Text

Exam Rept 50-412/88-06OL on 880223-24.Exam Results:One Senior Reactor Operator Candidate Passed Both Written & Operating Exams
	ML20154B226
Person / Time
Site:	Beaver Valley
Issue date:	03/30/1988
From:	Eselgroth P, Yachimiak E; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20154B190	List: ML20154B185; ML20154B226;
References
	50-412-88-06OL, 50-412-88-6OL, NUDOCS 8805170116
	Download: ML20154B226 (126)
	v • d • e

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U. S. NUCLEAR REGULATORY COMMISSION REGION I

OPERATOR LICENSING EXAMINATION REPORT

EXAMINATION REPORT NO. 50-412/88-06 (OL)

FACILITY DOCKET NO. 50-412

FACILITY LICENSE NO. NPF-73

LICENSEE: Duquesne Light Company

Post Office Box 4

Shippingport, Pennsylvania

15077

FACILITY: Beaver Valley Unit 2

EXAMINATION DATES:

February 23-24, 1988

CHIEF EXAMINER:

f~/m

30 W$<77~

p dwar7 Yachimiak, Operations Engineer, Dits

Date

APPROVED BY:

[.[d

_

pr%, y f

' Peter W. Eselgroth, Chir _, PWR Section

Date

N,,0pera(tionsBranch,DRS

SUMMARY: One Senior Rear. tor Operator (SRO) candidate was administered written

and operating examinations.

Both parts of the examination were

completed successfully and a license was issued,

i

C 8 0 5 1 7 C ; 1'6' 8 8 0 5 0 9' .

PDR ADOCK 05000412

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F1-------------------- --fMel--.--------------.-.----

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REPORT DETAILS

TYPE "F EXAMINATION:

Replacement

EXAMINATION RESULTS:

One (1) SR0 candidate passed :.oth the written and

operating portions of the examination.

CHIEF EXAMINER AT SITE:

E. Yachimiak, NRC

OTHER EXAMINERS:

R. Temps, NRC

Personnel Present at the Exit Meeting

NRC Personnel

R. M. Gallo, Chief, Operations Branch

R. Temps, Operations Engineer

E. Yachimiak, Operation Engineer

Facility Personnel

A. J. Morabito, Manager, Nuclear Training

T. W. Burns, Director, Operations Training

T. D. Noonan, Plant Manager

T. E. Kuhor, Nuclear Operations Instructor

Attachments:

1.

SR0 Written Examination and Answer Key

2.

Facility Comments on the Written Examination

3.

NRC Response to Facility Comments

o

.

U.

S.

NUCLEAR REGULATORY COMMISSION

SENIOR REACTOR OPERATOR LICENSE EXAMINATION

>

-

FACILITY:

@EAyER_yALLEY_2__________

REACTOR TYPE:

PWR-WEC3_________________

DATE ADMINSTERED:

ggfggdg3_________________

EXAMINER:

YACHIMIAK

E.

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CANDIDATE

_

___

_ _ _S TR_ _UC_T I O_ N S_ _TO_ _C A_ N_D_I D A _T E_ _:

IN

_

_ _ __

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U M separate

paper for the answers.-

Write an'kers onione side only.

s

St c p l e. 4 question sheet

on top of.the answer

sheets.

Points for each

quantion are indicated in parentheses after the questi on.

The passing

grade requi res at least 70% in each category

and a final

grade of at

least 8 0". .

E>: a mi n a t i on papers will be picked

up six (6)

hours after

e examinat:cr sta-tv.

% OF

ATEGORY

% OF

CANDIDATELS

CATEGORY

l

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___SCQ3E___

_y@6UE__ ______________C@lEGQ5Y_____________

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24.C0 *

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._25.99

_______5.

THEORY Or NUCLEAR POWER D L o h' ?

-______ ___

OPERATION. FLUIDS.AND

7HERMODYNAMICS

d 4,60 4

3 22__ _25 99

6.

PLANT SYSTEMS DESIGN. CONTGOL.

__,________

________

AND INSTRUMENTATION

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PROCEDLIRES - NORMAL, A9 NORMAL,

EMEHbENCY AND RADIDLOGICAL

CONTROL

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_: . : : _ _ _ _ 2 9 3. 2 L

.

_ _ _ _ _ _ _ _ _ . . .

___ _

9.

ADMINIS'TRATIVE PROCEDURES,

CONDITIONS, AND LIM]1ATIONS

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Totals

. _ _ _ _ . _ _ .

L inal beace

Al1 wart cono er t92s n a.ntnatien a ". mv cwn.

I have nelther given

ce receiseo ata.

--_________ _______________ _______

.

Candidate's Signature

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NRC RULES AND GUIDELINES FOR t.ICENSE EXAMINATIONS

.

,

aring the administration of this examination the following rules apply:

-

Cheating on the examination means an automatic denial of your application

and could result in more severe penalties.

i

Rectroom trips are to be limited and-only one candidate at a time may

leave.

You must avoid all contacts with anyone outside the examination ~

room to avoid even the appearance or possibility of cheating.

Use black irk or dark pencil only to facilitate legible reproductions.

'

Print your name in the blank provided on the cover sheet of the

excmination.

m

-

,*

W

%

'

.,

. Fill in the date on the cover sheet of the examination (if necessary).

Use only the paper provided for answers.

Print your name in thc upper right-hand corner of the f irst Dage of each

section of the answer sheet.

Consecutively number each answer sheet, write "End of Category __" as

appropriate, start each-category on a new page, write only on one side

l

of the paper, and write "Last Page" on the last answer sheet.

~l

Number each answer as to cat'qory and number.

for examcle.

1.4

6.3.

Skip at least three lines between each answer.

.

Separate answer cheets irc? pad a~d place finished answer sheets face

down on your desk or taole.

{

Use ab' brevi at t enc on!'

3

' - m,

are common!v used in facility -literatare.

The point value for each cuestion is indicated in parentheses after the

.

vestion and can be used as a guide for the dep,th of answer reoutred.

. Show all calculations. methods, or assumotions used to obtain an answer

to mathematical problems whether indicated in the question or not.

Partial credit may be given.

Therefore, ANSWER ALL PARTS OF THE

.

QUESTION AND DC, NOT- LE AVE ANY ANSWER ULANK.

-j

la parts of the ex am: u ; un +' e not clear as to intent, as4 Questicos of

the Okaminer only.

You must sign the statement on the cover sheet that indicates that the

.

work is your own and you have not received or been given assistance in

completing the examination.

This must be done after the examination has

been completed.

.

-

. . .

.

13. When you complete your excmination, you shall:

i

.

examination as follows:

l

e.

Assemble your

(1).

Exam questions on top.

(2)

Exam aids - figures, tables, etc.

(3)

Answer pages including figures which are part of the answer.

b.

Turn in your copy of the examination and all pages used to answer

the examination questions.

c.

Turn in al1 scrap paper.and the balance.of the-paper that you.did

-

not$ use f or answeringJ tho ' questions.

-

.

d.

Leave the examination area,~as defined by the examiner.

If after

l eavi ng , you are found in this area while the examination is still-

in progress, your license may be denied or revoked.

.

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4

FLUIDS AND_THERMgDYNAMICS

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,

)UESTION

5.01

(3.00)

a. Explain, in terms of neutron flux, WHY a dropped rod could be worth

approximately 200 pcm whereas a stuck rod could be worth 1000 pcm, even

though the same rod could be considered in both cases.

(2.00)

b. WHAT are TWO (2) reasons for having control rod bank overlap?

(1.00)

o

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s

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)UESTION

5.02

(3.00)

a.

If critical data is recorded with Control Bank D at 120 steps, Tave at

547

F.,

and beron concentration at 1000 ppm. WHAT would be the expected

SUR

.f

Bas D is

aised tc 152 steps?

A s s u m.e Bank D's cifierential' roc

wceth is 6 PCM/ step and lamoda is 0.1/sec.

Show all work and state all

assumptions.

(2.00)

b.

A control rod falls-into the core when reactor power is at 50% at DOL.

HOW (Higher, Lower, No Change) would the resultant steady-state Tave

been different if this event had occured at EOL?

Justi f y your answer.

Assume a reactor trio DOES NOT occur. reds are in M ANL'AL , and all ether

systems are in automatic.

NO calcul ati ons are necessary.

(1.00)

l

uESTION

5.03

(3.00)

hH of the c-ima v parar. tors listed below. state HOW (Increasec.

ror E

C

Decreases, Nr Changes and explain WHY an INCREASE in tnat parameter affects

the DNBR.

Assume the other parameters remain constant.

.

!

a.

Reactor Power

o.

Tave

c.

Core Flow

d.

Pressurtzer pressure

)

CEr!ON

5.04

(4.'

,

i

Given Attachment :

"Estimated Critical Posi ti on Calculation."

{

.

a.

Complete ALL blank spaces on the form.

(3.00)

b. Calculate WHAT rate (gpm) of boron addition would be needed to change

1

the Estimated Critical Rod Position from Bank D at 75 steps to Bank D at

228 steps, assuming the current time is 0800 on 2/23/88.

(1.00)

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D.!ECONY

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5.

THEORY OF NUCLEAR POWER PLANT OPERATION

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JUESTION

3.05

(2.00;

Answer the f oll owi ng statements concerning Heat Exchanger Operation by

responding TRUE or FALSE.

2

Once tu-bule^t

2 2 c ;-

1-

2 ' cat

xchanger nas b an aatablishad,

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bcccmcs

appr=vimatc'; 2 cd

lu=.

c.

- en; f. r :crz : 2.n::t ex c n a .ga.-

23 ngt cqoat:nt.tn;n j,,T

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' :2. r ':v:rd;:? t =p:::tur=, .; u ad ta

c;r 'e;i ;:1;ulat: th: "::t

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tr:r.:f:r

tc.

c.

The heat removal rate for a heat exchanger will increase if either of

the fluid flowcates through the heat exchanger is increased.

c.

The U-tutes at the steam gener atcrs can experience thermal shock if the

feedwater flowrate is tncreased rapidly,

a

JESTION

5.06

(2.50)

v)H A T are FIVE

(b- indications that natural circulation has been established

after a loss of offstte power occurs.

r7' ION

5.07

'.00)

,

Ouring norma.

.sl a n t ocerations. WHEN does the reactor vessel evoertence the

~1ghest Gtressa" 6'ND WHOT TWO

I2'

primary parameters can be centrollec to

limit these stresses?

,

qTrnN

5,09

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c <enon esci11atIcns conver ae

dampeni

"r0 -aptels at BOL

r-

r

'

Just1i, , r ..

,' % m

i e- me ne

eactsvtty e + + .yc t s .

c.

Would the magnituce anc + r ecuenc v of xenon oscillations be Less at 3 0 */.

power or 100* power ~

J u ., t iy your answer.

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Page

6

FLUIDS.AND THERMODYNAMICS

.

)UESTION

5.09

(2.00)

i

For EACH of the following statements below, state HOW (Increase, Decrease,

No Change) ACTUAL Shut Down Margin (SDM) would be affected.

d.

The plant is in Mode 5 when a charging pump is mistakenly started

resulting in the injection of 200 gallons of boric acid ~into the RCS.

b.

The plant is in Mode 3 when all the shutdown. bank. rods ars. withdrawn out

fo4 - the core.~'

'

c.

The plant status changes from Mode 5 to Mode 4

d.

A control red drops into the core with the plant in Mode 1 at 5 0V. ocwe .

The reacto- cces n c *. trip.

,

JUESTION

S.10

(1.00)

Explain HOW the Moderator Temperture Coefficient (MTC) can act to increase

reactc" pcwe- when turti me

team demand Increaser.

Assume the plant

.3

initially at 7 5'/. p o w e r with rod control in MANUAL and all ether systems ir

automatic.

UESTION

3.1.

( 1.50)

,

Answer the following statements concerning Pump Operation ey responding

J

TRUE or FALSE.

I

as If flow throue a pump IncreaGes or the temperature 04 the tlutc

I n c r' e a t e s , the Re?.1 red Net PCCitive Surtion Head (W EH) ni ll Increase.

b.

When a ru r : e

coneated

a*

R ~nu t . cnd:

+

ras: t i t i en

y,c i r .31 1 s W !, ;

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occur.

c.

Hun-out is ilmited Ds ector 5: tepower.

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PLANT' SYSTEMS DESIGN

L AND INSTRUMENTATION

Page

7

L CONTROL

------------

.

(O.

2UESTION

6.01

( 3. O M

The plant is operating at 50% power when a control system hot leg RTD f ails

high.

Does this failure INCREASE, DECREASE, or NOT AFFECT the following:

Consider each item independently.

Assume no operator action and that all

control systems are in automatic.

a.

affected channel overpower delta T trip setpoint

b.

steam bypass cooldown valves (first bank)

c.

charging flow

. i n i t i al l y _),

(

d.?contro1# rod banktposit' ion

G.

rod insertion limit setpoint

1

2'f orted

"2-

ri 2:tu21 2;crtcmpcr;turc dcita

andicaticr

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JESTION

e.02

(3.00)

a. WHAT is the purpose of the high positive AND negative rate reactor

protection trips, respec t i vel y?

b. WHAT TWO (2) reactor protection trips are automaticall y reinstated below

p-107

c.

In WHAT TWO (2) ways are the SRNI's affected when the logic for P-10 is

satisfied?

'

ESTION

6.03

'?.50)

Using Attachment

2.

CD.

Manual Fig. No.

13-2,

"Quench Scray System.

Identify the f oll owi ng components on the attachment as specified in each

part celow.

,

l

a.

Highlight the

"A"

quench spray pump recirculation flowpath back to the

RWST.

n+'t

1

b.

Circle the THREE

(3'

bui l di ng /ar ea ecundarles that the

"D'

conta:-ment

quench baras heacer pacces through.

10.75'

c. Cl-cle WHERE the +lourate tor the

'n'

chemical injection pump 1s

measured.

,J.bos

d. Circle the THREE (3) valves that realign when the RWST level reaches the

l evel setpoint for 20SS-LSKK100B-1.

(0.75)

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JUESTION-

6.04

(2.00)

a.

HOW (Increase, Decrease, No Change) will an INCREASE in the reference

,

junction temperature effect indicated thermocouple temperature?

b.

HOW (High, Low, As Is) will an RTD temperature indication fail if a

short circuit occurs across the RTD?

c.

WHAT is the major disadvantage of using a Thermowell RTD for RCS wide

range temperature measurement?

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.

d.

Given the graph shown in Attachment 3,

identif y the curve which

represents the calibration curve for a HOT calibrated instrument.

JESTION

o.05

52.00)

For EACH of the following radiation monitors, state the automatic actions

which occur, if any, when the monitors alarm HIGH.

a.

25WS-RQIl01 - Component Cooling Service Water

b.

2HVR*RQllO4A - Containment Purge

c.

2RMC*RQ2OI - Control Room Area

d.

2GWS-RQIlo2 - Air Ejector Delav Bldg Exhaust

.

JESTION

6.06

(3.'O)

buxiliary Feedwater System.

answer the following questions concerning the

a. WHAT are FOUR (4) conditions /stgnals (including applicable logic) that

an cause a Motre Dri ven Am 11 i ary roadwate- Patp (MDnrWP' to automat -

cally start?

Assume the following conditions have been met:

(1) CS in

-

,2i

NMEPT o'

Out u"cervoltage

(3) Normal p;wer supply breaker ti is closec

(2.00)

b. WHAT TWO (2) plant conditions / signals (including applicable logic) will

cause the Turbine Driven Auxiliary Feedwater Pump (TDAFWP) to automati-

cally start?

(1.00)

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...P.LANT SYSTEMS DESIGN------------------ L CONTROL------ L AND INSTRUMENTATION.

Page

9

5. . .

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JUESTION

6.07

(3.00)

A significant leak occurs in the reference leg capillary of 2RCS-LT459,

pressurizer (PRZR) level transmitter.

Assume the, plant is at 100% power

.and all control systems are in automatic with 2RCS-LT459' selected as the

controlling channel.

a.

State TWO (2) automatic actions which would occur because of this

failure.

(1.00)

^

b'.~ State -FOUR. (4) ~ control room indications that areTavailable to alert the.

operator of this failure.

.(2.00).

,

.E5' ION

1.59

( 2. :'O )

,

WHAT do the following Safety Inj ec ti on System interlocks prevent?

a.

Low Head Safety injection (LHSI) pump minimum flow recirculation

i sol at i on valve C2 SIS *MOV8890A] opens when LHS1 pump C2 SIS *P21A]

discharge flow is low.

o.

Safety injection accumulator discharge stop valve C2 SIS *MOV865A] coens

when its control switch is in AUTO and 2 out of 3 pressurizer pressure

channels are greater than 2,000 psig.

j

c.

If a valve receives an S1, CIA. or C1B signal, the motor thermai

ov,erload i n ter l oc k s are Dvpacted.

i

)

d.

uMSI dischar;e <alves (25;S*MOV9i3~.5c35

to charging pum

suction

,

'

neader cpens in AUTO only if:

1. LHSI pumo discnarge neader valve [251S*MOVB911A] is fully open, and

2.

Hi gh Wead 31 alt, mint ' l ow : < a: 3t:en valves L2CHSoMOv500A'P.!83A/B]

are shut, and

j

3.

a recircula+ien mode intt:ation signal 1m cro"ent.

'

_ESTION

e> . .

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>>

..

l

WHAT are FOUR (4? cou-ces

c+

Hydecgen in the containment ouilding'

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a i" s

o L'UN i I f JUL L.

UN Ntf s i , 4 . il eeeeoi

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Page 10

t

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JUESTION

6.10

(2,50)

The plant is stable in Mode 5 with the

"A" Residual Heat Removal System

(RHS) in service.

a.

At WHAT pressure (psig) will the RHS isolate from the RCS?

(0.50)

b.

WHAT is the design capacity of the RHS suction line relief valve

[2RHS*RV721A37

Include ALL applicable information.

(0.80)

ci. ~ Loss of ' primary'!ciwnpoNent"coolinig water cari' af f ect WHAT TWO (2) RHS

'

components, when operating?

(0.70)

bYW Q

d.

Failure of RHS Hx vflow control valve, C2RHS*FCV605A3 to the closed

position will result in a (Increase. Decrease. No Chance) to RCS

tempe-atv e7

, , 5 ;, ,

as

+

1

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PROCEDURES - NORMALt_ABNORMALt_ EMERGENCY

Page 11

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A_ N_ D _ R_ A_ _D _I O_ L O_ G_ _I C_ A L _ C_ O_ N_ _T R_ O_ L_

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2UESTION

7.01

(1.50)

For the f ollowing questions assume B.V.P.S.

- D.M.

51, Station Shutdown

Procedure, is in use.

a.

When using condenser steam dumps, WHAT operator action (s) must be taken

to cooldown the RCS below the Lo-Lo Tavg setpoint?

(0.50)

b.

When *he Residual Heat Removal System ( RHS.) is in operation, at.least

one reactor coolant; pump'must~ remain in service'until RCS. temperature 1s

~

less than 200 degrees F.

WHY?

(0.50)

c.

If minimum RCS flow requirements CANNOT be met while in Mode 4,

the

operator's immediate response is to refer to WHAT procedure?

(0.50)

,

UESTION

7.02

(2.50)

.

Answer the following statements concerning Refueling by responding TRUE or

FALSE.

a.

Incning can ONLY ce accomplished when the key is set in the "RUN"

position and the inching permissive light is illuminated.

t.

e

e ergency cul l out c a:l e it usec te cull

t"e transfer car tack into

containment after the conveyor motor is disengaged from the transfer

syste'.

.

c.

1+

the gripper 2s engaged wnlle holding a RCCA and the Dillon Load Cell

reads greate- than 1200 lbc.. actuation of the gripper interlock bypass

switch will allcw the gripper to be disengaged.

1.

DeJore +Jel ha

..no opo ations in the 4 uel cullcing can commerre. the

fuel building vent system shall be in service and discharging through

at least one t ain of SLCnS WEPA filters and charcoal advorbers.

c. If the operatcr moticon a larce unexplained change in laac on the Nllen

reacTut, he shoalc : w9ediate;

rnve-se c1rn tion.

.

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PROCEOURES - NORMAL _ ABNORMAL _ EMERGENCY

Page 12

7

t

A_ N_ D_ _R_ A D _I O_ L_ O_ G_I C_ A_ L_ _ C_ON_ TR_ O_ L_

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2UESTION

7.03

(2.00)

Answer the following questions concerning

B.V.P.S.

procedure AOP-2.1.3,

"Continuous Insertion of RCCA Control Bank."

a.

WHAT anticipated operational transient could cause a continuous bank

insertion of the controlling bank?

(0.50)

b.

If a malfunction cau~.es a RCCA control bank to insert past the Low-Low

i n ser t i ori 'l i mi t? WHAT [i mmed i a t e op er a t or action'is rehulred? ~ ~

- ( 0. 50 f '

c. If rod control is transferred to Manual and a continucats insertien

condition is still present, WHAT TWO (2) operator a c ti ons should be

performed?

(1.00)

JUESTION

7.04

(2.50)

-

Answer the f ollowing questions concerning B.V.P.S.

- O.M.

AOP-2.6.3,

"Loss

Reactor Coolant Flow."

a.

WHAT THREE (3) symptoms / indications would an operator visually identify

in the control room to verify that Annunciator A2-5E, "Reactor Coolant

Loop Flow Low," was in the alarmed condition?

t1.50)

b.

14 ,a

partial loss at reacter coolant

4 1cw is indicated. totween WHAT TWO

(2) RDS crctectivo intor' '-t

m t i r ~ '. u d : - ' rcte Intsi 1; it ccsoirle 4cr a

reactor trlp to oCCL" '

(1.00/

.

EETION

~ M

(!.u0)

Answer the 'oll owi nc quest t oms ccncerninq D . './ . P . S . - FOP FR C 1

't?ec ron ,o

to Nuclear Fower Generatton/4TWS."

.

WHAT are tne IWU (2/ Indicati;ms that a reactor trip hac NL I occurred?

.

1.00'

.

b.

WHAT are the THREE (3) operator actions that can be taken to shut down

the reactor if a reactor trip CANNOT ce verified?

(1.50)

,

c.

WHY is a turbine trip required during an ATWS event?

(0.50)

eie***

CoILGORv

/ LON!INULD UN NLii PAUL

4****1

1

-..

_

_,-

. - . .

__

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

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

_

. - - _ , __

-

_.

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.

7

PROCEDURES - NORMALt_ ABNORMAL _ EMERGENCY

Page 13

t

A_ N_ D_ _ R_ A_ D_ _I O_ L_ O_ G _I C_ A_ L_ _ C_ O_ N_ T R_ O_ L_

_

,

>UESTION

7.06

(1.50)

a.

WHAT procedure (by name) would you consult if annunciator Al-lE,

"Containment air part'al pressure hi gh-l ow, " alarmed?

b.

WHAT could cause containment pressure to slowly increase wi th little or

no humidity increase, and a possible decrease in temperature?

c.

If the plant is in Mode 2,

and containment pressure, temperature, and

humi di t y" ALL :begi n takincrease rapidly, WHAT. action ~ should the operator

take?

(3.0o)

TOTION

'.07

'.

A condition arises that requires entry into containment at 40% power.

The

operator entering containment needs to work in a gamma radiation field of

150 mrem /Hr for aprroxhamtely 2.0 hours0 days 0 hours 0 weeks 0 months .

The below candidates are

presented to your

Candidate

1

2

3

4

Sex

male

mal e

4emale

male

Age

27

38

24

20

Otr/ exposure

-

1000 mrem

500 mrem

1000 mrom

Life exposure

100C* rem

54730 meem

5200 mrem

c500 mrem

Remarks

quarterly

Form

3

-

history

NnC-4

months

'

.

unava:!able

eava ! at; o

.meegnant

l

Each candidate is technically competant and physi c al l y capable of

performing the task.

All candidates have a co7pleted Form NHC-4 and have a

,

documented current calender quarter exposure history, with the e.v c ep t J c n s

l

for those cand: dates stated above.

Emerge cv 1:mits do NOT apply.

r c-

EACH persen, indicate if you woulo ACCEPT or HEJECT that person to perform

the task based on EXPOSUAE PE DLii r'C"ENT S ONL Y .

Justi+v EACH answer AND

1

include ALL applicable limits.

EST1DN

7. vl3

(:.00)

I

Answer the following question concerning

B.V.P.S.

- O.M.

AOP-2.38.1,

"Loss

of 120 VAC Vital Bus."

1

!

WHAT are FOUR (4) automatic actions that an operator can vi suall y veri f y in

the control room i f power to 120 VAC Vital Bus 1 is lost?

ONLY consider

cafety system actuations.

'

t eee**

CoILGOHV

/ CON'1NULD L;N Nt sI i Wl,L

- e44)

.

.-

l

.

'.

PROCEDURES - NORMAL

ABNORMAL

EMERGENCY

Page 14

'

t

A_N_D__R_A__D_IO_L_O_G__IC_AL__C_O_N_TR_O_L_

i

_

,

i

1UESTION

7.09

(1.00)

WHAT are the normal expected values for Source Range (SR) AND Intermediate

Range (IR) Nuclear Instrumentation an operator would expect to see when

'

verifying that the SR has reenergized after a reactor trip from power?

.

>UESTION

7.10

(3.00.)a

-

.s,

s

4 >

,

-

'

m

-

.

4. . W7

M

_

Answer ' the f ollowing questions concerning B.V.P. S.

- O.M.

2.24.2,

"Steam

Generator Feedwater System."

a.

WHAT action must an operator take in order to prevent a reactor trip if

a Steam Generator ISGi ree: Pu m;; Auto-Step annunciator a.2-ms wttn the

plant at 75% power?

10.50)

b.

WHAT are FIVE (5) indications / conditions that an operator would verify

if a Hi-Hi SG level-trip occured with the plant at 40% power?

(2.00)

(I. So)

JESTION

7.11

'2."^;

Answer the fol1owing questions concerning L1 quid Waste System Operation.

'

,

M.

WirH TWO (2) flowrates (numerical values NOT required) are usec in

ca,1culating h + ? Cooling Tower D1owdown F1ow when the Unit 2

elowdown 'Ias tnutrument f

' '

w

e out et ser. ice. and a 11cu;e

waste discharge is to De made Dy way of the Una

Q tower

. .

b1owdown 11ne?

g

g NAu

o.

Before sampling the contents of the "A" waste drain tank. WHAT action

must be taken by the operator?

Inc'ade any applicable "recautiona v

setpoints or time related values.

( 1.00)

,

c.

