided on pages 7.4.-1, 7.4-2, 7.4-3, 9.3-2 and 9.3-3 (attached).

The SLCS P&ID has been updated to reflect the automatic start.

This P&ID, Figure 9.3-1, was transmitted under letter, R. C.

Mitchell to Robert C.

Pierson," Updated ABWR Piping and Instrumentation and Process Flow Drawings,' February 3, 1992, MFN No. 030-92.

The SLCS IBD has also been updated to reflect the automatic start.

This is GE Proprietary Information and is provided under separate. cover.

+mv v

ABWR nuiom

. - garidard Plant nry n

= 7.4 SYSTEMS REQUIRED FOR SAFE 7.4.1.2 standby uquid Control system.

~ SHUTDOWN Instrumentation and Controls 7.4.1 Description (1) Function or oMC C

This section examines and discusses the in-The instrumentation and ontrols for the strumentation and control aspects of the fol.

SLCS are designed to initicte an_d continue lowing plant systems and functions designed to injection of a liquid neut/on absorber into assure safe and orderly shutdown of the ABWR:

the reactos when manuallygalled upon to do so. This equipment also provides the neces.

(1) Ahernate rod insertion function (ARI) sary controls to maintain this liquid chemi.

cal solution well above saturation tempera.

(2) ' Standby liquid control system (SLCS) ture in readiness for injection.lThe system PalD is shown in Figure 9.34. The inter.

(3) Reactor shutdown cooling mode (RHR) lock block diagram (IBD) is shown in Figure 7.4 1.

(4) Remote shutdown system (RSS)

(2) Classification

" 4'"

gg)

See Subsection 7.1.2,4 which addresses the de-sign basis information required by Section 3 of The SLCS is a backup method to shut

- lEEE 279.

down the reactor to cold subcritical conditions by independent means other than

- 7.4.1.1 Alteriaate Rod Insettlon Function.

- the normal method by the control rod.

instrumentation and Controls system. Thus, the system is considered a safe shutdown system. The standby liquid I

The alterna'te rod insertion (ARI) function is control process equipment, instrumentation, accomplished independently and diversely from the and controls essential for injection of the reactor protection system (RPS), Independent neutron absorber solution into the reactor sensors (i.e., ECCS sensors) provide reactor trip are designed to withstand Seismic Category I signals, via the _ recirculation flow control earthquake loads. Any nondirect process system (RFCS), both to ARI valves (part of the equipment, instrumentation, and controls of C

control rod drive system) and to the rod control the system are not required to meet Seismic J

and information system (RC&lS). The ARI valves, Category I requirements; however, the local (seperate from the scram valves), cause reactor and control room mounted equipment is shut down'by hydraulic scram of the control located in seismically qualified panels.

rods._ The RC&lS, acting upon the same ARI signals _that are provided to ARI _ valves, cuases (3) Power Sources.

reactor shut down by electromechanical (i.e.,

through the usage of FMCRD motors) insertion of The power supply to one motor-operated injec.

controt rods, tion valve, storage tank discharge valve, and injection pump is powered from Division The RCalS, including the active run.in I,480VAC. The power' supply to the other function of the FMCRD motors and the ARI valves motor operated injection valve, storage tank are not required for safety, nor are these outlet valve, and injection pump is powered components qualified in accordance with safety from Division II,480VAC. The power supply criteria. However, the FMCRD components to the tank heaters and heater controls is associated with hydraulic scram are qualified in connectable to a standby AC power source, accordance with safety criteria.

The standby power source is Class 1E from an onsite source and is independent of the The inherent diversity of ARI provides offsite power. The power supply to the main mitigation of the consequences of ATWS controt room benchboard indicator lights and (anticipated transient without scram) events.

the level and pressure sensors is powered from a Class 1E instrument bus.

Amendment 10 74-1 t

emWR -

usswcar Standard Plant nrv 4

, function of the SLCS. It is included for a the inboard valve from Division Il logic.

number of special consideration events:

(6) Redundancy and Diversity (a) Plant capability to shut down the reactor without control rods from normal Under special shutdown conditions, the SLCS operation (Chapter 15).

is functionally redundant to the control rod drive system in achieving and maintain-(b) Plant capability to shut down the ing the reactor subcritical. Therefore, reactor without control rods from a the SLCS as a system by itself is not re-transient incident (Chapter 15).

quired to be redundant, although the actise componer s and control channels are redun.

Although this system has been designed to a dant ior serviceability.

high degree of reliability with many safr.y system features, it is not required to.neet The SLCS provides a diverse means for shut-the safety design basis requirement. of the ting down the reactor using a liquid neu-safety related systems.

tron absorber in the event of a control rod drive system failure.

(5) Initiating Circuits

' g,4. -

,.u oll The method ofidentifying redundant power The standby liquid controljis[nitiated in) cables, signal cables, and cable trays and the main control room by turning a keylock-the method of identifying non safety relat-ing switch for system A or a different key.

ed cables as associated circuits are locking switch for system B to the RUN posi-discussed in Subsection 8.3.1.3.

tion.

(9) Actuated Devices (6) Logic and Sequencing A

When the SLCS is initiated to inject When one division of SLCS is initiated, one liquid neutron absorber into the reactor, injection valve and one tank discharge valve the following devices are actuated:

start to open immediately. The pump that has been selected for injection will not (a) one of the two injection valves is start until its associated tank discharge opened; valve is at the fully open position. In order to provide maximum MOV availability (b) one of the two storage tank discharge when the SLCS is in normal standby read-valves is opened; iness, the overloads for the storage tank outlet valves are bypassed by a contact from (c) one of the two injection pumps is a test switch in its NORMAL position. When started; and the TEST position is selected, the overload short is removed, thus allowing motor (d) one of the reactor water cleanup protection during test operation of the isolation valves is closed.

valves.

