Massachusetts Institute of Technology Department of Nuclear Science and Engineering

22.0 6 Engineerin g o f Nuclea r Systems

Dynamic Behavior of BWR

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The control system of the BW R cont rols th e re ac tor pre ssure , power and water level, as f o llows:

• The reactor pressure is controll ed by the tu rbine c ontrol v a lve s

• The reac tor power (rea c tivity ) is controll ed by the recirculation pum p s (via void reactivity feedback) and the Control Rods (CRs)

• The level is contro lled b y the Feed W a ter (FW ) flow. W hy do we want to control level?

A schem a tic of the BW R control system is shown below.

System

Sensed Level

Sensed Steam Flow

Feedwater Control System

Sensed Feedwater Flow

Feedwater Pump

Sensed Turbine Inlet Pressure

Pressure Control System

Bypass Valves

Turbine Valves

Load Demand Error

Load Reference

Steam Rate

Manual Controls

Turbine

Sensed Flux

Control Rod

Sensed Flow

Recirculation Flow Control System

Core Flow

Manual Controls

Rod Drive

Control Rod Drive System

Manual Controls

Pressure Control

Courtesy of GE Hitachi Nuclear Systems. Used with permission.

How does the BW R respond to a change in CR position (e.g. CR withdrawal)?

P (p r e ssu r e)

 (d e n si ty )

Fue l

q ( pow e r )

C o nt r o l r od w i t h dr a w F e w e r ab so r p ti o n s

H i ghe r pow e r q hi ghe r he a t f l ux

D e n s i t y d e cr ease 

l e s s mode r a t i on m o r e l e akag e

Co n t r o l r o d m ( f l o w ) … reac hes new s t e ady s t at e

h ( s ubc ool i ng)

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Courtesy of GE Hitachi Nuclear Systems. Used with permission.

How does the BW R respond to a change in core flow (e.g. flow increase)?

P ( p r essu r e ) F l o w i n cr ease m

 ( d en si t y )

Fu el

q ( p ow e r )

m o r e h eat r e m o v a l D e n s i t y i n cr ease 

be t t e r mode r a t i on l ess l e akag e

H i ghe r pow e r q hi ghe r he a t f l ux

D e n s i t y d ecr ease 

l ess mode r a t i on C o n t ro l ro d m ( f l o w ) m o r e l e akag e

h ( s ubc ool i ng)

… r eac he s new s t e ady s t at e

Courtesy of GE Hitachi Nuclear Systems. Used with permission.

In both cases the void reactivity feed back takes the system back to a (different) s t ead y- state condition.

Predict the response of the BW R to:

(i) an increase in coolant su bcooling at the core inlet (for exam ple this could be caused by a decrease in th e feedwater tem perature)

(ii) an increase in reactor pre ssure (for exam ple, this could be caused by partially closing the turbine valve)

(iii) an increase in turbine load. W ithout th e control system , woul d the BWR naturally f o llow the tu rbine load ?

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IAEA Simulator Quick Reference Guide

The IAEA B W R si m u lato r si m u lates an Advanced Boiling Water Reactor (A B W R)

ABWR has n o external recirculation loops and pum ps; uses internal pum ps instea d. The LB - LOCA is eli m inated by design.

BW R/6 A BW R

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SIMULATOR STARTUP

• Sel ect progra m ‘BWR’ for execution - the executable file is BWR.exe

• Cli c k any w here on ‘BWR si m u lator” screen

• Cli c k ‘OK’ to ‘Load Full Power IC?’

• The si m u l a t o r will displa y the ‘Plant O v erview ’ screen with all param e t e rs initialized to 100% Full Power

• At the bottom right hand corner click on ‘Run ’ to start the sim u lator

SIMULATOR INITIALIZ A TION

If at any t im e y ou need to r e turn the simulator to one of the stored i n itialization points, do the following:

• ‘Fre ez e’ the si m u lator

• Click on ‘I C’

• Click on ‘L oad IC’

• Click on ‘F P_100.IC’ f o r 100% full power initial state

• Click ‘OK’ to ‘Load C:\B WR_Sim ulator\FP_10 0.IC ’

• Click ‘YES’ to ' Load C:\BWR_Sim ulator\FP_100 .I C’

• Click ‘Return’

• Start the si m u l a tor operating by select ing ‘Run’ .

LIST OF BWR SIMULAT O R DISPLAY SCREENS

(1) BWR Plant Overview

(2) BWR Control Loops

(3) BWR Power/Flow Map & Controls

(4) BWR Reactivity & Setpoints

(5) BWR Scram Par a meter s

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(6) BWR Turbine Generat o r

(7) BWR Feedwater & Ext raction Steam

(8) BWR Trends

GENERIC BWR SIMUL A TOR DISP LAY COMM ON FEATURES

The generic BWR si m u lat o r has 8 interactive disp lay screens or pa ges. Each scr een has the sa me inform ation at the top and bottom , as follows:

• The top of the screen contains 21 plant alar ms and annunciations; t h ese indicate i m portant status change s in plant para m e t e rs that require operator actions;

• The top right hand corner shows the si mulator status:

□ The window under ‘Labvi ew’ (this is the proprietary graphical user interface software that is used to gen e rate th e screen displa y s ) h as a counter that is increm enting when Labview is running; if La bview is frozen (i.e. th e display s cannot be changed) t h e counter will not be incrementing;