WHAT action shoula an operator take

1+

1 oc al -1 i qui d waste process

1

effluent [2SGC-ROIlOO3 hiah alarm actuates AND is vertf led to be in the

alarme: ccrdition"

tv.5v'

i

teee** LND UF CAIL6UHy

1 ee4e*)

,

.

-

.

3.

ADMINISTRATIVE PROCEDURES _CONDITIONSt

Pcgo 15

t

AND LIMITATIONS

.

JUESTION

8.01

(3.00)

Using Attachment

4,

classify the following events in accordance with BV-2

EPP/I-1, Recognition and Classification of Emergency Conditions, AND

justify your answer and any assumptions.

Consider each case separately.

a.

B.V.P.S.

EOP E-1,

"Less of Reactor or Secondary Coolant," is in use.

Pressurizer level in :f f-scal e low and RCS pressure is 1500 psig and

decreasing.

The rea 'or was manually tripped because pressurizer level

could not be maintained.

~ ~

-

b.

A turbine trip from 75% power occured and the reactor did not automati-

cally trip (ATWS).

The reactor remained critical until an operator man-

ually inserted control rods.

c.

A truck carrying Ammonia gas is involved in a collision at the the plant

main entrance.

Gas is leaking from the truck.

d.

An earthquake is registered on-site with the plant in Mode 1.

The

severe ground motion results in the generation of a missile in the

turbine building from the detachment of a LP turbine blade.

(a .co)

JESTION

9.02

'T

~~

using

B.V.P.S.

- Unat 2 Technical specl+tcations. list ALL apolicable

act:cm statements. by ramber. for EACH C 4

the f ol l owi ng ecu1Dment failures.

C o n c 2 'd e r EACH i o : '. u r e

tadesc~~~-tiv.

a.

The 4uel oil trans4er pump for Dtesel Generater 21 has been 4cund to be

inoperable.

A reactor startup is in progreps with reactor power at 1%

and increasing,

b.

RHS Heat Excnanger outlet thermocouples, TE60eA and

B,

have been +ound

to bo inoperabl?.

c.

Centre! room bottled air system pressure :s found to be at 1500 octo.

t REST I ON

8.03

(2.00)

Diesel Generator (DG) 21'.s operability load test is scheduled for today.

The last THREE (3) tests were completed 35, 69 and 102 days ago

respectively.

The plant is at 100% power.

Using B.V.P.S.

- Unit 2

Technical Sp eci f i c at i ons, are DG 21's operability requirements being met?

Explain WHY and/or WHY NOT.

ieeeee IATEGuay

a coNTINULD UN NLLT W4G L

ee***)

_ -

_ - _ _ _ _ _ - _ _ - _

.

$t__@Qdlyl@l6@llyg_E@gggpuB@@t_ggypillgypt

Pego 16

,

A_ N_ D _ L_ _I M_ _I T A_ _T _I O_ N_ S

_

.

i

,

1

1UESTION

8.04

(2.00)

Answer the f ollowing statements concerning Clearances by responding TRUE or

FALSE.

a.

The NSS, NSOF, and the STA (or NCO) all must sign the "Authorization for

Rurmova l From-Service" lines of the Emergency Safeguards Equipment

Clearance Checklist.

~.v,<.

,

.

b. ' Onl yfthe 'NSS needs to sign the Equipment / Radiation Clearance Log for the

cl earanc e to become ef f ective.

,

c.

A Master Clearance can be used to cover maintenance that requires

eculpment to be operated in order to cerform the necessary work.

i

d.

A Caution Tag may be removed by Test Group Personnel without obtaining

the NSS/NSOF's permi ssi on.

>

.!UESTION

8.05

(2.00)

In accordance with B.V.P.S.

Site Administrative Procedure (SAP) 3B,

"Reporting Requirements," utilize tne Code of Federal Regulations provided

to you to determine whether the NRC should be notified within ONE (1) hour

or FOUR (4) hourr ANL 19dicate WHV by specifying

appecpriate sect:en

- -

numbers / letters.

example:

10 CFR xx.xx

(1) (i) (a)

4

l

a.

A

r-trolled l l e _.

e+': uca*

- ' care war deter.inec'tu rac accured at 5

!

~

times the Maximum Permiss1 Die .oncentration (MPC).

!

o.

An Unusual Event is declared in accordance With the Emergency 31an,

j

c.

During a refueling outage, several pipe snubbere that were attached to

the RCS cold legs were found to be Inoperable.

d.

While the plant was in Mode 3.

a Safety Injection signal was generated

,

and an estimated 2000 qallons of RWST water was injected inte ;% core,

r

l

6

l

t*****

CA1,LGul4Y

H CONIINULD UN NLt1 P AUl. 9ees4)

l

.

-

.

. . - .

-.

_

.

..

ADMINISTRATIVE PROCEDURES

L CONDITIONS--------- t

Page 17

9.

AND LIMITATIONS

t

1UESTION

8.06

(2.00)

In accordance wi th B. V.P. S.

O!1 2. 48. 2 Proc edure C,

"Adherence and

Familiarizatinn to Operating Procedures,"

-

~a.

WHEN can an operator take action that departs from a license condition

or Technical Specifications?

(0.75)

f

b.

WHOM, as a minimum, must approve the above actions to be taken?

(0.50)~

..n:

n-

z.

-'

_

c.

Do .non-licensed personnel ever have the authority to take independent

l

action (s) that they deem necessary to place the plant in a safe

,

condition?

Justify your answer.

(0.75)

UESTION

8.07

(2.00)

Match EACH of the f ol l owi ng statements (a-d) with the most appropriate

report listed

(1-4).

a. The oncoming NSS sians this recort signifying that he is assuming

responsibility for the station.

b.

This report contains, in chronol og i ca l order, the times when the

'

Emergency Plan is i mp l e. men t ed and/or rad 10 active effluents are "eleaGod.

'

l

c. This report is signed by the oncoming Nuclear Operatcr 21gni3 ying that

i

n e' I s assuming rosconsit: 11tv +or 51s area c4 duties,

i

d. This report is reviewed to ensure familiarity with plant opera ti ons

!

during times of watch relic + or vacation.

,

Report Number

i

1.

Sh18t Doerating Report

2.

Nuclear Shatt Operting Foreman's Report

'

3.

Nuclear Control Operator Report

Nuclear Operator's Report

-.

.

i

(...**

WTEGCHY

H CONI J NUL D ON Nil t i P%,L

4*e**i

J

1

)

_

.

. _ .

.

,

'

e

.

3.

ADMINISTRATIVE PROCEDURES _ CONDITIONS

t

Pego 18

A_ N_ D_ _ L _I M_ _I T A_ _T _I O_ N_ S

7

_

,

i

I

,

l

2UESTION

8.08

(2.50)

Utilizing B.V.P.S.

- Unit 2 Technical Specifications, state the Containment

Isolation Valve surveillance requirements (including ALL applicable time

<

contraints) for EACH of the highlighted portions of the systems indicated

below.

Assume the Plant is in Mode 3.

a.

Chemical and Volume Control System (Attachment 5)

(1.10)

{

b. Recilircul ati on Spray' System "( Attachmenti's)

(0.70)

'

c.

Containment Area Ventilation System (Attachment 7)

(0.70)

_lETION

8.09

(2.25>

I

Use B.V.P.S.

- Unit 2 Technical Specification Table 3.3-6 and determine

WHAT SEVEN (7) Area or Process radiation monitoring instruments must be

,

f unc t i onal f ollowi ng a LOCA.

!

i

5

i UESTION

B.10

(3.00)

i

!

re- EACH of the J ol l o.si ng statements, determine e ether the Reacter

l

Coolant System (RCS> chemistry has been maintained within the requirements

t

!

apecified by Techn::al Ecec1 4 stations.

Justify ycur answer.

'

-

4,

l

,

a.

ine plant has peer at 100;. power for 21 daye.

Chemistry notifies you

j

that RCS fluoride concentration has increased to 0.20 ppm.

c.

The plant is in Mcce t and 1G being prepared for a startup.

Cnemistry

i

notif tns you that ACE chlor ice concentratton is 2.00 ppm.

I

c.

The elant is stable in "ode

2, when Chemistrv nett 4 i e e.

veu t h.i t

l

dissolved oxygen concentration is 1.50 ppm.

<

c.

ine plant

aa in "cee ; een inemistry notified seu at 100v en 2 / 2 2. - u O

!

)

tnat chlori de cenrentr at i on was 0.25 ppm.

Attempts to reduce the

l

t

chlorice c oncentr at i on were unsuccessful.

The plant was shutdown at-

!

1900 on 2/24/88.

I

1

.

J

1

(8****

LMi GOHY

0 LUN1INUED ON'NE1,i f'Af>L

e****)

4

l

)

-

- -

- - - - - - -

- - -

- - - - - - - - - - -

.

h__B90101916BI1YE_E5gcgggggS _cgyp111gggt

Pcgo 19

t

A_ N_ D_ L_ _I M_ _I T_ A_ _T _I O_ N_ _S

_ ,

,

JUESTION

8.11

(1.25)

c. WHAT is the FULL Technical Specification Basis for the RCS operational

l eakage limi t stated in 3.4.6.2c.?

'O.00)

b.

WHAT Technical Specification (state by number) addresses the surveill-

ance program establi shed to prevent the leakage limits in 3.4.6.2c.

from

bc.ng becoming an operational concern.

[}s

.

l

<**eee

f. N D (2F C O T E (;UU y

U eeoet)

( *********4

END OF f.*AMIts.41IUN *4644444et>

.

,

- .

.

i. .

THEORY OF NUCLEAR POWER PLANT OP

Page 20

ERATION

L

FLUlpS @NQ_IbER[QQyN@[lCS

j

t

,

i

4NSWER

S.01

(3.00)

I

a.

Rod. worth is a function of the ratio of local flux to average flux

(squared). C0.50]

If a rod is dropped with all other rods withdrawn,

'

the dropped rod depresses the local flux relative to the rest of the

core so that its worth is small

(* 200 pcm). CO.753

When a rod is stuck

with all other rods inserted, the tip of the stuck rod is exposed to a

much higher local flux than the rest of the core causing its worth to

increase (*

1000.pcm). CO.75]

'

b. - to malntain' 'a' ~more ~ uni f orm di f f erenti al rod worth

- minimize the possibility of creating a positive delta I

C2 X O.503

- to ensure that any control rod motion will have some

effect on total core reacti vi ty

~rERENCE

B.V.P.S.

LP-RT-8 Enabling Objectives 2,9,11

i

B.V.P.S.

Reactor Tneory Text Cnapter 8 pages 20-24,27

K/A 001000 KS.02 3.4

-

K/A 001000 A2.03 4.2

OO1000A203

OO1000KSO2

..(KA's)

NSWER

5.02

(3.00)

a.

rho = (IS2 steps - 120 steps) > o ocm/ step CO.60]

'

102 oce CO.

'

=

,

sUR

2e

Lrne

lambda / LBeft - rhot] Ev.60]

=

x

-

1

- 26

CIG2 pcm

0.1/sec / (660 ccm - 1925 3 CO.40)

x

4

1.07 DPM (0.20]

=

.

D.

Tave would be nigrer (0.30] since MTC CO.30] is larger CO.202

at EOL CO.20)

ErERENCE

j

B.V.P.S.

LP-RT-5 Enac1:nq Objectives

e.'

B.V.P.S.

L P -R T- 5 cage a

V/A 102003 kl.ve

...

N/A 000003 Ekl.le 3.2

OOOOO3K116

192OO3KlO6

. . t v: A ' c )

.

N

i

j

(*****

C 4 I Lfa O R Y

b LUNIINUED UN NEti Phi >L

!

. - -

.

. - . . - - - - - - - - - - - -

.

. _ .

.

-

.

. .

. _ _ _ ._

.

..

i.

THEORY OF NUCLEAR POWER PLANT OPERATION

Pego 21

t

F_L_U_ _I D_ S _. A_ N_ D_ _T_ H_ E R_M_O_ DY_ N_ A M_ _I C_ S

_

,

1

t

.

s

TNSWER

5.03

(3.00)

a. . Decreases CO.353 because' raising power increases the heat-flux on the

fuel rod, reducing the DNBR CO.403

b.

Decreases CO.353 because the.subcooling margin decreases CO.403

c.

Increases CO.353 because more heat can be absorbed by the water [O.40]

,

d.

Increases (0.353 because the subcooling margin increases CO.40]

,

<EFERENCE

,

B.V.P.S.'LP-TMO-7 Enabling Objectives 11,12

.

B.V.P.S.

LP-TMO-7 pages 21,21,23,26

2

K/A 193008 K1.05 3.6

193OOOK105

..(KA's)

i

l

I

,N ERJ ER

5.04

(4.00)

1

Y

a.

See attached ECP calculation,

b.

Rod worth of Bank D at 75 steps

-910 ccm CO.253

=

boron ch ange recui red

CIO ( a c .? ) /

-9.9

Cp:-/ ppm)

=

93 ppm (0.253

=

Usi ng nomogr apn CB-31:

PPM boron in coolant = 993 ppm

PPM boron accit!on =

72 pen

boric acid volume

'700 gallons (0.25]

=

r e,cu i r ed rate r

700/2/60 e S.O gpm [0.253

OR Using nonograp- CB-32:

!

DRM ooron in cool 2,t

993 ppm

=

baron addit:en rate =

93/2 = 46.5 ppm /hr CO.25)

boric acid flow

5.5 apm (0.25]

=

.

m "ENCE

,

t.V.P.S.

LP-RT-G Ema"ling OD!octive

-

D.v.P.5.

OM :.5v.4

-

t. / A 192008

).t

a. c

r/4 194001 A1.00 0.1

i

19aOOIA108

102000r.1<.

..(An's,

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LAIEGORY

5 CONilNUED ON NEAl P4bL e44et#

,

i

1

.

. - _ . . - , . -.

.

. - ,

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

- - , - - - . - - - . . , - ,

-

. - . - . . .

.,

.

ATTACHMENT 1

,

.

3.V.P.S. - 0.M.

1.30.4

,

F.

EST.ATED CRITTCAL POSTTIcN CAIcALTION

FORM ECP-1 (Page 1 of 7)

1

NOTI:

Reference Guide in Chapter 49, Section 6, Procedure M.

Tav: Assumed to Equal 5477 2 IF at startup

j

A.

CRITICAI. DATA

.

F1101 TO SEUTDOWN

EXPECIID CRITICAI.

-

Date* h 3 J Time

0400

Data l j 23/ 88 Tim. J 00

I

)

3eren Cone

900

pp pever 100

seren Cene.

m

i

Zanen eauil.

Xenon

%

5amarinn eqyiI.

5amarina

%

l

Control Rod Position:

A

228

C

228

a

228

C 223

1

246

D

426

3

228

D

75

(

L.

3.

REaguv A u IAI.ANCI

I

II

III

-

.

Reactivity

Prior to

expewted at

Difference

Defect.s

Shutdeva

, criticality

I-II

0

pca

(5)-1400

Pc2

1. ?cwer (Tis.

1400[0.2y

30-7)

[0.20]

.

2. Control Rod.s

H tr. 50-8)

or Boren (Fig.

0

Pc8

910

Pc2

(2) +910

Pc2

-

"

30*10)

[0.25]

[0.20]

J"

2300

P*"

(*) -550

P**

3. Ianon

2850(0.25

f0.251

f0.201

610(0.25j

740[0.25]

)+130[0.2[

6. samarium

r

.

5.

Reactivity Chang. (sus of 1 4) =

(t) -910

P**

.

ISSUI 2

'

REVISION 1

-35-

__

.__.

_ _ _ _ . _ _ . _

._

, . .

. . .

.

..-

.

._ _

.

_ - _ ._

_

_ _ _

.

,

.

'

3.V.P.S.

0.3.

1.30.4

F.

ESTI'd.ATED CRITICAL PCSITION c1Lct:I.AT!0N (continued)

FORM ECP-1 (Page 1 of 7)

NOTE:

If Reactivity Changs 1.s greater than 2500 pcz. perforz I/M plot.

Table 30-1.

.

'

C. * CRITICAL 30R(34 CONCINI1ATION (Use if critical baron concentration is

desired.)

.

'

!

_!!.

121

17 ~

Y

,

toastiytty

Seron Worth

Baron Change

Aeron conc

loros conc

,

Change (3-5)

(Fig. 30-10)

I divided by II

at shutdova

for startup

, III + IV

(2)-910

PC*

-9.8

EE () +93 00.27

900 P

993[0.2bf

.

O.

C15TICAL 10D POSITION (Use if critical rod posindm is desired)

l

'

,

.

'

I

II-

III

IV

Ratetivity

Raactivity due to

Reac:tvity for

C:stical Rod

Change (3-5) Rod Prior to Shutdown

Criticality (!+II)

Position (Tig. 30-8)

,

(Tig. 30-8)

.

-

(:)

pc=

pcz

(:)

pc:a

steps

'

-

N/A

i

e

)

.

'

.

! $ 5t.T. 2

.

RIVISICS 1

-36-

- - - - -

-

,.- -,,

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5.

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~ ~ ~ E_ E_0_i_ B_ E_ E_ F_

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ANSWER

5

ANSWER

'2.005

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-

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v1 ' :

5

'

the ci

t.

X O.503

b.

50% pt

-

c.

TRUE

m.,

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+

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xenon

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'.,

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be led REFERENCE

,

1EFERENCE p

= B. V. P. S.1 IJh-TMO-3 Enabl i ng . ObJ ec t i ves 4,7

^

"

B.V.P.S. LP-THO-3'pages 8,12

,. . . _

B . V . P . S .~ .

4 K/ A :191006 $1' 03n2. lea

t,,

.

B.V.P.S.

K/A 191006 K1.04.2.7..

K/A 0010i

K/A 191006 K1.07 2.6

M/A 1920(

191006K107

191006K104

191006K103

..(KA's)

192OO6K1(

ANSWER

5.D6

(2.50)

INSWER

5.

1) core exit TCs - stable or decreasing

C5 X

,

a.

Incred

2) RCS hot leg temperatures - stable or decreasing

b.

Decrea

3) RCS cold leg temperatures - at saturation for existing S/G pre ,

c_

r CN

4) RCS subcooling (based on core exit thermocouples) - greater th '

u

d. No Chi

subcooling par attachment (7)

5) S/Q oressuree - stable

c-

decreasing

i EFERENCE

NMT

M h 01 6 9&1h *Io 60 ' h

REFERENCE

B . V . P '. S .

B.V.P.S.

LP-TMO-7 Enabling Object 2ve 16

K/A 192O(

B.V.P.S.

EOP ES-0.1

"Reactor Trip Response." Attachment 5

192OO2K1)

K/A 193008 K1.22 4.2

193OOBK122

..(AA's)

'

'JSWER

5.

ANSWER

5.07

(1.00)

increasec

core coo:

curing plant cooldown Co.50]

neutron r

temocratore OR cooldoven rate L .25:

ancrease

pressure LO.25)

REFERENCE

B.V.P.S.

Unit 2 Technical Specifications page B 3/4 4-7

-

K/A 193010 K1.07 4.1

193010K107

..(KA's)

(*888*

CATEbuRY

S LONIINUED UN ,NExT PAGE *eeee)

-

.

tt__IHEg61_gF_ NUCLE @g_EgWE3_EL@N1_g[EB@IlgN t

Pago 24

F_ L_ U_ _I D_ S_. A_ N_ _D _ _T H_ E_ R_ M_ O_ D_ _YN_ A_ M_ _I C_ _S

_

,

<EFERENCE

B.V.P.S.

LP-RT-6 Enabling Objectives 2,9

B.V.P.S.

LP-RT-6 page 6

K/A 192004 K1.13 2.9

192OO4K113

..(KA's)

4NSWER

5.11

(1.50)

a.

TRUE

b. FALSE C3 X O.50]

c.

TRUE

EFERENCE

B.V.P.S.

LF-TMO-4 Enabling Oojective 8

9.V.P.S.

LP-TMO-4 pages 6. 7

K/A 191004 Kl.11 2.4

K/A 191004 K1.12 2.7

-

K/A 191004 K1.15 2.8

191004M115

191004K112

191004K111

..(kA's)

.

l

t ee***

f.f JD Ur

ie6'!ttDfY

b 8880ei

..

_

.

,

PLANT SYSTEMS DESIGN

L CONTROL------ L AND INSTRUMENTATION

Page'25

S. .

.

-k

~

(2.6o)

q

)NSWER

6.01

'!.00?

O.

NOT AFFECT

b.

NOT AFFECT

c.

INCREASE

o. DECREASE

CO.50 X 63

'O.

INCREASE

!

'. NOT ^TT CT

Tf

,

?EFERENCE

a

,t,.

g.

B.V.P.S.

2LP-SQS-1.1 Enabling Objective 6

,

B.V.P.S.

2LP-SQS-1.3 Enabling Objective 10,12

i

B.V.P.S.

2LP-SQS-7.1 Enabling Objective 7

,

B.V.P.S.

2LP-SGS-21.1 Enabling Obj ec ti ve 4

B.V.P.S.

- O.M.

2.01.I pages 12,20; 2.7.1 page 3S;

3

2.21.1 page 22:2.6.1 page 64

6

!

D.V.P.S.

- Unit 2 Technical Speci+1 cations table 2.2-1

K/A 001050 K5.01 3.6

I

K/A 004010 A1.01 3.6

-

K/A 041020 A3.02 3.4

!

041020A302

OOa010A101

OO1050KSO1

..tkA's)

i

!

l

i

!

-NSWER

6.02

(3.00)

'

i

,

i

a.

the high positive trip protects against a roc ejection accident (0.50)

'

sm;.1e t e ntg-

e;ative trip c,oter:c age "st a 'oc drec accice-t L:.bO2

I

,

RP

~

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S]

a m -* ?R

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1 0 , rf l rec

0"

r;; . i, ' . $ ;- )

t

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ca tor trt

signal ts automat 1:2.

v olocked LO.50] are tre Sh

'

n i c. n vol tage is crevented 4r0* enerc2:tng CO.SO)

+

,

j . ' EilL N C.E

'

i

!

f.). V . P . S .

LP-SCE . 2 EnablIra 00tectave 4

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)

11. V . D . S . ELD 5 ~."

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%/4 0150v0 N1.Oi 4.1

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t

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c.03

(2.50s

1

4

a.

C0.503

b.

CO.75)

Use attaened drawing

c.

(0.303

as answer key

I

d.

CO.753

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PegG 26
.
.
lEFERENCE
- '
B.V.P.S.
2LP-SOS-13.1 Enabling Objectives 2,4,5
'
B.V.P.S.
OP. Manual Fig. No. 13-2
K/A 194001 A1.07 3.2
'
194001A107
..(KA's)
!
4NSWER
6.04
(2.00)
'
.
c. Decrease CO.503
b.lloW CO.503
,
c.
Thermowell RTDs has e a relatively long response time CO.503
d.
A CO.503
ECERENCE
B.V.P.S.
LP-TMO-7 Enabling Objective 5
B.V.P.E.
LP-TMO-7 page 11
,

N/A 191002 K1.13 2.0
'
K/A 19100'. K1.14 2.9
-
.
191002K114
191002K113
..(KA's)
4
NSWER
6.05
(2.00)
g
Als folit
c2$4
a.
none
c.
closes 2HVR* MOD 23A and 2HVReMODa4G ' pp
cc'-
.;1.z na as -mmms.abic:-
,
c.
a c,t u a t e s control room pressurt:atten
c.
nmc
ta o o.503
,
-:FERENCE
i
B.V.P.S.
2LP-SQ5-43.1 Enabling Objective 4
,
D.V.P.G.
2LP-505-42.1 pages 16.21.24,'9
"
N/A 072000 GO.04 3.7
.
072OOOGOO4
(*A'>'
.
4
J
a
1
!
s t.,E R
6.06
(3.vv>
a.
2 S/Gs (0.25] at low-icw level CO.2b]
S1 signal present (0.503
TDAFWP running CO.253 and prossure low CO.153 after T/D CO.103
both MFWPs not running CO.253 and either of the NFWPs control switches
in Afterstart to.253
b. 2/3.CO.253 RCP bus undervoltage CO.253
2/3 CO.103 detectors in 1/3 S/Gs (0.153 at the low-low level C O. 25.1
i
iss***
Lt4 I E GOR Y
6 CONilNULD UN NEA1 t 's t bl
4e***
.
-
-
-
-
-
-
-
-
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.
.
-
-
-
.
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. -
-
-
-.
..
.,
t__PL8Ml_@l@ LEOS,QESigyt_Cgyl69Lt_gyQ,1GS169dEGl@llgy-
Pcgo 27
.
..
7EFERENCE
B.V.P.S.
2LP-SQS-24.1 Enabling Objective 16
B.V.P.S.
2LP-SOS-24.1 pages 16,18
K/A 061000 K4.02 4.6
K/A 061000 K4.06 4.2
061COK406
061000K402
..(KA's)
iNSWER
6.07
(3.00)
cuta-actions
backup heaters turn ON CO.503
'
"" '
flow control valve 2CHS-FCV122 goes to i ts mini .num open
posi t i on CO.503
indications:
level devi ati on al arm
-

one
( .)
channel of ' S tr:c status lignis tor hign
FRZR level light
- comouter alarm
- high level trip alarm annunciator
C4 X O.50J
- high kevel indication on level meter
- high level indication on level recorder
flow control valve 2CHS-FCV122 indication at minimum open
-
nesstier
EFERENCE
B.V.o.S.
2LF-005-6.2 E"3bli"9 Objectives
9.I'
B. V. P. S.
- OM e.e.1 page o3 5 figures 6-39 e-34
/A 011000 Kl.O! 3.9
E/A O'11000 V1.04
7. 0
K/A 011000 K3.01 3,4
K/A 011000 KS.13 3. 4
h/A 011000 A2.10
I. e
011OQOA210
v11000*513
v1;o00K301
Oi2000&104
011000K101
..(KA's)
SWER
5.08
t ?.00)
a.
prevents pump c a v i *. a *.2 c a
4 rom occuring L (' . .o v )
c.
prevents accumulators trom ceang anoperable L L' . 50 ]
c. prevents the val ve 's motor operator from trippina on thermal overload sc
that the valve will reach its designated safe position CO.50]
d. prevents pumping contaminated sump water into the RWST CO.503
.e**ee
im ! t v ass
i; t UN T I Nt.'L L' UN NLsi I i eu t.
eei
'
.