Additionally, the pressure and tank level (7) Bypasses and Interlocks sensing ' quipment indicates that the SLCS e

is pumping liquid into the reactor.

Pumps are interlocked so that either the storage tank discharge valve or the test (10) Separation tank discharge valve must be fully open for the pump to run. When the SLCS is initiated The SLCS is separated both physically and to inject the neutron absorber into the electrically from the control rod drive reactor, the outboard isolation valve of the system. The SLCS ciectrical contral reactor water cleanup system is automatic-channels are separated in accordance wn.

ally closed from the Division I logic and the requirements of Subsection 8.3.1 A.

d codrol s3 54em

% e. shedh h }h Ve d g u s m q is

^mendment 2

@ % 4 u.M m

on ci pa act h %: 4 m 46,3

~

cm s t r om ( ATWS) n apo l.

q

l N S 6 kT 9

When the SLCS is automatically initiated to inject a liquid neutron absorber into the reactor, the following devices are actuated:-

(a).the two injection valves are opened; (b) the two storage tank'ddscharge valves are opened; (c) the two injection pumps are started; and (d) the reactor water cleanup 1 solation valves are closed,

sm,

'N ' - LAB M mooxr

= Standard Plant arva 7

(11) Testability (2) = Status Lights

- The SLCS is capable of being tested by man-(i)

Pump or storage tank outlet ual initiation of actuated devices during _

valve overload trip or power

-normal operation.- In the test mode, demin-loss; eralized water is circulated in the SLCS

loops rather than sodium pentaborate, Dur-(ii)

Position ofinjection line ing reactor shutdown, demineralized water manual senice valve;

-: may be injected into the reactor vessel for -

,the injection test mode, i(ili) Position of storage tank outlet valve and in-test

_ (12) Environmental Considerations ~

status; The environmental considerations for the in."

(iv)' Position of test tank strument-and control portions of the SLCS discharge manual senice are the same an'for the' active mechanical valve;

'o components of the system (Section 3.11).

The instrument and control portions of the.

(v)

SLCS manually out of senice;

'SLCS are seismically' qualified not to fait during, and to remain functional following, (vi) Pump auto trip.

a safe shutdown earthquake (SSE) (see _

Section 3.10 for seismic qualification hakj3) Annunciators bau4omabceB

-aspects).

mi"^#

f**""

The SLCS annunciators indicate:

37* b '" *

(13) OperationalConsideratioEs (i)

Manual or automatic out-j The control scheme for t le SLCS can be found of senice condition of SLCS in the' interlock bio k diagram (Figure A and/or B due to:

7.41)'. The SLCS is anually initiated in the' control room _byJnserting the key in the operation of manual out of.

A' or B keylocking switch and turning it to-

_ senice switch; the ' pump run' position; It will take bet.

-ween 50 and 150 ' minutes to complete the

- storage tank outlet valve injection and for.the storage tank. level" in test status; or sensors to indicate that the-storage tank is dry. When the injection is completed, the

< overload trip or power loss -

systenunay be manually turned off by turning in pump or storage innk out.

.the keylocking switch counterclockwise to?

let valve controls;

'the S1QP position, w is adwohcoO MM

- ow \\ ow +ava \\ tv i0" -

(ii)

Standbyliquid storage tank -

- (14): Reactor Operator Information high or low temperature; -

(a) The following items are located in the (iii) Standbyliquid tank high er control room for operation information:

low level;

^

(1) AnalogIndication (iv) Standby liquid pump A (B)

. auto trip.

.(i)

Storage tank level and temper-ature; (b) The following items are located locally at the equipment for operator utili-(ii)

System pressures; zation:

74'3 Amendment 2

AME ux6 worn Standard Plant RTV B 9.3.3 Equipment and Floor approaching actual use requirements, Drainage System Demineralized water, rather than the actual neutron absorber solution, can be injected The system which collects and transfers all into the reactor to test the operation of radioactive liquid wastes is discussed in all components of the redundant control Subsection 9.3.8 The drainage systems for system.

non. radioactive liquid wastes are not discused because they are not a part of the ABWR Standard (5) The neutron absorber shall be dispersed

Plant, within the reactor core in sufficient quantity to provide a reasonable margin for 9.3.4 Chemical and Volume Control leakage or imperfect mixing.

System (PWR)

(6) The system shall be reliable to a degree (Not applicable to a BWR) consistent with its role as a special safety system; the possibility of unintentional or 9.3.5 Standby Liquid Control System accidental shutdown of the reactor by this system shall be minimized.

93.5.1 Design Bases 9.3.5.2 System Description adosMo% mM*d

' orc" h 93.5.1.1 Safety Design Bases The S S (Figure 9.31) isananuglly initiated The standby liquid control system (SLCS) has a through3

' ogle keyboard switchin the main safety related function and is designed as a control room to pump a boron neutron absorber Seismic Category I system. It shall meet the solution into the reactor if the operator following safety design bases:

determines the reactor cannot be shut down or kept shut down with the control rods. Once the (1) Backup capability for reactivity control operator decision for initiation of the SLCS is shall be provided, independent of normal re-made, the design intent is to simplify the activity control provisions in the nuclear manual process by providing dual keylocked reactor, to be able to shut down the reactor switches, This prevents inadvertent injection if normal control ever becomes inoperative.

of neutron absorber by the SLCS. However, the insertion of the control rods is expected to (2) The backup system shall have the capacity for assure prompt shutdown of the reactor should it controlling the reactivity difference between be required.

the steady. state rated operating condition of the reactor with voids and the cold shutdown The keylocked control room switch is provided condition, including shutdown margin, to to assure positive action from the main control assure complete shutdown from the most room should the need arise. Procedural controls reactisc conditions at any time in core life, are applied to the operation of the keylocked control room switch.