□ The window display i ng ‘C ASSIM’ (this is th e proprie tary sim u lation engine soft ware that com putes the si m u lation m odel resp onses) will be green and the counter under it will not be incrementing when the sim u l a to r is frozen (i.e. the m odel program s are not executing), and will tur n red and the count er will increm ent wh en the sim u lator is runnin g ;

• To stop (freeze) Labview click once on the ‘STOP’ (red “Stop” sign) at the top left hand corner; to restart ‘Labview’ click on the  sy m bol at the top left ha nd corner;

• To start the sim u l a tion click on ‘Run’ at th e bottom right hand corner; to ‘Stop’ t h e si m u lation click on ‘Freeze ’ at the bottom right hand corner;

• The bottom of the screen s hows the values of the following m a jor p lant param e ters:

□ Reactor neutr on power (%)

□ Reactor ther mal power ( % )

□ Turbine gene rator output power (Gross) (%)

□ Reactor pres s u re (kPa)

□ Core flow (kg/s)

□ Reactor w a ter level ( m )

□ Balance of plant (BOP) ste a m flow (kg/s) — that mea n s stea m flow after the m a i n steam isolation valve

□ Feedwater flo w (kg/s)

□ Average fuel te m p erature ( D eg. C)

• The bottom left hand corner allows the initiation of t w o major plant events:

□ ‘Reactor trip’ or ‘reactor sc ra m ’

□ ‘Turbine trip’ These correspond to hardwired push buttons in the actual control room .

• The box ab ov e the trip butt ons shows th e display curr ently selected (i.e. ‘plant overview’); b y clicking an d hol ding on the arrow in this box t h e tit les of the oth e r display s will be shown, and a new one can be selected by highl ighting it;

• The rem a ining butt ons in t h e bottom right hand cor n er allow control of the sim u lation one iteration at a ti m e (‘Iterate’ ); the selection of initialization points (‘IC’ ); insertion of malfunctions (‘Malf’); and calling up t h e ‘Help’ screen (onli n e h y p e rlinked “Help” is not

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available y e t) . As a general rule, all dy na m ic display values shown in displa y bo xes on the screens fo llow the follo wing convent ions:

□ All pressure values are de signated as “ P ” next to the displa y box, and have units of kPa;

□ All te m p eratu re values are designated as “T” next to the displa y box and have va lues of deg. C;

□ All flow values are de signated as “F” n e xt to the dis p lay bo x and have values of kg/s;

□ Steam qualities are indicat ed as “X” next to the displa y box and ha ve % as units.

Valve status and pum p sta tus as shown by dynam ic e quipm ent sym bols ar e rep resented as follows:

• Valve status — red for valve fully open; gr een for valve fully closed; partial red and green indicates partial valve openin g ;

• Pum p status — red for run n ing; green for stopped.

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Abnormal Events Studied wi th th e IAEA BWR Simulator

 Inadvertent withdrawal of control rods

 Total lo ss of f eedwater f l ow

 Saf e ty Relie f Valve (SRV) f a ils ope n

 Reactor pu mp trip

 Steam line break

The IAEA BW R sim u lator has a m o re rea lis tic co ntrol system than PRISM. In particular, it tries to stabilize the reactor following an abnorm a l event without scram , whenever possible. W hy is scram generally undesirable?

Exercise 1: Inadv ert ent wi thdrawal of cont rol rods

Set reactor at 100% power in sim u lator

In the “Malf unction” m e nu select “Inadve rtent Withdrawal of One Bank of Rods” with 5 s delay

Run sim ulation and observe reactor power (h eat releas ed in th e fuel), steam quality, fuel tem p erature, pressure

Note the “high neutron power vs flow” warning sign, indicating operation in an abnorm a l region of the power/flow m a p

How does the contro l sys t em react? Observe con t rol rod reactivity and pu mp speed

Exercise 2: Total loss of f eedwat e r fl ow

Set reactor at 100% power in sim u lator

In the “Malf unction” m e nu select “L oss of Feedwater – Both FW Pumps Tripped” with 5 s delay

Do you expect the reactor to trip? If so, upon what signal?

Run the simulation and observe what happens. In particular observe reactor level and power. W h y does the power zig-zag up and down? How would the power vary with tim e, if the control rods did not m ove?

Go to the BW R Scram Param eters. W hy did the reac tor trip ? Does the turbine also trip eventually ?

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Exercise 3: Safety Relief Val v e (S RV) fails open

Set reactor at 100% power in sim u lator

In the “Malf unction” m e nu select “S RVs on 1 Main Steam Line Fail Open” with 5 s delay

Run the simulation and observe what happens. In particular observe reactor pressure, reactor level and power.

Would the power go up or down, if the control rods did not m ove?

The reactor reaches a new quasi-steady state w ith half of the steam flow rate going to the suppression pool. Is this an accep table situation in the long term ?

Exercise 4: Reactor pump trip

Set reactor at 100% power in sim u lator

In the “Malf unction” m e nu select “P ower Lo ss to 3 Reactor Internal Pumps” with 5 s delay

Run the simulation and observe the variati on of reactor pow er and steam quality. What do the control rods do?

Does the reactor trip ? Check out th e power/f low m a p. Is the reacto r in a s t able region of the m a p?

Exercise 5: Steam line break

Set reactor at 100% power in sim u lator

In the “Malf unction” m e nu select “Steam Line Br eak inside Drywell” with a 5 s delay Run the simulation.

W h at is the f i rst warnin g sign ?

How can we surely determ ine it was a break ? What does the ECCS do following the break?

Why does the Main Steam Isol ation Valve (MSIV) close?

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22.06 Engineering of Nuclear Systems

Fall 20 10

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