.
, . '_ _P_ L_ A_ N_ _T _ S_ Y S_ _T E_ M_ S_ _ D_ E S _I G_ N_ L _ C_ O_ N_ _T R_ O L_ t _ A_ N_ D _ _I N_ S_ _T R_ U_ M_ E_ N_ _T A_ T _I O_
Pc9o 28
-_
_
__
_
_
_
_
.
.
?EFERENCE
B.V.P.S.
2LP-SQS-9.1 Enabling Objective o
B.V.P.S.
- OM 2.11.1 pages 14,18,19,23
K/A 006000 K4.06 4.2
K/A 006000 K4.09 4.1
K/A 006000 K4.19 3.4
OO6000K419
OO6000K409
OO6000K406
..(KA's)
TNSWER
6.09
(2.00)
1.
metal-water reaction between the zirconium fuel cladding and the reactor
cool an t
2.
pressurizer gas space and RCS water
3.
radiolvtic drcomposition of water collecter om the containment floor
with a c c r r e s; Cn d i r. g gc9eration of oxyger
4
radiolytic cecompostion of water in the reactor core
[4 A O.50]
3.
corrosion of metals by solutions used +ce emergency coolire or
containment spray
.
-EFERENCE
B.V.P.S.
2LP-SOS-46.? Enar!ing Oc ecti e
l'
9.V.P.S.
- DM 2.46.1 cage 1
K/A 028000 K5.03 3.6
02GOOOK503
..(KA's)
'
GWEw
6.in
'2._-
a.
> 700 pslo CO.50]
o.
TWO (2s :
.;b] charging cumps tv.;b.
'
.
at the relle4 va;ve set D " c O "> u r P C O. b> J
c.
RHS neat exchanger CO.35]
RHS pump seal cooler [0.35]
d.
Decreace IO."12
'7EhENCE
Enaoling Coject
.nt L% vo ! L o BL E
E<. V. P. S.
- OM 2.iv.1 pages
1.
.b.o.Ju.21
A/A 000023 K1.O!
K/A 000025 K3.02
K/A 000025 A1.01
OOOO25A101
OOOO25K302
OOOO25K101
..(KA's)
..** i 1:D U F ,L e4i L L.U U v
e.
i
. -
-
.
.
7 .*
PROCEDURES - NORMAL t_A; NORMAL _ EMERGENCY -
Pcgo 29
t
AND RADIOLOGICAL CONTROL-
.
.
.
.
.
> r:%

,f-
ANSWER
7.01
(1.50)
6. place steam bypass l'nterlock' selection switch to the. DEFEAT TAVG
' -position CO.503
b. prevent rea: tor vessel void formation (maintain RCS subcooling) _ [0.503
c.
B.V.P.S.
-~E.O.P.
ES-0.2, "Natural Circul ation Cooldown" to.503
9EFERENCE
J S.lV.n S.,f2LP-5GS- 21'.1'55 nit'1's_ng .'Oslec t i vos '. 4's .
Y"***
' * -
.
2LP-SDS-SO.51.52.~.1: Enabling' Objectives 2,3
B.V.P.S.
- 0.M.
2.51.4 pages C9,D2,D4; 2.51.2 page 3
2.53C.4 page 3
K/A 005000 GO.10 3.5
K/A 005000 GO.15 3.9
K/A 041920 A4.09 3.1
041020A408
OO5000G015
OO5000G010
..(KA's)
a
ANSWER
7.02
(2.50)

a.
FALSE
b.
TRUE
c. FALSE
C5 X O.50]
d.
TRUE
e.
FALSE
1
'EFERENCE.
B.V.P.S. 2LP-FHP-1.0 Enabling Obj ec ti ves 8,9.12
)
B.V.P.S. 2LP-FHP-1.0 pages 4.28.31.44.47
K/A 034000 A1.01 3.2
K/A 034000 A3.01 3.1
.
K/A 034000 GO.07 3.7
034000 GOO 7
034000A301
034000A101
..tAA's)
l
l N5WER
7.03
(2.00)
a. turbine runback (OTdt or OPdt) OR loac rejection (0.50]
b. omergency boration OR baration at concentration and flowrate atleast
,
that as stated in Technical Specifications CO.50]
c. trip the reactor [0.503 and go to E-O CO.503

l
l
teeeea CATEGORY
7 CONTINUED UN NLAT PAGE 4eoe )
i
,
1
--
.
_ . . _ - . - _
1
'
-
+
.
i
?t.- 669EE996EE_: 09Bd86t 8BN9Bd861 50569E091
j
Pcgo 30
]
AND RLDIOLOGICAL CONTROL-

,------------

,

.g{
T
"1EFERENCE
'
B.V.P.S.
- O.M.
53C AOP-2.1.3 page 1
I
B.V.P.S.
- 0.M.
1 page AAM1
K/A 001000 A1.04 3.9
, i
s
K/A 001000 A3.02 3.6
s
'
OO1000A302
OO1000A104
..(KA's)
.
,g,
,
.
-
~ .

,
i
yep,
2
A
P
%,<,
y
,jK @
Iggg g
, , ,g egg
.
,
d. RCP A,B,
or C bright' lights illuminated
'
> ^
t
Low RCS flow indication in Loop A,B,
or C
C3 X O.503
high delta T
b.
P-B C O. 25 3 30*/. [0.253
'
P-7 CO.253 107. [0.253
.
1EFERENCE
B.V.P.S.
2LP-SOS-53C.1 4nabling Objectives 1,4
B.V.P.S.
- 0.M.
ADP-2.6.3 page 1
,
'
K/A 000015 GO.11 3.6
OOOO15G011
..(MA's)
i
,
45WER
7.05
(3.00)
a
a.
r o,d bottom lights NOT lit (0.503
neutron flux NOT decreasing CO.503
D.
manually trip the reactor
manual l y insert control rods (3 x 0.503
]
initiate emergency boration
.
c.
to prevent excessive cooldown of the RCS CO.503
<EFERENCE
B.V.P.S.
2LP-SOS-53A.1 Enabling Objectives 2.3
B.V.P.S.
- EOP FR-S.1 page 1
K/A 000029 EK3.06 4.3
'
K/A 000029 GO.11 4.6
OOOO29G011
OOOO29K306
..tAA's)
i
,
I
i
4
+=
.
i
9
-
a
.
J
'
!
t*****
CoTEGORY
/ CONIINUED ON NEAT PAGE 4eeee)
,
l
.
.
.
.
.
.
.
7'
' PROCEDURES - NORMAL d BNORMALa_E_MERGENCY
Pcgo 31
!
A_ N_ _D _L_ A_ D _I O_ L_ O_ G_ _I C_ A_L_ _ C_ O_ N T R_ O L_
_
__ _
me
p
,
TNSWER
7.06
(1.50)
c. Coss of Containment Vacuum (AOP-2.12.1) CO.503
J
b.'a breach'of (leakage into) containment CO.50)
c.
mcnually trip the reactor CO.503
.c.
REFERENCE
B.V.P.B.' 2LP-SQS-53C.If Enabling Objectives 1,3
'34Vp
"S.y ROiMi( AOP-2.~12/17 page"1#'
K/A.
29 EA2.01 4.3
K/A 000029 GO.11 4.2
OOOO69G011
OOOO69A201
..(KA's)
ddA " ^^)
(s oo
g . , ,ll w e
, m
mredccra Sc Wvd=
"="
01 - De tem CO.253 sinee h:
'2 : m: ;u
-t:-1,
" nt; y avc;Iabl: :-"
>s+4--
e"r rd t h e 20^ - r
'-t -
tr! r bcdy '. . -- ! : CO.503
C2 - REJECT CO.253 since he does not have a Form NRC-4 available and would
exceed the 1250 mrem /qtr whole boev li':
?. O . 5 0 3
C3 - REJECT CO.253 since she will exceed the allowable epoosure limit
,
during the term of her pregnancy CO.50]
Qc.o
e/(C 6Bti
/
04
^CCF"'
CO.253 since he
ill
et c :cer
'~r
au:-tr '
-
'
' "
N whol e body 1imit of 10000 .mee+
1.fet: e expe r e to.502
b
<EFERENCE,
Enabling Objectives UNAVAILABLE
B.V.P.S.
- R.C.M.
pages 5.6.7
K/A 194001 K1.03 3.4
'
,
194001K103
. . (K A's )
NSWER
7.00
(2.00)
l
1) atmospheric steam dump valves +a: 1 closen.

ren
2) letdown will isolate
id
v . 'z
l
j
>
3) PRZR heaters will deenergize
4) standby service water pump (25WC-P21A) autc starts, sf not alreaas
running
5) component cooling. water to containment instrument air compressor closes
6) primary component cooling water supply and return isolation valves
(2CCPaMOV175-1,176-1,177-1,178-1) close
(*****
CATEGORY
7 LONIINulD liN NL\\l 6 'e n d ' ******
-
- .
- - -
- -
-
-
- - - -
- -
-
-
-
-
-
-
- . -
- -
.
.
F i. ' -PRO'CEDURES - NORMAL _Ag@RM b ,, EMERGENCY
Pcgo 32
t
A,ND_R_A_D__IO_L_O_G__IC_A_L__C_O_N_TR_O_L_
__
_
,
.-3.,
fEFERENCE
>
B.V.P.S.
2LP-SQS-53C.1 Enabling Objective 5
B.V.P.S.
- 0.N.
AOP-2.3B.1 pages 1,2
K/A 000057 EA2.19 4.3
OOOO57A219
..(KA's)
,
,
-.m
n,
4NSWER
7.09
(1.00)
? g ;ci d F
,k mj'ggjpgg!gg,gj.
i&-
-
'
v
~
IR: J1E-10 T82-- O.SE 10) amp s - ( C O. 503
1
~
.EFERENCE
B.V.P.S.
2LP-SGS-2.2 Enabling Objective 4
B.V.P.S.
- 0.M.
2.2.4 pages B4,C1
K/A 000032 EA2.04 3.5
OOOO32A204
..(KA's)
NSWER
7.10
(3
gefuce. re.00)w
g" Q O
f
g x o.co]
[
m QW[
.
o. place the SG Startup Feedwater Pump i n servi ce {0. "O1
b.
- turbine trip
- main feedwater pump (MFWP) tripped
- MFWP discharge valves closed
C5 X O.503

- UFW Reg valves closed
- SG Bypass fIow con:rel valves e1c'oe
o R.
Feefwder-
r.s o d} ion
- MFW isolataon trip valves closed
5ERENCE
.
B.V.P.S. 2LP-SOS-24.1 Enabling ODJectives 7,9A(14)
B.V.P.S.
- 0.M.
2.24.2 pages AAE1
N/A 000054 GO.09 3.1
K/A 000054 GO.10 3.2
OOOO54G010
OOOO54 GOO 9
..(KA's)
,
l
,
,
(e****
CAILGORV
7 CONTINUED UN NExT PAGE
e)

i
- - - -
-
-
-
.
.
.
-
.
.
- - -
- -
-
-
- - -
- -
- -
-
-
-
-
-
- - -
-
-
,
.
.-.
.
.
7.*
PRdCEDURES - NORMAL _ABNORMA6_ EMERGENCY
Pcgo 33
t
A_ N_ D _ R_ A_ D_ _I O_ L_ O_ G_ _I C_ A_ L_ _C_ O_ N_ _T R_ O_ L_
_
,
.
I . SDb
,
ANSWER
7,11
(2.50'

,30,- M
>
n- a , . . , - ,
,c--31_;
.c,_.s
y 2,
,,= :n
e .=::
U-it 1 err!' ; t:.;;r 512.d: -
'!:.
0. ,0 ;
.
.
b. recirculate the tank C0.503 for a minimum of TWO (2) tank volumes OR 8.5
hours CO.503
c. verify closed (C2SGC-HSV-1003) liquid waste EFF high rad isolation
valve CO.503
2.2 # L
~
?'" '
"" I'
'Y
' ' " '
IEFERENCE' '
'
'
'-
E
.
B.V.P.S. 2LP-SOS-17.1 En4bling Objectives 2d,9,5e
B.V.P.S.
- 0.M.
2.17.2 page 1,
2.43.4 page AEE1
K/A 000059 EA2.02 3.9
4 000059 EA2.05
3. 9
OOOO59A205
OOOO59A202
..(KA's)
,
.
.
4
f
a
j
(**eee END OF CATEGORY
~ eeeoe)
/
.
,
_ _ - . - . _ _
,
_ _ _
- _ .
. . . - - - .
. - - - -
- - , _ . _
._
. _ _ _ - _ _ _ _ _ _ _ _ _ _ _
"
.
.
, -3. . ' ADMINISTRATIVE PROCEDURES

L CONDITIONS--------- t

Pcgo 34
AND LIMITATIONS
,---
y----------
4
4NSWER
G.01
(3.00)
,
,
c. CITE AEIA CO.403
TAB 5 -- RCS/ Containment leak exceeds make-up capacity CO.353
b. ALERT CO.403 TAB 14 -- Reactor not suberitical,.after valid scram
.
signal (s) CO.353
c. Unusual Event (0.403
TAB 18 -- Toxic gas nearby release potentially harmf ul CO.353
d.
ALERT CO.403._..
.
,_
.
,
.
., ,, , -

_
,
JTASf28;-- Turbine'ruptuFe causing casing pen,etratt .on CO.353-
-
-
.n,s
graders award 1/2 credit if event is classified more' conservatively
award full credit if clasification is also properly justified
.
iFERENCE
Enabling Objectives UNAVAILABLE
B.V.P.S.
Unit 2 Implementing Procedures BV-2 EPP/I-1 Table 1
-
K/A 194001 A1.16 4.4
-
194001A116
..(KA's)
<
o't, . 6 0
NSWER
8.02
-f3.vv;-
'
g
3,
7.g_:
. ra e
g ,
gms
,n
en,
em * n -
_
r1,mmet
enti- g ct,-t
7
c,-
g
y:- ;;r :st ct: g
m;;r; c ,-
c 1, i , y
.mi.m.,
m . w,,m. , u ; CO.50]
.,
b.
3 . ,3 . 3 . 5 . (remote shutdown monitoring) C1.OOJ
c.
3.7.7.1.b
(control root habitattitty:
4.7.7.2.a specifies cressure
requirement c4 1825 pstg) L1.00]
{
EFERENCE
l
.
Enabling Objectaves UNAVAILABLE
B.V.P.S.
- Unit 2 Technical Specifications
B.V.P.S.
- O.M.
2 page 2.10.1
N/A 0e2000 GO.05 3. 8
K/A 016000 GO.OS 3.b
,
016000G005
062000G005
. . t h .1 ' s )
i
i
,iNSWER
8.03
(2.00)
I
1
.
q
N3 CO.503 each test is within 25% of the required time interval CO.753 but
tho THREE (3) consecutive combined test intervals exceed 3.25 of the
roquired interval LO.753
I

(*****
CATEGORY
B CUNTINUED ON NEAT PAGE eoeso)
'
J
f
,
..
-
--
_ , _ ,
..
. . - _ . _ . . _ _ _
,,
. - . . - ,
_ _ _ ,
-c
, _
.
.

').'
ADMINISTRATIVE _PRQCEDURE@t_CQNDITigNgt
Pcgo 35
ANQ_Llgl!QIlgNS
,

-
)
tEFERENCE
Encbling Objectives UNAVAILABLE
B.V.P.S.
Technical Specifications 4.0.2
K/Q'0640OO.GO.05 3.8
03C050GOOO
..(KA's)
.
4NSWER
B.04
(2.00)
QQ Q3pqugfiT.)?'
-
l'~

'
N
'
'
~
'-

,
~P
' b . -. F AL SE
.
c. FALSEc
[4 X O.503
d.
FALSE
EFERENCE
,
Encbling Objective UNAVAILABLE
,
l
DLC SAP Chapter 41 pages 17,47,50; Chapter 42 page 6
K/A 194001 K1.02 4.1
-
194001K102
..(KA's)
.N5WER
8.05
(2.00)
a.
4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 30.72 (b) (2) (l v) (B) CO.502
4
b.
I hour 50.72 (a)(i) (0.50)
c. 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 50.72 (b) (2) (1 ) CO.503
d.
I hour 50.72 (b?(1) (2v) C0.502
dFERENCE
.
Enabling Objective UNAVAILABLE
!
B.V.P.S SAP 3B page
4,
Appendix E
10 CFR 50.72
,
v/A 194001 A1.06 3.4

194001A106
..(K4's)
NSWER
8.06
(2.00)

c. during an emergency CO.253 when this action is immediately needed to
protect the public health and safety CO.503
-
b,
a licensed senior operator [0.503
c. yes CO.253 but only in the event of an emergency or casualty not covered
- by an. approved procedure tO.503
~
1
l
1
9
]
(***ee
CATEGORY
8 CONTINUED,C' NEAl PAbt ee4e*i
l
,
4
-
- -
. _ .
_
_
_
_
_
-
.
.
j
,3.'
ADMINISTRATIVE PROCEDURES
CONDITIONS
P go 36
t
t
-AND LIMITATIONS

.
.
a
-
. . .
~:,
$IEFERENCE
'
i
Enabling Objectives UNAVAILABLE

8.V.P.S.
- OM 2.48.2 page 8
,
.
K/A 194001 A1.02 3.9
f

. v.
1940014102
..(KA's)
i
i
- 4NSWER
B.07
(2.00)
-
3. ff,kg,!$g ,,,yy
liy
' %sy 6:1 '
NN
WQlg& lff
Q;- 9W"k&%
0,
'
1C4 X O.503
b.
i
c.
4
d.
2.
'
EFERENCE
,
i
Enabling Onjectives UNAVAILABLE
8.V.P.S.
- OM 2.48.5 Procedure A,
"Logs and Reports," pages 3.5,6
,

K/A 194001 A1.06 3.4
-
194001A106
..(K4's)
,
I
J
i
!
.NSWER
8.08
( 2. 5ut
'
a. cycl e val ve 2CHLb A0V204 CO.20] in less than 60 seconds CO.203 anc val ves
2CHSsADV2OO,A.B.C CO.2OJ an 10 seconds CO.20] through one complete cycl e
of full travel CO.20]
j
b.
c y'c l e val ves 1 SS * MOV 155 A.156 A CO.2OJ to the open cosition CO.20] in
less than 60 seconds CO.20]
c. cycl e val ves 2HVR* MOD 234,8 CO.20] through one complete cycle of full
.,
I
travel CO.233 in 10 seconds CO.203
'
j
.
ALL the above valves must be cycled aeleast once per 92 day 2 CO.3v]
EPERENCE
'
l
i
j
Enabling Objective UNAVAILABLE
j
B.V.P.S. - Unit 2 Tecnnical Soccafications Section 3/4.6.3 Table 3.e-1
'
t/A 103000 K4.06 3.7
'
N/A 103000 GO.05 4.1
'
103000G005
103OOOK406
..(KA's)
i
1
!
'
\\

,
i
1
I
4
4
4
(s****
CATEGORY
8 CONTINUED ON NEAT PAGE
)
.

4
l
1
.
.
.
-
--
-
.. . _ - _ __
.
.

L CONDITIONS--------- t

ADMINISTRATIVE PROCEDURES
PC93 37
1. . *
,b9 LIM 1101190E
@
r
ANSWER
8.09
(2.25)
b
,-
Go7
(
,
'
2RMR-RQ395*,206 (Containment Area) CO.SO3
2HVS-RQ109C (Mid Range Noble Gas) CO.50]
2HVS-RQ109D (High Range Noble Gac) CO.503
2 MSS-RQ101A,B,&C (Main Steam Discharge) CO.753
?EFERENCE
'
"
R
15.,V.P.S. 2LP-SQS-43.1 Enabling"Dbjective'4:
.
..
s f ,7
1
B.'V. P. S.
- Unit 2 Technical Speci fi cati ons Table 3. 3-6 ' Action- 36

'
K/A 016000 GO.04 3.4
016000 GOO 4
..(KA's)
.
b[
NSWER
8.10
(3.00)
g
g
a . 44Mr C O. 25 3 b ec aus e BCS fluoride concentration is Prss-than the 6"=m=irnt
limit CO.50]
i
b.
Yes CO.25J because the Technical Specification is not applicable in
'
Mode 5 CO.50]
'
c. No CO.25] because the concentration exceeds the transient limit CO.503
.
,
d. No 00.253 because the LCO action statement was not met CO.503
'
ECERENCE
'
Enabling Objective UNAVAILABLE
!
B . V . P '. S . - Unit 2 Technical Specif ications Section 3/4.4.7
I
/A
194001 A1.14 2.4
194001A114
.. WA's)
i
.
75WER
B.11
(1.25)
>
r
a.
ensure that the cosage contributton CO.20] from the tube leakage will be
j
1.mited to a small 4raction of the 10 CFR Part 100 limits (0.203 in the
(
event of either a steam generator tube rupture CO.20] or a steam line
-'
break [0.203
b.
3/4.4.5 Co.45]
1
1
!
'ses**
CATLGORY
O CONTINUED ON NEAl PAGE eeoe: )
,

i
l
.
.
.
.
.
. .
.
.
.
.
.
.
.
.
..
.
-
.
-

-.
.
., 3.'
ADNINISTRATIVg_ PROCEDURES , CONDITIONS
P go 39
l
g
t
AND LIMITATIONS
i
4
!
..
4
.
g
(EFERENCE
!
'
!
Enabling Objective UNAVAILABLE
'
,
B.V.P.S. Unit:2 Technical Specifications Section 3/4.4.6.2,'3/4.4.5
i
K/A.000037 Go.04 3.9
.
0000370004
..(KA's)
<
i
,
-
1

.
I'
!
5
.-

3
m,s . N; s
.,
. . . . ,.
-
.+:
.

,
...
y
.
i
".
?
-
1

1
e
i
j
.
i
,
,
t
-
,
i
'

,
,
!
'
.
t
i
i
l
i
.
i
i
'(
l
1
i
I
1
q
i
I
.I
l

,
.
j
l
i
l
i
(e**eo END OF CATEGORY
B eeoee)
L**eoeeeeee END Or ERAMINATION eeeeeeeeee)
!
-
,

.
I
1
._.
,
.
.
.
.
.. .
,
,
f = ma
v L's'/ O y'
'. '
Cy'ct'e'effic1hei = Y(het wori' 90 '
out)/(Energy in)
'
2
'
w = ag
.s = V ,t + 1/2 at
E = sc
A * A * xt
.
KE = 1/2 av
a=(Vf - Y )/t
A = 1M
o
g
PE = agn
v = e/t
A=.sn2/h/2=0.493/t1/2
Yf = V, + at
t
eff = C(tw;)(ts))
-
-
'""'#*
8j
-
ijg
[(h/t * IhI
I
A=
-
.
,
,
.
.
'E * I
"
A
-DC
m = Y,y ,
j,
~*
Q = kpat
-*
I * I ' ux
-
j = UAaT
o
I = 1,10'*O
p, = w th
7
TYL = 1.3/u
-
sur(t)
HYL = -0.693/n
P = P 10
p = p e /T
t
o
SUR = 26.06/T
SCR = S/(1 - K,ff)
CR = S/(1 - K,ffx)
x
SUR = 26p/t= + (a - p)T
CR)(1 - K,ffj) = CR (I ' Ieff2)
2
T = ( t*/a ) + (( a - o V Ia ]
M = 1/(1 - K,ff) = CR)/G,

T = V(o - s)
M = (1 - K ,ffe)/(1 - Kefft)
T = (o - o)/(Io)
SDM,= (
- K ,ff)/K,ff
e - (X,f f-1)/K,ff = AK,f f/K,ff
t' = 10
seconds
I = 0.1 seconds'

o = ((L'/(T Keff)] + (I,ff (1 + IT)3
/
il
Ig j = 1 d2 ,2 2
P = (t*V)/(3 x 1010)
g4
gd
1
22
2
I = oN
R/hr = (0.5 CE)/d (ceters)
R/hr = 6 CE/d2 (f,,g)
Watec Parametars
y
Miscellaneous Conversions
-
1 gal. = 8.345 lem.
1 curie = 3.7 x 1010dps
1
a; . = 3.78 liters
" kg = 2.21 les
'
I
t* = 7.48 g al ,
hp = 2.54 x 10 Stu/nr
Oensity = 62.4 lbm/f t3
1 m = 3.41 x 10 6tu/hr
Oensity = 1 go/c.M.
lin = 2.54 cm
UO Itu/1tn
F = 1/5'C + 32
HC3C of v4cortration
a
Heit of fusion . 12 3:u/1t,
C = 5/9 (*F.32)
1 A t.m = 14. 7 051' = 29.9 in. 99
1 BTU = 778 ft Ibf
1 f t . H 0 = 0. 4 3 3 5 I t.f / i n .
7
.
. - -
_ _ _
._.
__

.
>!
ATTACHMENT 1

t
i
,

3.Y.F.3.
0.M.
1.30.6
.'
i
F.
ESTIMATED CRITICAL POSITION CAIcJLATION
CtP
roRM ECP-1 (Page 1 of 7)
NOTE:
Reference Guide in chapter 69, Section 4, procedure M.
$1
Tavs Assumed to Equal 347T 2 LF at Startup
[
%7
[kw .'
-
s.
en m en. a m
.
,ua To _a
- cameu.
-
-
-
-
.
,
Date*h3_88 Time
0400
Date D h .88. Tim. 1000
'
i
.
Beren conc
900
ppe Power 100
seron cone.
ope

.
"
Zanon equil.
Xanon

.
.
'
5amarium equi 1.

5amarium

.
.
~
Control Rod Positioni
control Rod Position:
A
228
C
228
A
228
C 223
1
446
D
246
3
228
D
75
.
(
.
.
-
3.
REA uvA u IAI.ANCI
I
II
III

Reactivity
Prior to
axpewted at
Differesen
Oefects
Stutdown
criticality
I II
pcs
pcs
($)
pcs
1. Pcvar (Tig.
-
-
30 7)
2. Control Rods
Qlt. 50-8)
pcm
pcm
(2)
pcs
or Baron (Tig.
-
-

30 10)
pcm
pcm
(2)
pea
3. Iaaen
-
-
pcm
,-
pcm
(2)
pcm
6. Samarine
-
.
3.
Reactivity change (sus of 1-4) =
(t)
pea
'
-
.
.
,
e
155ct 2
RIVIS:CN 1
i
'
-35-
l
.
.
.
--
.
. .
._
- , - - - ,-,
-
.
- . - - .
.__ _
-
.
. - - _ .