(3) The time required for actuation and effectiveness of the backup control shall be The SLCS is required only to shut down the consistent with the nuclear reactivity rate reactor and keep the reactor from going critical of change predicted between rated operating again as it cools.

and cold shutdown conditions. A fast scram of the reactor or operational convol of fast The SLCS is needed only in the improbable reactivity transicats is no, specified to be event that not enough control rods can be accomplished by this system.

inserted in the reactor core to accomplish shutdown and cooldown in the normal manner.

(t) Means shall be provided by which the functional performance capability of the The boron solution tank, the test water tank, backup control system components can be the two positive displacement pumps, the two serified periodically under conditions motor operated injection valves, the two motor-I I

Amendmtni 6 9M

~

  1. BWR -

m-i Standard Plant p1x n

',~

o'perated pump suction valves, and associated Each positive displacement pump is sired to local valves, panel, and controls are located in inject the solution into the reactor in 60 to the secondary containment outside the drywell and 150 minutes, independent of the amount of wetwell. The liquid is piped into the reactor solution in the tank. The pump and system vessel throughout the high pressure core flooder design pressure between the injection valves and (HPCF)1ine downstream of the HPCF inboard check the pump and system design pressure between valve.

relief valves are approximately 1560 psig. To prevent bypass flow from one pump in case of The boron absorbs thermal neutrons and thereby relief valve failure in the line from the other terminates the nuclear fission chain reaction in pump, a check valve is installed downstream of the uranium fuel.

cach relief valve line in the pump discharge pipe.

The specified neutron absorber solution is sodium pentaborate (Na2B oOl6 10H O).

The SL_CS is, actuated by either of two i

2 It is prepared by dissolving stoichiomett!c cylocked, sprinbreturn switches on the control quantities of borax and boric acid in deminera. room console. This assures that switching from lized water. An air sparger is provided in the.

the OFF position is a deliberate act. Changing tank for mixing. To prevent system plugging. the either switch status to RUN starts an injection tank outlet is raised above the bottom of the pump, opens one motor operated injection valve, tank.

opens one pump suction motor operated valve, and closes both of the reactor cleanup syste m At all times when it is possible to make the outboard isolation valves to prevent lo<.s of reactor critical, the SLCS shall be able to boron.

deliver enough sodium pentaborate solution into 3*

b

\\

the reactor (Figure 9.3 2) to assure reactor A light in the control room indicates that shutdown. This is accomplished by placing sodium ;

power is available to the pump motor contactor pentaborate in the standby liquid control tank and that the contactor is deenergized (pump not and filling it with demineralized water to at running). Another light indicates that the least the low level alarm point. The solution contactor is energized (pump running).

can be diluted with water to within 14 inches of the overflow level volume to allow for Storage tank liquid level, tank outlet valve evaporation losses or to lower the saturation position, pump discharge pressure, and injection temperature.

' valve position indicate that the system is functioning. If any of these items indicates The minimum temperature of the fluid in the that the liquid may not be flowing, the operator tank and piping shall be consistent with that shall immediately change the other switch to the obtained from Figure 9.3 3 for the solution RUN mode, thereby activating the redundant train temperature. The saturation temperature of the of the SLCS The local switch will not have a l recommended solution is 59 F at the low level STOP position. This prevents the isolation of alarm volume and a lower temperature at 14 inches the pump from the control room. Pump discharge below the tank overflow volume (Figures 9.3-2 and pressure and valve status are indicated in the 9.3 3). The equipment containing the solution is control room.

installed in a room in which the air temperature is to be mair.:ained within the range of 50 to Equipment drains and tank overflow are not l 100oF. An electrical resistance heater piped to the radwaste system but to separate system provides a backup heat source which containers (such as 55 gallon drums) that can be maintains the solution temperature at 750F removed and disposed of independently to prevent (automatic operation) to 85oF (automatic any trace of boron from inadvertently reaching shutoff) to prevent precipitation of the sodium the reactor, pentaborate from the solution during storage.

High or low temperature, or high or low liquid Instrumentation consisting of solution tem-level, causes an alarm in the control room, perature indication and control, solution lesel w ko ke ci cG% rsm f %40nohcai\\cM m pM kess ed m 4k+ s em ( ATw s)

o. e t(3 ys, l oc ccm be mcWolj

^"' * " a 12 m

s...

INSGRT

-An'ATWS condition exists when either of the following occurs (a) High RPV pressure (1125 1psig) and average power range monitor (APRM) not down scale for 3 minutes, or (b) Low-RPV level (Level 2) and APRM not down scale for 3 minutes.

4

4-RESPONSE TO AGENDA ITEM 6 Since ABWR will incorporate an automatic ADS inhibit following an ATWS, there is no operation action required.

Therefore, the 29 second delay in the actuation of ADS following a Low Water Level i signal is acceptable.

_This delay is used to confirm that Low Water Level 1 signal is present and is consistent with the low pressure ECCS pump start-up time.

- m

6 RE S PONSE To A GCh>0 A ~I TGM ~7 An 8 minute high drywell pressure bypass timer has been added to the ADS initiation logie to address TMI action item II.K.3.18.

This timer will initiate on a Low Vater Level 1 signal.

When it times out, it bypasses the need for a high drywell signal to initiate the standard ADS initiation Icylc.

For all LOCAs inside the containment, a high drywell signal will be present and ADS will actuate 29 seconds after a Low Vater Level 1 signal is reached.

All LOCAs outside the containment become rapidly isolated and any one of the three high pressure ECCS can control the water level.

The_high drywell-pressure bypass timer in the ADS initiation logic will only affect the LOCA response if all high pressure ECCS fail following a break outside the containment.

For this case the ADS will automatically antiate within 509 seconds (8 minute timer plus 29 second standard ADS logic delay) following.a Low Vater Level 1 signal.

An analysis was performed to evaluate the adequacy of the 8 minute bypass timer.

As discussed above this analysis involves multiple failures.

Therefore it is part of the PRA success criteria realistic LOCA evaluation.

The main difference between the input assumptions for this case and-those'in Section 6.3.3 of the SSAR is the use of a realistic decay heat curve (i.e. ANS 5.1).

A complete circumferential break of the main steamline outside the containment (which is representive of those LOCA cases where no high drywell signal is present.) was evaluated.

The key results from this analysis are given in Figures 7-1 through 7-7.

The peak' cladding temperature (PCT). refer to Figure 7-7, for this case is less than 1100'T which is well below the 2000

  • T licensing limit. This PCT is due to the initial voiding in the core at the beginning of the event as in the case of the design basis evaluation.

The heatup caused by the core uncovery late in the event is negligible.

These results confirm the acceptability of the 8 minute high drywell pressure bypass timer.

bdc6 ci d a cLo cl.

rke c o mspeaol (po $os 7 3-s,g7.3-6 47.3 -7 ) 34 s au l

l l

l

..y 1

.g.~

4

-Figure 7 ' WATER LEVEi? IN FUEL CHANDELS FOLLOUING A MAIN STEAHLINE j :-

BREAK OUTSIDE CONTAINMENT, NO HIGH PRESSURE ECCS AVAILABLE

-(REALISTIC ANALYSIS ASSUMPTIONS)

' 60.

5 HOT CHANNEL 2 AVERAGE CFANNEL

- s TOP OF ACT IVE. FUEL -

40.

n 1

l' 2

2 2

2 1

1 2

v M

f p

3

\\> f J

\\

w>

W 20-

_J O'

Ld l-

<C3

~

'I'

I 0.

O.

0. 3.

O. 6.

O. 9

1. 2> 10' l,7 b j; =,s m s TIMEL L(SECOND)

r Figurc 7-2 WATER LEVEL INSIDE 3RROUD FOLLOWING A MAIN STEAMLINE BREAK OUTSIDE CONTAINMENT. 110 RIGH FRE55URE ECC5 AVAILABLE (REALISTIC ANALYSIS ASSUMPTIONS) 60.

UPPER PLENUM 2 BYPASS 3 LOWER PLEilUM GUIDE TUBES 4

  1. A12

(

i v%'> -

J>\\'

I 2

u-y

'VV(y(~

i

_J w>

Ld 4"O-j 43 43 43 43 43 43 43 43 L

Y LU i-43 M

0.

O. 3

0. 6
0. 9 1.2 10'
"lp mms TIME (SECOND)

~

Figure "I-3 WATER LEVEL OUTSIDE SHROUD FOLLOWING A KAIN STEAffLINE -

BREAK OUTSIDE CONTAINMENT, NO HIGH PRESSURE ECCS AVAILABLE (REALISTIC ANALYSIS ASSUMPTIONS) 60.

~

REGION 6 i

2 REGION 7 s TOP OF ACT IVE FUEL 2

40. j, 5

1

^

2 l-g 2,

v 3

2, 1

_J w

W 20-

_J E

LLI l-

<C 3

I'

0.

O.

O. 3

0. 6
0. 9 1.2 10'

= mos9 TIME (SECOND) 011392 1828.5

li:

~

'0 1

=2 1

EL B

AL I

A 8

VA E

SC R

8 C

f E

t E

9 ER NU E

N 0

I S LS D

ME AR EF L

TSH M

8 G

NI)

I HS k

A N

MOO NI A

T

,F GTM A

NNU M

1 ES

)

uMS D

t oNA LI 6

LAS OTI N

FNSO Y, 0O ECI C

R A

UEN SDA E

SI ESC S

RTI PUTOS

(

L I

EKL SA A SEE ER R VB(

E M

4 3I J'

T N

O cr ug i

F 1

W I

O t

s.a 2a i

.O" 8

4 0

0 0

1 T

x L

ACUOv LgDy)g) LEO J

J t

t i

1

, 'i i!

!;)I

^~

~.

Figure 7 -FLOV OUT OF VESSEL FOLLOWIffG A MAIN STEANLINE-BREAK OTJTSIDE CONTAINMENT, NO MIGH PRESSURE ECCS AVAILABLE

-(REALISTIC ANALYTIS ASSUHPTIONS) 3.

BREAK 3

x10 2 ADS s STEAMLINE SRV'

-4

-(

n O

.w 2.

2 (A

N 2

CD g

v 2

U l-4 1.