'
.
.
'

3.V.P.S.
0.3.
1.30.4
,,
,
T.
ESTIMTTD CRITICAL PC3m0N CALCUL4 TION (constaned)
FORN ECP-L (Fase 1 of 7)
,
'
WII:
If Reactivity Change is greater than 2500 pcm. perfors 1/M plot,
Table 30-1.
,
I
C. * CRITICAL SQRCM CCHCINERATION (Use if critical boron concentration is
'
desired.)
.
I
II
!!!,
17
Y
, ,
a.
-v
m.
<
,
Change (3-3)
(713 30-10)
I divided by II .
Beren Cens
Baron Cosa
Raastiyity
'Beros Worth.
Beres Change
,
at shutdown
for startup
'
III + IV
,
(2)
pca
ps (2)
ppe
ppe
ppa
ppe
-
' O.
CliTICAL ROD POSITION (Use if critical rod posit m is desired)
!
d
,
'
'
!
II-
!!!
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(
Reactivity
Rasctivity due to
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Change (3-5) Rod Prior to Shutdeva
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Wm CRITICE P051 TION CECUIaTICN (constanad)
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ACD LD1IT5
TORM ECP-1 (Page 3 of 7)
1
II
III
IT
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' Espected Red
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Rod Position
Aod Position
Defect as cris
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for III from
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Fig. 50-4
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MAIIBCM RCD EIGHT Toit CRITICALITY (Itea E IV) = 3ank N/A at
N/A
steps

Mnt=n= Rod Height, insert rods
It' criticality is not achieved by/A
NOTE:
to (Item E-7) bank
N/A as
N
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MINIMCM RCD E(G37 TCR C1ITICEITY = 3ask C at 115
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'
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ATTACHMENT 3
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.,
.
Er*/Iy Implementing Procedure
Ett/I-1
,
Accognition and Classification
-
or [mergency Conditions
,

ACflott t[ VEL CRITElllA FOR CI AS$1flCAT10ll of LIEW Coel180lls
INITIATING
UNUSUAL
ALERT
S
M
CONDITION
EVENT
E
Easti
Y
Orr-normal Events
Events ldhich Involve
Events idhich
Events a re in progress
Which Consid Indicate
or have occurred which
Actual or LIWy
involve Actual
e Potential Degrad-
Involve an scLuel or
or Fallures d
er feelnut
potentist substantial
Plant functiu s W
Matential Com
ation or the level
W m tectiu M N
kg Mat h w M e
or Safety of the
degradation or the

Plant-
M ilc.
ig W th Potuttal
level or sarety of
the plant.
for Less of
%, g
g g
,ggy,
,
,!
!
1
Cadioactive [ffluent
Unplanned airborne
Unplanned airborne
Release Correspeede
Redfelegical effluent
release gives orrsite
release gives erralte
to >20 aree/hr. et
-
release results In
ApollC! big _j
dose rate greater
dose rate greater
Site SounderY
effelte dose preJacted
M ga3r to
nt a
than 0.5 mrem /hr.
than 2.0 mRee/hr.
-er-
te exceed 1 roe
gnd Result ne free
Offsite Dese Due
to the nStole Sedy
any initletine [ vent
to Event is
or 5 rem to the
-or-
-o r-
Projected to Exceed
Chlid Thyrold.
Unplanned liquid
unplanned liquid
170 orea to ldhoto
release in excess
release results in
Sody or Child Thyreld.
or MPC limits,
downstrese community
water redleectivity
, greater then 12 times
and/or
[PA standards.
TAB 1
Celease or toss of
fuel Handling Acel-
Plejor Desege te Spent
Redlelegical effluent
i
Control or Radioactive
dont Resulting in
fuel Due to Fuel
corresponds te greater
'
Materias Within the
Release or Radlemativ-
IlandlingtAccident
~
body dose rate er
then 125 enee/hr. whole
Plant.
Ity to Occupied
-or-
.
Areas Sucn That the
Uncontrolled Decrosse
6430 ellas/hr. child
Direct Radiatten
in Tuel Fool lister
thyreld at ties site
I
tevels in the Areas
to Selow, Level
beesadery.
increase by a factor
of fuel.%

L
7
or > 1000
1;
"
-or-
Other Veri fied, lIncen-
~
troIIed Events idlich
l
-
Result in en Unexpected
4
.
tocrease or in-Plant
D. rect Radletten Levels
by a f actor or > 1000.
,
i
-
.
6
.
.
.- -
_ . _ -
_ _ _
.
--.
- .
.
[PF/IP taptementing Procedure
(PP/I,
Recognition and Classa rication
~f
'
of toergency Conditions
-
ACTION LEVEL CalTERIA FOft CLASSIFICATieu of ElmiLEGY MITIGES
INITIATING
UNUSUAL
ALERT
SJ _
-
CONDITIO7L
EVENT
EME
Y
,
_
Reactor Coolant system
Below Tech spec
Lees er 2 or 3 Fission
(RCs)
t)=iting Conditions
Predeact sorriers With a
temperature Low
for operation (tCO)
Potential tess er Ihlrd
terrier.
usen,_se amqnitinies
ROS Pressure High
[ =cceds LCO Limi t
t May Legg ta
'
l'"
TAB 4
-or-
Asur Initleting Events, from
,.
RCS/ Containment teak
facceds LCO
[=ceeds 50 gpo
Exceeds Make-up Cepeelty
18 estover Source thst Makes
Rolesse er Large Amounts
i
.
TAB
5
of Radleectiv8.ty In a
sn.rt use. Pro .mi..
PCS/ Secondary Leak
[=cceds LCO
> 200 gpa
350 spe w/ MSL
TAB
6
-o r-
are.k w/indic. tion
i. toCA with r. :=r. .r (CCs.
Main 5tese Line
>10 gpo w/ ptSL Break
of fuel failure
2.
LOCA Wlth Smit 8elly Succ-
areak er Aspid
Depressurization
-o r-
-o r-
essful ECCS. Subseaguent
'
Faifure er Isost Aemove8
'
or Secondery Side
F
)
MSL Break w/ IWlV
> 1000 gpa
Systems with Likely -
Fallure er Containment.
s
' ' " " ' *
TAB 7
~
s.
to.s er Aii onset. .nd
Of f s i te Powe r Concurrent
13f
fues Cladding
RCS Activity [wceeds
RCS
1 Activity
Degraded Core-Pesslble
With Total toss er
Degradation
100
-or-
> 300 uCl/ge
toss or Coolable
[me rgency f eedwa te r.

Reactor Cooient
Geoestry.
Monitor Als ra, or
4.
Less er feedweter and
or analyses 1 uCl/
Condensate followed by
ga, 5teady State
failure er Emergency
TAB 8
r dwet.e systes.
RCs serety or
t eak t =ceeds 100 ar
S.
Reacter Protectlee Systee
neller valve
Valve Inoperable
Falls to In8tlete er
Complete a postul red Scree,
j
isiture
,
rollowed by Less er Core
TAB 9
Co iin
.nd n.ke-u, syst.e
-or-
<
RCS f eeperature High
t=ceeis LCO
-
Less of Plant Control Occu
!
TAB 3
-
i
PCS Pressure Low
Be lo.a 100
TAB 4
~
-
.
{
!
~
7
1
. -
-
-
-
-
.-- --_. -
-
.
- .
[FP/IP Implementing Procedure
EPP/l-1
fleregnition end Classificetten
"
.
cr toergency Conditions
-
-
.
ACTIoe LEVEL CRS TERS A FOM CLAS$1f_JCATIOLOF tMLP. ft%v C0308 TIOst}
-
INITI
URQ0UAL
N $ ALERT
SITE AREA
GENERAL
COND
EVE.R
/
EMERGENCY
EMERGENCY
- e
snitiation er ECCS
Volld Seroty Circuit
". ~W
toss or 2 or 3 rission
j
rruduct Barriers with a
Trly or lessessary
o
flottese l Initiation.
<
8'otential Loss er Third
TAB 10
- '~
ea r ri e r.
CJ7&d*
egem sniana ,Amr leitiatin>
cco Poe,r.Ilure
'
'"' to
TAB 11
f r. i s ure
[;iN,jffga
"
,
-or-
,
tocs er Centalsesent
Requiring Shutdowes
Containment P re s su re-
Any Initleting [ vents, free
Integ ri ty
by LCO
M ar <8 $ psig
Wha teve r Source tha t Ma kes
Release of La rge Amour.ts
TAB 12
_
or Radio.ctivity In a
short Time Prebeble,
tocs er Engineered
ftequirlseg Sleestdown
f or [xample:
Safety or fire
by LCO
Protection restures
1
LOCA With failure or ECCS.
,
TAB 13
2.
toCA With initi.fiy Socc-
'
essful ECCS.
Subsequent
Folture of Roseter
-
Reactor Not
Fallure or 60est Removal
L
Subcritical arter
Systees with likely
Protection System
r
Valid Scree
Isilure of Containment.
to f altiete or
,
Complete e Scree
l
Signaits).
TAB
14
-
3.
Loss or All Onsits and
,
"
orrsite P-r Concurrent
With Total Less or
Locs er Plant
Loss of CopeblIlty
loss or Capability
Emergency reedwster.
L
to Achieve Cold
to Achieve flot
Control /3efanty
7
Simstdown
Shutdovre
as . Loss o.7 f eedwa te r a nd
Cyctees
Condensate rellowed by

TAB 15
r.i sure er toer,ency
Icedvetor System.
Loca er Indicators,
Less en Process
Loss or All
Loss or At8
annuncletors or
er Errlueset Pers-
Alarms ( Anttuncle-
Alarms 1$ min
S.
Reactor Protection System
c.scres
meters, Regesiring
tors) Sustained rer
with Ptsnt Not in
Is8ts to Inftlete or
Shutdowet by LCO
3 ntns.
Cold S/D
Complete a Requi red Scras,
-o r'
f oI loved by Los s or Core
Plant Transient
Cooling and Stoke-up Systems
Occurs Whlle Afi
-o r-

Ala rms a re Lost.
Loss of Plant Control Occurs,
TAB 16
1
Control Room
Reeguired or Antl~
Required. Shutdown
I
tvecuation
cipated. Control or
System Cont'ol at
'
Shastdown Systees
'itemote Shutdown ranel
'
"
Established at
- Not [stablished
Reseote Shutdown
Within 15 min.
"*'-
'
_ JAB 17
- -.
.
- - - _ _ _ _ _
_ _ _ - - - - _ -- ---
i
.
EPP/IP Implementing ProcCCuro-
t r P/ 8 - :

Recognition and Classification
-
'
or Emergency Conditions
'
ACTION LEVEL CRITERI A FOR CLMSIFICATION Of ENERGENCY CMITIOlli
i
~
INITIATING
UNUSUAL
ALERT
SITE AREA
GENERAL
CONDITION
EVENT
EMERGENCY
EMERGENCY
-
4

i
Toxic or Flame-
Nea r-by or On-Si te
Enters Foollity.
Enters Vital Areas
Loss of 2 of 3 Fassion
k
able Cases
Release Potentially
Potential Hobit-
onel Restricts
Product Sarriers With a
i
Hereful.
ability Problems.
Necessary Access.
Potential Loss or Third
TAB 18
a. rri e r.

Security Compromise
in Accordance with Security Plans
Ismeinent Loss or
Ugg JL lo Any inlded.cg
i
Physical Control
gg, & t May Lesd tg
!
or Plant,
s
gg.
TAB 19
'
-or-
1
Any Initiating Events. from
l
oss or On-Site
Loss:or Capability
Whatever Source that **mkos
!
Release of Large Amounts
1
AC Power
of Radioactivity in a
.
Short flee Probable.
TAB 20
TeePorary toss or
toss or soth
ror E as,ie:
'
Loss or All Orr-
Upon Occurence
1.
LOCA With Failure or ECCS.
Sito Power
i
2.
LOCA With initia.lly succ-
TAB
20
.s.rui rCCS.
S.
s.quent
Failure or Mesi Removal
I
Loss or All On-
,-
Power For More
Fallure er Containment.
,
Upon Occurrence
Loss or Vital DC
Systees With Likely
I
Site DC Power

than 15 eins.
I
TAB 21
3.
toss er Ait ansit. and
Orraito Power Concu'rrent

Tornado or Other
Warning. Probable
Strikes Vital
Winds in Excess
With Total Loss or
.-
High Wind
Errect on Station.
Plant Structures.
or Design Levels
Emergency feedwa ter. ,
'

II . Loss or Feedwater and
!
Flood or Los
Flood <705 root
Flood >705 feet MSL
Flood > 735 feet MSL
Condensate followed by
l
Water
MSLi Requiring
-or-
Failure of Emergency
5/D. Law Water
Dasege to Vital
Feedwater System.

<LCO.
I
Equipment.
.
i
!
-

TAB 23
3.
n
ter crot.ction syst.e
.
_
.
- Falls Le lattiste or
l
Earthquake
Detected on Site
Creater then OSE
Cruiter thah SSE
Complete a lloquired Scree,
i
Solsmic Instrumenta-
Occurs
Occurs'
Followed by Loss of Core
'
tion.
Cooling and Make-up Systees
TAB 24
,
-o r-
i
Loss of Plant Control Occurs,
Fire
Fire within protected
Potentially Affecting
Arrncting Safety
area ' lasting more
Safety Systems.
Systems Required
'
,
i
than 10 minutes,
for Shutdown.
.
1
.
'
TAB 25
.
s

.
l
9
'
.
.
,
i

.
-
.-- .
. - - -
.
-
_ _ - - _ _ . -_--
'I
l
.
.
EPP/IP leptements- Procedure
E PF/ l
-
i
Recognition and C .sst rication
-
or Loorgency Conditions
IN!TIATING
UNUSUAL
ALERT
S "E ARE A
GENERAL
j
CONDITION
EVENT
EN El1GENCY
EMERGENCY
I
to
Severe Damage to
Loss of 2 er 3 Fisslode
l
-
Explosion
near or On-site
^
Mnown Da y, fracting
Facility
Sare Shutdown
Product Barriers With a
l
Explosion Potential
Sigelficant Damage
Ope ra tion.
Equipment.
Potentles less or Ihlrd
1

TAB 26
'
e. r r l . r.
*? '" f/, C J. ., M 8 *4 h 1Mc*
Musclel! AckifttyW w CfAl d i
M. <wi CrYoh .Arrect s ,VI,ta l .
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_Lged to

Over Facility
f rom Whatever Source
Structures by

l
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this gge
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Aircrert Crashes
cantly Degredes a
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Station Safety
Any taltisting Events. True
Wha teve r S,enerce tha t Ma ke s
!
L
Structure.
j
TAB 27.
.
..le.se .
t. ,,e
o..ni s
or Radi ctivity in a
1
Short ilme Probattle,
j
i
train
Derelleent in ensite
For Ememple:
Areas

1.
LOCA With Fal!ure or ECCS.
q
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Stribes intake
ess fiel ECt.S . Subsequeent
l
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f a l lsare of Hea t R':mova l
In Flow Reduction
systems with likrty

TAB
28
re s i re of Contelament.
,
Contmeinsted
Tradsportation of
3.
toss of' All Onsite and
~
!
Injest y
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Offsite Npwor Concurrent
j
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to GFfsite Hospital.
.
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4.
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Rep @te of Pipe-
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Falleste of Emergency
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5.
neoct r fret.etlen syst
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I*I'* ** I"'"3*** *'
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Tustine rotatino
Turtine 'fellure
j cesponent rallure
causing casing
Q uplete a_
Ired scree *
lurbine Rupture
i
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rapid plant
penetratten
felleved by Le . or Core
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Coollag and peake-up systems
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g gm
g
rall'urs of onc t/G
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toss of Plant Control Occur
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S/C Tube failure
.
l
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Tube with lede or
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c
Orrsite peuer.
with loss of
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n

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,
Nuclear Group
P O Bon d
Shippengport. PA K07 7-0004
Februa ry 24, 1988
'!D 2 V PN : 5350
'!.
G . i ! l .. .
'li.
h.i.
-
t i.i.
Ope t a t i ci,
' ranch
.
I; ; s .
o,
i neac:
a i t e t '.
l! . 5.
Nuc lea r Regula tory Commission
Fea:on !
-
Allendale
ld
ir
ine ef Prussia, PA
l ' < 4 ti
-
Rett renc,
Beaver i.' 1 1 ' . v Po .' r S t a t i e ". , l'n i t u2
l b s c 'r et
,i_,
i 1 ; '. t ' ' .' i f 74
'
Ijr e'
'x4-in1:
Repor
>
ir Mr. G.i l ' t
,
Please find enclosed c ome.e n t s generated by cur Training Section
as3ectated -
r
the aritten examination adninisterce et bruary 23, 1988 at
Jur !
If you have any questians concerning this report please contact Mr.
T.
W. Burns at (412) 393-5751.
-
i
Very truly yours,
,
[ /\\ ,/ \\ ,h .
g.
D.
S iebe r
_
{
/Vice President Nuclear
l
JDS/ cal
Enclosure
cc:
T. W Burns
Central File (2)
.
.
.
._ _
(*
,-
H

.
.
j
'
QULS110N~ 5.03
(3.00)
for EACH of the primary parameters listed below, state'll0W (Increases,
Decreases, No Change) and explain WHY an INCREASE in that parameter
affects the DNBR. Assume the other parameters remain constant.
a.
Reactor Power
b.
Tave
-
c.
Core flow
'
d.
Pressurizer Pressure
,
' ANSWER 5.03
(3.00)
,
a.
Decreases (0.35) because raising power increases the heat flux on
'
the fuel rod, reducing the ONBR (0.40)
'
b.
Decreases (0.35) because the subcooling margin decreases (0.40)
c.
Increases (0.35) because more heat can be absorbed by the water
(0.40)
d.
Increases (0.35) because the subcooling margin increases (0.40)
.
REFERENCE
B.V.P.S. LP-TM0-7 Enabling Objectives 11,12
B.V.P.S. LP-TM0-7 pages 21, 22, 23, 26
K/A 193008 Kl.05 3.6
193008K105
..(KA's)
!
COMMENT:
4
5.03.c
!
Better heat transfer should also be an acceptable reason for DNBR
increasing with increasing core flow, as can be seen by equation from
Attachment 5.03.c with * increasing.
l
i
-
-
- -
- -
-
-
-
-
. -
-
-
.
.
-
-
- -

.
-
,
,
.1
'
,
a
L ,
I

>
'
= s/t'
Cycle efficiency = (net work
i = .tle
v
out)/(Energy in)
Z
,,q
s = V,t + 1/2 at
Z
E = ac
A " # * At
-
'
A * 18
0
KE = 1/2 av
a = (Vf - V )/t
o
" **"
. ,e
. . ./t
i = ta2/ti/2 = 0 683/t1/2
yf,
2
1/28#I * b(t1/~'}( h}
W = v aP.
NO
(( g ,2 } + (t ) 3
g,
1
b
e.E = 931 m
$=Y,yAo
,ge
-Ex
b
. ux
I=Ie
6 = M4 T
g
I=I
10-*
Por = W .h
o
)
f
TVL = 1.3/u
j
p = P 10 ,. ( )
Hyt = -0.693/u
sw
,
o
p = p e /'
t
SCR = S/(1 - Keff}
SUR = 26.06/T
CR = S/(1 - Keffx)
x
SUR = 26o/t* + (s - o)T
CR)(1 - Keff1) = G (1 - eff2
2
T = (t=/s) + ((8 - oV Isl
M * I/II - Keff) = CR /G
j
o
T = 1/(o - a)
M " (I ~ K$ffo)/II ~ Keffi}
T = (s - o)/(Io)
SOM = (
- K ,ff)/Keff
10
a = (Keff-1)/Xeff = dXeff/K,ff
{=0.1secondsmondj
.
A =

eff (1 + IT))
o = (( t=/(T Keff)3 + (I
/
Ibl1*Id2 ,2 2
P = (tav)/(3 x 1010)
Id
gd
j3
22
2
R/hr = (0.5 CE)/d (meters)
g , og
R/hr = 6 CE/d2 (f,,g)
Miscellaneous Conversions
water Parameters
s' -
-
10eps
](curie =3.7x10
1 gal. = 8.345 lem.
kg = 2.21 lbm
1 gal. = 3.78 liters
3 8t /hr
1 ft4 = 7.48 gal.
1 no = 2.54 x 10
Oensity = 62.4 lem/ft3
1 m = 3.41 x 106 5tu/hr
Oensity = 1 gm/cr3
lin = 2.54 cm
j
Heat of vacornation = 970 Itu/lem
F = 9/5'C + 32
,
seit of fusion = 14 3:u/lem
'C = 5/9 (*F-32)
t A c = 14. 7 o s i = 29.9 i n .
.-q .
1 Biu = 778 f t-lbf
1 ft. H.0 = 0.4335 lbf/in.

.
.
.
.
.
QUESil0N 5.05
(2.00)
Answer the following statements concerning Heat Exchanger Operation by
responding TRUE or FALSE.
a.
Once turbulent flow in a heat exchanger has been established, VA
becomes approximately a fixed value.
b.
If the AT across a heat exchanger is not constant then
AT , the median (average) temperature, is used to accurately
caiculatetheheattransferrate.
c.
The heat removal rate for a heat exchanger will increase if either
of the fluid flowrates through the heat exchanger is Increased.
d.
The U-tubes of the steam generators can experience thermal shock if
the feedwater flowrate is increased rapidly.
ANSWER 5.05
(2.00)
a.
TRUE
b.
FALSE
(4 x 0.50)
c.
TRUE
-
d.
TRUE
REFERENCE
B.V.P.S. LP-TMO-3 Enabling Objectives 4,7
B.V.P.S. LP-TMO-3 pages 8, 12
K/A 191006 Kl.03 2.3
K/A 191006 Kl.04 2.7
_
K/A 191006 Kl.07 2.6
191006K107
191006K104
191006K103
..(KA's)
COMMENT:
Part a. asks if VA will vary or not for a heat exchanger with turbulent
flow. Operators monitor flow, pressure, and delta T for heat
exchangers.
VA is not something that can be monitored, nor is whether a
heat exchanger has laminar or turbulent flow.
This question goes beyond
the knowledge required of an operator.
K/A 191006 Kl.03 requires
knowledge of "Basi
heat transfer in a heat exchanger". We ask that this
question be withdrawn.
The statement is also incorrect since fouling
would cause UA to vary once turbulent flow is established.
Part b. tests the knowledge of the proper name for the symbol AT
This is a minor point.
The accepted method of calculating heat bansfer
across a heat exchanger such as a steam generator is to use the average
temperature (i.e. Q = UA (T
ThisisaccuraEOenoug$)forourpurposes.
a
st
, not the log mean
-T
'
This knowledge
temperature.
is not a good measure of an operator's ability to safely operate the
,
plant and we ask that the question be withdrawn.
- - -
-
- - -
.'

.
.
.
l)UIS110N 5.06
(2.50)
WilAT are FIVE (5) indications that natural circulation has been
established after a loss of offsite power occurs.
ANSWER 5.06
(2.50)
1)
core exit TCs - stable or decreasing
(5 x 0.50)
2)
RCS hot leg temperatures - stable or decreasing
3)
RCS cold leg temperatures - at saturation for existing S/G pressure
4)
RCS subcooling based on core exit thermocouples) - greater than
subcooling per attachment (7)
5)
S/G pressures - stable or decreasing
REFERENCE
B.V.P.S. LP TMO-7 Enabli,9 Objectives 16
8.V.P.S. E0P ES-0.1, "Peactor Trip Response," Attachment 5
K/A 193008 Kl.22 4.2
193008K122
.(KA's)
COMMENT:
Less than or equal to 60 F temperature difference hot leg to cold leg
should also be an acceptable indication of natural circulation in the
answer key, as can be seen in attachment 5.06 E0P background document on
natural circulation.
~
-
-
.
.
.
We#".
,e
I
.Ikk

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n
<
.
BVPS - E0P
2.53B.5
-
Executive Volume
[
Natural Circulation
'
'
include natural circulation verification.
The steps that verify natural
circulation flow are included in the E0Ps after SI flow is terminated.
If
the SI system is in operation, natural circulation flow is not verified
since with SI on there are more important steps to be taken and the SI
flow
may affect the indications used to confirm natural circulation.
If natural circulation flow based on the symptoms listed in the attachment
is not verified, then the E0Ps direct the operator to increase steam dump
flow to tt, to establish verifiable natural circulation flow
The following symptoms are used in the Natural Circulation attachment to
verify natural circulation flow:
A.
RCS subcooling based on core exit TCs should be greater than instrument
inaccuracies.
B.
The core exit TCs, RCS hot leg temperatures and SG pressures should be
,
decreasing slowly with time, as core decay heat falls off.
C.
Vith
SG
pressures held relatively c o ns t ant ,
the RCS cold
leg
temperatures should remain relatively constant at or slightly above the
i
saturation temperature for the SG pressures being maintained.
.w-
~ , . . . ,
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~

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h ~%ertot-uM1Rmeg- semperature-dif ferenco -sbould'bi~~ rBritaTEYf,
""** effet *td"the *ftM 14eweO fsidaeCJconendtidssteperatureif
Tee M e
B.
The core exit average temperature
(core exit TCs averaged reading)
should be higher than the average cold leg t'esperature. This avera:ed
reading should also decrease as core decay heat falls off, in step with
core exit TC, hot leg temperature, and SG pressure readings in all
active loops.
,
To facilitate the verification of transient equilibrium attainment in the
'
natural circulation process, the operators should start to record these
parameters at regular intervals beginning as soon as instructed in the E0Ps.
The continuous recording will provide trending information on the parameters
of importance in order to eliminate the effects of pointwise variations in
the readings and minimize the chance of misinterpretation of any one set of
readings.
Variations in discrete readings, and between the same parameters
Ln different loops, can result from several causes*
Asymmetry in the heat transfer and heat transport processes between

loops.
Instrument inaccuracies.

Difference in instrument sensing element placement between loops.

Variations in feed flowrates to steam generators.