O'

~

2

~

4444444444444 44444 O

1

]Y l

2 f

5 et s

e e e e

e sie e

4 8,1 e 34 34 y

a t

3 e

o e

a e

e e a la e

l '

-l l

0.

0.

O. 3 -

0. 6
0. 9
1. 2=10'

}Lyos9 TNE

. (SECOND) m3

Figure 7-6 FLOW INTO VESSEL FOLLOUING A MAIN STEAMLINE BREAK OUTSIDE CONTAINMENT, N0'HIGH FRESSURE ECCS AVAILABLE' (REALISTIC ANALYSIS ASSUMPTIONS)

1. 5 x 10' FEEDWATER i.

2 LPFL n

O Ltj 1.

U3

\\

1 CD

._I v

2 0R O'

~

2 3

O

_J Lt_

~

2 1

2 1

2

~1 2

1 2

e 2

I'1 O.

O.

O. 3

0. 6
0. 9 1.2 10 2

l>L3" mms TIME (SECOND)

Figure 7-7 PEAK CLADDING TEMPERATURE FOLLOli1NG A MAIN STEAMLINE BREAK OUTSIDE CONTAINMENT, HO HIGH PRESSURE ECCS AVAILABLE fREALISTIC ANALYSIS ASSUMPTIONS)

2. 4 x10' i PEAK. CLAD TEMPERATURE

,m J-

1. 6 0

, : uj O

' Ltj i

M i

!D lH

0. 8 i

<C m

i Ltl i

i O_

[

'1 LJ l--

i l

'I'

0.

O.

O. 3

0. 6
0. 9
1. 2=10'

' 7# b e " 2828.3 TIME (SECOND)

ABWR msima Standard Plant nrv 4

^ '-

Table 6.31 SIGNIFICANT INPUT VARIAHLES USED IN THE LOSS OF COO! ANT ACCIDENT ANALYSIS (Continued)

Yariable L' alls Yalue Initiating Signals

~

ft above TAF 10.

M Watrr Leve1~ ~

and

- -p ' High Drywell Pressure

(

__ __ psig, _

_.t2.0j j

DelayThee from Alllaitiating sec 5 29 Signals Completed to the Time J

Valves are Open C. FUEL PAMMETERS Yariable 1,'Alla Yalus g

FuelType Initial Core Fuel Bundle Geometry 8x8

~ ~ * -

[J([jCe C

Number of Fueled Rods 62

.)

-~~

Peak Technical Specification kw/ft 13A Linear Heat Generation Rate laitial Minimum Critical '

1.13 Power Ratio '

Design Axial Peaking '

1A0 Factor I ~ &nal #

- kH4*tka ia ater 1 eve) 55 ft a bove. TA F 0G ant drywell pressu% d F8'3 iLo (1)$

ig re or High drywell pressure bypass y e c._

e:. y yo timer timed out O h b.

6 kl8

w-_

v ABWR usuu Slandatil Plant krv A ways; they will relieve pressure by actua.

drywell at 70% of design gage itessure tion with electrical power or by mechanical following failure of the pneumatic supply to I

actuation without power. The suppression the accumulator. Sensors provide inputs to pool provides a heat sink for steam reliesed local multiplexer units which perform signal by these valves. Relief valse operation may conditioning and analog.to digital be controlled manually from the control room conversion. The formatted, digitized sensor to hold the desired reactor pressure. Eight inputs are multiplexed with other sensor of the eighteen SRVs are designated as auto.

signals over an optical data link to the matic depressurization system (ADS) valves logic processing units in the main control and are capable of operating from either ADS room. All four transmitter signals are fed logic or safety / relief logic signals. The into the two.out of.four logic for each of safety / relief logic is discussed in Para.

two divisions either of which can actuate graph (4). Automatic depressurization by the ADS. Station batteries and SSLC power the ADS is provided to reduce the pressure supplies energize the electrical control during a loss.of coolant accident in which circuitry. The power supplies for the the HPCF and/or RCIC are unable to restore redundant divisions are separated to limit vessel water level. This allows taskeup of the effects of electrical f allutes.

core cooling water by the low pressure Electrical elements in the control system makeup system (RHR/LP flooding mode).

energire to cause the relief valses to open.

(2) Supporting System (Power Supplies)

(a) ADS laitiating Circuits Supporting systems for the ADS Cal include Two ADS subsystems for relief valve the instrumentation, logic, control and actuation, ADS 1 and ADS 2 are provided.

motive power sources. The lastrumentation (see Figure 7.3 2). Sensors from all and logic power is obtained from the SSLC four divisions and division I control Division I and 11,120.VAC buses F1 and 01.

logic for low reactor water level and The control power is from the Division I and high drywell pressure initiate ADS 1, 11,125.VDC battery buses F and G (see Fig.

and sensors from all four divisions and

-i ute 8.31). The motive power for the elec.

division il control logic initiate ADS trically operated gas pilot solenoid valves

2. The division I logic is mounted in a is from local accumulators supplied by the different cabinet than the division 11 high pressure nitrogen gas supply systems logic.

-(Divisions I and II) (see Chapter 6).

The reactor vessel low water level' (3) _ Equipment Design initiation setting for the ADS is selected to depressurire the reactor The autornatic depressurization system (ADS) vesselin time to allow adequate cooling consists of redundant trip channels arranged of the fuel by the RHR (LP flooding in two separated logics that control two mode) system following a loss.of coolant separate solenoid. operated gas pilots on accident in which the HPCF and/or RCIC each ADSpalva. Either pilot valve can f all to perform their functions Ql*"g operate its associated ADS valve. These adequately, Timely depressurization of pilot valves control the pneumatic pressure the reactor vessel is provided if the

, It applied by accumulators and the high reactor water level drops below pressure nitrogen gas supply system. The acceptable limits together with an !

operator can also control the SRV's indication that high drywell pressure {

m a n u a lly. Separate accumulators are has occurred, which signifies there is a i included with the control equipment to store loss of coolant into the_ containment pneumatic energy for relief valve operation, with insufficient high pressure makeup to maintain reactor water level.

The ADS accumulators are sired to operate Reactor isolation occurs on loss of the safety relief valve two times with the coolant outside the containment.

Amendment 2 W

gr

-<,y--

ar-g-

v>-e em v tv--1 v

'ns--ew w

y m

- + * --~m r4www-4 Mw-p-e-e--

y w

+y--*y

--*7--w

ABWR mainw Standard Plant uv a The HPCF and RHR.LPFL discharge (b) Logic and Sequencing pressure settings are used as a permissive for Jepressurization and are Two parameters of initiation signals are sslected to assure that at least one of used for the ADS: drywell high pressure, the three RHR pumps, or one of the two and rector vessel low. low water level HPCF pumps, has received electrical (Level 1). Two out.of.four of each set power, startsd and is capable of of signals must be present throughout delivering water into the vessel. The the timing sequence to cause the safety /

pressure setting is high enough to relief valves to open. I!ach parameter assure that the pump will deliver at or separately seals itself in and annuncl.

near rated flow without being so high as ates following the two.out.of four logic to f all to show that the pump is confirmation. Low water level 11: the actually running.

final sensor to initiate the ADS.

The !cvel transmitters used to initiate A perminive signal of RHR (LP flooder one ADS logic ate separsted from those mode) or HPCF pump discharge pressure is used to initiate the other ADS logic, also used. Discharge pressure os any Reactor vessel low water levelis de.

one of the three RHR pumps or one of the tected by eight transmitters ihat two HPCF pumps is sufficient to give the measure differential pressure. Drywell permissive signal which permits high pressure is detected by four pres.

automatic depreuurization when the RHR sure transmitters. All the vessellevel or HPCF systems are operable.

and drywell high pressure transmitters,g are located la the@rimarv containmenf wig After receipt of the initiation signals' outside the drywell. The drywell high and after a delay provided by time delay pressure signals are arranged to elements, each of the two solenoid pilot

  • seal.in' the control circuitry. They las valves is energized. This allows must be manually reset to clear, pneussatic pressure froam the accumulator to act on the gas cyllader operator.

Time delay logic is used in each ADS The gas cylleder operator opens and control division. The tinse delay holds the relief valve open. Lights la setting before actuation of the ADS is the mala control room indicate when the long enough that the HPCF and/or RCIC Jolenold. operated pilot valves are have tinse to restors water level, if energized to open a safety / relief capable, yet not so long that the RHR valve. Linear variable differential (LPFL asode) is unable to adequately cool transformers (LVDT's) mounted on the the fuelif the HPCP falls to prevent valve operators verify each valve low water level. As assusciator la the posillos to the performance monitoring control rooms is actuated when either of and control systems (PMCS), and the the timers la timing. Rosetting the ADS annunciators, initiating sissala has no effect on the timers If the laillating signals are The ADS Divislos I control logic l still present.;

actuates a solenoid pilot valve on each ADS valve. $1milarly, the ADS Division if the reactor level is restored 11 control logic actuates a second sufficiently to reset the previous separate solenoid pilot valve on each actuation setpoints before the timer ADS valve. Actuation of either times out, the timer automatically solenoid. pilot valve causes the ADS resets and auto depressurization is valve to open to provide depressurl.

aborted. Should additionallevel dips ration.

occur across the setpoints, the timer resets with each one.

Manual reset circuits are provided for Amendment 7 734

)

l

ABWR nAstooAr SIAndudPlani nw 0 the ADS initiation signal and the two (c) Bypases and f alerlocks paranneter sensor loput logic signals.

An atteropted reset has no effect if the Before the ADS timers titue out, it 153 two.out of.four initiation signals are possible for the operator to manually still present from each parameter (high delay the depressurizing action by the drywell prenure and low. low reactor manual inhibit switches (although the water level). liowever, a keylocked time for this is very short, i.e., only inhibit switch is provided for each 29 seconds). This action resets the division which can be used to take one time delay logic to zero seconds and ADS division out of service for testing prevents depressurization for another or maintenance during plant operation, timer cycle. The operator would make g

This switch is ineffective once the ADS this decision based on an assessment of timers have timed out and thus cannot be other plant conditions. The primary ',#

used to abort and reclose the valves purpose of the inhibit switch is to once they are signalled to open. The remove one of the two ADS logic and h" "

inhibit mode is continuously annunciated control divisions from service for la the main control room.

testing and realotenance during plant,

operation. Automatic ADS is interlocked Manual actuation pushbuttons are pro.

with the llPCF and RHR by means of vided to allow the operator to initiate pressure sensors located on the ADS immediately (no time delay) if discharge of these pumps. Manual ADS required. Such initiation is performed bypasses these interlocks and the timers by first rotating the collars surround.

and immediately opens the ADS valves, ing the pushbuttons for each of two The rotating collar permissives and channels within one of the two divi.

duality of button sets combined with sions. An annunciator will sound to annunciators assure manual laitiation of warn the operator that ADS la armed for ADS to be a deliberate act.

s that division. If the two pushbuttons are then depressed, the ADS valves will (d) Redundancy and Dhrrrity open, provided the ECCS bunp(s) running permissives are present. Though such The ADS is initiated by high drywelli,MS manual action is immedicte, the rotating pressure and low reactor vessel water collar permissives and duality of button level. The initiating circuits for each P"

s sets combined with annunciators assure of these parameters are redundant as ( W*1 -

manual initiation of ADS to be a desuibed by the circuit description of hn "I deliberate act.

this section. Diversity is provided by.