!
i
PAGE 3 0F 8
ISSUE 1
REV 0
_
_

.
.
.
.
(JUISTION 5.08
(2.00)
a.
Do xenon oscillations converge (dampen) more rapidly at BOL or

E0L? Justify your answer in terms of reactivity effects,
b.
Would the magnitude and frequency of xenon oscillations be less at
50% power or 100% power? Justify your answer.
ANSWER 5.08
(2.00)
a.
E0L {0.25) the negative power coefficient of reactivity tends to
dampen the oscillations (0.50).
1his coefficient is more negative
at E0L (3 25).
h.
50% power (0.25) the lower neutron flux at 50% power does not
produce xenon as fast as at 100% power (0.50) since the rate at
which xenon is produced is slower, the magnitude and frequency of
the oscillations will be less (0.25)
-
,
REFERENCE
B.V.P.S. LP-RT-7 Enabling Cbjectives 5,6
.
B.V.P.S. Reactor Theory Text Chapter 6 page 51; Chapter 7 pcge 17
K/A 001050 A2.06 4.0
K/A 192006 Kl.06 3.4
192006K106
001050A206.
..(KA's)
COMMENT:
When xenon oscillations occur at Beaver Valley, the oscillations are
plotted for a period of time, usually 24-hours before an attempt is made
to dampen them.
This is done to determine the frequency and magnitude of
the oscillation.
See attached procedure 2.49.4.G.
As is apparent from
the procedure, operators do not have to estimate magnitude or frequency
on their own at different plant conditions.
Since-these oscillations are
slow, and since they are plotted to determine their magnitude and
frequency, the knowledge asked for in this question is not a good measure
of whether or not an operator has enough knowledge of xenon oscillations
to control them. We ask that this question be deleted from your exam
bank and replaced with a more operationally oriented question.
i
i
- -
- -
- -
-
-
- -
-
- -
-
.
\\ _ c .V

.
.
it . V . P . S . - 0.M.
2.49.4
Issue 1/ Revision 1
,
-(
Pan., G1 of 2
G.
DAMPENING AXIAL XENON OSCILLATIONS
.
PI'RPOSE
The purpose of this procedure is to provide a means of dampening
ax2a1 xenou est:!;n tens ar.d thus help stabiliac the reauter cere
.
AL CONDITIONS

%t Appliuble
oJM:TTIONS
CAUTION:
u
IT APPEARS THAT THE TECHNICAL SPECIFICATION LI.MIT ON
~ ' ' ' ~ . ROD INSERTION L!"2TS OF AX:AL FLC. DIFFERENCE
KILL BE VIOLATED.
THEN IT VILL EE NECESSAFJ TO REDUCE
1
,-
._
..
-..w-
m,e
r. . m. . ,.- - -- . .: --
u
. . . . .
.
. .
.
NGTE-
Utiline the delta
flux
(40) channel that is the most
restr:ct:ve channel w::t respect to targe.
f l u:: when
pictting xencn oscillations.
When any channel is within
1. 5 *. of target
limit,
al
four channels
must
be
cont: .;cus:. monitered.
Plot deltc flux (40) vs.
Time for approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or
.
for a time pericd sufficient to determine the frequency of
the axial estillatien and the midpoint about which the oc
escillates.
(Eest results are achieved Jhen rods are > 215
steps on D1.
1.
JSing the
C u r '. * .
predict when the pe3k at the ICp cf Core
will occur.
5 - ._ 2 :
.t i gu r r. ~9-5
for an exarple.
3.
At approximately
1-1/2 hours before tne peak at the top of
the core, record 40 and commence
inserting control
rods.
When a ac corresponding to the cidpoint of the oscillation is
achieved, then maintain constant control rod position.
4
Find the difference between the 60 recorded in step 3 and the
midpoint determined in step 1.
This difference is called
E.
5.
Delta Flux will now drift towards the bottom of the core.
i
When it reaches a value of E below the cidpoint, withdraw
control
rods
to achieve the midpoint oc again.
The
oscillations should new be dampened.
1
e.
Ccnt :.ae
te riot 40 for several hours.
If the oscillatiens
resumes as determined by a repeat of stop
1 above,
repeat
steps
through 5 until the oscillation has been reduced to
acceptable limits.
i
i
j
_
_
- -
_
_
_.
__
_ _ _
r
.
.
[
.
.
.
.
b . \\' . I ' . S . - U . fl .
.'.49.4
1:
.u.-
1, in v i:. inn 1
.
i
P.ini G2 of 2
G.
DAMPENING AXIAL SENON OSCILLATIONS (Continued)
REFERENCES,
.:
NOTT. -
All
references used prior to March 10, l'18 6 a r e !ccated
in Section 5.
1
ills
U."CN 2-87-cOS (Kev 1)
!
i
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e
h
.
!
3
V
k
,
-. _ _,-_..- .-. - ..
_ --.._ _... _ .- .,_~....- - _ _..~.__ _ -
.
-
.
.
.
.
.
.

.
.
'
.
QUI Si l0N 5.09
(2.00)
for LACH of the following statements below, state HOW (Increase,
Decrease, No Change) ACTUAL Shut Down Margin (SDM) would be affected.
a.
The plant is in Mode 5 when a charging pump is mistakenly started
resulting in the injection of 200 gallons of boric acid into the
RCS.
b.
The plant is in Mode 3 when all the shutdown bank rods.are
withdrawn: cut of the core.
c.
The plant status changes from Mode 5 to Mode 4.
d.
A control rod drops into the core with the plant in !!ade 1 at 50%
power.
The reactor does not trip.
ANSWER 5.09
(2.00)
a.
increase
b.
Decrease
(4 x 0.05)
c.
No Change
d.
No Change
1
'
REFERENCE
B.V.P.S. LP-RT-9 Enabling Objectives 2,4
B.V.P.S. LP-RT-9 pages 4-7
K/A 192002 Kl.14 3.9
_
192002K114
..(KA's)
C0KMENT:
5.09.b
The answer for 5.09.b could be No Chanae if the BVPS Technical
Specification (T.S.) definition of Shutdown Margin is applied.
Refer to
Attachment 5.09.b & c for T.S. definition.
Please consider possible
interpretations of the question during grading.
5.09.c
The answer key should be INCREASES since the mode change from 5 to 4
would result in a higher moderator temperature.
With~a negative
moderator temperature coefficient, this would result in the addition of
negative reactivity.
Refer to Attachment 5.09.b & c for T.S. definition
of Shutdown Margin.
-
- -
-
- -
-
--
t
1
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l
IId!!h)dP71555'25""I
'
y,w
,a y _r.
s*-vs
^1-e t.$ n.%b.:*- :me ---- . - ~ _
.
se.c - .
.ws
_e
.
_ _ .
- ., ::iu:s.-
i, r- : v r
- . 4 ,g .
.
__
.
Clcsec ey manual valves, blinc 'langes, or ceactivatec aut:-
matic valves secured in their closcc positions, except as
previcec in Tabie 3.e-1 of Speci fication 3.6.3.1.
' . . E. :
C'
vi: ment natches are :lesed anc sealec.
,
.
1
's $$551E'.E !u :a:*t *
3:eci#':a;# r 3. i. ' . 3. ,
..;.
E a
'. * * ' : s
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_.e...
e a s .2 . -
.
....
.-c.
e
'3
-
.
.
. .. .
.
3CeCifiCatien 3.5.1.2.
1.3.E
'he sealing me:nanism ass:ciate: witt ea:n eenetratier (e.g.,
u.ic:, re'ics:, -- :--ir;s '
's C? E I E '_ E .
....N.............,.
.- .:
.~. .:-- ..
1.9
A ".MANNE' CA'.:E:. ~ :S stall Di .ne a:j stce-., as
ecessary.
O'
tre ::arne
.
..:;; s :
ini.
't
es:cr.:s .i -
= *e:es s ary
r ;e an: at:ura:y 1: ( c.
-
v a '. v e :" tre ci are .e
":r *.r4 :na- ei Oc-it: 3
The CHASSE. CALIS ATION
-
^
e :: :2 : /
e-.-
s : a-
'_:;-- *-
. ser Erl ciarr ar.$ c- t-ir
't :."
i, 2 .: : a'i irci.:e :na C'-25NE. ;cs:TIchAL TEST
The CHANSEL
C A.!E RAIIC N may b e p e r f o rme c
b,.
ar.. s e r i e s
f sequentiai, everlapping, Or total
l
.
rarne' st :s sa:
.'at the e tire :nannel is c 'it
ed.
CHAhNEL CHECK
,
i
1.1:
A .4ASNE CHECK shai' te t ? Oualitati ve assessne-t of channel behavier
curir; cperati:n ty observation. This cete nination shall include, where
pcssiDie, cce:aris: 1 of the :hannel indicat:en and/cr status with other indi-
-
ca* ions and/cr s.a:us derive. frem indeperdent instr. ment channels measuring the
same para.eter.
CHANSEL FUN *TIO!'t TEST
.
C n. . h 3
t
rps.i.vh..n.
.. sna ,, t
t,. injecticn c,.
a simulatew signa,, int:
-
.7
s
. .
s
, . ,
.
,
i
.
n .e
... n
e
the channel as clcse to the primary sensor as practicable to verify OPERABILITY
including alar:i and/or trip functions.
CORE ALTERATION
l
1.12 CORE ALTERATION shall be tl.a
.vemer,t or nanipulatien of any compenent
l
witFin tr.e reactcr pressura vessel with the vessel im:ad removed and fuel in the
vessel.
Suspension of CORE AllERATIONS shall not p eclude completion of move-
ment of a ccmponent to a 3afe conservative position.
,
t
g.
_ .
.
.
.
..
.-
.
. .~ 3
-..
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_ - _ . ,
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58
.b
.
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-
. . - . _ _ . . . . _ _ , , . _ _ _ . . - . . . _ _ _ . _ . . _ _ _ _ _ _ _ _ _ . - . _ _ . , _ _ .
_ _ _ _ _ _ _ . , . _ _ . . . , _ _ . _ _ _ _ . .
-
.
.
l

.
.
.
.
QULS110N 6.01
(3.00)
The plant is operating at 50% power when a control system hot leg RTO
fails high.
Does this failure INCREASE, DECREASE, or NOT AFFECT the
following: Consider each item independently.
Assume no operator action
and that all control systems are in automatic.
a.
affected ct:r:ml overpower delta T trip setpoint
b.
steam bypass cooldown valves (first bank)
c.
charging flow (initially).
d.
control rod bank position
e.
rod insertion limit setpoint
f.
affected channel actual overtemperature delta T indication.
ANSWER 6.01
(3.00)
a.
NOT AFFECT
b.
NOT AFFECT
c.
INCREASE
(0.50 x 6)
d.
DECREASE
e.
INCREASE
f.
NOT AFFECT
REFERENCE
B.V.P.S. 2LP-SQS-1.1 Enabling Objective 6
B.V.P.S. 2LP-SQS-1.3 Enabling Objective 10,12
B.V.P.S. 2LP-SQS-7.1 Enabling Objective 7
B.V.P.S. 2LP-SQS-21.1 Enabling Objective 4

B.V.P.S. - 0.M. 2.01.1 pages 12,20; 2.7.1 page 35
2.21.1 page 22; 2.6.1 page 64
B.V.P.S. - Unit 2 Technical Specifications Table 2.2-1
X/A 001050 K5.01 3.6
i
K/A 004010 A1.01 3.6
K/A 041020 A3.02 3.4
'
041020A302
004010A101
001050K501
..(KA's)
COMMENT:
'
6.01.f
The question asked for affected channel (control) actual overtemperature
delta T indication for a failed high hot leg RTD.
At BVPS, there are no
indicators for actual overtemperature delta T.
There are delta T
indications for Control delta T, Protection delta T, and overtemperature
delta T setpoint.
The candidate should be given full credit for the
following answers depending on which indicator he believes the question
to be asking him about:
1.
Control delta T - increases
2.
Protection delta T - increases
3.
Overtemperature delta T setpoint - decreases
See ettachment 6.01.f, pages 1-5 for Vertical Board indicators.
- -
-
-
-
-
-
-
- -
- - -
-
-
-
-
-
-
- - -
a
.
.
.
f
.
.
.
F.Y.P.S.
-
0. !; .
..t. .)
'
SPECIFIC INSTRU?!ENTATION AND CONTROL
_2RCS*TX410A
Type
Westinghouse 7300 NLD isolator
funct:en:
l'revide _ :gnal to RCP .'1 A thermal
overload t: :p circuit
-.t
>
'
.
.
4:.;,.
'w e : a ngheus.- 7300 signal ccn.parator
r
'
t re . : d,-
7
n
cc pu -
' T2 E F 2 'i
-
n:
,
ACCT hCS TRAIN b TE!P KCTelu and
A::* m : :e- W: 4:w No. A_-DA
, . _ ,
:. 3.s a t a t. . . . . . .. Lt ,n_.
h m. - ..:
, . .

.

. _....
. . .

i
.
ype
}'
e**-
1 60
tMt,
..
s
m. s . .
s.
ru u
ha:w
-7tJ
t
,
,
a
t o c,
., ,, ... - 2 L. . .
-
--
-
.u-
r
l
2RCS-TI 10E
~cm
K..s t :n;nc :< .
i.~X 2 5 .'
,y
o.--.r
==..p-
F u r.c t i c n :
Gived
c-
% .:cid leg t orpe r a t u re
i n d i c a t :c r.
at the Alternate
S hut dow:: Panel (ASP).
~
.. =.
.w.s:
---
. Sensing-:Peiht >P -f Loop A hot leg manifold '
ETD
3 ,e4,,f:120e
. w ,, , T ,,,.Ti d,d,r.....de
,
hc
-.,
. ~.
Type:
?!c
.
Range:
2 ; - 700:-
Funct:en:
Provide signal to [2RCS-TU411E and TUJ.110)
2RCS-TU411C
T pe:
Westinghcuse 7300 NSA su=ing aeplifier

Functicr
Receives signal frca [2KCS"TEJ.11b and
tee 11C} and develops a Loop A average
te ; era.ure signal which is passed
to (2RCS-TUe:Ej via the TAVG DEFEAT
switch located at Vertical Board - Section E
and alsc providos Locp A average temperature
gna; tc the following-
100
1,'s ti.
h!. , ,:
.
. . - - -
.-
- - -.
.-

i
.
l

e
l
l
.
.
.
.
l! . V . I' b
- 0 ..*:.
..e,1
SPECIFIC INSTRt?!ENTATION AND CONTROL
l
l
2KCS-TX408B
t
t
Type:
Westinghouse 7300 NLP isolator
'
Function: Provide auctioneered high Tavg
signal to [2RCS-TX40SC)
i
i
~ . p. , o s .
u
I
iy; .
L 3: :nghcuse 'W
'J
s o l a :.c r
J
Fu. . i c:: :
l' r ov ide signal t.o i LKC5-JGCSJ j
1
7 oup
J. Ud Centrol
. .
-;_3.-
_
l
l'o : n t
Lce; S c c.l c J eg ta: if oi .
I
S.
'
,
~
EdF Plat :nur KTD ':.xe'
..s
.
'
5lO-e50P
'
"re.

.
'5-Tt J.;'"
ar
Ti.* 11 *r ar
~T,~ ; 1 C :
%_ ~q
-.
..
,
2 : :._ch
Type
Westinghcuse 7300 NC: cc: puter input
Functicn:
Provide sip al te ce = uter {T0402A]
RCLA CC:.? TT"
'[
Type:
Westin-b

? summin a
14 "
-
f
Functicn:
e
.
2RCS TZ4113
(
T'/pe :
Vestinghouse 7300 NCI c aputer input
Function:
Provide signal to computer [T0404A),
RCLA CONTROL DT
l
Type:
Vestinghouse \\'X-23;
Range:
0-150*.
,n..
, q n
.e
.
re..
s
>s!
A
A f. \\
J'N
.
.
- - . , , - - - - - ,,_.- - - _ ., - ,. . ..
- , - - - - . . - - , , - . . . , . - - . - - - - - . . - -
. _ . , _ . _ , - . , - - _ . . _ - _ . .
.. .
- ..--- - , ~
e
.
-
.
a
'I a
!
[
'j(
2,
s.
e
B.V.P.S.
- 0 M.
2.6.1
.
SPECIFIC INSTRUMENTATION AND CONTROL
l
'
.
i

2RCS*TE411D s
3
Sensing Point:
Loop A hot leg manifold
,
,
,
'
Type:
RdF Platinum RIT) Model 21204
Range:
530-650F
Function:
Installed spare
,
4 Di~:; -
, , , r ~: ?. r ~ ~ ~ - - - -
'*
SeERW.
.
}L*&
,
Type:
inum
D Mo e-
1204
(
Range:
530-650F
i
Function:
Provide signal to [2RCS-TT412H]
I
2RCS-TT412 H
Type:
Westinghouse 7300 NRA RTD amplifier
Functicn:
Provide signal to (2RCS-TU412J and
TG12K]
2RCS-TU412J
Type:
Westinghouse 7300 NSA summing a:plifie-
'
Function:
Provide signal to the following:
(
2RCS-TX412s
Type:
Westinghouse 7300 NLP isolator
Function: Supply signal to [2RCS-TSH412B
-
,
and TSH412C] and to (2RCS-TX412A)
2RCS-TX.12A
Type:
Westinghouse 7300 NLP isolator
Function:
Provide signal to the following:
2RCS-TZ412A
(
'
%
Type:
Westinghouce 7300 NCI computer input
Function: Provide signal to computer (T0403A),
RCLA PROTECTION DT
k'
106
ISSUE 1 REY 2
. -
-
-
-
-
- -
-
-
-
-
-
- -
- - -
-
-
- - -
- -
-

.

.
.
.
,
f
.
.
N
'
'
,

B.V.P.S.
- 0.M.
2.6.1
. ,.
(
,
SPFCIFIC INSTRUMENTATION AND CONTROL
Type:
Westinghouse VX-252
Range:
0-150*
Y
)
2RCS-TR412
'
Type:
Westinghouse Hagan Optimac '!cdel 102, 2 Pc:.
hange:
0-15C*.
Functien: Record Loop A Delta T (Pen 1), Lcop
i
Overteeperature Delta T Set Point (Pen 2), and
i
Loep Overpower Delta T Set Point (Pe: 3)
at Vertical Board - Section B.
Lccp
i
selected by REC LOOP SELECTOR sw:tch at
I'enchbeard - Sect ion E , pcsit::: '
{
are leap A - Loop F
loop C
-
2RCSfTE4120
Sensing Point:
Loop A cold leg manifold
Type:
RdF Platinum RTD Model 21204
Range:
510-630F
Funct en-
Installed spare
2RCS*TE412D
.
Sensing Point:
Loop A cold leg manifold
Type:
RdF Platinue RTD Model 21204
Range:
510-630F
Functicn:
Previde
'
signal to [2RCS-TT412]
2RSC-TF412
Type:
Westinghouse 7300 NRA RTD amplifier
Function:
Provide signal to [2RCS-TU412J and
TV412K]
2RCS-TU412K
Type:
Westinghouse 7300 NAS sum. ming a:plifier
Function:
Provide signal to the following:
2RCS-TX412L
Type:
Vestinghouse 7300 NLP isolator

Function:
Supply signal to (2RCS-TU412G}
107
ISSUE 1 REY 2
--
..
.,
. - _ _ -
-
._
.
.-
-.

.s
.
.
.
h
.
s
li . V . I' . S . - 0. !! .
2.o.1
SPECIFIC INSTRUMENTATION AND CONTROL
2RCS-TSH412C
Type:
Westinghouse 7300 NAL signal comparator
Function:
Receive signa 1 from [2RCS-TU412F and
TX412SJ and provide signal to [OVER-
'
POWER DT REACTOR TRIP)
_2_KC S -TX4_12 C
,
.
Type:
'mestinghouse 73L NLP isolator
Functicn
f rov 2 d.. signal to {2RCS-TRe12] via
REC LOOP SELECTOR switch and to the
o'.y.,u ; c. m .es:
o
,3C$-TE&ffW

Type:
K,-
. ; & as.
VX-2~2
bd%** ion:r/lil.,: :
_ .
t.sg
, m~
t--- m
.
icate
F.u.nct
'
4. .s , b .a
, : . 3. o.swa 4 c A4,-
'
2RCS-T2412C

Type:
We s t : ngh : a r...
300 NCI cc:::puter input
F unc t i e r. : F: c c :c. .:gna; to cerputer [T0410A),
RCLA OVERTT.':P DT I SP

6
i
2RCSoTE413
,,
Sensing Point:
Lecp A het leg
l
Type:
P3s::nur RTP FDF Corp. P/N 21205
Rcnge:
C
00F
Function:
Frcvide s . gr.a ; tc [2RCS-TXI.22A], to
[2RCS*PS403A} for [2RCS*PCV456) interlock, and
,
'
to the following:
i
I
)
2RCS-TI413A
\\
Type:
Westinghouse VX-252
Range:
0-700F
Function:
Indicate Loop A hot leg teep
at Emergency Shutdown Panel (SDP)
2RCS*T1413
,
Type:
Westingheuse VX-252
Range:
0 700F
Function:
Ind:cate uccp is hot leg tecperature
(PAM 1J at Vertical Board - Section A
k
o
sscE : Rtv 2
-
- --
-
-
-
-
-
-
-
-
-
a
J
S--
c
i
. -
,

,
.
.
.
QULSi!ON 6.03
(2.50)
Using Attachment 2, Op. Manual Fig. No. 13-2, "Quench Spray System,'
identify the following components on the attachment as specified in each
part below,
a.
Highlight the "A" quench spray- pump recirculation flowpath back to
the RWST.
(0.50)
b.
Circle the THREE (3) building / area boundaries that the "B"
containment quench spray header passes through.
(0.75)
c.
Circle WHERE the flowrate for the "A" chemical injection pump is
measured.
(0.50)
d)
Circle the THREE (3) valves that realign when the RWST level
reaches the level setpoint for 20SS-LSKK100B-1.
(0.75)

ANSWER 6.03
(2.50)
a.
(0.50)
b.
(0.75)
Use attached drawing
(0.50
5'
c.
(0.50)
as answer key
-
,
d.
(0.75)
REFERENCE
i
B.V.P.S. 2LP-SQS-13.1 Enabling Objective 2,4,5
'
B.V.P.S. Op. Manual Fig. No. 13-2
K/A 194001 A1.07 3.2
194001A107
(KA's)
,
. .
COMMENT:
,
6.03.d
Question should have included written description of name for
.
2QSS*LSKK10081 (RWST Level Extreme Low-Low).
(See attachment for
'
,
i
6.03.d.)
3
.
T
i
-
- .
. .
.
- -
. - .
-
. -
-
.
-
- -
-
.
- -.
. --
.- .
-. .- - -
-
- - - - -
-
.
. -
- - -
- - -
. - - -
-
- .
- - - - -
. - . -
-
- _ _ ____ _ - _ _
_ _ _ _ _ _ _ _ _ _ _
,
.
!! '. . ! ' . S . - U.M.
.13.]
O
S I'EC I F I C INSTRUMENTATION AND CONTROL
2QSS-LZ100A
Type:
Westinghouse 7300 Series Signal Comparator
Function:
Provides signal to computer point {LO500A),
RWST Level
. , a.w..,,s.
.
..
4.
%:
'
n.
. ::
.r
. , r a g..
ta:S
+
,
Type
. n gh o u., e level ... :cator, model VX-2',2
a.
,
g m ,s ,,
n
- :,
<.t,.,s
Function:
indicate h EFUE', .cTh STOR TK LEVEL on Vert ical Board -
.
,
.
w ' ~ 1 o c t:
.,w.
>r
-
< . . . .s : u; !i :c
.
.
t a:s
p,
y-
-
,,
i r r.s m i t . - -
'133bS5PA
.
hange;
. c
7'
.
.e 3 of water
l
rur.c:icn
n r:s
!"
=_.. :: :N fclicwing
s - ~:>-
2QSS-LSK10CE
Type
4.> s - -
.s o ';
NAL: dual signal ccersrator
l
Functien:
Fravide le'.". s:gnals to ccmputer poin: [LO517Dj,
-
.m..
_
1
n:.r :. . a.a d.v
.s ~ uv a - LC,. . that corresponds to .sSS pumps
a
to ccic leg recirculation switchover and to
Annunciator Windcw No. A6-lD, REFUF.L WlTER
- . - - - .
. . . . , . .
1
5 i u a s G . w. . - . i.u :. u
.N.C R.yA,u
i

r,
r
.
A
l
4
i
Type:
West:ngb.cuse ~3CC N;.1.2 dual signal comparator
l
Functicn:
Provide signal tc cceputer point LO51SDJ,
b
-
7
FUEL WATER
.~
m.
.uo..-
..%.
STORAGE TASK LEVEL OFF NORMAL
20SS*LSKK100B1
Type:
Westinghouse 7300 NAL1 Single Comparator
Function:
Provides a leve: 3 ;gnal :o [2QSS*LYKK1C ^ 31] for the
following components. 2 CSS *SCV1005,
QSS*SOV1013, and
2 CSS *S^*.1023 to init i ne chemical injection valve swn.chover
O
. - . ,
.. ,
.
a%r
pk
.
_ , __ -
-.
- - - ,
. - _ __._-
. - .
-,. . _
,.
. - . ,
__
.
.-
.
.
.
)
.
QUES 110N -6.04
(2.00)
a.
HOW (Increase, Decrease, No Change) will an INCREASE in'the
reference junction temperature effect indicated thermocouple
temperature?
b.
HOW (High, Low, As Is) will an RTO temperature indication fail. if a

short circuit occurs across the RTD?
c.
WHAT is the major disadvantage of using a Thermowell RTD for RCS
wide range temperature measurement?
d.
Given the graph shown in Attachment 3, identify the curve which
represents the calibration curve for a HOT calibrated instrument.
ANSWER 6.04
(2.00)
a.
Decrease (0.50)
b.
Low (0.50)
c.
Thermowell RTDs have a relatively long response time (0.50)
,
d.
A (0.50)
,
REFERENCE
B.V.P.S. LP-TM0-7 Enabling Objective 5
'
B.V.P.S. LP-TM0-7 page 11
K/A 191002 Kl.13 2.8
K/A 191002 Kl.14 2.9
,
191002Kil4
191002Kil3
..(KA's)
COMMENT:
The answer to part d. is incorrect.
For a given delta P, a cold
calibrated channel should indicate a lower level than a hot calibrated
channel due to the density difference.
Therefore, the correct answer
should be that curve B is the hot calibrated instrument.
i
i
I
8
O
.
.
.
.

ATTACHMENT 3
.
A
B
100*
t ,c
( ,i w ,*
'
1
\\
0%
\\
4P
.
1
i
-
-
- - . .
- - - . - -
- - - .
)
.
+
.
.
.
.
.
!
.
QUE5110N 6.05
(2.00)
,
for EACH of the following radiation monitors, state the automatic actions
which occur, if any, when the monitors alarm HIGH.
a.
25WS-RQIl01 - Component Cooling Service Water
b.
2HVR*RQIl04A - Containment Purge.
c.
2RMC*RQ201 - Control Room Area
d.
2GWS RQIl02 - Air Ejector Delay Bldg. Exhaust
ANSWER 6.05
(2.00)
a.
none
(4 x 0.50)
b.
closes 2HVR*M0023A and 2HVR*M0023B (applicable valve names
!
l
acceptable)
-
c.
actuates control room pressurization

d.
none
REFERENCE
i
B.V.P.S. 2LP-SQS-43.1 Enabling Objective 4

8.V.P.S. 2LP-SQS-43.1 pages 16,21.24,39
!
X/A 072000 G0.04 3.7
'
072000G004
...(KA's)

COMMENT:
,
6.05.b
The answer key incorrectly identifies 2HVR*M00238 as an auto action for
2HYR*RQ104A. The correct aamper is 2HVR*M0025A as seen in attachment
i
6.05.b page 1.
The referenced lesson plan had a typographical error (see
attachment 6.05.b, page 2). This has been entered into the Training
Department's Action List and shall be corrected in the near future.
Additionally, the applicable damper names have been provided for use as
identified on the answer key.
(See attachment 6.05.b, page 3.)
l

e,
,

.