HPCF, A control switch is available in the main control room for each safety / relief (e) Actuated Devices valve including the ones associated with the ADS. Each switch la associated with Safety / relief valves are actuated by any one of the four electrical divisions and one of four methods, maintains electrical separation consis-tent with the requircJ operability (1) ADS Action though its function is not required for S

safety. The switches are three position Automatic action after high drywell keylock type, OFF. AUTO OPEN, located pressure followed by 29 seconds at 3,., I t on the main control board. De OPEN po-low water level (L1), plus makeup l[', ' h sition is for manual safety / relief valve pumps running, resulting from the 3

operation. Manual opening of the relief logic chains in either Division I or valves provides a controlled nuclear Division II control logic actuating;,

system cooldown under conditions where the normal heat sink is not available.

Amendment 8 717

Make the following changes to Chapter 7, Section 7.3.1.1.1.2 " Automatic Depressurization System Instrumentation and Controls.

4 1.

The second paragraph in Subsection (a) " ADS Initiating Circuits

  • In Sub Section (1)
  • System identification
  • In Section 7.3.1.1.1.2
  • In Chapter 7 of the United States (US) Advanced Boiling Water Reactor (ABWR) Certification Program's Standard Safety Analysis Report (SSAR) reads as follows:

The reactor vessel low water level initiation setting for the ADS la selected to depressurize the reactor vesselin time to allow adequate cooling of the fuel by the RHR (LP flooding mode) System following a loss of-coolant accident in which the HPCF and/or RCIC fall to perform their functions adequately.

Timely depressurization of the reactor vesselis provided if the reactor vessel water level drops below acceptable limits together with an indication that high drywell pressure has occurred, wh.lch signifier. there is a loss of coolant into the containment with insufficient high pressure makeup to maintain reactnr water, level. Reactor isolation occurs on loss of coolant outside contalryment.

Change this paragraph to read:

The reactor vessel low water level Initiation setting for the ADS is selected to depressurize the reactor vesselin time to allow adequate cooling of the fuel by the RHR (LP flooding mode) System following a loss-of-coolant accident in i

which the HPCF and/or RCIC fail to perform their functions adequately.

Timely depressurization of the reactor vesselis provided if the reactor vessel water level drops below acceptable limits together with an indication that high drywell pressure has occurred, which signifies there is a loss of coolant into the containment with insufficient high pressure makeup to maintain reactor water level. For br==M nadalda the containment ilmalv danraamuriration 6f the remedar v

' is orovidad if the r= war vammal water level drons below amentahla limits for A time oeriod sufficient for the ADS hioh drywell oressure bynaam ilmer and the ADS timer to time-out. Reactor isolation occurs on loss of coolant outside containment.

92UNBSLO1.JKS

2.

Subsection (c)

  • Bypasses and interlocks" In Sub Section (3)
  • In Chapter 7 of the US ABWR Certification Program's SSAR reads as follows:

(c) Bypasses and Interlocks.

Before the ADS timers time out, it is possible for the operator to manually delay depressurization action by the manualinhibit switches (although the time for this is very short, i.e., only 29 seconds). This action resets the time delay logic to zero and prevents the depressurization for another time cycle.

The operator would make this decision based on an assessment of other alant conditions. The primary pur)ose of the inhibit switch is to remove one of 11e two ADS logic and control d visions from service for testing and maintenance during plant operation. Automatic ADS ls interiocked with the HPCF and RHR by means of pressure sensors located on the discharge of these pumps.

Manual ADS bypasses these interlocks and the timers immediately opens the ADS valves. The rotating collars permissives and dualityof buttons sets combine with annunciators assure manualinitiation of ADS to be a deliberate act.

4.

Change this paragraph to read:

(c) Bypasses and Interlocks.

There is one mantini ADS Inhibit switch in the control room for each ADS logic and control divialen which will Inhibit ADS Initiation. If ADS has not initiated.

The primary purpose of the inhibit switch is to remove one of the two ADS logic and control divisions from servios for testing and maintenance during plant operation. Automatic ADS is interlocked with the HPCF and RHR by means of pressure sensors located on the discharge of these pumps. Manual

- ADS bypasses these interiocks and the timers immediately opens the ADS valves. The rotating collars permissives and duality of buttons sets combine with annunciators assure manual Initiation of ADS to be a deliberate act.

92UNBSLO1.JKS

3.

Subsection (d) ' Redundancy and Diversity

(d) Redundancy and Diversity.

The ADS is initiated by high drywell pressure and low reactor water level. The initiating circuits for each of these parameters are redundant as described by the Crcuit description of this section. Diversity is provided by the HPCF.

Change this paragraph to read:

(d) Redundancy and Diversity.

The ADS is initiated by high drywell pressure andistt low reactor water level.

The initiating circuits for each of these parameters are redundant as described by the circuit description of this section. Diversity is provided by the HPCF.

4.