,
.
.
.
B.V.P.S. - 0.M.
2.43.1
.
SPECIFIC INSTRUMENTATION AND CONTROL
[2HVL-DAU112]
Typc~:
G.A. Technologies, RM-80
'
Range:
Particulate:
10(-10) to 10(-5) pC1/CC
,
Gascous:
10(-7) to 10(-2) pC1/CC
Function:
Inputs to DRMS and to Ann. Vindows A4-5A RADI ATION
MONITORING SYSTEM TKOUBLE. A4-5C RADIt.TICN MONITGKING
LEVEL HIGH and provides local alarm and indicatien.
1_2HYL-VP112J
Type:
KVRZ Model 455
hange:
0-20,000 CFM
Function:
Senses Flow rate in the Condenstae
Polishing ' Stack SKID and inputs to
[ 2 HVI.-:a 12'; l
7.I2EVR*M/ M '
,
Sensing Point:
Centainment purge exhaust
Ty;:.
Be t a S c int i l : a*.c r , i:. lit.e duct =td gas
Function:
Inputs to [2HVR*DAU104A]
12HVR*DAU104A]
Type :
G.A. Technologies, RM-60
Range:
10(-e) to 10(-1) pCi/CC
Tunction:
Inputs to ADDS (Annunciation and Digital Display System)
Vindows A4-5A RADIATION MONITORING SYSTEM TROUBLE, A4-5C
RADI ATION MONITORING LEVEL HIGH and
local
alare and indication as well as
[
tand;
25A}{en high radiat
oa
..
...
..
,.
-end'-en'RM-23 in the control room
[2HVR*RQ104B]
Sensing Point:
Containment purge exhaust
i
Type:
Beta Scintillator, inline duct etd gas
'
Function:
Inputs to [2HVR*DAU104B]
1
<
1

77
ISSUE 1 REV 1
,....
.-
. - - - . - . . - - . . - -
-
-=
. . . - . . . . .
. . - ~
-
- - - .
=-
2'
'-*-"""'"==':=:-----
t
-
['
'
'
b)
Mgnit;or_ Items - C/S (Check:ourco) actuation
,
.
used in conjunction with th" chinnel.
i
.,
b.
Power Sup_p_1_ies - 120 VAC is supplied to tim nonitor
'! iTE :
RD-25A uses 257. .*-
where it converts it to four 24 VDC eutputn, three 5
1.750 VDC.
VDC outputs, two 8 VDC outputs and a 3 VAC output.
A
424 VDC output is supplied to two liigh Voltage Pouer
Supplies (adjustable from 500 to 1250 VDC or 700 to
2000 VDC).
c.
Lio_ del 3101 Mark !4 umbers
(RD-25A)
.
1)
2IIVR* ROIL 04 A - Conta_inment_Ilurge
TF-13
-
Location - 780' Containment
-
Func_ tion - Monitors containment activity during
initial purge and refueling operations
-
Power Supply - 120 VAC [PNL-AC2-E7; Bkr E7-10
-
uto Actionn - Closes [ 211VR* MOD 2 3 A ] and
on high rad.
2HVR*DAU104A
-
Location - 730' Service Building
2)
_21IVIi* RQJ 104 B - Cont a i nmen t Purgo
-
Locatior! -
780' Containment
-
Function -
Same 'as 211VR*RQIO4A
-
Power Supp_ly - 120 VAC [PNL-AC2-E8] Bkr E8-9
Auto Actions
0525Af and
[ 2HVR*!40D25B]
_211VR * DAU104 B
-
J4 cation - 730' Service Building
.P-SQS-43.1
- 24 of 48 -
- -
,
._
i
.
.
-.
B.V.P.S.
- 0.M.
2.43.4
7
Issue 1/ Revision 1
Page ACX1 of 1
,
ACX. LOCAL CONTAINMENT PURGE [2HVR*9Q104A(B)] HIGH ALARM LEVEL
l
Ann. Window No. N/A
Setpoint
Device
.t e r
l 2H\\M AU10 A(b > 1
PRTBABLE CAUSE
Radioactive gases or particulates in Containment.
ID. E E ECTI VE AC_T! ONS..
-
--
1.
Ver:fy the alarmed' condition at the operators censole.
2.
Close
c- v+ ri ty close the i c i l e.c : ::: rator cperated darrers a-
- -
Euilding ' n ue ea:.c ;
EA
E
"
.
hY/
b.
(2HVR* MOD 233) CNMT Purge Exhaust Isolation Daeper.
Afhapperft
d.
[2HVE* MOD 25B} CNMT Purge Air Supply Isel Darper.
3.
Evacuate the Centainment,
-
'
1
e.
Instruct
all personnel in the Containment to report te RadCen for
dose assessment and possible decontaminatien.
'
.
Sottfy AAEC^N,
have them ccnduct surveys, if possible to locate
j
'
the source cf the activity.
'
6.
Take re=edial actions as necessar" to reduce the activity.
l
REFERENCES
NOTE:
All
references used prior to 4-13-87 are located in Section
l
l
S.
!
1.
EVPS-2 OMCN 2-87-23 (Rev 1)
,
i
t
a
$
.
4
,i?-
- - -
,-
~
-
, - -
y
.
~.,..s
.
s
.-
,.a.
. = n as
-
&
._.a
a
-
+
.a
a
&
.

,
.
- *
.
,
.
'
j
.
QULST10N 6.10
(2.50)
The plant is stable in Mode 5 with the "A" Residual Heat Removal System
i
(RHS) in service.
l
a .'
At WHAT pressure (psig) will the RHS isolate from the RCS? (0.50)
i
- i
b.
-WHAT is the design capacity of the RHS suction line relief valve
(2RHS*RV721A]?
Include ALL applicable information.
(0.80)
c.
Loss of primary component cooling water can affect WHAT TWO (2) RHS
r
i
components, when operating?
(0.70)
i
d.
Failure of RHS Hx flow control valve, [2RHS*FCV605A] to the closed
position will result in a (Increase Decrease, No Change) to RCS
'
temperature?
(0.50)
l
2
ANSWER 6.10
(2.50)
.

!
a.
>700 psig (0.50)
j
,

b.
TWO (2) (0.25) charging pumps (0.25)
j
at the relief valve set pressure (0.30)
-
1
c.
RHS heat exchanger (0.35)
.
RHS pump seal cooler (0.35)
-
,
a
i
d.
Decrease (0.50)
j
REFERENCE
)
Enabling Objectives UNAVAILABLE
i
B.V.P.S. - 0.M. 2.10.1 pages 1,2,5,6,20,21
!
a
)
K/A 000025 Kl.01

K/A 000025 K3.02
"
,

1
K/A 000025 A1.01
000025A101
00025K302
000025K101
..(KA's)
'
.
I
COMMENT:
1

i
6.10.d
l
.
i
The correct nomenclature for [2RHS*fCV605A] is provided in attachment
j
j
6.10.d.
The nomenclature used in the question is incorrect.
i
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StCI10N SEVEN GENERAL COMMLNT:
-Beaver Valley Power Station's administrative procedures on adherence
to operating procedures state, "When extensive operations, infrequent
operations or any operations requiring documentation are to be performed,
the operating procedure must be present and followed."
Because of this,
there is no need to, and operators are not trained to, memorize procedures
with the exception of the immediate action steps of the E0P's.
Section 7 of this examination contains seven questions (over 25% of
the section) that require the candidate to repeat from memory information
contained in Normal, Abnormal or Alarm Response Procedures that are
required to be "present and followed" when these operations are
performed.
These questions are not a measure of an operators ability to'
'
do his job and should not be used to evaluate whether or not he should be
,
granted a license.
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-.
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.
.
.
QUISil0N 7.01
(1.50)
'
for the following questions, assume B.V.P.S. - 0.M. 51. Station Shutdown
Procedure, is in use.
a.
When using condenser steam dumps, WHAT operator action (s) must be
taken to cooldown the RCS below the Lo-Lo Tavg setpoint?
(0.50)
b.
When the Residual Heat Removal System (RHS) is in operation, at
lea. . one rector coolant pump must remain in service until RCS
.
tempt /ature is less than 200 degrees F.
WHY?
(0.50)
c.
If minimum RCS flow requirements CANNOT be met nhile'in Mode 4, the
operator's immediate response is to refer to WHAT procedure? (0.50)
ANSWER 7.01
(1.50)
,
a.
place steam bypass interlock selection switch to the DEFEAT TAVG
position
(0. 50-)
b,
prevent reactor vessel void formation (maintain RCS subcooling)
(0.50)
_
c.
B.V.P.S. - E.0.P. ES-0.2, "Natural Circulation Cooldown"
(0.50)
REFERENCE
'
B.V.P.S. 2LP-SQS-21.1 Enabling Objectives 4:
2LP-SQS-50.51.52.1 Enabling Objectives 2,3
B.V.P.S. - 0.M. 2.51.4 pages C9,02,04; 2.51.2 page 3; 2.53C.4 page 3
K/A 00.c's00 60.10 3.5
K/A Or 000 G0.15 3.9
e
K/A 0 41020 A4.08 3.1
04iC20A408
005000G015
005000G010
..(KA's)
"
'
C0KMENT:
The answer to part c. is contained in cautions in the Normal Operating
Procedures for Station Shutdown.
It is not an immediate action in an
emergency operating procedure and, therefore, is not required to be
'
committed to memory by Beaver Valley's administrative procedures
l
(Operating Manual 1/2.48.2.C.3). We ask that the question be withdrawn.
'
1
I
[
s
e
t
,
~ --
_ _ . . .-
_.
.
.
.
-.
.
.. -
,
_

-
..
,
-
.
.
QULST10N 7.03
(2.00)
1
Answer the following questions concerning B.V.P.S. procedure A0P.2.1.3,
,
"Continuous Insertion of RCCA Control Bank."
!
.
a.
WHAT anticipated operational transient could cause a continuous
bank insertion of the controlling bank?
(0.50)

.
b.
If a malfunction causes a RCCA control bank to insert past the
Low-Low insertion limit, WHAT immediate operator _ action is

'
required?
(0.50)
,
c.
If rod control is transferred to Manual and a continuous' insertion
condition is still present, WHAT TWO (2) operator actions should be
performed?
(1.00)
ANSWER 7.03
(2.00)
l
J
a.
. turbine runback (OTdt or OPdt) OR load rejection (0.50)
i
i
b.
emergency boration OR boration at concentratics and flowrate at
least that as stated in Technical Specifications (0.50)
.
c.
trip the reactor (0.50) and go to E-0 {0.50)
j
.
'
REFERENCE
{
B.V.P.S. - 0.M. 53C A0P-2.1.3 page 1
'
i
B.V.P.S. - 0.M. I page AAMI
l
X/A 001000 A1.04 3.9

X/A 001000 A3.02 3.6
_
-
001000A302
001000A104
..(KA's)
!
COMMENT:
l
,
1
!
j
The answers to parts b. and c. are contained in an Alarm Response
!
}
Procedure and an Abnormal Operating Procedure respectively.
Neither of

these procedures are required to be memorized. We ask that these
!
questions be withdrawn.
I
i
i
I
!
!
.
1
l
i
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J

i
,

i
.
.

s
..
,
QUESTiott 7.06
(1.50)
'
a.
WHAT procedure (by name) would you consult if annunciator Al-lE,
"Containment Air Partial Pressure High Low." alarmed?
b.
WHAT could cause containment pressure to slowly increase with
4
little or no humidity increase, and_a possible decrease in
'
temperature?
c.
If the plant is in Mode 2, and containment pressure, temperature,
i
and humidity ALL begin to increase rapidly, WHAT action should'the
operator take?
ANSWER 7.06 (1.50)
a.
Loss of_ Containment Vacuum (A0P-2.12.1) (0.50)
!
i
b.
a breach of (leakage into) containment (0.50)
j
c.
manually trip the reactor (0.50)
r
REFEREliCE
-
B.V.P.S. 2LP-SQS 53C.1 Enabling Objectives 1, 3
!
B.V.P.S.
0.M. A0P-2.12.1 page 1
,
K/A 000029 EA2.01 4.3
'
K/A 000029 GO.ll 4.2
000069G0ll
000069A201
..(KA's)
i
COMMENT:
,
4
l
When an annunciator alarms, as given in part a., the Alarm Response
Procedure should be consulted.
The Alarm Response Procedure will give
i
'
!
direction for responding to the situation including references to other
procedures,
in this particular case, the Alarm Response Procedure does
i
not reference the A0P for loss of containment vacuum.
Since this
'
annunciator is a symptom in the AOP, the AOP could also be consulted for
j
guidance. We ask that either the Alarm Response Procedure or the A0P be
i
an acceptable answer for full credit. We have submitted paperwork to the
4
procedures group to change the Alarm Response Procedgre so that it
references the AOP.

i
The answer to part c. is contained in an A0P.
It is not an immediate
<

action and is not required to be memorized.
We ask that this question be
'
withdrawn.
I
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e

i
I
1
'
I
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--
-
-
-
- -
- - - - - - - -
-
- - -
- - - -
-
--
-
- -
- -
-
- - - -
--- :
_
_
.

.'
OVEST10N 7.07
(3.50)
A condition arises that requires entry into containment at 40% power.
The operator entering containment needs to work in a gamma radiation
field of 150 mrem /hr for approximately 2.0 hours0 days 0 hours 0 weeks 0 months . The below candidates
,
are presented to you:
,
Candidate
1
2
3
4
Sex
male
male
female
male

Age
27
38
24
20
Qtr/ exposure
1000 mrem
500 mrem
1000 mrem
-.
Life exposure
1000 mrem
54730 mrem
5200 mrem
9500 mrem

Remarks
quarterly
form
3 months
!
history
NRC-4
pregnant
,,
unavailable
unavailable
Each candidate is technically competent and physically capable of-
performing the task.
All candidates have a completed Form NRC-4 and have
a documented current calendar quarter exposure history, with the
exceptions for those candidates stated above.
Emergency limits do NOT

apply.
For EACH person, indicate if you would ACCEPT or REJECT the
person to perform the task based on EXPOSURE REQUIREMENTS ONLY.
Justify
EACH answer AND -include ALL applicable limits.
.
ANSWER 7.07
(3.50)
,
1 - REJECT (0.25) since he has no quarterly history available and would
exceed the 200 mrem /qtr whole body limit (0.50)
2 - REJECT (0.25) since he does not have a Form NRC 4 available and
would exceed the 1250 mrem /qtr whole body limit (0.50)
3 - REJECT (0.25) since she will exceed the allowable exposure limit
.
during the term of her pregnancy (0.50)
!
84 - ACCEPT (0.25) since he will not exceed the quarterly limit (0.50)
5
i
j
or the whole beiy limit of 10000 mrem lifetime exposure (0.50)
J
-
REFERENCE
Enabling Objectives UNAVAILABLE
4
B.V.P.S. - R.C.M. pages 5, 6, 7
K/A 194001 Kl.03 3.4
i
,

19400lK103
..(KA's)
i
j
COMMENT:
Candidate #1 would be accepted based on 10 CFR 20 limits of
1250 mrem /qtr.
This is the reference the candidate was give to use.
Candidate #3 cannot be evaluated with the given information.
Exposure
,
limits for pregnant women are in Reg. Guide 8.13 which was not
!
available.
We ask that the key be changed to accept candidate #1 and
j
that candidate #3 be deleted from the question.
j
1
<
l

.
.. .
.
QULSil0N 7.08
(2.00)
Answer the follow;ng question concerning B.V.P.S
0.M. A0P 2.38.1
"Loss
of 120 VAC Vital Bus."
WHAT are FOUR (4) automatic actions that an operator can visually verify
in the control room if power to 120 VAC Vital Bus 1 is' lost? ONLY
consider safety system actuations.
ANSWER 7.08
(2.00)
1)
atmospheric steam dump valves fail closed, if open
2)
letdown will isolate
(4 x 0.50)
3)
PRZR heaters will deer.rgize
4)
standby service 4ter pumps (2SWE-P21A) auto starts, if not already
running
5)
component cooling water to containment instrument air compressor
closes
.
6)
primary component cooling water supply and return isolation valve]
(2CCP*MOV175-1.176-1.177-1,178-1) close
REFERENCE
!
B.V.P.S. - 2LP-SQS-53C.1 Enabling Objective 5
B.V.P.S. - 0.M. AOP-2.38.1 pages 1,2

K/A 000057 EA2.19 4.3
-
,
000057A219
..(KA's)
COMMENT:
The reference from the K/A catalog states "Ability to determine or
interpret the plant automatic actions that will occur on the loss of a
vital AC electrical instrument bus".
The way to determine the automatic
actions that will occur is to consult the AOP for loss of that vital
,
bus. There are 17 different automatic actions that could occur depending
on which vital bus is lost.
It is not necessary to rely on an operator's
memory to verify the correct auto actions for the correct vital bus
failure.
This is why procedures are required to be present and followed
for infrequent operations such as responding to loss of vital bus. .We
j
ask that the question be withdrawn.
.
=
.-
. - .
.
.
.
.
.
QUES 110N 7.09
(1.00)
,
WHAT are the normal. expected values for Source Range (SR) AND
Intermediate Range (IR) Nuclear Instrumantation an operator-would expect
to see when verifying that the SR has reenergized after a reactor trip
from power?
ANSWER' 7.09
(1.00)

,
.;
,
SR:
IE+5 (+/--2.5E+4) counts /second (0.50)
IR:
lE-10 (+/- 0.5E-10) amps (0.50)
'
1
REFERENCE
B.V.P.S. - 2LP SQS-2.2 Enabling Objective 4
B.V.P.S. - 0.M. 2.2.4 pages B4, C1
K/A 000032 EA2.04 3.5
000032A204
..(XA's)
,
1
i
COMMENT:
,
5
1 x 10 cpsistheSourceRangeHi-FluxTripsetpoint.
The actual
reading would be less than 10 cps.
We ask that the answer key be
changed accordingly.
!
'

,
!
l
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!
'
,
4
i
!
l
1
,
k

'
4
1
2
,
.
.
.
. m
.
. .
-
. . .
.
)- ( 'f

.
!! . V . I' . S . - 0. t! .
2.01.2
~f
Sr.T POINTS
Reactor Protection System
(Steamline Isolation)
High-2 containment pressure
3 PSIG
fReactorProtectionSystem(ReactorTrip)
^' ~% s
i

I
Source R..nge high level
10 E+5 Cis
l
,
- - - - .__ _ __
_
___
. .
..
Intermediate Range high level
Current
equivalent to
25* of full
PCwCr
Powe: Ra..;;e , h:gh range, high lete:
109*. of full
pcwer
l'en r h 1:.
..;;c
. . .d ) " . '
5

: - ;
cwe-
Fower hange., h:ga neutron flux rate (pcst:2re;
+ 5*. of rated
therra: tcwrr
with a timer
22 secends censtant
Pcwcr Range, high neutren flux rate (negative
3'. cf rated
-
thermal power
with a timer
22 secends constant
H:gn p:essuriner pressure
2355 PSIG
High pressuriner water level
92*. o f span
Lc. pressurtcer pressure
19-5 FSIG
Lead ::te c:nstant
10 secends
Lag time constant
I seccnd
Loss of Primary Coolant Flow
Low flow
9 0'.
j
Low frequency
> 37.5 Hz
Low voltage
2750 VAC
Undervoltage time delay
0.5 second
Reactor coolant pu=p trip
2/3 pu=p
breakers
auto trip
Low-Low steam generator water level
15. 5*, o f s p an
Coincident low level and stear /feedwater
Steamflow 40', greater
'
flow mismatch
than feed flow with
25',of SG water level
4
1
!
j
11
ISSLT 1 REY 2
i
i
.
.
-
.
-
. . . .
..
.
.
'
.
.
,
l
.
QUES 110N 7.10
(3.00)
'
Answer the following questions concerning B.V.P.S. - 0.M. 2.24.2, "Steam
-

Generator feedwater System."

a.
WHAT action must an operator take in order to prevent a reactor
-
trip if a Steam Generator (SG) Feed Pump Auto Stop annunciator

clarms with the pl&nt at 75Y. power?
(0.50)
>
!
b.
WHAT are FIVE (5) indications / conditions that an operator would
verify if a Hi-Hi- SG level trip occurred with the plant at 407,
!
,
power?
(2.00)
l
,
ANSWER 7.10 (3,00)
j
a.
place the SG Startup Feedwater Pump in service-(0.50)
!
b.
turbine trip
-
main feedwater pump (MFWP) tripped
!
-
-
MFWP discharge valves closed
(5 x 0.50)

MFW reg valves closed
-
4
1
-
SG bypass flow control valves closed
}
MFW isolation trip valves closed
.
-
REFERENCE
B.V.P.S. - 2LP-SQS-24.1 Enabling Objectives 7, 9A (14)
,
B.V.P.S. - 0.M. 2.24.2 page AAEl
'
,
K/A 000054GO.09 3.1
l
l
K/A 000054 G0.10 3.2
l
000054G010
000054G009
..(KA's)
!
COMMENT:
The answer to part a. is the second corrective action of an Alarm
Response Procedure and is not required to be memorized. .We ask that this
question be withdrawn.
I

I
l
!
1
l
1
I

1
_
.
. .
QUEST 10N 7.11
(2.50)
Answer the following questions concerning liquid Waste System Operation.
!
a.
WHICH TWO (2) flowrates (numerical values NOT required) are used in
i
calculating the-Unit 2 Cooling Tower Blowdown Flow when the Unit 2
t
blowdown flow-instrument (2CWS-FR101] is out of service, and a
liquid waste discharge is to be made by way of the Unit 1/2 cooling
-
tower blowdown line?
-(1.00)
[
b.
Before sampling the contents of the "A" waste drain tank, WHAT
-
'
action must be taken by the operator?
Include any applicable
precautionary setpoints or time related values.
(1.00)
c.
WHAT action should an operator take if local-liquid waste process

effluent [2 SCC-RQl100] high alarm actuates AND is verified to be in
'
the alarmed condition?
(0.50)

I
ANSWER 7.11
(2.50)
i
'
a.
Unit I and 2 (cooling tower) flow (from [FR-CW-101)) (0.50)
Unit I cooling tower bicwdown flow (0.50).
,
b.
recirculate the tank (0.50) for a minimum of TWO (2) tank volumes
OR 8.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months (0.50)
i
f
c.
verify closed ((2SGC HSV-100]) liquid waste EFF high rad isolation
'
valve (0.50)
}
REFERENCE

j
.
B.V.P.S. - 2LF SQS-17.1 Enabling Objectives 2d, 9, Se
I
B.V.P.S. - 0.M. 2.17.2 page 1, 2.43.4 PAGE AEEl

K/A 000059 EA2.02 3.9
.'
K/A 000059 EA2.05 3.9
4
000059A205
00059A202
..(KA's)
i
COMMENT:
1
The answer to part a. is contained in a Normal Operating Procedure and is
not required to be done from memory. We ask that the question be
withdrawn.
4
i
l
3
- -
-
-
.
.
.
.
QUIS110N 8.01
(3.00)
Using Attachment 4. classify the following events in accordance with BV-2
EPP/1 1, Recognition and Classification of Emergency Conditions AND
,
justify your answer and any assumptions.
Consider each case separately.
a.
B.V.P.S. E0P E 1, "Loss of Reactor or Secondary Coolant," is in
use.
Pressurizer level is off-scale low and RCS pressure is 1500
psig and decreasing.
The reactor was manually tripped because
pressurizer level could not be maintained,
b.
A turbine trip from 75". power occurred and the reactor did not
automatically trip (ATWS).
The reactor remained critical until an
l
operator manually inserted control rods.

c.
A truck carrying A=onia gas is involved in a collision at the
plant main entrance.
Gas is leaking from the truck.
,
,
d.
An earthquake is registered on-site with the-plant in Mode 1.
The
severe ground motion results in the generation of a missile in the
turbine building from the detachment of a LP turbine blade.
,

ANSWER 8.01
(3.00)
-
a.
SITE AREA (0.40)
.
TAB 5 -- RCS/ Containment leak exceeds 'ike up capacity (0.35)
l
b.
ALERT (0.40) TAB 14 -- Reactor not subcritical after valid scram
signal (s) (0.35)
f
c.
UNUSUAL EVENT (0.40)
TAB 18 -- Toxic gas nearby release potentially harmful (0.35)
d.
ALERT (0.40)
l
TAB id -- Turbine rupture causing casing penetration (0.35)

i
l

REFERENCE
,
Enabling Objectives UNAVAILABLE
B.V.P.S. - Unit 2 Implementing Procedures CV-2 EPP/I-l Table 1
'
K/A 194001 A1.16 4.4
i
'
194001All6
..(KA's)
f
COMMENT:
'
It was required of the candidate to utilize a copy of the BVPS 2 EAL Tab
,
Matrix for the purpose of classifying an event.
This comment is written
I
i
to inform you that the matrix is not to be used for classification
j
purpose but only as a guide to choose the most appropriate Tab.
For

future testing, please include a copy of both the EAL Tab Matrix
!
<
(Attachment 1) and the EAL Tabs (Attachment 2) if it is required for a
candidate to classify an EPP event.
(Refer to enclosed references.)
4
,
,
.
.
.
.
-
,
_
-_-____ --_____
.
.
.
t
.
EPP/Implemanting Procedure
EPP/I-1

Recognition and Classification
of Emergency Conditions
A
.
'
s
2.2
Emergency Action Levels
(EAL's)
are specified in the TABS
(Attachment 2) to this procedure.
EAL TAB reference guides
are contained in Attachment 1 to this procedure and in the
BVPS Emergency Plan,. Table 4.1.
\\
'
NOTE.
-
'
The EAL's in the TABS to this procedure have precedence ove.
j
other EAL's matrix which should be used for-a quick reference
I
to the TASS and not for classification purposes.
____________
___________________
2.3
EAL's will be triggered by the results of offsite dose
'
projections and/or assessment of plant status by the onshift
operating staf f or the Emergency Organization, if activated.
In many cases. the proper classificatien will be
im ediately
apparent.
In other cases,
more
extensive
assessment is
,
necessary
to
determine
the
applicable
emergency
classification.
. 4 ,,. g s _ -e v ; .. , . g m g .,-
,- yg--
.
,-
,
-
,
2
sais, - ts ' s

.
wm
'
,
'"
u 4, . .eemuna '
asse e s .
..m
- . -
-
- ci,
.-
f L '.
'E.
M
-
~
'f
'
"y
- q - ,
. . , _
r. k
$
. -'-
unu ww ' ..xn.-
--
-
s
,
..
U
Wi
N @ r M k a 'i P_ N Iti.. i
'#l
M.'W **' '
I
"
4
.
..
.,_c
.... .
3
with
. ' . L .' 1

';
^
. L. . ' ?. ! ".*

'
a.ldentified-
each EAL is a representative listing of the
.
. , - .
various
instruments
and other
indicators which may
be
sy=ptoms of an emergency, which should be used to assess and
classify the condition.
Symptoms should not be confused with
the actual EAL criteria.
2.6 The EAL's have been developed to provide adequate response to
postulated emergency conditions that could exist
at Beaver
Valley.
Emergency conditions could arise, however, that are
not covered by a specific EAL.
In these cases emergency
,
conditions should be classified by the general definitions of
the emergency classes which are:
a.
Unusual Event - Off-normal events which oculd indicate a
potential degradation of the
level of safety of the
plant.
'
2
Icsue S. Rev. 2
.
--- -
,
- , , , - , - - , - , , - - - , - - - - - - - - - -
- - - - - , , ,
,w,-,.--
- - - , , - .
,,,-,.__,,,,,-_a
_w.,
,,-._,,,,,,-,,v,-.,_
w___,a-o
,,
o
,
.
EPP/ Implementing Procedure
EPP/1-1
"
Recognition and Classificat ton
,
of Emergency Conditions
--
Events are in progress or have occurred which
b.
Alert
-
involve an actual or potential subst antial degradation
of the level of safety of the plant.
Events which !.nvolve actual or
c.
Site Area Emergency
-
likely major failures of plant
funecions needed
for
protection of the public.
d.
General Emergency - Even:s which involve actual or
imminent substantial core degradation or melting with
potential for loss of containment integrity.
E.
Procedure
1.0
Verify
1.1
Upen receipt of an initial indication (alarm, surveillance
report, Observation, etc.) that an emergency condition may
exist.
2n assessment shall
be
initiated
verify the
v a lid it:. of the indica ica and whether the EAL criteria have
been met.
This may be parformed by comparisen to redundant
instrument channels,
ccmparisen to other
related
plant
parameters, physical observations and field measurements.
1.2
If this
assessment cannot be completed within 15 minutes of
the initial indication,
it
shall be considered that the
emergency condition indicated does exist
and apprcpriate
emergency acticns shall be initiated in accordance with the
applicable emergency i=plementing instructions or precedures .
- , n . ~n_
72;9'i_ Class.ificatl,er?
-~
- . - .
'h'
b
Wa
.
either a single plant parameter (ie., RCS Pressure Hi/ Low) or
parameters (ie. , Fuel Clad Degradatien).
2.2 Determine
the
appropriate
e=ergency classification by
comparing the verified plant parameters to the EAL's for each
emergency condition.
l
1
1
<
l
3
Issue 5. Rev.
.
.
1
. . .
.
6 / I rl / Il- I mg. l cme n e l py re eere.f es s e
g ,c
i l'f*/ 8 - 1
ps- # 9
esce e.epel a leese nes.1 (;1o s si 8 le a t lene

.
ef E mes ep sery E:nsest14 loses
a
.
'
b F_0!! CIASSIFICATJRN Or FNfHQLtiCV CONCITf0 % .
'
INITIATING
ALERT
SITE AREA
GENERAL
CONDITION
T
EMERGENCY
EMERGENCY
, __
sit t -seeee ena l 8 vesel s
iversts are in progress
! vents W!alch involve
Events Wielcle
Wiele:le Cieutal leselle nt n
os havn occurrod wielch
Actual or t.lkoly
involve Actual
a ro t eest i n g leng s e.1-
involvo ese actual or
itajor railures or
or Imminent
al lene of t ien l eevn t
Ieutuest i n t sotss te n t i a l
Plant functions Noeded
Substantisi Core
of Safety eel t ien
einge nsbe t ion of the
for Erotection of the
Degradation or me t t-
I Inset.
I n vee l or seroty of
Puti l l c .
Ing wi tte Potentia l
t im pineet,
for loss of
Containment Integrity.

,
5:
staelloart ive a s e leeeent
tinplasseeerd n i e t orem
Eins.Innnad a l tborne
Ho los se Corrot, ponds
fladiologica l effluent
,
evicase gives utszita
e r. l o n e.es gives os'rsite
to 20 mreia/hr. a t
release results in
As. . f l e . ele i c to Any
sin se ente gecatne
ein ses sato gecater
Site Doundary
orrsite dose proJocted
ste s cas se. f oleet( s)
o fmn II.Te m.fie e=/ h r .
t iease ?.O m. Item /ler.
-o r-
to eeceed I rea
.eeed lie sul t ise 1 l e uan
-oe-
-or-
Orrsite 00s0 Due
to the Wisole Body
j
Ae y f ee s t la t t en1 E ve?nt
If en. I n nesed 11gesid
De q. t a nnod liquid
to rvent is
or S rem to the
Projected to rwcoed
Cielld thyrold.

seecase les cwnss
solosso results in
110 enrem to W1eolo
""'"#

e. f ill:: Ilmi t s.
einwns t e oam cosemuni ty
llody cf Chi ld t hyrold.
wales endlosctivity
staes iolog ica l e r r i escent
g e c.e t o r than 12 times
corresponds to grea ter
l l' A s t a tedn e d s,
than 125 mitem/ter, wieule
~
T@l
taeee's doso rate or
600 settess/ter. chiid
see i ce.. e u s leess ut
l em l Itasedi leen Acci-
theJor Damage to Spent
tleysold at tieo sito
.
e .e.e e t e n i os Itmole.netese
sinn, etn see l t l e.9 in
luol Duo to Fool
1:oesoda ry,
sletce tal Wittelee sim
I'" l o n so or fladioactiv-
Ilandling Accident
ssane,
fly ten occupled
-or-
Asces Sucle lhet the
Uncontrolled Decrease
1eleort findistion
l *e luol Pool Water
l ove 16 Ise the Areas
to Bolow Lovel
lesee caso ley a f actor
or fuel,
of
1000
-ur-
tit im e Veelfled, Uncose-
I
teolled tvents Whee.Is
l'eseelt les noe Useewpected
s nc e r -e .se ear i n-I l ant
Inf e e t le.eill se t I nse ievels
1. ) es l ate teer of
100d.
_____
TAO.2
._. _.
.
.
_
- -
-
- -
.
. r . 't l" " .
i
.
te
riu^ imi ^sstoc^ mon or (nutcuicy Co'*olitolls
w
b!
SITE AREA
GENERAL
. . SATIN G
UNU
r
CONDITION
g'"" * "
EMERGENCY
EMERGENCY
"
- _
. . -
.m
. . . . .
. . .
W:o7N &p.".'p " M,
toss or 2 or 3 rission
Initi. tin.. or fres
valid Savoiy Clicult
- , "" O M
MFf.x : ;,d" Q f.g
.Q *;
I roduct Darriers Wi th a
. 7;
~
, g g.,G ,
Is Ip or tiecessas y
.;
-
5
.f
O ,;
Potential toss of Third
ttn eena l l ea l t l a t line.
.
. . , ,
.
. , . ; 7 . ./ -
n.
s c,v
TAE 1L
ew=j-
"
+tu -
-
ea rri e r.
.an.m-f(:n n
g-d , 4 ,
f f fd
opt.Ilcable.t9_Any_!altia gng

it< *1 a.esemp Sa l zure
h
nes rua.p salinen
_
/
Indnd rotor) Inading
p~ wlp . ;,
;.*J .4 :
Lvent_put_l4av tcad to
TAE 12_
- lo
_
m cynintion.
.
m -"
rnoi r.Ilur.
-or-
'1( b '
f* *
Con t a l snmn t Pressure
Any ini tia ting Events, from
.'
..
I n .s
..I
c..ntain ent
itequi rlog Shistd..wie
- x
r
inteneley
l y 100
, < *? m, , , Q~ e a 't *
5 ac
fa's psig
Whatever Source that Makes
-*
!
nelease or Largo Amounts
h-
bpm
-*
z
.a
or nadioactivity in a
TAB 13
.__.2.m
-
Short ilme Probable *
y,f.' . 7,f 4cw$ % ,
i . . . . .
.
h y..d M Q
.- 7
, i-[ p
.j
f or E) ample:
i n .s er l ugleicce ed
Itcqui ring sleuliinwn
.
.
ggf.. ; ,n * j W w #g, rg .g.gf
.
2
t
. ;; -
p; p.;y e-
.
s.1ety n.
I1.e
by Ico
-
q,a .m.g
d
1.
LOCA With failure or [CCS.
r. ni cc i In.. le.tn.as
A d a * c
'e 'EWO o?2x C2 *>*r
TM M
.
2.
toc 4 with initiaisy Socc-
.
-"
'
-
J g'p.!f.% - s5 . gEpfym.4-
estrul ECCS.
Subsequent
%.o
K@h
failure or lleat Reseovat
s. .c .h M ;-@M?i ~*jJ
unnetor tiot
l ailin e of Itcac t n e
systems with til.ely
-
1
suin:e l t ica l attor
re ni er t loes system
,.
7 > K.d d,g29 M c' 4.. ;y~g'.yl g
yf
f ailure or Conta irment.
Velid Scean
TQ:
an initintn or
nb
a9~
run. pinto a Scene
slquel(s).
.
9
3.
Loss or All Onsite and
';"4 ,' " h
,,'
orrsite rover Concorrent
'
TAB 15_
With Total Loss or
In'sest Calability
loss or Capability
Eme rgency f eedwa te r.
leets us risest
to Ae:hleve Cold
in Achinve Ifot
L net s o f /Sa f ot y
,
Lt.n t elnwe s
Shutdowse
Is . loss of reedwater and
~
Sy s t eins
Condensate rollowed by
TAS 16
_
reedwa ter System.
raiiur. or t.,e,g.,ncy
l'oss of All
loss of All
In.. ud lentira tne s,
loss on renross
A..senne l a t n e s or
ue Eiftncut Insm-
A l s u.s (Anunnelm-
A l m s ins
I's min
S.
Reactor Protect ion System
Ainems
nnters, licqists inne
tus ;) Sustalised for
with Plant flot in
rails to Initiate or
l
'fentalown by leu
. mins.
-
Cold S/D
Complete a pequired Ser m ,
-or-
rollowed by toss of Core
Plant Isansient
Coollog and Hake-up Sy ', t c M
Occurs While All
-or-
Ala reas a re lost.
Loss or Plant Control Occ a rt, .
TAB _17_
_ _
_ _.
,
t ..n; e n t I t n. .
Ite.gu t e nit .sr Anti-
licqis i s ed . Slus t down
..etent.
Coset e o g or
System Cisntrol at
.
s varesas lene
St.n .ilnwn Sys tenas
itevanto Shutdown Panel
-
" y[./
l a.t el.i l steud a t
tio t i s t .a ts l i shed
lena.ol o Shutdown
Within 15 sein.
""""
. .. . . . . TAB.18 -
.-
. . -
_
p* ,.
-
o
.
. .
.
.
( I'l / I P lepicoenst lug 1 s en:ottess e
(l' P / 8 - 8
Hoe nyeei t t ene snel I;1assil lra t ione
.
-
nl 8.cevancy con.Iltluns
! Aptt 8.1 - ^CI Dyl I t y[l, _ chi]QLA f 98_Cl ASSI f ICAIIOf( OF IM[lTC[f4rY COf401 TIOf(5
1
INITIATING
UNUSUAL
ALERT
SITE AREA
GENERAL
CONDITION
EVENT
EMERGENCY
EMERGENCY
I .s. sus a no
Iso s e o r one- mit a
~
henwn esamage tu
>o w n s o Isamago to
toss of 2 or 3 fission
t =selos toss rn e ceil l ee t
l en I l l t y, Affecting
Salo Steu tdowse
Product Da rriers Witte a
E qui enieset.
Potential Loss of Ilo i rd
l'a mn is n
8'posation.
TAB _2L _ Sisseelficant
l
oa r rie r.
A s e e s ee r t
unn sise l Astivlsy
A l e c e .e l t ne fli s s i l o
t.e a sle Allects Vital
Appl {calsjo_tq_pny,8nigDJ gl
ownr fac48 sly
l a ne tilsa tnver Sons te
St s nectur ns ley
L v e s.e Lif ea t _s4a y t e a d 3.o
-os-
!.te!Fue and Signifi-
Impact or f i re.
18:15 Qqngl_ tion.
Aleceert cessloon
rantly i r;rades a
-or-
(Pes s i t e
S t a t line Saloty
Any ini tia tlng Events, frue
Sl e nc t is e n.
%dlaa tever Source Elia t F1Jbos
Release of Large Amounts
TAE 28'
or nadioactivity in a
". g, . ~
Slio r t I 4 me Protsab lo,
lealn
liceallmont in nuslen
s
3.a g:p+ c
y. i.y, gf..*N ',,'
f or Esample:
A e s.e s
.r-
v
~ .
"h
8 ? f.
- r"
1.
(_OCA Witte Failure or [CCS.
m@' s Cf;*f -
N} ' WM*a cT;
,4
TAB
29
~ u
'
o
~
is@N ' Z'.o" V,
2.
LOCA Witle initially Succ-
@~%e[Na'a',l'
M
u F N 4*y*f' >
I.
.
essful [CCS.
Suleseque .t
61.e l e e s s n i t
S t r il<c s l es t alio

m
. f d ,, c.uM A..
4
'
Systems witie t.llcIy
5 t s uc tose en, lioso f t le.q
Q;o p*. s g
[k '
TAB
29
.
~ :'c ;,*.* ./ AM 1.7 ' ? W %u4$g* > ' *g' ;$.:
/
Failure or lleat Remov.e l
, ,
in i Iow fledeset leen
'
a -c
m
d +.e
ra nure or Conta ise.ent.
e ent me sea s.~s
-
I s asespue tss t iene oi
3.
toss of All Onsite arid
. up.n+gfg .gy ; ',;
',r/ t v' m m W'iY
/'J
'm
d, . g- T % '
. .
iQ'4
Of fsite Power Concurrent
lee lne y
lesJnseni and Cuntam-
~ J,:4F' g 4 :
. l
.
4
-
Wi ti, totai toss or
lesa t ed Ind a v lelua ll s )
to offsito linspflat.
' . 6.,- a. g7;-
' [R ,,$*[-[nO
[mergency Feedsater.
d
-
-
'

TAB
29
-
-
s-
-
--
-
.
e
- h M .p g ..
Acc lets nt at s'.it r
lacqui s es re ns et t iva

-
I,M
84 . Loss of Icedsater anJ
.3+b y-
. M. yp g;@f g' . f. v -
'
q.7 7 ,;
Corydensate followed Isy
gg
..g A

-
Fasture of foergency
Actions at liv PS
.g .. ,
4. . xj _ --
a.
TAB
29.
..
,
.mw
reedwat.c syste..
,, .jg*g(
nes : s pe+ s e no
Huptune eel 8 Isen-

T i
.Y '
N
5.
scactor Protection systen
. av1
'
s
,
'
f
K't
$
--
n a .{(,[{
Falls to initiate or
H q.t ese n
18 eso Onsi te w/
. > M f $,'{ s '., ,-
'; p 3 89% - ?G
. .//e
m. 4 y <( e e -

!, j ' .,
"t
M,
Complete a Required Scrai.
or w/o IIee
Jn -
-
29
H
.
, . ,g
.
,
.
N
Followed by Loss or Core
,
,
a..
.
.
lesel.seen lesep s es e n
l es e le l un entalin3
t en t. lsen failure
--!*-
-
,s.. . s
Cooling and Pialae-up Systcss
. A y.7;
.'
-
-or-
t ermseonoset f .e l l ess o
e ssessinq cas tseg
-
Loss of Planet Control Occurs.
o n ess l eng s a p f el ge l se n t
peesio t e n t ion
. ,
'
1 4 "c.' "* '
'
sinstelnwse.
B
29
_

-
.
& M W^~0(WT'
., mmm
.,,.
.
-
.
11C"T uR PCM 7
j
v t..
-
e
./
. .
.
.
.
.
..
.
.
_
.
.
_.
_
_
,
.
.
.
,
h
.
QUESTION 8.02
(3.00)

,
,
l
Using B.V.P.S. - Unit 2 Technical Specification, list ALL applicable
'
action statements, by number, for EACH of the following equipment
,
]
failures.
Consider EACH failure independently.-
,
!
a.
The fuel oil transfer pump for Diesel generator 21 has been found
'
to be inoperable. A reactor startup is in-progress with reactor
j
power at 1% and increasing,
i
b.
RHS Heat Exchanger outlet thermocouples TE606A and B, have been
l
found to be inoperable.
c.
Control room bottled air system pressure is found to be at 1500.
,
psig.
.
E
ANSWER 8.02
(3.00)
!
a.
3.8.1.1 (A.C. Sobrces) (0.50) AND 3.04 (cannot continue startup
>

since you cannot change modes by relying on action statements

(0.50)
!
b.
3.3.3.5.
(remote shutdown monitoring) (1.00)
'
c.
3.7.7.1.b (control room habitability; 4.7.7.2.a specifies pressure

3
requirement of 1825 psig) (1.00)
]

].
REFERENCE
l
Enabling Objectives UNAVAILABLE

B.V.P.S. - Unit 2 Technical Specifications
-
B.V.P.S. - 0.M. 2 page 2.10.1
K/A 06200 G0.05 3.8
.
K/A 016000 G0.05 3.5
!
016000G005
062000G005
..(KA's)
i.
!
COMMENT:
!
j
8.02.a

The question asks to list ALL applicable action statements for the
i
equipment failures listed.
Part a of the question states that a fuel

oil transfer pump for Diesel Generator 21 had been found inoperable. The
j
answer stated that the Diesel Generator was inoperable by T.S. 3.8.1.1
1
4
due to part b.3.
This is incorrect.
The Diesel Generator is still
operable, since it has 2 fuel oil transfer pumps and would still meet the
requirements of the Technical Specification LCO.
Therefore, the correct
answer is - NONE - No Technical Specifications applicable. (See attached
i
,
j
references.)
'
i
!
-
--
. , , -
- ~,
.
.
.
1
.
-
li . V . P . S . - 0.M.
2.36.1
i
MAJOR COMPONTNTS
1
Bore and stroke (inches)
15.7% x 16.11
Total displacement In. (3)
42,324
Brake horsepower
5,899
operating speed, RPM
514
Alare protection energized, RPM
360
Ccmpressien ratio
13:1
'
Lub.; c:. system capacity, Gal.
,- .>
Cm!:ng system capacity, Ga:
c '. c
tclosed leep)
ci. day ta s . Gal.
1,.
.
Fe. 4
'.s : : s t art in;; sys tem
supply pressure, TSIG minicum
2;U
The two emergency units are located in the Diesel Generator Building and
I
are physically and electrically isolated f ree each cther.
Each unit
is capable of carrying the
required emergency lead en its
respective bus during step leading and steacy state
follcwing a
loss of
preferred AC pcwor to the 4 KV eeergency buses.
~
Diesel engine supporting systems ccnsist of a fuel cil system, a starting
system, a cooling system, a lubricating oil
syster,
a turbocharger,
and
engine tretective devices.
Each emergency diesel
generator
is equipped with an overspeed governor
which shuts off the injection of fuel to the cylinders
when the engine
,
exceeds a speed of 565 to 576 rpm.
(
)
'
Fuel Oil System
A&ss &
e
M
1:sekase m % dilemmenneca,
als
W
h14tintYamfet'N, a 1,100 ga1len
!
e er
diesel generater iuel oil day tank, an engine driven fuel pump,
i
and an elect:1c driven fuel priming pump.
'
l
l
-5-
!$2s:
FEV J
t
. . - _ _ .-
- - _ _ . , ,
. -
. - -
.
- - - - - - - - .
- - - - - -
- - - - . - - - - , - - - - - - -
l
kp
'
.
3/4.8 ELECTRICAL POWER SYSTEMS
(
3/4.8.1
A.C. SOURCES
.
OPERATING
LIMITING CONDITION FOR OPERATION
__
i
T iT1 3 As a minimum, the folic ing A.C. electrical power sources shall be
' OPERABLE:
a.
Two physically independent circuits between the offsite transmission
'
network and the onsite Class IE distribution system, and
% Two separate and independent diesel generators each with:
1.
Separate day tank containing a minimum of 350 gallons of fuel,
2.
A separate fuel storage system containing a minimum of 53,225
1
gallons of fuel,
E 3./ 1000erste .fue{$trQ giump T
3
4.
Lubricating oil storage containing a minimum total volume of
504 gallons of lubricating oil, and
5.
Capability to transfer lubricating oil from storage to the
diesel generator unit.
APPLICABILITY:
MODES 1, 2, 3 and 4.
!
ACTION:

i
a.
With either an offsite circuit or diesel generator of the above re-
quired A.C. electrical power sources inoperable, demonstrate the
OPERABILITY of the remaining A.C. sources by perfoming Surveillance
Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.5 within one hour and at
least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least two offsite cir-
cuits and two diesel generators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or
be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
b.
With one offsite circuit and one diesel generator of the above re-
1
quired A.C. electrical power sources inoperable, demonstrate the
OPERABILITY of the remaining A.C. sources by perfoming Surveillance
Requirements 4.8.1.1.1.a and 4.8.1.1.2.a.5 within one hour and at
least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the in-
operable sources to OPERA 8LE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or b6 in COLD
SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
Restore at least two offsite
.
circuits and two diesel generators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 
i
i
from the time of initial loss or be in COLD SHUTDOWN within the next
'
36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
.
BEAVER VALLEY - UNIT 2
3/4 8-1
'
__
_
.
.
_ _ _ _
-
_
_-
<
.
.
I
ELECTRICAL POWER SYSTEMS
{
LIMITING CONDITION FOR OPERATION (Continued)
-
c.
With two of the above required offsite A.C. circuits inoperable,
demonstrate the OPERABILITY of two diesel generators by performing
Surveillance Requirements 4.8.1.1.2.a.5 within one hour and at least
once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter, unless the diesel generators are already
operating; restore at least one of the inoperable offsite sources to
OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STANDBY within
the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With only one offsite source restored, restore at
least two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from
time of initial loss or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
d.
With two of the above required diesel generators inoperable, demon-
strate the OPERABILITY of two offsite A.C. circuits by performing
Surveillance Requirement 4.8.1.1.1.a within one hour and at least
once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the inoperable
diesel generators to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in COLD
SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
Restore at least two diesel gen-
erators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss
or be in COLD SHUTOOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .
SURVELLLaNCE REOUIREMENTS
4.8.1.1.1.
Two physically independent circuits between the offsite trans-
mission network and the ensite Class 1E distribution system shall be:
a.
Determine OPERABLE at least once per 7 days by verifying correct
breaker aligneent, indicated power availability, and
b.
Demonstrated OPERABLE at least once per 18 months c; transferring
(manually and automatically) unit power supply frotu .he unit circuit
to the system circuit.
4.8.1.1.2
Each diesel generator shall be demonstrated OPERABLE:
a.
At least once per 31 days on a STAGGERED TEST BASIS by:
1.
Verifying the fuel level in the day tank,
2.
Verifying the fuel level in the fuel storage tank,
j
.
BEAVER VALLEY - UNIT 2
3/4 8-2
.
.
.
kD
.
-
ELECTRICAL POWER SYSTEMS
,
SURVEILLANCE REOUIREMENTS (Continued)
-
3.
Verifying that a sample of diesel fuel from the fuel storage
tank is within the acceptable limits specified in Table 1 of
ASTM 0975 when checked for viscosity, water and sediment,
r.pm ,*v y, , , gp u -
p p wywN
'y
.
,
,
hM,A
L am
n, .
s- -
nua ..
..
-
5.
Verifying the diesel starts from ambient condition,
6.
Verifying the generator is synchronized, loaded to > 4,238 kw,
and operates for at least 60 minutes, and
7.
Verifying the diesel generator is aligned to provide standby
power to the associated emergency busses.
S.
Verifying the lubricating oil inventory in storage.
b.
At least once per 18 months during shutdown by:
1.
Subjecting the diesel to an inspection in accordance with pro-
cedures prepared in conjunction with its manufacturer's recom-
mendations for this class of standby service,
,
2.
Verifying the generator capability to reject a load of g 825 kw
without tripping,
3.
Sirulating a loss of offsite power in conjunction with a safety
injection signal, and:
,
a)
Verifying de-energization of the emergency busses and load
shedding from the emergency busses.
i
b)
Verifying the diesel starts from ambient condition on the
auto-start signal, energizes the emergency busses with per-
manently connected loads, energizes the auto-connected
emergency loads through the load sequencer and operates for
> 5 minutes while its generator is loaded with the emergency
Toads.
4.
Verifying that on a loss of power to the emergency busses, all
diesel generator trips, except engine overspeed, generator
,
differential current, and generator overexcitation are
automatically disabled.