Paragraph (1) " ADS Action"in Subsection (e)" Actuated Dey!ces" in Sub Section (3) " Equipment Design"in Section 7.3.1.1.1.2 ' Automatic Depressurization System Instrumentation and Controls"in Chapter 7 of the US ABWR Certification Program's SSAR reads as follows:

(1),.DS Action.

Automatic action after high drywell pressure followed by 29 seconds at low water level (L1), plus make up pumps running, resulting from the logic chains in either Division I or Division ll control logic actuating;.

Change this paragraph to read:

(1) ' ADS Action.

Automatic action after high drywell 3ressure followed by 29 seconds at low water level (L1) or low water level ( _1) for 8 minutes (ADS hich drvwell oressure bvomu timer) and 29 seconds (ADS timer). plus make up pumps running, resulting from the logic chains in either Division i or Division ll control logic actuating;.

92UNBSLO1.JKS L

i HESPONSF TO DISCUSSION ITEtt 17 The attached text modifications clarify the effective break areas for the maximum vessel bottom head drain line break and the maximum RHR shutdown auction line break, i

,-.-.,..-..-._,..,.,,-c.

,-.~~.,-.n.

-, -, -, -,,- - +,, - - -,.,

,-~.-w-w

ABWR nuima Sandard Plant arv c j

i..-

(9) peak cladding temperature as a funetton of dralnline break case. Therefore,Ihe dralnline time.

break analysis is also bounding for any credible break within the reactor laternal pump-l A conservative assumption made in the analysis recirculation system and its associated motor i

is that all offsite AC power is lost simul.

housing and cover.

i taneously with the initiation of the LOCA. As a i

further conservatism, all reactor internal pumps As expected, the core flooder line break is were assumed to trip at the start of LOCA event the worst break location in terms of minimum i

even though this in itself is considered to be an transient water level in the downcomer, in accident (See Subsection 15.3.1). The resulting elevation it is the lowest break on the vessel rapid core flow coastdown produces a calculated except for the drainline break. Furthermore, the departure from nucleate boiling in the hot worst break /fallure combination leaves the bundles within the first few seconds of the fewest number of ECC systems remaining and no i

transient, high pressure core flooder systems, LOCA analyses using break areas less than the maximum l

, LOCA analyses using break areas less than the values were also considered. The cases analyzed g

maximum values were also considered for the are lodicated on the break spectrum plot (refer steamline, feedwater line, and RilR shutdown to Figure 6.310). From these results it is

-i suction line locations. -The cases analyzed are clear that the overall most limiting break In' i

indicated on the break spectrum plot (refer to terms of minimum transient water levelin the Figure 6.310). In general, the largest break at downcomer,is the maximum core flooder line each location Is the worst in terms of minimum break case.

transient water levelin the downcomer.

6JJ.7.7 Line Breaks Outside Contalement 6J.3.7J Intermediate Line Breaks inside Containment This group of breaks is characterized by a g

rapid isolation of the break. Since a maximum r7 For this case the maximum RHR PFL injection steam line break outside the containment pro-l 2

line break (0.221 ft ) was analyzed. Important duces more vessel inventory loss before isola-variables from this analysis are shown in Figures tion than other breaks in this category, the re.

~

6.3 37 through 6.3 43, suits of this case are bounding for all breaks-i in this group. Important variables from these i

6.3J.7.6 Small Line Breaka laside Containment analyses are shown in Figure 6.3 60 through 6.3 66.

For these cases the maximum high presure core 2

flooder line break (0.099 ft ) and the maximu As disc in Subsection 6.3.3.7.4, the i

2 bottom head drain line break (0.0218 ft ).ere trip of all reactor internal pumps at the start l

g analyzed. A mportant variables from these of the LOCA produces a calculated departure from l

analyses are shown in Figures 6.3 44 through ' nucleate boiling for all LOCA events. Further.

7 6.3 59.' A break in a reactor internal pump more, the high void content in the bundles fol.

would involve either the welds or the casing, lowing a large steamline break produces the if the weld from the pump casing to the PRV stub earliest times of loss of nucleate boiling for t

tube breaks, the stretch tube will prevent the any LOCA event. Thus, the summary of results in pump casing from moving; The stretch tube Table 6.3 4 show that, though the PCTs for all clamps the diffuser to the pump casing, where its break locations are similar, the steamline nut seats. The land is located below the casing breaks result in higher calculated PCTs and the attachment weld and therefore the stretch tube outside steamline break is the overall most

_ forms a redundant parallel strength path to the limiting case in terms of the highest calculated-pump casing restraint _ which is designed to PCT. Results of the analysis of this break will

- provide support in the event of weld failure, be provided for each bundle design for In case the pump casing and the stretch tube information by the utility referncing the ABWR break, the pump and motor will move downward design.

vatil stopped by the casing restraints. The pump is part of the stretch tube, la cither case the break flow would be much less than the Amendment t$

6.3-12 A

T l

iwcsv.xs eoe. t tsm s7 l

Since the bottom head drain line ties into the RHR shutdown suction A

line, the total break flow for the maximum RHR shutdown suction line break includes flow from the vessel through RHR shutdown suction vessel nozzle as well as through the bottom head drain line.

6

, based on a 2 inch penentration in the vessel bottom head since the bottom head drain line ties into the RHR shutdown suction C

line, the total break flow for the maximum bottom head drain line break includes flow from the vessel through the bottom head drain line penetration as well as through the RHR shutdown suction line.

4

,

Retrieved from "https://kanterella.com/w/index.php?title=ML20092F460&oldid=4332854"