'
l
5.
Verifying the diesel generator operates for at least 60 minutes
while loaded to 1 4,238 kw.
'
BEAVER VALLEY - UNIT 2
3/4 8-3
_
_ _ . - - . _ _ _
___
.
.
.
4
.
$
.
ELECTRICAL POWER SYSTEMS
SURVEILLANCE REQUIREMENTS (Continued 1
-
6.
Verifying that the auto-connected loads to each diesel generator
do not exceed the 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months rating of 4.535 kw.
7.
Verifying that the automatic loao sequence timer is OPERABLE with
each load sequence time within 110% of its required value.
c.
Check for and remove accumulated water:
1.
From the day tank, at least once per 31 days and etter each
operation of the diesel where the period of operation was
greater than I hour, and
2.
From the fuel oil storage tank, at least once per 92 days,
d.
At least once per 92 days and from new fuel oil prior to its addi-
tien to the Storage tanks by verifying that a sa ple obtained in
accordance with ASTM 0270-1975 meets the following minimum require-
.
ments anc is tested within the specified time limits:
1.
As soon as a sample is taken (or prior to accing new fuel to
the storage tank) verify in accordance with the tests specified
- I
in ASTM D975-1977 that the sample has:
a)
A water and sediment content of less than or equal to 0.05
volume percent;
b)
A kinematic viscosity at 40'C of greater than or equal to
1.9 centistokes, but less than or equal to 4.1 centistokes;

c)
An API Gravity of within 0.3 degrees of 60*F, or a specific
gravity of within 0.0016 at 60/60'F, when compared to the
supplier's certificate or an absolute specific gravity at
60/60*F of greater than or equal to 0.83 but less than or
equal to 0.89, or an API Gravity of greater than or equal
}
to 27 degrees but less than or equal to 39 degrees; and
2.
Within one week after obtaining the sample, verify an impurity
level of less than 2 milligrams of insolubles per 100 milliliter
is met when tested in accordance with ASTM D2274-1970.
The
analysis on the sample may be performed after the addition of
new fuel oil.
3.
Within two weeks of obtaining the sample, verify th&t the other
properties specified in Table 1 of ASTM 0975-1977 and Regulatory
Guide 1.137 Position 2.a are met (when tested in accordance with
ASTM 0975-1977).
An analysis for sulfur shall be performed
in accordance with ASTM 0129, ASTM D1552-1979 or ASTM D2622-1982.
BEAVER VALLEY - UNIT 2
3/4 8-4
_
'
.
.
.
ELECTRICAL POWER SYSTEMS
(
SURVEILLANCE REOUIREWENTS (Continuedi
.
e.
At least once per 10 years or efts * any modifications which
could affect diesel generator it
rdependence by starting ** both
diesel generators simultaneously, during shutdown, and verifying
that both diesel generators accelerate to at least 514 rpe in less
than or equal to 10 seconds.
.
f.
At least once per 10 years by:
1)
Oraining each main fuel oil storage tank, removing the accumu-
lated sediment, and cleaning the tank using a sodium hypochlorite
solution or other appropriate cleaning solution, and
2)
Perfcrmitig a pressure test, of those portions of the diesel fuel
oil system designed to Section III, subsection ND of the ASME
Code, at a test pressure equal to 110% of the syste'a design
pressure.
,
,
!
,

l
i
I
.
l
This test shall be conducted in accordance with the manufacturer's recommen-
dations regarding engine prelube and warmup procedures, and as applicable
regarding leading recommendations.
4
i
BEAVER VALLEY - UNIT 2
3/4 8-5
-- -
-
-
..
. _
.
_ _
_.
. _
._.
_ _ _ _
.
.
<
i
,
,
i
!

'
i
QUESil0N 8.09 (2.25)
i
i
'

Use B.V.P.S. - Unit 2 Technical Specification Table 3,3 6 and determine
,
WHAT SEVEN (7) Area or Process radiation monitoring instruments must be

j
functional following a LOCA.
1

!
ANSWER 8.09 (2.25)
j
2RMR RQ205,206 (Containment Area (0.50)
(

2HYS RQ109C (Mid Range Noble Gas) (0.50)
t
2HVS RQ109D (High Range Noble Gas) (0.50)
-
,

2 MSS-RQ101A,B,1C (Main Steam Discharge (0.75)
,
'
!
,
j
REFERENCE
i
B.V.P.S. 2LP-SQS-43.1 Enabling Objective 4
B.V.P.S. - Unit 2 Technical Specifications Table 3.3 6 Action 36
!
K/A 016000 G0.04 3.4
-
1
016000G004
..(KA's)
,
COMMENT:
-
4
8.09

The question asks to determine what SEVEN (7) Area or Process radiation

monitoring instruments must be functional following a LOCA (using Table
3.3-6).
However, Table 3.3-6 does not address a condition of
'
<
operability, for the raotation monitors listed, following a LOCA.
They
must only be OPERABLE per their applicable modes as outlined in the

t abl e.
However, it may be interpreted, that with the conditions stated

.
l
(i.e., use of Table 3.3 6 and following a LOCA), the plant could possibly
E
be in Modes 1 4.
At that time, all 11 of the following radiation
!
monitors must be operable:
!
I
'
Containment Area (2RMR RQ206,207)
-
Control Room Area (2P30-RQ201,202)

j
-
RCS Leakage Detection Gas & Particulate (2RMR RQ303A,B)

i
-
Mid Range Noble Gas (2HVS-RQ109C)
j
-
,
High Range Noble Gas (2HVS RQ1090)
J
-
i
Main Steam Discharge (2 MSS RQ101A,B&C)
-
,
.

Therefore, the correct answer should be (n.y 7 of the 11 radiation
{
monitors listed above.
(See attached references.)
!
i
i
4
s
i
i
I
]

i
i
a
I
. .
.
. . ,
0
' Y ./ k. ' '
.
.

,
1
-
.
w
.
INSTRUMENTATION
3/4.3.3 MONITORING INSTRUMENTATION
RADIATION MONITORING
LIMITING CONDITION FOR OPERATION
. $ % %.*%,*',':M W .0.^ M M 's,th w'<'A'_ & ' W
l' '**5 %& . '
..
'
'~
~ m l _L % : : '
-- -

=-.--1
n, :. . w s .
.
..
..
.MPLICA31%- Esti$i45157A.n J.F6[
~
W
..
-
-- n
.
ACTION:
a.
With a radiation monitoring channel alarm / trip setpoint exceeding the
value shown in Table 3.3-6, adjust the setpoint to within the limit
within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.
oring# 5aYnI [irioperabli'."t&helt.hi '
' 6. ' , -
de
n
c
w n w
r~.,.+nwe.e.wk
,
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
_ SURVEILLANCE REOUIREMENTS
4.3.3.1
Each radiation monitoring instrumentation channel shall be demonstrated
OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and
CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies
shown in Table 4.3-3.
.
een,co uniev . nun o
2,a 2.so
. - -
-- -
_
_
.-
-_ -
.
-
.
.
.
.
. _ .
-
_ _ _ _ _ _ _ _ _ _ _
..
.. ..
.
.
.

.
to
m
. -
E
RADIATION MONITORING INSTRtl MENTATION
m
<
MINIMUM
d'
CilANNEls APPLICABLE
MEASUREMENT
E
IN!TRUMENT
OPERABLE
MODES
SETPOINI
RANGE
ACTION

1.
ARIA MONITORS
C
l
a.
Fitel Storage Poal Area
1
< 75.8 mR/hr
10 1 to 104 mR/hr'
19

(2RMF-RQ202)
l
~
R/hr l'.to 10 'lhbruf B M h 9
2
n,
<T3.29IiO3
7
r
-
.
-
--
-
,
T O) 76[mR/hr 3 5 to?,1
2.
PROCESS MONITORS
. . ,
2
a.
Containment
y
. . , .
.
7-
-- ~ z

..
_
ii.
-
ggc ?

' NS$Y$Y1PR[
.
'
b.
Fuel Building Vent
-
1.
Gaseous Activity (Xe-133) 1
<7.82x10 6 pCi/cc 10 6 to 10 1 pCi/cc 21
"
(2RMF.-RQ3010)
~
-
With fuel in the storage pool or building
With irradiated fuel in the storage pool
Above background
Doring movement of irradiated fuel
4
,
I
a
-
_
_ _ _ _
_ _ _
- _ _ - - _ _ _ _ _
.

~
"
.
RADIATION MONITORING INSTRUMENTATION
l
'
-
MINIMUM
2
CHANNELS APPLICABLE
MEASUREMENT

}.
INSTRUMENT
OPERAME
MODES
SETPOINT
RANGE
ACTION
, 2. PROCfSS MONITORS (Continued)
.
i
j
i
ii.
Particulate (I-131)
1
<6.70x10M pCi/cc 10 80 to 10 5 pCi/cc 21

(2RMF-RQ301A)

3
.
,
c.
Noble Gas and Effluent Monitors
I
i.
Supplementary Lcak
i
Collection and Release
l
System
"
2A~
-
l
I)
h.,' .
'.T 5'JE 1 " N .10 [ $ % .
2)
,.
!
ii.
Containment Purge Exhaust 1
6
< 3 x background 10 8 to 10 2 pCi/cc 22
(Xe-133)(2HVR-RQ104A & B)
,
], yQ
%25[f[i01 tow
g
iii.
r
,

Above background
i
i
a
v
l
'
.
i
)
!
i
!
i
'

i

I
l
'}
y
o
f
.L
y-
,
,
(
TABLE 3.3-6 (Continued)
ACTION STATEMENTS
-
With the number of channels OPERABLE less than required by
ACTION 19
-
the Minimum Channels OPERABLE requirement, perfom area sur-
veys of the monitored area with portable monitoring
instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
ACTION 20
-
With the number of channels OPERABLE less than required by
the Minimum Channels OPERABLE requirement, comply with the
ACTION requirements of Specification 3.4.6.1.
With the number of channels OPERABLE less than required by the
ACTION 21
-
Minimum Channels OPEPABLE requirement, comply with the appli-
cable ACTION requirements of Specifications 3.9.12 and 3.9.13.
ACTION 22
-
With the number of channels OPERASLE less than required by
the Minimum Channels OPEPABLE requirement, comply with the
ACTION requirements of Specification 3.9.9.
ACTION 36
With the number of OPERABLE channels less than required by
-
'
the vinimum Channels OPERABLE requirement, either restore
the inoperable channel (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ,
or:

1)
Initiate the preplanned alternate method of monitoring
^
the appropriate parametar(s), an6
2)
Prepare and submit a Special Report to the Commission
pursuant to Specification 6.9.2 within the next 14 days
following the event outlining the cetion taken, the
cause of the inoperability and the plans ana schedule
-
'
for restoring the system to OPERABLE status.
ACTION 43
With the number of OPERABLE channels less than required by
-
the Minimum channels OPERABLE requirement, either restore
the inoperable channel (s) to uPEDABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ,
or:
'
1)
Initiate the preplanned alternate method of monitoring
the appropriate parameter (s), and
2)
Return the channel to OPERABLE status within 30 days or
explain in the next Semi-Annual Effluent Release Report
why the inoperability was not covered in a timely manner.
ACTION 46
With the number of OPERABLE channels one less than' required by
-
the Minimum Channels OPERABLE requirement, either restore the
inoperable channel to OPERABLE status w. thin 7 days Or close
the control room series normal air intake and exhaust isola-
tion dampers.
At. TION 47
With no OPEkABLE channels either restore one inoperable channel
-
to OPFOABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months cr close the centr 01 rcom
series nomal air intake and exhaust isolation dampers.
AF AVFD VAll FY - IINIT ?
3/4 3-49
_
.
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.
a
QUESTION 8.11
(1.25)
a.
WHAT is the FULL Technical Specification Basis for the RCS
operational leakage limit stated in 3.4.6.2c?
(0.80)
b.
WHAT Technical Specification (state by number) addresses the
surveillance program established to prevent the leakage limits in
3.4.6.2c. from becoming an operational concern?
1
ANSWER 8.11
(1.25)
a.
Ensure that the dosage contribution (0.20) from the tube leakage
will be limited to a small fraction of the 10 CFR Part 100 limits
(0.20) in the event of either a steam generator tube rupture (0.20)
or a steam line break (0.20)
b.
3/4.4.5 (0.45)
REFERENCE
Enabling Objective UNAVAILABLE
B.V.P.S. - Unit 2 Technical Specifications Section 3/4.4.6.2, 3/4.4.5
-
K/A 000037 G0.04 3.9
000037G004
..(KA's)
COMMENT:
8.ll.a
The question asks to state the FULL Technical Spec 3fication Basis for the
RCS operational leakage limit stated in 3.4.6.2c.
The answer given is
stated as it appears in the Bases Section and assigns point values for
certain portions of the statement.
By using the word FULL in the
question, it implies that only this answer is acceptable.
It is
requested that other answers be accepted which convey the intent of the
basis even though it may not be written exactly as it appears in the

BVPS-2 Technical Specifications.
(See attached references.)
<
i
I
.i
.
.
.
. .
.
.
I
.
l
'
REACTOR COOLANT SYSTEM
I
OPERATIONAL LEAKAGE
LIMITING CONDITION FOR OPERATION
Reactor Coolant System leakage shall be limited to:
a.
No PRESSURE BOUNDARY LEAKAGE,
b.
1 GPM UNIDENTIFIED LEAKAGE,
j
E
d.
10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and
e.
28 GPM CONTROLLEC LFMAGE at a Reactor Coolant System pressure of
2235 2 20 psig.
APPLICABILITY:
MODES 1, 2, 3, and 4.
-
ACTION:
l
a.
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY
(
within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,
b.
With any Reactor Coolant System leakage greater than any one of the
above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage
rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at_,least HOT STANDBY
~ within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following
~30 hours.
SURVEILLANCE RE0_VIREMENTS
4.4.6.2
Reactor Coolant System leakages shall be demonstrated to be within
each of the above limits by:
a.
Monitoring the containment atmosphere particulate and gaseous
radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,
b.
Monitoring the containment sump discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
c.
Measurement of the CONTROLLED LEAKAGE to the reactor coolant pump
seals when the Reactor Coolant System pressure is 2235 1 20 psig
at least once per 31 days with the modulating valve full open.
d.
Performance of a Reactor Coolant System water inventory balance at
l
least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during steady state operation, and
(
BEAVER VALLEY - UNIT 2
3/4 4-19

- _
__ _
_ _ _ . _ ,
. - .
. .
. .
__
__
.
. .
. _ . _
.
y'..^ju . . '
REACTOR COOLANT SYSTEM
BASES
.
3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE
3/4.4.6.1 LEAKAGE DETECTION SYSTEMS
The RCS leakage detection systems required by this specification are pro-
vided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.
j
These detectiun systems are consistent with the recommendations of Regulatory
'
l
Guide 1.45, "Reactor Coolant Pressure Boundary leakage Detection Systems."
3/4.4.6.2 OPERATIONAL LEAKAGE
,
1
Industry experience has shown that while a limited amount of leakage is
expected from the RC5, the unidentified portion of this leakage can be reduced
to a threshold value of less than 1 GPM.
This threshold value is sufficiently
low to ensure early detection of additional leakage.
The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited
amount of leakage from kno n sources whose presence will not interfere with the .
detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
The CONTROLLED LEAKAGE limitation restricts operation when the total flow
supplied to the reactor coolant pump seals exceeds 28 GPM with the modulating
valve in the supply line fully open at RCS pressures in excess of 2235 psig.
This limitation ensures that in the event of a LOCA, the safety injection flow
,
will not be less than assumed in the accident analyses.
PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptablo since it may
be indicative of an impending gross failure of the pressure boundary.
Should
PRESSURE BOUNDARY LEAKAGE occur through a component which can be isolated from
the balance of the Reactor Coolant System, plant operation say continue
provided the leaking component is promptly isolated from the Reactor Coolant
System since isolation removes the source of potential failure.
-
3/4.4.6.3 PRESSURE ISOLATION VALVE LEAKAGE
The leakage from any RCS pressure isolation valve is sufficiently low to
ensure early detection of possibic in-series valve failure.
It is apparent
that when pressure isolation is provided by two in-series valves and when
failure of one valve in the pair can go undetected for a substantial length of
time. verification of valve integrity is required.
Since these valves are
BEAVER VALLEY - UNIT 2
B 3/4 4-4
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i
ATTACHMENT 3
NRC Response to Facility Examination Review Comments

SECTION 5
'
5.03c.:
Comment accepted.
5.05a.:
Comment accepted. The question was deleted from the examination,
reducing the value of Section 5 by 0.50 points.
,
,
5.05b.:
Comment accepted.
The question was deleted from the examination,
reducing the value for Section 5 by 0.50 points.
5.06:
Comment accepted.
5.08:
Comment not accepted. Both of the enabling objectives require the
candidate to fully understand xenon oscillations.
LP-RT-7 Enabling
Objectives (E0) 5 states.
"Describe Xenon oscillations," and EO 6
states, "Discuss Xenon oscillation dampening at both BOL and E0L."
The question war not intended to measure the candidate's ability to
control xenon oscillations, which is not addressed by the enabling
objectives, but s'.mply to measure his understanding of WHEN and HOW
xenon oscillations could affect plant operations.
5.09b.:
Comment noted.
'
5.09c.:
Comment accepted.
t
SECTION 6
6.01f.:
Comment accepted.
The question was deleted because the question did
.
not accurately describe the indications available to an operator at
BVPS, thus resulting in the possibility for more than one correct
answer. The value of Section 6 was reduced by 0.50 points.

6.03:
Comment noted.
.
6.04:
Comment not accepted. The calibration curves show that if the delta
P correction for a hot calibrated and cold calibrated instrument are
the same at 0% level, then the hot calibrated instrument, because of
its lower fluid density, needs more delta P correction than the cold
calibrated instrument to balance the reference leg pressure at 100%.
6.05b.:
Comment accepted.
In addition, the applicable valves names were
l
added to the answer key.
6.10:
Comment noted.
j
1
. . -
.- .
-. , - .
. . .
_
.
._.
. . - _ ..
.
-
._
a
-
.
2
RESPONSE TO SECTION SEVEN GENERAL COMMENT
NRC NUREG-1021, "Operator Licensing Examiner Standards," Chapter ES-202 and
ES-402 state that each candidate must demonstrate complete knowledge and
understanding of symptoms, automatic. actions and immediate action steps
specified in abnormal and emergency procedures.
For a candidate to demonstrate
a complete knowledge and understanding of abnormal and emergency procedures, he
must be able to recognize an abnormal condition, and be able to take knowledge
based actions to mitigate the con:equence of the abnormal condition. This is
the position stated by Beaver Valley Power Station in its letter from J. D.
Sieber to S. J. Collins, of the NRC, dated July 28, 1987.
Section 7 of this
examination contains questions which require the candidate to demonstrate his
knowledge of abnormal procedures by examining his knowledge and understanding
of inter-system relationships and time dependent, knowledge based operator
actions.
SECTION 7
7.01:
Comment not accepted.
This question calls for the candidate to
recognize an abnormal condition ("minimum RCS flow requirements
CANNOT be met while in Mode 4") and to demonstrate the necessary
knowledge to proceed with the mitigation of the consequences of the
abnormal condition (knowing the correct procedure to implement,
Emergency Operating Procedure (EOP) ES-0.2, "Natural Circulation
Cooldown"). The facility's Lesson Plan (LP) for the E0Ps does not
address the requirement for an operator to know the conditions for
entry into the E0Ps. Without a facility LP enabling objective, the
K/A catalog rating is used to determine the safety significance of
,
the question.
The K/A catalog importance rating for this question
is 3.9 (on a scale of 1.0 to 5.0).
,
i
7.03:
Comment not accepted.
The comment presented by the facility for
parts b. and c. of this question does not address the issue
concerning memorization of knowledge based operator actions. A
senior operator who does not possess the prerequisite knowledge to
direct an operator to Emergency Borate, does not possess the required
level of knowledge necessary to ensure the plant will remain within
its design basis envelope. Additionally, an operator who could not
,
recognize the need to manually trip the reactor, under the conditions
of the question, would not possess the. level of knowledge required to
ensure safe operation of the plant.
7.06a.:
Comment noted.
7.06c.:
Comment not accepted. The knowledge level required to ensure safe
operation of the plant should encompass the immediate actions an
operator would take to place the plant into a safe condition (ie.
manually trip the reactor).
i.
,
n -
-
-
..
b
.
.
3
7.07:
Comment noted. The answer key has been modified as follows:
1 - Accept [0.25] since he.is allowed 1250 mrem /qtr [0.50]
2
... AS PER ANSWER KEY ... [0.75]
3
... AS_PER ANSWER KEY ... [0.75]
4 - Reject [0.25] since he has already exceeded his whole body _
limit of 10000 mrem lifetime exposure [ exposure [0.50]
This revision resulted in a reduction in the value of_Section 7 by
0.50 points.
7.08:
Comment not accepted. The question addresses system design
interactions and relationships.
An operator should not need to
consult the A0P if his level of knowledge met the requirements of the
facility training program, as discussed in the lesson plan
referanced in the answer key (Attachment 1) regarding each of the
. systems affected by a loss of 120 VAC Vital Bus 1.
7.09:
Comment accepted.
,
7.10a.:
Comment not accepted, per the AOP, the loss of a steam generator feed
a
pump requires that prompt, time dependent operator action to be
taken.
If an operator possessed an acceptable level of knowledge
concerning SG water level control and feedwater system operation, he
would be able to correctly identify the need to place the SG startup
feedwater pump into service.
However, after further review of the
question and its answer, the following alternate correct response was
identified:
-
"reduce power to within the capacity of 1 MFWP"
This response was added to the answer key.
7.11a.
Comment accepted. The question was deleted reducing the value of
Section 7 by 0.50 points.
SECTION 8
8.07:
Comment noted.
'
B.02a.:
Comment accepted. The answer key was modified and the value of
Section 3 reduced by 0.50 points.
8.09:
Comment not accepted. Only the seven (7) monitors listed in the
answer key have the required range to allow for their operation in a
post-LOCA environment.
8.11:
Comment noted.

1
.-
.
.
.
.
-
..
. -
-
- . - _ - - .
b --
8
r
i
ATTACHMENT 3
NRC Response to Facility Examination Review Comments
SECTION 5
5.03c.:
Comment accepted.
5.05a.:
-Comment accepted. The question was deleted-from the examination,
r educing the value of Section 5 by 0.50 point s.
5.05b.:
Comment accepted.
The question was deleted from the examination,
reducing the value for Section 5 by 0.50 points.
5.06:
Comment accepted.
5.08:
Comment not accepted. Both of the enabling objectives require the
candidate to fully understand xenon oscillations.
LP-RT-7 Enabling
Objectives (EO) 5 states.
"Describe Xenon oscillations," and EO 6
states, "Discuss Xenon oscillation dampening at both'BOL and E0L."
The question was not intended to measure the candidate's ability to
,
control xenon oscillations, which is not addressed by the enabling
objectives, but simply to measure.his understanding of WHEN and HOW
xenon oscillations could affect plant operations,
j
5.09b.:
Comment noted.
5.09c.:
Comment accepted.
SECTION 6
!
6.01f.:
Comment accepted.
The question was deleted because the question did
not accurately describe the indications available to an operator at
BVPS, thus resulting in the possibility for more than one correct
answer. The value of Section 6 was reduced by 0.50 points.
6.03:
Comment noted.
6.04:
Comment not accepted. The calibration curves show that if the delta
P correction for a hot calibrated and cold calibrated instrument are
the same at 0% level, then the hot calibrated instrument, because of
its lower fluid density, needs more delta P correction than the cold
calibrated instrument to balance the reference leg pressure at 100%.
6.05b.:
Comment accepted.
In addition, the applicable valves names were
added to the answer key.
4
6.10:
Comment noted.
-
-
1
- - _ ,
.
_ , - - _ _ _ _
_
. _ _ . , - . . _ , , _ _ _ _ . _ . ~ _ . . - _ - . _ _ _ _ , _ . _ . . _ , . . - , . - . _ _ - - - , . , _ . . . _ . - _
,
- -
-
.
-
.
-
.
b
.
.,
2
RESPONSE TO SECTION SEVEN GENERAL COMMENT
NRC NUREG-1021, "Operator Licensing Examiner Standards," Chapter ES-202 and
ES-402 state that each candidate must demonstrate complete knowledge and
understanding of symptoms, automatic actions and immediate action steps
specified in abnormal.and emergency procedures.
For a candidate to demonstrate
a complete knowledge and understanding of abnormal and emergency procedures, he
must be able to recognize an abnormal condition, and be able to take knowledge
based actions to mitigate the consequence of the abnormal condition.
This is
the position stated by Beaver Valley Power Station in its letter from J. D.
Sieber to S. J. Collins, of the NRC, dated July 28, 1987.
Section 7 of this
examination contains questions which require the candidate to demonstrate his
knowledge of abnormal procedures by examining his knowledge and understanding
of inter-system relationships and time dependent, knowledge based operator
actions.
SECTION 7
7.01:
Comment not accepted. This question calls for the candidate to
recognize an abnormal condition ("minimum RCS flow requirements
CANNOT be met while in Mode 4") and to demonstrate the necessary
knowledge to proceed with the mitigation of the consequences of the
abnormal condition (knowing the correct procedure to implement,
Emergency Operating Procedure (E0P) ES-0.2, "Natural Circulation
Cooldown").
The facility's Lesson Plan (LP) for the E0Ps does not
address the requirement for an operator to know the conditions for
entry into the E0Ps. Without a facility LP enabling objective, the
K/A catalog rating is used to determine the importance of the
question. The K/A catalog importance rating for this question is 3.9
(on a scale of 1.0 to 5.0).
7.03:
Comment not accepted. The comment presented by the facility for
parts b. and c of this question does not address the issue
concerning memorization of knowledge based operator actions. A
senior operator should possess the required level of knowledge
necessary to ensure the plant will remain within its design basis
envelope by directing emergency boration. Additionally, an operator
should recognize the need to manually trip the reactor, under the
conditions of the question.
7.06a.:
Comment noted.
7.06c.:
Comment not accepted. The knowledge level required to ensure safe
operation of the plant should encompass the immediate actions an
operator would take to place the plant into a safe condition (ie.
manually trip the reactor).
,
.,.,n
r,
-,m.-
-
, - - . - - , , , . _ .
. - , , .
.v.,
, , . - , , , , , _ , - _ , . ,
n-.---_.,,n,,,
~
-o

.
3
7.07:
Comment noted. The answer key has been modified as follows:
1 - Accept [0.25] since.he is allowed 1250 mrem /qtr [0.50]
2
... AS PER ANSWER KEY ... [0.75]
3
... AS PER ANSWER KEY ... [0.75]-
.
4 - Reject [0.25] since he has already exceeded his whole body
limit of 10000 mrem lifetime exposure [ exposure [0.50]
This revision resulted in a reduction in the value of Section 7 by
0.50 points.
7.08:
Comment not accepted.
The question addresses. system design
interactions and relationships. An operator should not need to
consult the A0P if his level of knowledge met the requirements of the
facility training program, as discussed in'the lesson plan
referenced in the answer key (Attachment 1) regarding each of the
systems affected by a loss of 120 VAC Vital Bus 1.
7.09:
Comment accepted.
7.10a.:
Comment not accepted.
Per the A0P, the loss of'a steam generator
i
feed pump requires that prompt, time _ dependent operator action to be
taken.
Concerning SG water level control and feedwater system
operation, an operator should be able to correctly identify the need
to place the SG startup feedwater pump into service. However, after
further review of the question and its answer, the following
alternate correct response was identified:
"redur? power to within the capacity of 1 MFWP"
This response was added to the answer key.
7.11a.
Comment accepted.
The question was deleted reducing the value of
Section 7 by 0.50 points.
SECTION 8
8.01:
Comment noted.
8.02a.:
Comment accepted. The answer key was modified and the value of
Section 8 reduced by 0.50 points.
8.09:
Comment not accepted. 'Only the seven (7) monitors listed in the
answer key have the required range to allow for their operation in a
post-LOCA environment.
8.11:
Comment noted.

ML20154B226

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